es-401 bwr examination outline form es-401-1es-401 bwr examination outline form es-401-1 facility:...

ES-401 BWR Examination Outline Form ES-401-1

Facility: Hatch Date of Exam: July 2016 Tier Group

RO K/A Category Points SRO-Only Points

K1

K2

K3

K4

K5

K6

A1

A2

A3

A4

G* Total

A2 G* Total

1. Emergency &

Abnormal Plant Evolutions

1 3 3 3

N/A

4 3

N/A

4 20 4 3 7

2 1 1 1 1 2 1 7 2 1 3

Tier Totals 4 4 4 5 5 5 27 6 4 10

2. Plant

Systems

1 3 2 3 2 2 2 3 3 2 2 2 26 3 2 5

2 1 1 1 1 2 1 1 1 1 1 1 12 0 2 1 3

Tier Totals 4 3 4 3 4 3 4 4 3 3 3 38 5 3 8

3. Generic Knowledge and Abilities Categories

1 2 3 4 10 1 2 3 4 7

2 3 3 2 1 2 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and

SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the “Tier Totals” in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category.)

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by ±1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories. 7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be

relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As. 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics’ importance ratings

(IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

ES-401 2 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function K1

K2

K3

A1

A2 G* K/A Topic(s) IR #

295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4

X

X

AK3.04: Knowledge of the reasons for a Reactor SCRAM as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION AA2.02: Ability to determine and/or interpret Neutron monitoring as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION

3.4

3.2

295003 Partial or Complete Loss of AC / 6 X

X

AK2.03: Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C. POWER and the A.C. electrical distribution system AA2.04: Ability to determine and/or interpret System lineups as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER

3.7

3.7

295004 Partial or Total Loss of DC Pwr / 6 X

X

AA1.03: Ability to operate and/or monitor A.C. electrical distribution as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER G2.1.20: Ability to interpret and execute procedure steps as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER

3.4

4.6

295005 Main Turbine Generator Trip / 3 X

X

G2.2.37: Ability to determine operability and/or availability of safety related equipment as they apply to MAIN TURBINE GENERATOR TRIP G2.2.44: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions as they apply to MAIN TURBINE GENERATOR TRIP

3.6

4.4

295006 SCRAM / 1 X AK1.02: Knowledge of the operational implications of Shutdown margin as they apply to SCRAM

3.4

295016 Control Room Abandonment / 7 X AA2.01: Ability to determine and/or interpret Reactor power as they apply to CONTROL ROOM ABANDONMENT

4.1

295018 Partial or Total Loss of CCW / 8 X AA1.01: Ability to operate and/or monitor Backup systems as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER

3.3

295019 Partial or Total Loss of Inst. Air / 8 X

X

G2.1.32: Ability to explain and apply system limits and precautions as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR G2.4.41: Knowledge of the emergency action level thresholds and classifications as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR

3.8

4.6

295021 Loss of Shutdown Cooling / 4 X AA1.06: Ability to operate and/or monitor Containment/ drywell temperature as they apply to LOSS OF SHUTDOWN COOLING

2.8

295023 Refueling Acc / 8 X AK1.01: Knowledge of the operational implications of Radiation exposure hazards as they apply to REFUELING ACCIDENTS

3.6

295024 High Drywell Pressure / 5 X EK2.13: Knowledge of the interrelations between HIGH DRYWELL PRESSURE and Suppression pool spray

3.8

295025 High Reactor Pressure / 3 X EA2.01: Ability to determine and/or interpret Reactor pressure as they apply to HIGH REACTOR PRESSURE

4.3

295026 Suppression Pool High Water Temp. / 5

X EK3.05: Knowledge of the reasons for Reactor SCRAM as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE

3.9

295027 High Containment Temperature / 5

295028 High Drywell Temperature / 5 X G2.4.31: Knowledge of annunciator alarms, indications, or response procedures as they apply to HIGH DRYWELL TEMPERATURE

4.1

295030 Low Suppression Pool Wtr Lvl / 5 X

X

EK3.01: Knowledge of the reasons for Emergency depressurization as they apply to LOW SUPPRESSION POOL WATER LEVEL EA2.03: Ability to determine and/or interpret Reactor pressure as they apply to LOW SUPPRESSION POOL WATER LEVEL

3.8

3.9

295031 Reactor Low Water Level / 2 X

X

EK2.11: Knowledge of the interrelations between REACTOR LOW WATER LEVEL and Reactor protection system EA2.01: Ability to determine and/or interpret Reactor water level as they apply to REACTOR LOW WATER LEVEL

4.4

4.6

295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1

X EA1.10: Ability to operate and/or monitor Alternate boron injection methods as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN

3.7

295038 High Off-site Release Rate / 9 X EK1.02: Knowledge of the operational implications of Protection of the general public as they apply to HIGH OFF-SITE RELEASE RATE

4.2

600000 Plant Fire On Site / 8 X AA2.14: Ability to determine and interpret Equipment that will be affected by fire suppression activities in each zone as they apply to PLANT FIRE ON SITE

3.0

700000 Generator Voltage and Electric Grid Disturbances / 6

X G2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES

4.5

K/A Category Totals: 3 3 3 4 3/4 4/3 Group Point Total: 20/7

ES-401 3 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

E/APE # / Name / Safety Function K1

K2

K3

A1

A2 G* K/A Topic(s) IR #

295002 Loss of Main Condenser Vac / 3

295007 High Reactor Pressure / 3 X AA2.02: Ability to determine and/or interpret Reactor power as they apply to HIGH REACTOR PRESSURE 4.1

295008 High Reactor Water Level / 2 X AA2.02: Ability to determine and/or interpret Steam flow/feedflow mismatch as they apply to HIGH REACTOR WATER LEVEL

3.4

295009 Low Reactor Water Level / 2 X G2.4.30: Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator as they apply to LOW REACTOR WATER LEVEL

4.1

295010 High Drywell Pressure / 5 X AK3.01: Knowledge of the reasons for Drywell venting as they apply to HIGH DRYWELL PRESSURE 3.8

295011 High Containment Temp / 5

295012 High Drywell Temperature / 5

295013 High Suppression Pool Temp. / 5

295014 Inadvertent Reactivity Addition / 1

X

X G2.1.28: Knowledge of the purpose and function of major system components and controls as they apply to INADVERTENT REACTIVITY ADDITION AA2.05: Ability to determine and/or interpret Violation of safety limits as they apply to INADVERTENT REACTIVITY ADDITION

4.1

4.6

295015 Incomplete SCRAM / 1

295017 High Off-site Release Rate / 9 X AA1.05: Ability to operate and/or monitor SPDS/ERIS/CRIDS/GDS as they apply to HIGH OFF-SITE RELEASE RATE

2.7

295020 Inadvertent Cont. Isolation / 5 & 7 X AK2.01: Knowledge of the interrelations between INADVERTENT CONTAINMENT ISOLATION and the Main steam system

3.6

295022 Loss of CRD Pumps / 1

295029 High Suppression Pool Wtr Lvl / 5

295032 High Secondary Containment Area Temperature / 5

X EA2.01: Ability to determine and/or interpret Area temperature as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE

3.8

295033 High Secondary Containment Area Radiation Levels / 9

295034 Secondary Containment Ventilation High Radiation / 9

295035 Secondary Containment High Differential Pressure / 5

295036 Secondary Containment High Sump/Area Water Level / 5

500000 High CTMT Hydrogen Conc. / 5 X EK1.01: Knowledge of the operational implications of Containment integrity as they apply to HIGH CONTAINMENT HYDROGEN CONCENTRATIONS

3.3

K/A Category Point Totals: 1 1 1 1 2/2 1/1 Group Point Total: 7/3

ES-401 4 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K1

K2

K3

K4

K5

K6

A1

A2 A3

A4

G* K/A Topic(s) IR #

203000 RHR/LPCI: Injection Mode

X K5.01: Knowledge of the operational implications of Testable check valve operation as they apply to RHR/LPCI: INJECTION MODE

2.7

205000 Shutdown Cooling X

X

K3.02: Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on Reactor water level K6.04: Knowledge of the effect that a loss or malfunction of Reactor water level will have on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE)

3.2

3.6

206000 HPCI

X

X G2.4.50: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM A2.11: Ability to (a) predict the impacts of Low reactor water level on the HIGH PRESSURE COOLANT INJECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

4.2

4.2

207000 Isolation (Emergency) Condenser

209001 LPCS X A3.02: Ability to monitor automatic operations of the LOW PRESSURE CORE SPRAY SYSTEM including Pump start

3.8

209002 HPCS

211000 SLC X G2.2.36: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations as they apply to STANDBY LIQUID CONTROL SYSTEM

3.1

212000 RPS X

X

K1.03: Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and Recirculation system A2.05: Ability to (a) predict the impacts of Nuclear boiler instrument system failure on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

3.4

3.7

215003 IRM X K2.01: Knowledge of electrical power supplies to IRM channels/detectors

2.5

215004 Source Range Monitor X K3.04: Knowledge of the effect that a loss or malfunction of the SOURCE RANGE MONITOR (SRM) SYSTEM will have on Reactor power and indication

3.7

215005 APRM / LPRM X K1.08: Knowledge of the physical connections and/or cause-effect relationships between AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM and the Display control system

3.0

217000 RCIC X K6.01: Knowledge of the effect that a loss or malfunction of Electrical power will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC)

3.4

218000 ADS X

X

A2.06: Ability to (a) predict the impacts of ADS initiation signals present on the AUTOMATIC DEPRESSURIZATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations A3.09: Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including Reactor vessel water level

4.2

4.1

223002 PCIS/Nuclear Steam Supply Shutoff

X

X

A1.04: Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including Individual system relay status A4.04: Ability to manually operate and/or monitor System indicating lights and alarms in the control room

2.6

3.5

239002 SRVs X

X

A2.06: Ability to (a) predict the impacts of Reactor high pressure on the RELIEF/SAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations G2.4.31 Knowledge of annunciator alarms, indications, or response procedures as they apply to RELIEF/SAFETY VALVES

4.1

4.1

259002 Reactor Water Level Control

X K4.14: Knowledge of REACTOR WATER LEVEL CONTROL SYSTEM design feature(s) and/or interlocks which provide for Selection of various instruments to provide reactor water level input

3.4

261000 SGTS X A4.01: Ability to manually operate and/or monitor Off-site release levels in the control room

3.2

262001 AC Electrical Distribution

X A1.04: Ability to predict and/or monitor changes in parameters associated with operating the A.C. ELECTRICAL DISTRIBUTION controls including Load currents

2.7

262002 UPS (AC/DC) X

X

K4.01: Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) design feature(s) and/or interlocks which provide for Transfer from preferred power to alternate power supplies G 2.2.36: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations as they apply to UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.)

3.1

4.2

263000 DC Electrical Distribution

X A1.01: Ability to predict and/or monitor changes in parameters associated with operating the D.C. ELECTRICAL DISTRIBUTION controls including Battery charging/discharging rate

2.5

264000 EDGs X

X

K1.05 Knowledge of the physical connections and/or cause-effect relationships between EMERGENCY GENERATORS (DIESEL/JET) and Emergency generator fuel oil supply system K5.05: Knowledge of the operational implications of Paralleling A.C. power sources as they apply to EMERGENCY GENERATORS (DIESEL/JET)

3.2

3.4

300000 Instrument Air X

X

K2.01: Knowledge of electrical power supplies to the Instrument air compressor A2.01: Ability to (a) predict the impacts of Air dryer and filter malfunctions on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation

2.8

2.8

400000 Component Cooling Water

X

X

K3.01: Knowledge of the effect that a loss or malfunction of the CCWS will have on Loads cooled by CCWS A2.04: Ability to (a) predict the impacts of Radiation monitoring system alarm on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation

2.9

2.9

K/A Category Point Totals: 3 2 3 2 2 2 3 3/3 2 2 2/2 Group Point Total: 26/5

ES-401 5 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO / SRO)

System # / Name K1

K2

K3

K4

K5

K6

A1

A2 A3

A4

G* K/A Topic(s) IR #

201001 CRD Hydraulic X K6.06: Knowledge of the effect that a loss or malfunction of Component cooling water systems will have on the CONTROL ROD DRIVE HYDRAULIC System

2.8

201002 RMCS X A2.04: Ability to (a) predict the impacts of Control rod block on the REACTOR MANUAL CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

3.2

201003 Control Rod and Drive Mechanism

X G2.1.30: Ability to locate and operate components, including local controls as they apply to CONTROL ROD AND DRIVE MECHANISM

4.0

201004 RSCS

201005 RCIS

201006 RWM

202001 Recirculation

202002 Recirculation Flow Control

204000 RWCU X K4.07: Knowledge of REACTOR WATER CLEANUP SYSTEM design feature(s) and/or interlocks which provide for Draining of reactor water to various locations

2.9

214000 RPIS X A2.03: Ability to (a) predict the impacts of Overtravel/in-out on the ROD POSITION INFORMATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

3.6

215001 Traversing In-Core Probe

215002 RBM X K1.06: Knowledge of the physical connections and/or cause-effect relationships between ROD BLOCK MONITOR SYSTEM and Control rod selection

3.0

216000 Nuclear Boiler Inst.

219000 RHR/LPCI: Torus/Pool Cooling Mode

223001 Primary CTMT and Aux.

226001 RHR/LPCI: CTMT Spray Mode

230000 RHR/LPCI: Torus/Pool Spray Mode

233000 Fuel Pool Cooling/Cleanup X G2.2.22: Knowledge of limiting conditions for operations and safety limits as they apply to FUEL HANDLING EQUIPMENT

4.7

234000 Fuel Handling Equipment X A3.01: Ability to monitor automatic operations of the FUEL HANDLING EQUIPMENT including Crane/refuel bridge movement

2.6

239001 Main and Reheat Steam

239003 MSIV Leakage Control

241000 Reactor/Turbine Pressure Regulator

X A4.08: Ability to manually operate and/or monitor Control/governor valves (operation) in the control room

3.5

245000 Main Turbine Gen. / Aux.

256000 Reactor Condensate

259001 Reactor Feedwater X A1.04: Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including RFP turbine speed: Turbine-Driven-Only

2.8

268000 Radwaste

271000 Offgas

272000 Radiation Monitoring X K5.01: Knowledge of the operational implications of Hydrogen injection operation's effect on process radiation indications as they apply to RADIATION MONITORING SYSTEM

3.2

286000 Fire Protection X K2.02: Knowledge of electrical power supplies to Pumps 2.9

288000 Plant Ventilation

290001 Secondary CTMT X A2.04: Ability to (a) predict the impacts of High airborne radiation on the SECONDARY CONTAINMENT ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

3.7

290003 Control Room HVAC X K3.02: Knowledge of the effect that a loss or malfunction of the CONTROL ROOM HVAC will have on Computer/instrumentation

3.3

290002 Reactor Vessel Internals X K5.04: Knowledge of the operational implications of PCIOMR as they apply to REACTOR VESSEL INTERNALS

3.1

K/A Category Point Totals: 1 1 1 1 2 1 1 1/2 1 1 1/1 Group Point Total: 12/3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3

Facility: Date of Exam:

Category K/A # Topic RO SRO-Only

IR # IR #

1. Conduct of Operations

2.1.44 Knowledge of RO duties in the control room during fuel handling such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation

3.9

2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc 2.9

2.1.3 Knowledge of shift or short-term relief turnover practices 3.9

Subtotal 2 1

2. Equipment Control

2.2.38 Knowledge of conditions and limitations in the facility license 3.6

2.2.40 Ability to apply Technical Specifications for a system 3.4

2.2.43 Knowledge of the process used to track inoperable alarms 3.0

2.2.20 Knowledge of the process for managing troubleshooting activities 3.8

2.2.5 Knowledge of the process for making design or operating changes to the facility 3.2

Subtotal 3 2

3. Radiation Control

2.3.11 Ability to control radiation releases 3.8

2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc

2.9

2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions 3.5

2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc

3.7

2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc

3.8

Subtotal 3 2

4. Emergency Procedures / Plan

2.4.17 Knowledge of EOP terms and definitions 3.9

2.4.42 Knowledge of emergency response facilities 3.8

2.4.27 Knowledge of “fire in the plant” procedures 3.9

2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions 4.3

Subtotal 2 2

Tier 3 Point Total 10 10 7 7

ES-301 Administrative Topics Outline – RO DRAFT Form ES-301-1

Facility: PLANT E. I. HATCH ILT 10 Date of Examination: 06/20/2016 Exam Level: RO �� SRO-I �� SRO-U �� Operating Test No.: 2016-301

Administrative Topic (see Note)

Type Code*

Describe activity to be performed

Conduct of Operations Admin 1

M, R Determine if SBLC tank meets requirements of Tech Specs per Operations Table 3.1.7-1 & Table 3.1.7-2. (G2.1.25) (ALL)

Equipment Control Admin 2

N, R Determine the effect of the failure of a relay on system performance. (G2.2.41) (ALL)


D, R

Determine Bulk Average DW temperature per 34SV-SUV-019-1, and then determine if any additional DW cooling requirements. (G2.1.7) RO ONLY

Radiation Control Admin 5

P, R Evaluate a Radiation Work Permit (RWP) and Survey Map. (G2.3.7) ALL

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

* Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (≤ 3 (1) for ROs; ≤ 4 for SROs & RO retakes) (N)ew or (M)odified from bank (≥ 1) (2) (P)revious 2 exams (≤ 1; randomly selected) (1)

ES-301 Administrative Topics Outline – SRO-I DRAFT Form ES-301-1



Type Code*



M, R Determine if SBLC tank meets requirements of Tech Specs per Operations Table 3.1.7-1 & Table 3.1.7-2. (G2.1.25)


N, R Determine the effect of the failure of a relay on system performance. (G2.2.41)

Conduct of Operations Admin 3 D, R Verify Fuel Movements G2.1.35


P, R Evaluate a Radiation Work Permit (RWP) and Survey Map (G2.3.7)

Emergency Procedures/Plan Admin 6

M, S, R Given Plant Conditions, Determine the Emergency Classification and complete NMP-EP-110 Checklist 1 (G2.4.29)


* Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (≤ 3 for ROs; ≤ 4 (1) for SROs & RO retakes) (N)ew or (M)odified from bank (≥ 1) (3) (P)revious 2 exams (≤ 1; randomly selected) (1)

ES-301 Administrative Topics Outline – SRO-U DRAFT Form ES-301-1



Type Code*



M, R Determine if SBLC tank meets requirements of Tech Specs per Operations Table 3.1.7-1 & Table 3.1.7-2. (G2.1.25)


N, R Determine the effect of the failure of a relay on system performance. (G2.2.41)

Conduct of Operations Admin 3 D, R Verify Fuel Movements G2.1.35


P, R Evaluate a Radiation Work Permit (RWP) and Survey Map (G2.3.7)

Emergency Procedures/Plan Admin 6

M, S, R Given Plant Conditions, Determine the Emergency Classification and complete NMP-EP-110 Checklist 1 (G2.4.29)


* Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (≤ 3 for ROs; ≤ 4 (1) for SROs & RO retakes) (N)ew or (M)odified from bank (≥ 1) (3) (P)revious 2 exams (≤ 1; randomly selected) (1)

ES-301 Control Room/In-Plant Systems Outline – RO DRAFT Form ES-301-2


Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function

CR/SIM 1 JPM 01.10 – Withdraw control rods (Rod Drifts)(Alt Path) A, D, S

SF-1 Reactivity Control

201003A2.03

CR/SIM 2 – JPM 5.05D - Perform RC-2, HPCI Injection (Alt Path) A, EN, L, M, S

SF-2 Reactor Water Inventory Control

206000A2.14

CR/SIM 3 – JPM 14.14 - Open the MSIVs In An Emergency D, L, S

SF-3 Reactor Pressure Control

239001A4.01

CR/SIM 4 – JPM 20034A - Press Control EHC Fail (Alt Path) A, D, L, S

SF-4 Heat Removal From Reactor Core

239001A2.01

CR/SIM 5 – JPM 7.01A – Initiate Drywell Sprays (Alt Path) A, D, L, S

SF-5 Containment Integrity

226001A4.03

CR/SIM 6 – JPM 02711-18.2 - Transfer an Emergency 4160 VAC Bus from Emergency to Normal Power Supply

D, EN, S SF-6

Electrical 264000A4.04

CR/SIM 7 – JPM 25032 - Conduct A RWM Functional Test (Failure) D, L, S

SF-7 Instrumentation 201006A3.02

CR/SIM 8 – Perform 34IT-T45-001-2, Sump Isol Surv N, S

SF-9 Radiation Release

268000K1.04

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

PLANT 1 – JPM 10-18A – U1 From Outside the MCR, Insert a Manual Reactor Scram (Alt Path) A, D, E, R

SF-1 Reactivity

295006AA1.05

PLANT 2 – JPM 7.19 - From the RSD, Start RHR and Inject Into the Reactor D, R


203000A1.01

PLANT 3 – Transfer Unit 2 Vital AC from Alternate to Normal D, R

SF-6 Electrical

262002A4.01

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

* Type Codes Criteria for RO / SRO-I / SRO-U

(A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams (R)CA (S)imulator

4-6 (5) / 4-6 (x) / 2-3 (x) ≤ 9 (9) / ≤ 8 (x) / ≤ 4 (x) ≥ 1 (1) / ≥ 1 (x) / ≥ 1 (x) ≥ 1 (2) / ≥ 1 (x) / ≥ 1 (x) (control room system) ≥ 1 (5) / ≥ 1 (x) / ≥ 1 (x) ≥ 2 (2) / ≥ 2 (x) / ≥ 1 (x) ≤ 3 (0) / ≤ 3 (x) / ≤ 2 (x) (randomly selected) ≥ 1 (1) / ≥ 1 (x) / ≥ 1 (x)

ES-301 Control Room/In-Plant Systems Outline – SRO-I DRAFT Form ES-301-2




CR/SIM 1 JPM 01.10 – Withdraw control rods (Rod Drifts)(Alt Path) A, D, S

SF-1 Reactivity Control

201003A2.03



206000A2.14

CR/SIM 3 – JPM 14.14 - Open the MSIVs In An Emergency D, L, S

SF-3 Reactor Pressure Control

239001A4.01

CR/SIM 4 – JPM 20034A - Press Control EHC Fail (Alt Path) A, D, L, S

SF-4 Heat Removal From Reactor Core

239001A2.01



226001A4.03

CR/SIM 6 – JPM 02711-18.2 - Transfer an Emergency 4160 VAC Bus from Emergency to Normal Power Supply

D, EN, S SF-6

Electrical 264000A4.04



268000K1.04


PLANT 1 – JPM 10-18A - From Outside the MCR, Insert a Manual Reactor Scram (Alt Path) A, D, E, R

SF-1 Reactivity

295006AA1.05

PLANT 2 – JPM 7.19 - From the RSD, Start RHR and Inject Into the Reactor D, EN, R


203000A1.01


SF-6 Electrical

262002A4.01




4-6 (x) / 4-6 (5) / 2-3 (x) ≤ 9 (x) / ≤ 8 (8) / ≤ 4 (x) ≥ 1 (x) / ≥ 1 (1) / ≥ 1 (x) ≥ 1 (x) / ≥ 1 (2) / ≥ 1 (x) (control room system) ≥ 1 (x) / ≥ 1 (4) / ≥ 1 (x) ≥ 2 (x) / ≥ 2 (2) / ≥ 1 (x) ≤ 3 (x) / ≤ 3 (0) / ≤ 2 (x) (randomly selected) ≥ 1 (x) / ≥ 1 (2) / ≥ 1 (x)

ES-301 Control Room/In-Plant Systems Outline – SRO-U DRAFT Form ES-301-2






206000A2.14



226001A4.03



268000K1.04


PLANT 1 – JPM 10-18A - From Outside the MCR, Insert a Manual Reactor Scram (Alt Path) A, D, E, R

SF-1 Reactivity

295006AA1.05


SF-6 Electrical

262002A4.01




4-6 (x) / 4-6 (x) / 2-3 (3) ≤ 9 (x) / ≤ 8 (x) / ≤ 4 (3) ≥ 1 (x) / ≥ 1 (x) / ≥ 1 (1) ≥ 1 (x) / ≥ 1 (x) / ≥ 1 (1) (control room system) ≥ 1 (x) / ≥ 1 (x) / ≥ 1 (2) ≥ 2 (x) / ≥ 2 (x) / ≥ 1 (2) ≤ 3 (x) / ≤ 3 (x) / ≤ 2 (0) (randomly selected) ≥ 1 (x) / ≥ 1 (x) / ≥ 1 (2)

Name: ________________________________ ILT-10 SRO NRC EXAM Form: 0

Version: 01. 201001K6.06 001

Unit 2 was operating at 100% RTP when a loss of RBCCW occurs.

A manual scram is inserted which results in the following:

o Reactor power is 4% RTP o None of the RBCCW pumps can be started o Main Condenser Vacuum is 26.5 inches Hg Vacuum

Based on the above conditions and IAW 34AB-P42-001-2, Loss of RBCCW,

The CRD pumps __________ .

are NOT allowed to operate regardless of their pumping medium temperature

are NOT allowed to operate because the CRD Pump Room Area Coolers have lost theircooling medium

are allowed to operate provided their pumping medium remains less than a MAXIMUM of105°F

are allowed to operate provided their pumping medium remains less than a MAXIMUM of250°F

A.

B.

C.

D.

Description:

The RBCCW system is a closed loop system that removes heat from the reactor auxiliaryequipment it services and transfers the heat to the Plant Service Water System via one of twoheat exchangers. The pumps take a suction on the outlet of the heat exchangers and discharge toa common header. The discharge header branches to supply cooling water to three separateloops. Normally two pumps are in service with the third pump in standby. The water passingthrough the three loops picks up heat from the various loads and returns by a common header tothe heat exchangers. The loads cooled by the RBCCW System are connected to one of the threeloops.

Loop I loads are located inside the Drywell. Loop II includes the CRD pump coolers and Reactor Building drain sump coolers. Loop III includes other major heat loads located outside the Drywell.

Tuesday, April 19, 2016 3:14:37 PM 1

ILT-10 SRO NRC EXAMExamples of effects due to loss of RBCCW

a. Recirc pump bearings and seals overheat.b. CRD pumps overheat and will trip on overload.c. RWCU pumps will overheat and trip.d. RWCU will isolate on high temp NRHX outlet.e. Fuel pool temperature will increase:

IAW 34AB-P42-001-2, step 4.7:

States IF any of these conditions exist, RBCCW flow CANNOT be re-established,RBCCW suction temperature reaches 105°F as indicated on 2P42-R600, 2H11-P650,any temperature on Recirc System reaches its alarm setpoint.

Then perform the following:

4.7.1 Enter 34AB-C71-001-2, Scram Procedure, AND SCRAM the reactor.

4.7.6 UNLESS they are needed for Control Rod Insertion AND their pumping medium is lessthan 250°F, trip the Control Rod Drive (CRD) pumps.

With an ATWS in progress and their pumping medium less than 250°F (Main CondenserVacuum is 26.5 inches Hg Vacuum which equates to less than 126°F) suction temperature, theCRD pumps are allowed to remain in service to provide motive force for control rod insertion.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know that the CRD pumpsmust be secured; Effect to the CRD Hydraulic System (pumps) with the loss of the RBCCWSystem (Component cooling water systems).

The "A" distractor is plausible since it would be correct if an ATWS did not exist.

The "B" distractor is plausible because the cooling medium has been lost to the CRD PumpRoom Area Coolers and there are times listed in 34SO-C11-005-2 when the CRD pumps are notallowed to be operated with the area coolers unavailable.

The "C" distractor is plausible because with a loss of RBCCW, if RBCCW suction tempincreases to 105°F, a reactor scram is inserted and RBCCW heat loads are secured. The value of105°F does not apply to the pumping medium of the CRD pumps.

"D" is correct.

Tuesday, April 19, 2016 3:14:37 PM 2

ILT-10 SRO NRC EXAMReferences:NONE

K/A:

201001 Control Rod Drive Hydraulic System

K6. Knowledge of the effect that a loss or malfunction of the following will have on theCONTROL ROD DRIVE HYDRAULIC System : (CFR: 41.7 / 45.7)

K6.06 Component cooling water systems: Plant-Specific . . . . . . . . . . . 2.8 2.8

LESSON PLAN/OBJECTIVE:

C11-CRD-LP-00101, Control Rod Drive System, Ver. 9.1, EO 200.045.A.03 P42-RBCCW-LP-00901, Reactor Building Closed Cooling Water, Ver. 3.0, EO 200.014.A.04

References used to develop this question:

34AB-P42-001-2, Loss Of Reactor Building Closed Cooling Water, Ver. 2.6 34SO-C11-005-2, Control Rod Drive Hydraulic System, Ver. 32.2

Tuesday, April 19, 2016 3:14:37 PM 3

ILT-10 SRO NRC EXAM2. 201002A2.04 001

Unit 1 is operating at 15% RTP and is performing a reactor startup.

All control rods in the currently latched RWM step are at their Insert Limit.

o The Withdraw Limit for the latched step is position 08 o Control rod 30-31 is currently selected

When Control rod 30-31 is withdrawn, the rod begins DRIFTING OUT.

Based on the above conditions,

When control rod 30-31 passes position 10, the Rod Movement Control "Rod Out" whitelight will be __________ .

IAW 34AB-C11-004-1, Mispositioned Control Rods, a reactor scram __________ required.

ILLUMINATED; is

ILLUMINATED; is NOT

EXTINGUISHED; is

EXTINGUISHED; is NOT

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:37 PM 4

ILT-10 SRO NRC EXAMDescription:

When Control rod 30-31 reaches position 10, a rod out block signal will be active since the rod ispast its Withdraw Limit of position 08. The white "Rod Out" light on panel H11-P603 will beextinguished because of the rod block out signal. IAW 34AB-C11-004-1, Mispositioned ControlRods, Table after 4.4 when less than the LPSP of 20.6% RTP and one control rod drifting out,entering the scram procedure is required.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to predict the status of theRMCS Rod Out "White" light when a Control Rod Block is received and then mitigates thisabnormal condition using the table inside 34AB-C11-004-1.

The "A" distractor is plausible because the indicating lights associated with control rodmovement (Rod In, Rod Out & Rod Settle) only illuminate when an active insert, withdraw orsettle signal is present. With no active signal present, these indicating lights will beextinguished. The Rod Out block light will be illuminated with no active rod out block signalpresent, and extinguished with an active rod out block signal. The second part is plausible sinceit is correct.

The "A" distractor is plausible because the indicating lights associated with control rodmovement (Rod In, Rod Out & Rod Settle) only illuminate when an active insert, withdraw orsettle signal is present. With no active signal present, these indicating lights will beextinguished. The Rod Out block light will be illuminated with no active rod out block signalpresent, and extinguished with an active rod out block signal. The second part is plausible because the RWM is required by TS to be operable < 10% RTP (applicability). IAW34B-C11-004-1, if a rod is drifting out and power above the LPSP, section 4.5 is entered and areactor scram is NOT required.

"C" is correct.

The "D" distractor is plausible since the first part is correct. The second part is plausible because the RWM is required by TS to be operable < 10% RTP (applicability). IAW34B-C11-004-1, if a rod is drifting out and power above the LPSP, section 4.5 is entered and areactor scram is NOT required.

Tuesday, April 19, 2016 3:14:37 PM 5


K/A:

201002 Reactor Manual Control System

A2. Ability to (a) predict the impacts of the following on the REACTOR MANUALCONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct,control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.04 Control rod block . . . . . . . . . . . . . . . . . 3.2 3.1


C11-RMCS-LP-05401, Reactor Manual Control System, Ver. 6.0, EO 200.091.A.03


34AB-C11-004-1, Mispositioned Control Rods, Ver. 4.0 Bank question from HLT Database which was used on 2011 Hatch NRC Exam Q#1

Tuesday, April 19, 2016 3:14:38 PM 6

ILT-10 SRO NRC EXAM3. 201003G2.1.30 001

Unit 2 is operating at 100% RTP with the 2C11-F002A, Flow Control Valve, in service.

Subsequently, Drywell pressure increases to 2.5 psig.

One (1) minute later and based on the above indications/conditions,

2C11-F002A __________ responding correctly.

2C11-F002A indicating LIGHTS are located on __________ .

is; 2H11-P603 and locally at the 2C82-P001, Remote Shutdown Panel

is; 2H11-P603 Panel ONLY

is NOT; 2H11-P603 and locally at the 2C82-P001, Remote Shutdown Panel

is NOT; 2H11-P603 Panel ONLY

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:38 PM 7


The cause of the Scram can effect how the system responds. On a NON-LOCA SCRAM, theCRD pumps will remain running and the system will respond as follows: The flow control, valveF002A/B, senses maximum flow and closes. Due to valve construction, 5 gpm is maintained tothe CRD system to prevent thermal shock when the scram is reset and flow is returned to thenormal flow path. When the Scram is reset, the accumulators will recharge, the Flow ControlValve will open, and the system returns to normal conditions.

The System response for a Scram with a LOCA signal (RWL -101 inches or DW pressure 1.85psig) is similar except the CRD pumps will trip and the flow controller will sense NO flow,causing the Flow Control Valve (F002) to open fully.

2C11-F002A/B indications are located on 2H11-P603 only.

K/A JUSTIFICATION:

This question satisfies the K/A statement by requiring the applicant to know the effect of aLOCA scram has on 2C11-F002A, Flow Control valve (assists in motive force for CRDMechanism) and know where the indications exist in the MCR for the flow control valve.

The "A" distractor is plausible since the first part is correct. The second part is plausible sinceother control rod drive components are located on 2C82-P001, Remote Shutdown Panel (2BCRD pump).

"B" is correct.

The "C" distractor is plausible since it would be correct on a reactor scram (without a LOCA)which normally results in the flow control valve sensing a high flow condition from chargingwater header flow resulting in the flow control valve fully closing due to this high flow. Thesecond part is plausible since other control rod drive components are located on 2C82-P001,Remote Shutdown Panel (2B CRD pump).

The "D" distractor is plausible since it would be correct on a reactor scram (without a LOCA)which normally results in the flow control valve sensing a high flow condition from chargingwater header flow resulting in the flow control valve fully closing due to this high flow. Thesecond part is plausible since it is correct.

References:NONE

Tuesday, April 19, 2016 3:14:38 PM 8

ILT-10 SRO NRC EXAM

K/A:

201003 Control Rod and Drive Mechanism

G2.1.30 Ability to locate and operate components, including local controls. (CFR: 41.7 / 45.7) . . . . . . . . . . . 4.4 4.0


C11-CRD-LP-00101, Control Rod Drive System, Ver 9.1, EO 001.013.A.07


34SO-C11-005-2, Control Rod Drive Hydraulic System, Ver 32.2, EO 001.013.A.07

Modified from HLT Database Q#295006G2.1.31-001 which was used on the 2015 NRC ExamQ#41

Original Question

Unit 2 is operating at 100% RTP with the 2C11-F002A, Flow Control Valve, in service.

Subsequently, a transient occurs resulting in RWL lowering to -10 inches (lowest RWL reached)before being restored to normal with Feedwater.

Tuesday, April 19, 2016 3:14:38 PM 9

ILT-10 SRO NRC EXAM

One (1) minute later and based on the above indications/conditions,

2C11-F002A __________ responding correctly.

2C11-F002A indicating LIGHTS are located on __________ .

A. is; 2H11-P603 and locally at the 2C82-P001, Remote Shutdown Panel

B. is; 2H11-P603 Panel ONLY

C. is NOT; 2H11-P603 and locally at the 2C82-P001, Remote Shutdown Panel

D. is NOT; 2H11-P603 Panel ONLY

Tuesday, April 19, 2016 3:14:38 PM 10

ILT-10 SRO NRC EXAM4. 203000K5.01 001

Unit 2 has experienced a LOCA.

o RHR Loop 2B is injecting in the LPCI mode to maintain RWL o A leak develops on 2E11-F017B, RHR Injection valve o When both RHR Loop 2B pumps are secured, 2E11-F050B, RHR Injection Check valve,

sticks in the OPEN position


The water leaking from 2E11-F017B will be leaking into __________ Containment.

Closing 2E11-F015B, RHR Injection valve, __________ isolate the reactor water leakingfrom 2E11-F017B.

Secondary; will NOT

Secondary; will

Primary; will NOT

Primary; will

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:38 PM 11


This question tests whether the applicant understands the RHR system discharge valvearrangement and location. The applicant is being examined on the following valves:

o Air operated check valve (2E11-F050) o Inboard Discharge Valve (2E11-F015) o Outboard Discharge Valve (2E11-F017)

F050 is located INSIDE the DW, downstream of both F015 and F017.F015 is located OUTSIDE the DW, between F050 and F017.F017 is located OUTSIDE the DW, upstream of both F015 and F050.

With a leak on on F017, the water will be going into the Reactor Building/SecondaryContainment due to the valves location.

When F015 is closed, water from the reactor through F050 is isolated from F017 due to the valvearrangement (described above).

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine where the waterleaking from the 2E11-F017B will leak into (operational implications) after 2E11-F050B,Testable Check valve, has failed open and if isolating 2E11-F015B will isolate reactor waterflowing through this failed open check valve.

The "A" distractor is plausible because the first part is correct. The second part is plausiblebecause on the Core Spray System (ECCS), the outboard isolation valve has the lower MPLnumber (E21-F004 is the outboard valve & E21-F005 is the inboard valve). For CS, if a leakexists on the higher numbered valve (F005), isolating the lower numbered valve (F004) will notisolate the leak.

"B" is correct.

The "C" distractor is plausible because on HPCI System (ECCS), the inboard isolation valve(E41-F002) is located inside Primary Containment. Any water leaking from the valve will beinto Primary Containment. The second part is plausible because on the Core Spray System(ECCS), the outboard isolation valve has the lower MPL number (E21-F004 is the outboardvalve & E21-F005 is the inboard valve). For CS, if a leak exists on the higher numbered valve(F005), isolating the lower numbered valve (F004) will not isolate the leak.

The "D" distractor is because on HPCI System (ECCS), the inboard isolation valve (E41-F002)is located inside Primary Containment. Any water leaking from the valve will be into PrimaryContainment. The second part is correct.

Tuesday, April 19, 2016 3:14:38 PM 12

ILT-10 SRO NRC EXAMReferences:NONEDO NOT PROVIDE INBOARD/OUTBOARD INFORMATION TO THE STUDENTFOR THIS QUESTION (#4 - 203000K5.01-001). KNOWING WHICH VALVE IS THEINBOARD/OUTBOARD IS PART OF THIS EXAM.

K/A:

203000 RHR/LPCI: Injection Mode (Plant Specific)

K5. Knowledge of the operational implications of the following concepts as they apply toRHR/LPCI: INJECTION MODE (PLANT SPECIFIC) : (CFR: 41.5 / 45.3)

K5.01 Testable check valve operation . . . . . . . . . . . 2.7* 2.9


E11-RHR-LP-00701, Residual Heat Removal System, Ver. 9.2, EO 006.001.A.01

Reference(s) used to develop this question:

34SO-E11-010-2, Residual Heat Removal System, Ver. 41.0 Bank question from HLT Database which was used on 2011 HLT-6 NRC Exam Q#5

Tuesday, April 19, 2016 3:14:38 PM 13

ILT-10 SRO NRC EXAM5. 204000K4.07 001

Unit 2 is operating at 20% RTP with the following conditions;

o RWCU System operating in Blowdown To The Main Condenser o 2G31-FCV-F033, Dump Flow Control Valve, Enable and 20% Open o 2G31-F034, Disch to Main Cndsr, Open o 2G31-R602, Dump Flow Indicator, 50 gpm


2G31-FCV-F033 will automatically close if the __________ .

RWCU pump flow decreases to 70 gpm

2G31-F033 upstream pressure decreases to 3 psig

2G31-F033 downstream pressure increases to 90 psig

RWCU differential flow increases to 41 gpm and remains there for 15 seconds

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:38 PM 14


IAW 34SO-G31-003-2,

IF the RWCU F/D vessel flow drops to from 77 GPM to 41 GPM in less than 15 seconds,the "F/D Vessel 2G31-D002A Effluent Flow Low-Low - Backwash Required" alarm will bereceived and 2G31-F052A valve will go closed.

7.6.9.17, Throttle open 2G31-F033, using 2G31-R606, to achieve less than OR equal to amaximum dump flow rate of 150 gpm. (150 gpm corresponds to approximately 90 psigdownstream of 2G31-F033)

5.2.4, 2G31-C001A AND/OR 2G31-C001B, RWCU Pumps, trip under the followingconditions:

o High cooling water temperature, 140°F.o Low flow for the A RWCU pump, <60 gpm, is bypassed for 15 sec. following pump

start. Low flow for the B RWCU pump, <60 gpm, is bypassed for 30 sec. followingtaking the Pump Control Switch to START.

o 2G31-F001, RWCU Inboard Isolation, OR 2G31-F004, RWCU Outboard Isolation,NOT full open.

IAW 34AR-602-434-2, RWCU PUMP LOW FLOW, the RWCU pump common discharge lineflow is less than the setpoint of 60 gpm.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know the isolation designfeature/interlock for draining RWL to the Main Condenser via F033.

The "A" distractor is plausible because less 77 gpm will automatically start the Demin holdpumps.

"B" is correct.

The "C" distractor is plausible because 2G31-F033 will close if downstream pressure increasesto 140 psig. When RWCU is in blowdown mode, 2G31-F033 is throttled open to achieve < 150gpm. The max flow rate is 150 gpm. A downstream pressure of 90 psig corresponds to a flowrate of 150 gpm.

The "D" distractor is plausible because a RWCU system isolation will occur for Demin isolationvalves if flow through the Demins decreases from 77 gpm to 41 gpm in < 15 secs.

Tuesday, April 19, 2016 3:14:38 PM 15


K/A:

204000 Reactor Water Cleanup System

K4. Knowledge of REACTOR WATER CLEANUP SYSTEM design feature(s) and/orinterlocks which provide for the following: (CFR: 41.7)

K4.07 Draining of reactor water to various locations . . . . . . . . . . . . . 2.9 2.9


G31-RWCU-LP-00301, Reactor Water Cleanup, Ver. 5.2 EO 003.002.A.07


34AR-602-434-2, RWCU Pump Low Flow, Ver. 3.3 34SO-G31-003-2, Reactor Water Cleanup System, Ver. 43.5

Tuesday, April 19, 2016 3:14:38 PM 16

ILT-10 SRO NRC EXAM6. 205000K3.02 001

Unit 1 is in Mode 4 with both Recirc pumps off.

RHR Loop B is operating in Shutdown Cooling with a flowrate of 7300 gpm.

Subsequently, a tube rupture occurs in RHR Heat Exchanger 1B.

With the above conditions and NO operator actions,

RWL will start going __________ .

IAW 34SO-E11-010-1, RHR System, the MINIMUM listed corrected RWL that must bemaintained by the OATC is __________ .

UP; 34 inches

UP; 54 inches

DOWN; 34 inches

DOWN; 54 inches

A.

B.

C.

D.

Description:

Each heat exchanger is a vertical, U tube, single pass heat exchanger. There are two heatexchangers, one per RHR loop. The heat exchangers capacity is based on the most severe modeof operation, which is the SDC mode. It is designed to maintain reactor temperature less than orequal to 125°F, 20 hrs after all rods are fully inserted.

In order to prevent leakage from the RHR system into the RHRSW system, tube side pressure inthe heat exchanger is maintained higher than shell side pressure. The differential pressure ismaintained by controlling the position of 1E11-F068A/B.

If RHR (shell side) pressure was greater than RHRSW (inside tubes) pressure and a tube leakexisted, a possible radioactive release to the environment could occur (RHRSW discharge flowis directed to the flume).

Tuesday, April 19, 2016 3:14:38 PM 17

ILT-10 SRO NRC EXAM

A RHRSW (tube) side leak will cause RWL to go up since RHR is taking a suction off the RPVand discharging back to the RPV but now with RHRSW leaking into the shell side. If RHR wasinside the tube, the leak would be flowing from RHR to the shell side causing RWL to startgoing down.

IAW 34SO-E11-010-1, Section 7.2.3 Shutdown Cooling Mode (Hot Shutdown) CAUTIONbefore step 7.2.3.71 states

CAUTION: TO ENSURE ADEQUATE CORE CIRCULATION, CORRECTED RPV WATER LEVELMUST BE MAINTAINED AS FOLLOWS-> 53” RWL WITH SDC FLOW < 7700 GPM OR-> 33” WITH SDC FLOW > 7700 GPM OR->33” WITH AT LEAST ONE RECIRC PUMP RUNNING AT MINIMUM SPEED.

With RHR SDC flow 7200 gpm, corrected RWL must be >53 inches to ensure adequate corecooling.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant the effect that a malfunction ofSDC (tube leak) will have on RWL. In this question, the effect on RWL is that with the SDCtube leak, RWL will start going up from its current value. If RHR was inside the tube, the leakwould be flowing from RHR to the shell side causing RWL to start going down.

The "A" distractor is plausible since the first part is correct. The second part is plausible becauseif SDC flow is > 7700 gpm OR at least one Recirc Pump is running, corrected RWL is requiredto be maintained > 33".

"B" is correct.

The "C" distractor is plausible if the high pressure side of the heat exchanger is on the RHR sidetherefore RWL would lower due to leaking to the RHRSW side. The second part is plausiblebecause if SDC flow is > 7700 gpm OR at least one Recirc Pump is running, corrected RWL isrequired to be maintained > 33".

The "D" distractor is plausible if the high pressure side of the heat exchanger is on the RHR sidetherefore RWL would lower due to leaking to the RHRSW side. The second part is correct.

References:NONE

Tuesday, April 19, 2016 3:14:38 PM 18

ILT-10 SRO NRC EXAM

K/A:

205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

K3. Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLINGSYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: (CFR: 41.7 / 45.4)

K3.02 Reactor water level: Plant-Specific . . . . . . . . . . . 3.2 3.3


E11-RHR-LP-00701, Residual Heat Removal System, Ver. 9.2, EO 007.007.a.02


34SO-E11-010-1, Residual Heat Removal System, Ver. 44.4

Modified question from HLT Database which was used on 2015 ILT-9 NRC Exam Q#5

Original Question

Unit 1 is in Mode 4 with RHR Loop B Shutdown Cooling in service.

Subsequently, a tube rupture occurs in RHR Heat Exchanger 1B.


RWL will start going __________ .

IAW 34SO-E11-010-1, RHR System, the MINIMUM listed RHR to RHRSW differentialpressure allowed is __________ .

A. UP; 21 psid

B. UP; 31 psid

C. DOWN;

Tuesday, April 19, 2016 3:14:38 PM 19

ILT-10 SRO NRC EXAM21 psid

D. DOWN; 31 psid

Tuesday, April 19, 2016 3:14:38 PM 20

ILT-10 SRO NRC EXAM7. 205000K6.04 001

Unit 1 is in Mode 3 with RHR pump 1A in Shutdown Cooling (SDC).

o RPV pressure is 65 psig and slowly lowering

Subsequently, RWL lowers to -10 inches.

Three (3) minutes later and based on the above conditions,

1E11-F015A, RHR Injection Valve, will be __________ .

RHR pump 1A __________ be running.

closed; will

closed; will NOT

open;will

open;will NOT

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:38 PM 21


SDC Inboard/Outboard isolation valves, 1E11-F009 & F008, isolate if either of the followingoccur:

Reactor pressure exceeds 138 psig Reactor water level drops below 3 inches

1E11-F015A/B will auto close if a PCIS Group II signal is received (3 inches RWL or 1.85 psigDrywell pressure) while in SDC lineup. 1E11-F015A/B will automatically reopen if a LOCAsignal is present provided that the 1E11-F008 or 1E11-F009 is fully closed OR if reactorpressure is greater than 138 psig. F015A/B will not isolate on high reactor pressure.

Suction valve line up must have either the respective 1E11-F004 valve open OR the respective1E11-F006A valve open AND the 1E11-F008 and 1E11-F009 valves open. Without a suctionsource, the RHR pumps will receive a trip signal and will not auto start.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant the effect low RWL will haveon the SDC system.

The "A" distractor is plausible since the first part is correct. The second part is plausible becausethe Core Spray pumps (low pressure ECCS) do NOT have any "loss of suction" trips or valveinterlocks.

"B" is correct.

The "C" distractor is plausible because F008 & F009 will auto isolate on High Reactor Pressureof 138 psig, but F015A/B will NOT. Also, F015A/B will only isolate on Low RWL of 3" ifRHR is in a SDC lineup (i.e., F008 & F009 are OPEN). The second part is plausible because theCore Spray pumps (low pressure ECCS) do NOT have any "loss of suction" trips or valveinterlocks.

The "D" distractor is plausible because F008 & F009 will auto isolate on High Reactor Pressureof 138 psig, but F015A/B will NOT. Also, F015A/B will only isolate on Low RWL of 3" ifRHR is in a SDC lineup (i.e., F008 & F009 are OPEN). The second part is plausible since it iscorrect.

References:NONE

K/A:

205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

Tuesday, April 19, 2016 3:14:38 PM 22

ILT-10 SRO NRC EXAMK6. Knowledge of the effect that a loss or malfunction of the following will have on theSHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) : (CFR: 41.7 / 45.7)

K6.04 Reactor water level . . . . . . . . . . . . 3.6 3.6


E11-RHR-LP-00701, Residual Heat Removal System, Ver. 9.2, EO 006.008.A.03 & EO 006.008.a.04


34SO-E11-010-1, Residual Heat Removal System, Ver. 44.4 Modified question from HLT Database which was used on 2012 ILT-7 NRC Exam Q#6

Original Question

Unit 2 is shutdown with the following conditions:

o "A" loop of RHR is in Shutdown Cooling (SDC) o 2E11-F008, SDC Suction Vlv, is OPEN o 2E11-F009, SDC Suction Vlv, is OPEN o 2E11-F015A, RHR Inbd Inj Vlv, is OPEN o Reactor pressure is 125 psig and steady

o Subsequently, 2E11-F048A, Hx Bypass valve, fails full OPEN and cannot be re-closed o Reactor pressure starts increasing at 1 psig/minute

Without any operator actions and based on the above conditions, which ONE of the choicesbelow completes the following statements?

When Reactor pressure reaches 140 psig, RHR Valves, __________ will be CLOSED.

The RHR pumps _______ receive a trip signal as a result of these valve closures.

A. 2E11-F008 and 2E11-F009 (ONLY); will

B. 2E11-F008 and 2E11-F009 (ONLY); will NOT

Tuesday, April 19, 2016 3:14:38 PM 23

ILT-10 SRO NRC EXAMC. 2E11-F008, 2E11-F009 and 2E11-F015A;

will

D. 2E11-F008, 2E11-F009 and 2E11-F015A; will NOT

Tuesday, April 19, 2016 3:14:38 PM 24

ILT-10 SRO NRC EXAM8. 206000G2.4.50 001

Unit 2 HPCI is operating in the Pressure Control Mode IAW 34SO-E41-001-2, HPCI System.

At 10:00, the following alarm is received on 2H11-P602 Panel;

602-235

At 10:02, the following alarm is received on 2H11-P601 Panel;

601-117

At 10:03, the following exist:

o HPCI flow 4250 gpm in Pressure Control Mode o HPCI suction pressure 11.0 inches Hg Vac o Torus water level 151.0 inches and stable

Based on the above condtions and ARPs,

A HPCI pump low suction pressure trip setpoint __________ been EXCEEDED.

The NPO will confirm _________ .

has; 2E41-F004, CST Suction valve, is full open

has; 2E41-F041 & F042, Torus Suction valves, are full open

has NOT; 2E41-F004, CST Suction valve, is full open

has NOT; 2E41-F041 & F042, Torus Suction valves, are full open

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:39 PM 25


IAW 34SO-E41-001-2, Section 5.2, Limitations, HPCI will trip on Low HPCI Pump SuctionPressure of 10 inches Hg VAC.

IAW 34AR-601-117-2, the HPCI low pressure alarm setpoint is 8.0 inches Hg Vacuum and withHPCI suction pressure at 12.0 inches Hg Vac, the HPCI turbine should have alreadyautomatically tripped.

IAW 34AR-602-235-2, the high Torus level alarm setpoint is 149.5 inches and with Torus levelat 151.0 inches, the alarm is valid but has not reached the level where HPCI will transfer itssuction supply.

RCIC will transfer its suction supply when Torus level is 150.5 inches while HPCI will tranferits suction supply when Torus level is 152 inches.

IAW 34SO-C11-005-2, Section 5.2 Limitations, The CRD pump trips on Low Pump SuctionPressure of 12 in. Hg Vacuum (3 sec. time delay).

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to verify Torus high and HPCIsuction pressure low alarms (HPCI should have auto tripped) by using HPCI indications that arelisted in the ARPs (confirming valves 2E41-F004, F041 & F042).

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sinceRCIC will transfer their suction source from the CST to the Torus on a High Torus level of 150.5inches.

The "C" distractor is plausible since a CRD pump will trip on low suction pressure of 12 inchesHg vacuum (CRD pumps). The second part is plausible since it is correct.

The "D" distractor is plausible since a CRD pump will trip on low suction pressure of 12 inchesHg vacuum (CRD pumps). The second part is plausible since RCIC will transfer their suctionsource from the CST to the Torus on a High Torus level of 150.5 inches.

Tuesday, April 19, 2016 3:14:39 PM 26


K/A:

206000 High Pressure Coolant Injection System

G2.4.50 Ability to verify system alarm setpoints and operate controls identified in thealarm response manual.

(CFR: 41.10 / 43.5 / 45.3) . . . . . . . . . 4.2 4.0


E41-HPCI-LP-00501, High Pressure Coolant Injection (HPCI), Ver. 6.0, EO 005.005.a.05 & EO 005.002.a.03


34SO-E41-001-2, High Pressure Coolant Injection (HPCI) System, Ver. 30.0 34AR-601-117-2, HPCI Pump Suct Press Low, Ver. 2.2 34AR-601-127-2, Torus Level High, Ver. 4.3 34AR-602-235-2, Torus Water Level High / Low, Ver. 2.5 34SO-C11-005-2, Control Rod Drive Hydraulic System, Ver. 32.2

Tuesday, April 19, 2016 3:14:39 PM 27

ILT-10 SRO NRC EXAM9. 209001A3.02 001

Unit 2 is operating at 100% RTP.

At 13:00, 34SV-E21-001-2, Core Spray Pump Operability, is in progress with Core Spray pump 2A running with 4250 gpm flow.

At 13:01, a Loss of Offsite Power occurs to BOTH Units.

At 13:10, RWL is -103 inches and slowly lowering.

At 13:15, RWL is -115 inches and slowly lowering.


The EARLIEST listed time that Core Spray pump 2A will have automatically restarted is __________ .

twelve (12) seconds after 13:01

twenty four (24) seconds after 13:01

13:11

13:16

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:39 PM 28


IAW 34AB-R22-003-2, the following table indicates the start times for ECCS pumps (LOSP with concurrent LOCA);

TIME DG A DG B DG CStarts Starts Starts

(DG output0 sec brkr closes)

CS A starts CS B starts

0.5 sec LPCI C starts12 sec LPCI A starts LPCI D starts LPCI B starts

CS pump 2A will automatically restart when a LOCA condition (RWL -101 inches or DWpressure 1.85 psig) first exists. At 13:10, RWL is lower than -101 inches, therefore, CS pump2A will have already restarted at 13:11.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine the earliest timethat Core Spray pump 2A will automatically start.

The "A" distractor is plausible since the CS pump 2A will automatically start as soon as theEDG output breaker closes (LOSP with concurrent LOCA).

The "B" distractor is plausiblesince an ECCS pump will automatically start 24 seconds after theLOSP occurs (LOSP with concurrent LOCA - RHR pump 2A - 12 seconds for EDG to come upto speed & close output breaker then 12 seconds later RHR 2A will start).

"C" is correct.

The "D" distractor is plausiblesince the TS value for Level 1 (-113 inches) and since RWL is-115 inches at 13:16, therefore selecting this time as the earliest.

Tuesday, April 19, 2016 3:14:39 PM 29


K/A:

209001 Low Pressure Core Spray System

A3. Ability to monitor automatic operations of the LOW PRESSURE CORE SPRAYSYSTEM including: (CFR: 41.7 / 45.7)

A3.02 Pump start . . . . . . . . . . . . . . . . . . . 3.8 3.7


E21-CS-LP-00801, Core Spray System, Ver. 5.2, EO 008.006.a.01


34SV-E21-001-2, Core Spray Pump Operability, Ver. 22.0 34AB-R22-002-2, Loss Of 4160V Emergency Bus, Ver. 1.14 34SO-E21-001-2, Core Spray System, Ver. 25.2 U2 TS , LCO 3.3.5.1, ECCS Instrumentation, Table 3.3.5.1-1 (page 1 of 5) ECCSInstrumentation, Amendment 210

Tuesday, April 19, 2016 3:14:39 PM 30

ILT-10 SRO NRC EXAM10. 211000G2.2.36 001

Unit 1 is operating at 100% RTP with the following:

o 1C41-C001B, SBLC pump 1B, is DANGER Tagged out of service for a pump leak

Based on the above status of SBLC pump 1B,

If a troubleshooting activity resulted in tripping the feeder breaker to __________ , anadditional LCO entry would be required for the SBLC System.

1R24-S011, Rx. Bldg. MCC 1C

1R24-S012, Rx. Bldg. MCC 1B

1R24-S013, Rx. Bldg. MCC 1A

1R24-S014, Rx. Bldg. MCC 1D

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:39 PM 31


IAW 34SO-C41-003-1, SBLC System,

Reactor Bldg - 600V MCC 1C ESS. Div. 1 - 1R24-S011 130RHR13Frame 1A Standby Liquid Control Pump 1C41-C001A RACKED IN

(ON)

Reactor Bldg - 600V MCC 1B ESS. Div. 2 - 1R24-S012 130R2RHFrame 1A Standby Liquid Control Pump 1C41-C001B RACKED IN

(ON)

With SBLC 1B "Danger" tagged out of service and inoperable, a 7 day Required ActionStatement (TS 3.1.7 Condition B) exists. With Maintenance performing troubleshootingactivities on the feeder breaker for 1R24-S011, Rx. Bldg. MCC 1C, which 1R24-S011 is thepower supply to SBLC 1A, if this breaker trips, both Unit 1 SBLC Systems will be inoperable. TS 3.1.7 Condition C, Required Action C.1 would then be entered requiring an 8 hourcompletion time.

K/A JUSTIFICATION:

This question satisfies the K/A statement by requiring the applicant to determine the potentialeffect (analyze) of troubleshooting (maintenance activity) a power source on the LCO for theSBL System.

"A" is correct.

The "B" distractor is plausible since this is the power supply for the SBLC 1B however losingthis power supply would not affect the SBLC 1B since it is already INOPERABLE.

The "C" distractor is plausible since this is a 600 MCC located in the Unit 1 Reactor Building(same elevation as 1R24-S012).

The "D" distractor is plausible since this is a 600 MCC located in the Unit 1 Reactor Building(same elevation as 1R24-S012).

Tuesday, April 19, 2016 3:14:39 PM 32

ILT-10 SRO NRC EXAM

References:NONE

K/A:

211000 Standby Liquid Control System

G2.2.36 Ability to analyze the effect of maintenance activities, such as degraded powersources, on the status of limiting conditions for operations. (CFR: 41.10 / 43.2 / 45.13) . . . . . . . . 3.1 4.2


C41-SBLC-LP-01101, Standby Liquid Control, Ver. 6.1, EO 011.001.A.04


34SO-C41-003-1, Standby Liquid Control System, Ver. 12.3 A-10120, Load List for 1R24-S013, Ver. 4.0 A-10121, Load List for 1R24-S014, Ver. 1.0

Original concept from 2010 Oyster Creek NRC Exam Q#68 Modified question from HLT Database which was used on 2015 ILT-9 NRC Exam Q#69

Original Question

Unit 1 and Unit 2 are operating at 100% RTP. On your Control Room tour during turnover, younote the change in status in the Standby Gas Treatment System (SBGT) as shown below:

Tuesday, April 19, 2016 3:14:39 PM 33

ILT-10 SRO NRC EXAM

With the above status of SBGT, if a troubleshooting activity resulted in tripping the feederbreaker to __________ , an additional LCO entry would be required for the SBGT System.

A. 2R24-S011, Rx. Bldg. MCC 2C

B. 2R24-S012, Rx. Bldg. MCC 2B

C. 2R24-S013, Rx. Bldg. MCC 2A

D. 2R24-S014, Rx. Bldg. MCC 2D

Tuesday, April 19, 2016 3:14:39 PM 34

ILT-10 SRO NRC EXAM11. 212000K1.03 001


The following occurs at the listed time:

o 12:01 RPS MG Set 2A trips o 12:03 The Power Source Select switch (2H11-P610) is placed in the "ALT B" position


At 12:02, 2B31-R614, Recirc Drive Flow, recorder __________ indicate accurate flow.

At 12:05, Recirc Pump 2A Drive Flow will be __________ the Recirc Pump 2A Drive Flow at 12:00.

will NOT; less than

will NOT; approximately the same as

will continue to; less than

will continue to; approximately the same as

A.

B.

C.

D.

Description:

There are four differential pressure (dp) sensor/transmitters which monitor the flow in each ofthe two Recirculation System loops (Total of 8). Each APRM instrument powers andprocesses the signals for one sensor/transmitter monitoring loop A and for one sensor/transmittermonitoring loop B. Each sensor/transmitter generates an electrical current signal related to theamount of dP it senses. The APRM instrument converts the signal from each sensor/transmitterinto a voltage, digitizes it, and digitally filters the signal to remove noise. Each filtered dP signalis converted to a flow rate and scaled to a Recirculation Loop Flow signal. The twoRecirculation Loop Flow signals are then combined to obtain a Total Recirculation Flow signal.

The APRM A supplies the Loop A and Loop B Recirc flow inputs to 2B31-R614, Recirc DriveFlow recorder, located on the 2H11-P602 panel.

Each APRM has a dual power supply, one from RPS A and one from RPS B.

Tuesday, April 19, 2016 3:14:39 PM 35

ILT-10 SRO NRC EXAM

Since only one RPS Bus (initially) was lost no APRM will be deenergized, therefor the2B31-R614, Recirc Drive Flow indication will remain accurate.

Placing the Power Source Select switch (2H11-P610) in the "ALT B" position will initiallyde-energize RPS 2B (break before make) and then re-energize the bus back. RPS 2B will bepowered from Essential Bus 2B (normal backup). With both RPS buses de-energized, a fullscram will occur, causing total steam flow to decrease to less than 60% of previous 6 minutevalue. The reduction in steam flow will meet the condition for a speed limiter #1 runback,resulting in Recirc Drive flow decreasing.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine the effect oflosing an RPS Bus will have on the Recirc System flow transmitters and Recirc Drive flow.

The "A" distractor is plausible if the applicant remembers there are 4 flow transmitters on eachRecirc loop and since RPS Bus 2A is lost that Recirc Loop A flow signals (divisionalized) arelost resulting in an invalid Recirc Drive flow recorder value. The second part is plausible since itis correct.

The "B" distractor is plausible if the applicant remembers there are 4 flow transmitters on eachRecirc loop and since RPS Bus 2A is lost that Recirc Loop A flow signals (divisionalized) arelost resulting in an invalid Recirc Drive flow recorder value. The second part is plausible if theapplicant realizes that RPS Bus 2B is energized therefore the Recirc flow transmitters havepower and does not consider that both RPS Buses were momentarily de-energized causing a fullreactor scram, leaving Recirc Drive flow at minimum.

"C" is correct.

The "D" distractor is plausible since the first part is correct. The second part is plausible if theapplicant realizes that RPS Bus 2B is energized therefore the Recirc flow transmitters havepower and does not consider that both RPS Buses were momentarily de-energized causing a fullreactor scram, leaving Recirc Drive flow at minimum.

Tuesday, April 19, 2016 3:14:39 PM 36


K/A:

212000 Reactor Protection System

K1. Knowledge of the physical connections and/or cause effect relationships betweenREACTOR PROTECTION SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.03 Recirculation system . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4 3.6


C51-PRNM-LP-01203, Power Range Neutron Monitoring System, Ver. 9.0, EO 012.003.D.16


C51-PRNM-LP-01203, Power Range Neutron Monitoring System, Ver. 9.0 C71-RPS-LP-01001, Reactor Protection System, Ver. 8.2 B31-RRS-LP-00401, Reactor Recirculation System, Ver. 10.6 U2 TS Bases 3.3.1.1, RPS Instrumentation, 2.b, Page B3.3-8, Revision 21

Tuesday, April 19, 2016 3:14:39 PM 37

ILT-10 SRO NRC EXAM12. 215002K1.06 001


A central control rod has just been selected for withdrawal.


When the central control rod was selected, the RBM indication immediately went __________ .

After the RBM nulling sequence is complete, the RBM reading will be __________ .

downscale; 50%

downscale; 100%

to 100%; 50%

to 100%; 100%

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:39 PM 38


The RBM circuitry undergoes a nulling AND filtering sequence WHEN a rod is selected theRBM momentarily reads "0"and therefore a delay of at least 5 seconds must be allowedbetween selection and rod movement. The monitored LPRM detector signals are averaged andgain adjusted to obtain the RBM Average Flux value. The gain adjustment value is that valuenecessary to produce a RBM Average Flux value of 100%. The gain adjustment value is neverless than a value of one. In this way, if the local LPRM detector signals average to a valuegreater than 100%, the RBM Average Flux value is not scaled down to achieve 100% but ratheris kept at the naturally higher value for conservatism. Therefore, after the nulling sequence, theRBM reading will be "100" unless the local LPRM signal is > than 100.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine the effect on RBMwhen selecting a control rod.

The "A" distractor is plausible since the first part is correct. The second part is plausible thatRBM would normalize RTP back to 50% after the nulling sequence is complete since reactorpower is 50% RTP.

"B" is correct.

The "C" distractor is plausible since RBM indication does go to 100% at the end of the RBMnulling sequence; just not the beginning. The second part is plausible that RBM wouldnormalize RTP back to 50% after the nulling sequence is complete since reactor power is 50%RTP.

The "D" distractor is plausible since RBM indication does go to 100% at the end of the RBMnulling sequence; just not the beginning. The second part is plausible since it is correct.

Tuesday, April 19, 2016 3:14:39 PM 39


K/A:

215002 Rod Block Monitor System

K1. Knowledge of the physical connections and/or cause effect relationships between RODBLOCK MONITOR SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.06 Control rod selection: BWR-3,4,5 . . . . 3.0 3.1


C51-PRNM-LP-01203, Power Range Neutron Monitoring System, Ver. 9.0, EO 012.003.E.06


34GO-OPS-065-0, Control Rod Movement, Ver. 12.6 Bank question from HLT Database

Tuesday, April 19, 2016 3:14:39 PM 40

ILT-10 SRO NRC EXAM13. 215003K2.01 001

Unit 2 is starting up with Reactor power at 5% RTP.

24/48 VDC Cabinet 2A, 2R25-S015, de-energizes and can NOT be restored.


IRM Channels __________ will have lost their power supply.

2A & 2C

2A & 2D

2B & 2C

2B & 2D

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:39 PM 41


Intermediate Range Monitor Power Supplies:

IRM Channels A, C, E & G - 24/48 VDC Cabinet 2A, 2R25-S015 IRM Channels B, D, F & H - 24/48 VDC Cabinet 2B, 2R25-S016

Typically, power supplies are divisionalized where Div. 1 will be powered by the "A" side of apower system and Div. 2 will be powered by the "B" side of a power system. In some cases thisdoes not hold true, i.e. 2P41-F316A & F316D, PSW TB Isolation valves are powered by a Div. 1power supply (2R24-S025) and 2P41-F316B & F316C are powered by a Div. 2 power supply(2R24-S027).

K/A JUSTIFICATION:

This question satisfies the K/A statement by requiring the applicant to know the power supply toIRM Channels A and C.

"A" is correct.

The "B" distractor is plausible since there is a power arrangement where components "A" & "D"are powered by the same power supply (PSW 2P41-F316A & F316D) and applies this to IRMs2A & 2D.

The "C" distractor is plausible since there is a power arrangement where components "B" & "C"are powered by the same power supply (PSW 2P41-F316B & F316C) and applies this to IRMs2B & 2C.

The "D" distractor is plausible since this would be correct if 2R25-S016, 24/48 VDC Cabinet2B, is de-energized.

Tuesday, April 19, 2016 3:14:39 PM 42


K/A:

215003 Intermediate Range Monitor (IRM) System

K2. Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.01 IRM channels/detectors . . . . . . . . . . . . . . 2.5* 2.7


C51-IRM-LP-01202, Intermediate Range Monitors, Ver. 7.1, EO 012.003.C.15


A-20159, Plant Hatch Load List, 2R25-S015, Ver. 3.0 A-20160, Plant Hatch Load List, 2R25-S016, Ver. 2.0


Original Question

Unit 2 is starting up with Reactor power at 5% RTP.

24/48 VDC Cabinet 2B, 2R25-S016, de-energizes and can NOT be restored.

SRM Channels __________ will have lost their power supply.

A. 2A & 2C

B. 2A & 2D

C. 2B & 2C

D. 2B & 2D

Tuesday, April 19, 2016 3:14:39 PM 43

ILT-10 SRO NRC EXAM14. 215004K3.04 001

Unit 2 is in a Refueling outage (NO fuel removed yet).

The following conditions exits:

o Reactor Mode Switch REFUEL o SRM Shorting Links REMOVED

A Control rod located adjacent to SRM 2A is fully WITHDRAWN resulting in the followingSRM indications:

o SRM 2A 20 CPS o SRM 2B 7 CPS o SRM 2C 8 CPS o SRM 2D 7 CPS

Subsequently, a detector failure results in SRM 2D indication reading UPSCALE.


SRM 2A reactor power indication will __________ .

The SRM RPS SCRAM Setpoint is _________ .

reduce to a lower value; 7x104 CPS

reduce to a lower value; 3x105 CPS

remain approximately the same; 7x104 CPS

remain approximately the same; 3x105 CPS

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:39 PM 44


The Source Range Indication provides protective actions including a scram and rod blocks. Thetrip logic for this function is 1 out of 16; any trip on one of the four 2-out-of-4 Voter Modules,eight IRMs or four SRMs, will result in a trip. (This is also known as the non-coincident scramlogic). The protective actions are bypassed during certain operational conditions or by manuallybypassing the Neutron monitoring channel (SRMs, IRMs, or APRMs) on the P603 panel.

The UPSCALE TRIP SCRAM (setpoint 3 x 105 CPS) is used only during refueling operations. With the Shorting Links REMOVED, a single trip signal from any nuclear instrument channel (8IRMs or 4 SRMs) or a single 2/4 voter module will cause a full reactor scram. For a single 2/4logic module to trip, it must see a trip from at least 2 OPERABLE & UNBYPASSED APRMs. The Shorting Links are required to be removed by the Technical Requirements Manual in mode5 with any control rod withdrawn from a core cell containing one or more fuel assemblies andSDM not demonstrated per 42CC-ERP-010-0, Shutdown Margin Demonstration, for currentcore configuration.

The SRM UPSCALE ALARM (High Trip) rod block (setpoint 7 x 104 CPS) prevents rodwithdrawal due to excessively high power.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine the effect on SRMindication (reactor power) with a malfunctioning SRM.

The "A" distractor is plausible since the first part is correct. The second part is plausible since7x104 CPS is the setpoint for SRM Rod Block and applies this to the SRM SCRAM setpoint.

"B" is correct.

The "C" distractor is plausiblesince normally a single SRM will not cause a reactor scram andwould be correct if the SRM Shorting Links were installed. The second part is plausible since7x104 CPS is the setpoint for SRM Rod Block and applies this to the SRM SCRAM setpoint.

The "D" distractor is plausiblesince normally a single SRM will not cause a reactor scram andwould be correct if the SRM Shorting Links were installed. The second part is plausible since itis correct.

Tuesday, April 19, 2016 3:14:39 PM 45


K/A:

215004 Source Range Monitor (SRM) System

K3. Knowledge of the effect that a loss or malfunction of the SOURCE RANGEMONITOR (SRM) SYSTEM will have on following: (CFR: 41.7 / 45.4)

K3.04 Reactor power and indication . . . . . . . . 3.7 3.7


C51-SRM-LP-01201, Source Range Monitors, Ver. 7.1, EO 012.003.A.11


34FH-OPS-001-0, Fuel Movement Operation, Ver. 25.5

Tuesday, April 19, 2016 3:14:39 PM 46

ILT-10 SRO NRC EXAM15. 215005K1.08 001

Unit 2 is operating at 100% RTP with the following;

o APRM 2B INOP and BYPASSED

Subsequently, APRM 2C fails INOP.


The "background" for Reactor power on the SPDS "Primary Display" will be __________ .

green

yellow

orange

magenta

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:39 PM 47


Joe, this was a Pre-submittal RO question. Changes were incorporated based on yourES-401-9 comments.

If an instrument becomes INOP (Invalid), the background for the individual instrument turnsyellow. If too many instruments inputting to a parameter on the Primary Display become INOP,the background for the parameter turns yellow. A Primary display background will change toyellow dependent upon the number of inputs to the parameter and how many are off normal.(e.g. Rx power turns yellow with 2 APRMs off normal, INOP or logged out.)

"SCRAM" - orange when demand received and ten second timer running. Red if all control rods are not fully inserted after ten seconds. Message appears below thedrywell mimic.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know the response (causeeffect) between SPDS (Display control system) and APRMs due to APRMs becoming inop.

The "A" distractor is plausible since the "Primary Display" is normally green and with oneAPRM bypassed that the input for reactor power to SPDS is also bypassed, therefore whenAPRM 2C becomes inop, only one APRM is inop for reactor power indication (takes two)leaving the background green.

"B" is correct.

The "C" distractor is plausible since there are times when the "Primary Display" changes fromGREEN to ORANGE (orange when SCRAM demand received and ten second timer running)and since two APRMs are now inop a reactor scram should be in progress waiting on all rods tobe fully inserted.

The "D" distractor is plausible since there are times when a display will change to MAGENTA(Process Computer color display - Indications that one of the computers has stopped functioningproperly and MAGENTA data on the Heat Balance/Control Rod Position screens).

Tuesday, April 19, 2016 3:14:39 PM 48


K/A:

215005 Average Power Range Monitor/Local Power Range Monitor System

K1. Knowledge of the physical connections and/or cause effect relationships betweenAVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITORSYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.08 Display control system: Plant-Specific . . . . . . . . 3.0 3.0


X75-SPDS-LP-05601, Safety Parameter Display System, Ver. 6.1, EO 056.002.C.06


34SO-X75-002-2, Operation Of SPDS Equipment, Ver. 5.0

Tuesday, April 19, 2016 3:14:39 PM 49

ILT-10 SRO NRC EXAM16. 217000K6.01 001

Unit 2 was operating at 100% RTP when a transient occurs.

The following conditions exist:

o RWL -20 inches (lowest reached) o RCIC injecting at 400 gpm

Subsequently, 2R24-S021, 250V DC RX BLDG ESSEN. MCC 2A, DE-ENERGIZES.


RCIC will __________ injecting into the RPV.

If a steam line break occurs in the RCIC Pump Room causing an isolation signal, one (1) minute later, the RCIC Steam line _________ be ISOLATED.

continue; will

continue; will NOT

stop; will

stop; will NOT

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:39 PM 50


On a loss of 125/250 VDC Switchgear "A", 2R22-S016 (loss of 2R24-S021 & 2R25-S001),RCIC is not available and all indication for DC powered RCIC loads (valves and BarometricCondenser Vacuum and Condensate Pumps) are lost. Control power is also lost to the controlleralong with the Trip and Throttle Valve. Since only 2R24-S021 is lost, RCIC will continue topump at rated flow.

The Inboard Steam Supply Isolation Valve (F007) is located inside the drywell and is normallyopen. F007 is an AC MOV, powered from 2R24-S012B. F007 is AC powered since the valvewould require less maintenance during plant operations. This is important due to the valve'slocation in the Drywell.

The Outboard Isolation Valve (F008) is located outside the drywell and is normally open. F008is a DC MOV, powered from 2R24-S021. The function of E51-F007/E51-F008 is to isolate theRCIC system from the Main Steam System. These valves automatically close on a RCIC SystemIsolation signal.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to how a loss of electricalpower (2R24-S021) will effect RCIC (continue to run & will isolate).

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sincenumerous RCIC valves have lost power, including isolation valve, 2E51-F008, which will notchange position on the isolation signal and would be correct if isolation valve (F007) was alsopowered by this DC bus.

The "C" distractor is plausible since there is a time when a power loss will cause RCIC to stopinjecting (2R25-S001). The second part is plausible since it is correct.

The "D" distractor is plausible since there is a time when a power loss will cause RCIC to stopinjecting (2R25-S001). The second part is plausible since numerous RCIC valves have lostpower, including isolation valve, 2E51-F008, which will not change position on the isolationsignal and would be correct if isolation valve (F007) was also powered by this DC bus.

Tuesday, April 19, 2016 3:14:40 PM 51


K/A:

217000 Reactor Core Isolation Cooling System (RCIC)

K6. Knowledge of the effect that a loss or malfunction of the following will have on theREACTOR CORE ISOLATION COOLING SYSTEM (RCIC) : (CFR: 41.7 / 45.7)

K6.01 Electrical power . . . . . . . . . . . . . . . . 3.4 3.5


E51-RCIC-LP-03901, Reactor Core Isolation Cooling (RCIC), Ver. 6.1, LO H-OP-90000.001


34SO-E51-001-2, Reactor Core Isolation Cooling (RCIC) System, Ver. 27.0

Tuesday, April 19, 2016 3:14:40 PM 52

ILT-10 SRO NRC EXAM17. 218000A2.06 001

Unit 1 is operating at 100% RTP when a leak in the Drywell occurs.

Time Parameter

12:00 Drywell pressure is 2 psig 12:01 RWL is -102 inches 12:03 RWL is -114 inches 12:05 RWL is -116 inches


The EARLIEST listed time that SRVs will have automatically opened is ___________ .

IAW 31EO-EOP-015-1, CP-1, prior to the SRVs automatically opening, the NPO will __________ .

12:03; depress the ADS 102.5 second timer RESET pushbuttons

12:03; place the ADS INHIBIT switches to the INHIBIT position

12:05; depress the ADS 102.5 second timer RESET pushbuttons

12:05; place the ADS INHIBIT switches to the INHIBIT position

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:40 PM 53


ADS Initiation;

With a high Drywell pressure signal present (1.85 psig), the following must occur to initiateADS.

a. Low Reactor water level (Level 3) at +3.0 inches b. Low Reactor water level (Level 1) at -101 inches c. 102.5 second timer timed out d. CS Pump discharge pressure of 152 psig or RHR Pump discharge pressure of 127 psig. e. Once the 102.5 second timer has timed out, if RHR or CS Pump discharge pressure is

available, all 7 ADS Valves open.

IAW 31EO-EOP-015-1, CP-1, the NPO will be directed to place the "Inhibit" Switches in theINHIBIT position.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to predict when the ADS valveswill automatically open and then using a procedure (31EO-EOP-015-1) to mitigate thiscondition.

The "A" distractor is plausible since the first part is correct. The second part is plausible sincedepressing the pushbutton will reset the ADS 102.5 second timer which is normal when shuttingdown ADS.

"B" is correct.

The "C" distractor is plausible since Reactor Vessel Level 1 is 404 inches from the bottom of theRPV and Instrument zero is 517 inches from bottom of RPV and the difference results in -113inches; which was exceeded after 12:03, therefore selecting 12:05 as the earliest listed time. Thesecond part is plausible since depressing the pushbutton will reset the ADS 102.5 second timerwhich is normal when shutting down ADS.

The "D" distractor is plausible since Reactor Vessel Level 1 is 404 inches from the bottom of theRPV and Instrument zero is 517 inches from bottom of RPV and the difference results in -113inches; which was exceeded after 12:03, therefore selecting 12:05 as the earliest listed time. Thesecond part is plausible since it is correct.

Tuesday, April 19, 2016 3:14:40 PM 54


K/A:

218000 Automatic Depressurization System

A2. Ability to (a) predict the impacts of the following on the AUTOMATICDEPRESSURIZATION SYSTEM ; and (b) based on those predictions, use procedures tocorrect, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.06 ADS initiation signals present . . . . . . . . . . . . . . 4.2 4.3*


B21-ADS-LP-03801, Automatic Depressurization System (ADS), Ver. 4.1, EO 038.004.A.02


34SO-B21-001-1, Automatic Depressurization (ADS) & Low-Low Set (LLS) Systems, Ver. 12.11 31EO-EOP-015-1, CP-1, Ver. 9.0

Tuesday, April 19, 2016 3:14:40 PM 55

ILT-10 SRO NRC EXAM18. 218000A3.09 001

Unit 2 was operating at 100% RTP when a LOCA occurred:

o Drywell pressure 2.5 psig, slowly increasing o RHR/CS pumps RUNNING o AUTO BLOWDOWN TIMERS INITIATED, 602-306, has just ILLUMINATED o ALL SRVs are currently closed


AUTO BLOWDOWN TIMERS INITIATED, 602-306, will AUTOMATICALLY reset (turn GREEN) if ___________ .

RWL increases to -90 inches

Drywell pressure decreases to 1.5 psig

ALL of the RHR and CS pumps are secured

the ADS Inhibit Switches are placed in the "INHIBIT" position

A.

B.

C.

D.

Description:

34AR-602-306-2, Auto Blowdown Timers Initiated:

Indicates that all ADS permissive signals required are present to initiate the 102.5 secondtimer.

The following signals are needed to initiate the alarm:o RWL low (-101")o High Drywell Pressure (1.85 psig) or the RWL timer timed out (-101" for 11 min)o Confirmatory level (+3")

When the 102.5 second timer times out and if the RHR or CS discharge pressure permissive issatisfied, all 7 ADS valves will auto open.

Alarm 602-306 will clear (turn GREEN) when the 102.5 second timer is reset. The 102.5 secondtimer can be manually reset with ADS Logic Timer reset pushbuttons on panel 2H11-P602 atany time after they begin timing, but will restart if permissive signals are still present.

Tuesday, April 19, 2016 3:14:40 PM 56

ILT-10 SRO NRC EXAM

The 102.5 second timer will automatically reset anytime during timer operation if levelincreases above the Level 1 setpoint of -101". The 102.5 second timer will not automaticallyreset if Drywell pressure decreases to less than 1.85 psig or if the low level timer has timed out. The permissive signals for High Drywell Pressure and the low level timer are seal-in, and mustbe manually reset.

Once the 102.5 second timer times out, it can only be manually reset. If the timer is manuallyreset after the ADS valves have opened automatically, the ADS valves will close. Placing theInhibit Switches in INHIBIT will not reset the timers.

Placing the switch to INHIBIT will prevent/stop an ADS actuation but not reset the AutoBlowdown Timer. The RHR/CS discharge permissive only allows ADS to open SRVs, not stopthe ADS Blowdown Timer.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to monitor whichparameter/condition (RWL) that will automatically operate (reset) the ADS (Auto BlowdownTimer - 102.5 second timer).

"A" is correct.

The "B" distractor is plausible because Drywell pressure above 1.85 psig is a permissive signalfor alarm 602-306. The high Drywell pressure signal is seal-in and must be MANUALLY reset. Alarm 602-306 will clear if Drywell pressure decreases below 1.85 psig AND the High Drywellpressure permissive signal is MANUALLY reset.

The "C" distractor is plausible because the RHR/CS pumps discharge pressure permissive mustbe satisfied before ADS will automatically blowdown. If no RHR or CS pumps are running,ADS will NOT auto blowdown. If blowdown has already initiated and all RHR and CS pumpsare secured, all ADS valves will close. Securing the RHR and CS pumps will have no effect onthe 102.5 second timer.

The "D" distractor is plausible if because placing the INHIBIT switches to INHIBIT willprevent/stop an ADS actuation. Placing the INHIBIT switches to INHIBIT will have no effecton the 102.5 second timer.

Tuesday, April 19, 2016 3:14:40 PM 57


K/A:

218000 Automatic Depressurization System

A3. Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATIONSYSTEM including: (CFR: 41.7 / 45.7)

A3.09 Reactor vessel water level . . . . . . . . . . . . . . . . . . . . . 4.1* 4.2*


B21-ADS-LP-03801, Automatic Depressurization System (ADS), Ver. 4.1, EO 038.004.A.03 &EO 038.004.A.04


34AR-602-306-2, UTO BLOWDOWN TIMERS INITIATED, Ver. 3.1 34SO-B21-001-2, Automatic Depressurization (ADS) & Low-Low Set (LLS) Systems, Ver. 13.15

Tuesday, April 19, 2016 3:14:40 PM 58

ILT-10 SRO NRC EXAM19. 223002A1.04 001

Unit 2 is operating at 20% RTP when the following occurs:

At 10:00, 2B21-F022A, Inboard MSIV fully closes At 10:02, 2B21-F028B, Outboard MSIV fully closes At 10:04, 2B21-F022C, Inboard MSIV fully closes

Based ONLY on the above conditions with respect to MSIV position input to the RPS Logic,

At 10:03, the TOTAL number of 2C71-K14, RPS Scram Relays, that are ENERGIZED is __________ .

At 10:05, the TOTAL number of 2C71-K14, RPS Scram Relays, that are ENERGIZED is __________ .

four (4); zero (0)

four (4); four (4)

eight (8); zero (0)

eight (8); four (4)

A.

B.

C.

D.

Description:


The MSIVs position logic for RPS is a "Means" & "Extremes" arrangement (the two outsidelines are "Extremes", while the two inside lines are "Means").

Limit switches, mounted on the valve stems of all eight MSIVs, open their contacts when thevalves are less than 90% full open, giving a scram signal to RPS. All eight MSIVs provide inputto both RPS trip systems ("A" and "B"). Closure of either the inboard or outboard MSIV in agiven steam line trips the logic for that steam line. The MSIV scram logic is as follows:

a) Closure of any MSIV: NO ACTIONb) Closure of either MSIV in MSLs A & D (B & C): NO ACTION c) Closure of either MSIV in MSLs A & B (or C & D): HALF SCRAM CHANNEL "A" d) Closure of either MSIV in MSLs A & C (or B & D): HALF SCRAM CHANNEL "B"

Tuesday, April 19, 2016 3:14:40 PM 59

ILT-10 SRO NRC EXAM

e) Closure of any MSIV in three MSLs: FULL SCRAM

There are four sets of automatic scram logic (A1, A2, B1, and B2). Channel A1 receives itspower from the "A" RPS bus, and controls automatic scram relays K14A and K14E. ChannelA2 also receives its power from the "A" RPS bus and controls automatic scram relays K14C andK14G.

Channel B1 receives its power from the "B" RPS bus, and controls automatic scram relays K14Band K14F. Channel B2 also receives its power from the "B" RPS bus, and controls automaticscram relays K14D and K14H. A trip signal in any single channel will cause that trip system("A" or "B") to de-energize all 137 of its respective scram pilot solenoid valves.

For example, a trip signal into channel B2 will de-energize automatic scram relays K14D andK14H. Contacts will open in the scram actuation logic so that all 137 F118 valves will bede-energized (half scram condition).

Tuesday, April 19, 2016 3:14:40 PM 60

ILT-10 SRO NRC EXAM

K14D & H de-energize causing K14B & F to also de-energize resulting in a half scram conditionin the RPS Channel B ONLY.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine (predict) thechange in system relays based on MSIV (Nuclear Steam Supply) position changes.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sincethere is a combination of MSIV closures that will not result in a full RPS Channel actuation,only a half, and applies it to this combination of MSIVs.

The "C" distractor is plausible since there is a combination of MSIV closures that will not resultin a RPS Channel actuation and applies this to this combination of MSIVs. The second part isplausible since it is correct.

The "D" distractor is plausible since there is a combination of MSIV closures that will not resultin a RPS Channel actuation and applies this to this combination of MSIVs. The second part isplausible since there is a combination of MSIV closures that will not result in a full RPSChannel actuation, only a half, and applies it to this combination of MSIVs.

References:NONE

K/A:

223002 Primary Containment Isolation System/Nuclear Steam Supply

A1. Ability to predict and/or monitor changes in parameters associated with operating thePRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLYSHUT-OFF controls including: (CFR: 41.5 / 45.5)

A1.04 Individual system relay status . . . . . . . . . 2.6 2.8


C71-RPS-LP-01001, Reactor Protection System, Ver. 8.2, EO 300.008.A.02 B21-SLLS-LP-01401, Main Steam and Low Low Set, Ver. 10.0, EO 014.007.A.01

Tuesday, April 19, 2016 3:14:40 PM 61

ILT-10 SRO NRC EXAM


34AR-603-104-2, MSIVS NOT FULLY OPEN TRIP, Ver. 3.0 Modified from HLT Database Q#LT-300008-002

Original Question

Unit 2 is at 60% RTP when an electrical problems results in the following MSIVs going fullclosed:

o 2B21-F022A o 2B21-F028A o 2B21-F022B

Assuming no scram signals result from the pressure/power transient, complete the followingstatement regarding the designed RPS logic response?

Closure of the above combination of MSIVs will result in __________ .

A. RPS Full-Scram

B. RPS A Half-Scram

C. RPS B Half-Scram

D. Neither Half nor Full-scram

Tuesday, April 19, 2016 3:14:40 PM 62

ILT-10 SRO NRC EXAM20. 223002A4.04 001


At 10:00, an event occurred which resulted in the following conditions:

o RCIC Equipment Room, 160°F o Torus Area Ambient Temperature, 170°F

At 10:15,

o RCIC Equipment Room, 170°F o Torus Area Ambient Temperature, 175°F


The EARLIEST listed time that RCIC ISOLATION VLV F007/F008 NOT FULLY OPEN,602-336, alarm will be ILLUMINATED is __________ .

2E51-F007 valve position __________ be monitored on 2H11-P601 Vertical Display.

10:00; can

10:00; can NOT

10:15; can

10:15; can NOT

A.

B.

C.

D.

Description:

Any one of the following conditions will cause a Group 3 (HPCI) isolation:

1) HPCI Turbine Exhaust Diaphragm Press High (10 psig)2) HPCI Steam Line Flow High U2 202 in. H2O dp or -100 in. H2O dp)3) HPCI Steam Line Pressure Low 134 psig)4) HPCI Equipment (Pipe Penetration) Room Temp High (165°F)5) Suppression Chamber Area Air Temp High (165°F) (14 min TD)6) Suppression Chamber Area Diff Air Temp High (36°F) 14 min TD)7) Emergency Area Cooler Temp High (165°F)

Any one of the following conditions will cause a Group 4 (RCIC) isolation:

Tuesday, April 19, 2016 3:14:40 PM 63

ILT-10 SRO NRC EXAM

1) Suppression Chamber Area Air Temp High (165°F) (29 min TD)2) Suppression Chamber Area Diff Air Temp High (36°F) (29 min TD)3) RCIC Turbine Exhaust Diaphragm Press High (10 psig)4) RCIC Steam Line Flow High 143" H2O or -100" H2O 5) RCIC Steam Line Pressure Low (95 psig)6) RCIC Equipment Room Temp High (165°F)

With the RCIC Equipment Room temperature above 165°F, the RCIC system will receive anauto isolation signal without a time delay.

The 2H11-P601 Vertical Display has multiple system valve indications, but NOT all GroupIsolation Valves. 2E51-F007, Inboard Isolation valve, is one of the valves listed but 2E41-F041,Inboard Suction valve, is NOT indicated. Both are isolation valves and will close on theirrespective Group Isolaton signal.

K/A JUSTIFICATION:

This question satisfies the K/A statement by requiring the applicant to determine (monitor) theearliest time that a PCIS isolation signal will occur based on when (alarm) 602-336 will bereceived (result of 2E51-F007 valve closure) and if this valve position can be monitored on MainControl Room panel 2H11-P601.

The "A" distractor is plausible since there is an isolation signal that has a time delay of 29minutes and applies this to the RCIC Equipment room temperature isolation, which does nothave a time delay at all resulting in an isolation based on 170°F in the Torus Area at 10:00. Thesecond part is plausible since it is correct.

The "B" distractor is plausible since there is an isolation signal that has a time delay of 29minutes and applies this to the RCIC Equipment room temperature isolation, which does nothave a time delay at all resulting in an isolation based on 170°F in the Torus Area at 10:00. Thesecond part is plausible since there are some Isolation valves that are not on the Vertical Display(2E41-F041) and would be correct if asking for 2E41-F041.

"C" is correct.

The "D" distractor is plausible since the first part is correct. The second part is plausible sincethere are some Isolation valves that are not on the Vertical Display (2E41-F041) and would becorrect if asking for 2E41-F041.

Tuesday, April 19, 2016 3:14:40 PM 64


K/A:

223002 Primary Containment Isolation System/Nuclear Steam Supply

A4. Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

A4.04 System indicating lights and alarms . . . . . . . . . 3.5 3.6


T23-PC-LP-01301, Primary Containment, Ver. 7.5, EO 013.045.A.05


34AB-C71-001-2, Scram Procedure, Ver. 12.5 Modified from question which was used on 2015 NRC Exam Q#223002A4.01-001.

Original Question


At 10:00, an event occurred which resulted in the following conditions:

o HPCI Equipment (Pipe Penetration) Room, 170°F o Torus Area Ambient Temperature, 170°F

At 10:15,

o HPCI Equipment (Pipe Penetration) Room, 185°F o Torus Area Ambient Temperature, 185°F

With the above conditions,

Tuesday, April 19, 2016 3:14:40 PM 65

ILT-10 SRO NRC EXAM

The EARLIEST listed time that 2E41-F002, HPCI Isolation valve, should have received anautomatic isolation signal is __________ .

2E41-F002 valve position __________ be monitored on 2H11-P601 Vertical Display.

A. 10:00; can

B. 10:00; can NOT

C. 10:15; can

D. 10:15; can NOT

Tuesday, April 19, 2016 3:14:40 PM 66

ILT-10 SRO NRC EXAM21. 234000A3.01 001

Unit 2 is in a refueling outage with the following conditions.

o Rx Mode Switch ............ Locked in the REFUEL position o Control rods ................... One control rod is fully WITHDRAWN o Main Fuel Grapple ......... Lowered, but verified to be able to clear the cattle chute by 3 ft.

(Empty grapple move) o NO rod is selected

The Refuel Bridge is over the Spent Fuel Pool and is moving toward the core to pick up the fuelbundle at position 17-42.


The Refuel Bridge __________ move to the intended core location.

When the Refuel Bridge stops, if the Grapple Raise/Lower switch is placed in the "LOWER"position, the Main Grapple __________ lower.

will; will

will; will NOT

will NOT; will

will NOT; will NOT

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:40 PM 67


There are three types of refueling interlocks: o interlocks that prevent control rod motion, by causing rod blocks o interlocks that prevent the refueling platform from moving toward (over) the core, by

interrupting power to the motor that moves the refueling platform o interlocks that prevent hoist operation by interrupting power to the motor that moves the

hoist.

The refueling interlocks will not allow fuel to be moved in or near the core unless all controlrods are fully inserted. The refueling interlocks prevent the operation of loaded refuelingequipment over the core when any control rod is withdrawn. The refueling interlocks alsoprevent the withdrawal of any control rod when fuel is loaded on refueling equipment andoperating over the core. Fuel loaded is determined by a contact that opens when anything heavyenough to be fuel is loaded onto a hoist. In addition, when the reactor mode switch is inREFUEL, only one rod can be withdrawn. The selection of a second rod will initiate a rodblock. The Refuel Bridge will NOT move to the intended core location and the grapple willNOT be able to be moved.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to monitor if the Refuel Bridgemovement will be stopped according to Fuel handling equipment interlocks.

The "A" distractor is plausible since there is a time when the Refueling Bridge will move overthe core with not all rods in (Grapple Full Up). The second part is plausible since the FrameMounted/Trolley Mounted Hoist must be loaded, with not all rods in & over the core before theFrame/Trolley Mounted Hoist operation will be stopped and since the Main Grapple is notloaded applies this to the main grapple.

The "B" distractor is plausible since there is a time when the Refueling Bridge will move overthe core with not all rods in (Grapple Full Up). The second part is correct.

The "C" distractor is plausible since the first part is correct . The second part is plausible sincethe Frame Mounted/Trolley Mounted Hoist must be loaded, with not all rods in & over the corebefore the Frame/Trolley Mounted Hoist operation will be stopped and since the Main Grapple isnot loaded applies this to the main grapple.

"D" is correct.

Tuesday, April 19, 2016 3:14:40 PM 68


K/A:

234000 Fuel Handling Equipment

A3. Ability to monitor automatic operations of the FUEL HANDLING EQUIPMENTincluding: (CFR: 41.7 / 45.7)

A3.01 †Crane/refuel bridge movement: Plant-Specific . . . . . . 2.6 3.6


F15-RF-LP-04502, Refueling, Ver. 5.0, EO 045.018.A.01 & EO 045.019.A.02


34SV-F15-001-1, Refueling Interlocks and Hoist Limit Checks, Ver. 17.0 Bank question from HLT Database which was used on 2012 ILT-7 NRC Exam Q#24

Tuesday, April 19, 2016 3:14:40 PM 69

ILT-10 SRO NRC EXAM22. 239002A2.06 001

Unit 2 experienced an event resulting in the following conditions:

o Reactor Pressure 1135 psig, (HIGHEST reached) o MSIVs Closed o Torus temp 105°F, rising


The MAXIMUM number of SRV's that will have automatically opened is __________ .

IAW EOP PC flowchart, 31EO-EOP-012-2, the MINIMUM number of loop(s) of RHRREQUIRED to be placed in Torus Cooling is __________ .

eight (8); one (1)

eight (8): two (2)

four (4);one (1)

four (4);two (2)

A.

B.

C.

D.

Description:

LLS logic requires the following TWO initiating conditions:1) Reactor pressure > 1074 psig2) Any of the SRVs has opened AND has a tailpipe press 85 psig

On Unit 2, once initiated, LLS will control reactor pressure by cycling 2B21-F013B, D, F & G

On Unit 2, the following SRVs will automatically open if the pressure setpoint is reached:

(a) Electrical Setpoint 1120 psig:2B21-F013B2B21-F013D2B21-F013F2B21-F013G

Tuesday, April 19, 2016 3:14:40 PM 70

ILT-10 SRO NRC EXAM(b) Electrical Setpoint 1130 psig:

2B21-F013A2B21-F013C2B21-F013K2B21-F013M

(c) Electrical Setpoint 1140 psig:2B21-F013E2B21-F013H2B21-F013L

(d) Mechanical Setpoint 1150 psig for ALL SRVs.

IAW EOP PC flowchart, 31EO-EOP-012-2, an entry conditions is met once Torus watertemperature exceeds 100°F. If the WAIT UNTIL to maintain Torus water temperature below100°F can not be accomplished the following action block requires that ALL availiable Toruscooling be placed in service per 34SO-E11-0101-2, except RHR required for adequate corecooling. With Torus water temperature 105°F and rising this wait until has been met.

Per 34AB-T23-003-2, Torus Temperature above 95°F, subsequent operator action step 4.5 states"If Suppression Pool bulk average temperature exceeds 100°F START ALL available RHRloops in Suppression Pool Cooling and enter 31EO-EOP-012-2 and perform concurrently withthis procedure".

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to predict the response of SRVsto high RPV Pressure and then take action, per 31EO-EOP-012-2, to control the condition.

The "A" distractor is plausible because the 1st part of "A" is correct. The second part is correctper 34AB-T23-003-2 when Torus temperature is > 95°B but < 100°F or IAW 31EO-EOP-012-2if RHR is required for adequate core cooling in LPCI mode.

"B" is correct.

The "C" distractor is plausible because if RPV Pressure were to exceed 1120 psig but not exceed1130 psig, only 4 SRVs would open. The second part is correct per 34AB-T23-003-2 whenTorus temperature is > 95°F but < 100°F or IAW 31EO-EOP-012-2 if RHR is required foradequate core cooling in LPCI mode.

The "D" distractor is plausible because if RPV Pressure were to exceed 1120 psig but not exceed1130 psig, only 4 SRVs would open. The second part is correct.

Tuesday, April 19, 2016 3:14:40 PM 71


K/A:

239002 Relief/Safety Valves

A2. Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ;and (b) based on those predictions, use procedures to correct, control, or mitigate theconsequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.06 Reactor high pressure . . . . . . . . . . . . . . 4.1 4.3*


B21-LP-01401, Main Steam and Low Low Set, Ver. 10.0, EO 200.009.A.04 & EO 014.003.A.06


34SO-B21-001-2, ADS and LLS Systems, Ver. 13.1531EO-EOP-012-2, PC Primary Containment Control Ver 7.0 BWROG EPG/SAG REV 3 34AB-T23-003-2, Torus Temperature Above 95°F, Ver 2.4

Tuesday, April 19, 2016 3:14:40 PM 72

ILT-10 SRO NRC EXAM23. 241000A4.08 001


The Main Turbine Generator has just been tied to the grid (115 MWE).

Subsequently, VIBRATION ALARM, 650-136, is ILLUMINATED.

Investigation reveals that Turbine Vibration is 14 mils and has been at these values for the lasttwo (2) minutes.

Based on the above plant conditions and IAW 34SO-N30-001-2, Main Turbine Operation,

The Number 4 Turbine Control Valve will be __________ .

To CLOSE ALL of the Main Turbine Control & Stop Valves, the NPO will __________ onthe 2H11-P650 panel.

throttled open; select the "CLOSE VALVES" button on the "Control" "Speed" Screen

throttled open; depress the Turbine Trip pushbuttons

closed; select the "CLOSE VALVES" button on the "Control" "Speed" Screen

closed; depress the Turbine Trip pushbuttons

A.

B.

C.

D.

Description:

There are four Turbine Control Valves located between the turbine stop valves and the HPturbine first stage. Each valve has an internal pilot that opens just prior to the control valves.The internal pilot reduces the force against which the valve must be lifted.

Control Valves #1, 2, and 3 open simultaneously, and after 1, 2, and 3 are about 64% open(approximately 90% power) #4 TCV will begin to open as required by turbine load. This isreferred to as a Modified Partial Arc Admission.

Tuesday, April 19, 2016 3:14:40 PM 73

ILT-10 SRO NRC EXAM

Step 5.2.11 states: IF the following vibration limits are exceeded, the turbine must be shut down:

5.2.11.1 At speeds < 800 rpm, the maximum vibration allowed is 8 mils.

5.2.11.2 At speeds between 800 rpm and 1400 rpm, the vibration limit for any bearing is10 mils for 2 minutes. The maximum vibration allowed is 14 mils.

5.2.11.3 At speeds > 1400 rpm, the vibration limit for any bearing is 12 mils, unlessapproved by OPS Management and System Engineer. The turbine must beshutdown WHEN directed by OPS Management OR System Engineer.

5.2.15 The following electrical signals will trip the turbine:

5.2.15.2 High vibration (2 of 2 probes) 12 mils 5.2.15.7 Hign thrust bearing wear 35 mils

Since the vibration values have been in for 2 minutes, the Main Turbine has failed to auto tripand the NPO is required to depress both Turbine Trip pushbuttons.

Step 5.1.15 The "CLOSE VALVES" button closes the Stop Valves, Control Valves, &Intermediate Valves, ONLY. It has no effect on the turbine valves, WHEN theeither generator PCB is closed.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant how they will operate (close)the Main Turbine Control valves.

The "A" distractor is plausible because at this reactor power Control Valves 1, 2, and 3 will bethrottled open. The second part would be correct when the generator PCBs are open.

The "B" distractor is plausible because at this reactor power Control Valves 1, 2, and 3 will bethrottled open. The second part is correct.

The "C" distractor is plausible since the first part is correct. The second part would be correctwhen the generator PCBs are open.

"D" is correct.

References:Tuesday, April 19, 2016 3:14:40 PM 74

ILT-10 SRO NRC EXAMNONE

K/A:

241000 Reactor/Turbine Pressure Regulating System


A4.08 Control/governor valves (operation) . . . . . . . . . . . . . . . . . 3.5 3.4


N30-MTA-LP-01701, Main Turbine, Ver. 10.0, EO 200.088.A.01


34AR-650-136-2, Vibration Alarm (Main Turbine), Ver. 10.5 34SO-N30-001-2, Main Turbine Operation, Ver. 28.2 Modified from HLT Database Q#241000A2.23 001 which was used on 2012 HLT-7 NRC ExamQ#26

Original Question


The Main Turbine/Generator has just been tied to the grid (115 MWE).

Subsequently, VIBRATION ALARM, 650-136, is received.

Investigation reveals that Turbine Eccentricity is 3 mils and Vibration is 14 mils and has been atthese values for the last two (2) minutes.

Based on the above plant conditions,

Tuesday, April 19, 2016 3:14:40 PM 75

ILT-10 SRO NRC EXAMIAW 34SO-N30-001-2, Main Turbine Operation, the __________ is required to be used formonitoring Eccentricity on the Main Turbine.

To CLOSE the Main Turbine Control & Stop Valves, the NPO will __________ on the 2H11-P650 panel.

A. 2N32-R609, TURBINE METAL EXPANSION/TEMP, recorder; select the "CLOSE VALVES" button on the "Control" "Speed" Screen

B. 2N32-R609, TURBINE METAL EXPANSION/TEMP, recorder; depress the Turbine Trip pushbuttons

C. Human Machine Interface (HMI) Screen; select the "CLOSE VALVES" button on the "Control" "Speed" Screen

D. Human Machine Interface (HMI) Screen; depress the Turbine Trip pushbuttons

Tuesday, April 19, 2016 3:14:40 PM 76

ILT-10 SRO NRC EXAM24. 259001A1.04 001

Unit 2 is operating at 100% RTP with the following RWL indications:

o 2C32-R606A, GEMAC, indication: +36.9 inches o 2C32-R606B, GEMAC, indication: +36.0 inches o 2C32-R606C, GEMAC, indication: +37.9 inches

Subsequently, the REFERENCE leg for RWL instrument 2C32-R606A developes a leakequivalent to a 5 inch per minute change on 2C32-R606A.


INITIALLY, RFPT speeds will __________ .

Five (5) minutes later, FINAL RFPT speed will be equivalent to __________ .

DECREASE; windmilling speed

DECREASE; turning gear speed

INCREASE; windmilling speed

INCREASE; turning gear speed

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:40 PM 77


A leak in the reference leg causes 2C32-R606A & C to drift UP. Since one of these instrumentsis the Median signal (the level indication that is in the middle as determined by moduleC32-K648, the RWLC system senses that RWL is high (above the setpoint of the Master RFPcontroller). This causes an initial response of a RFP speed reduction, which causes actual RWLto decrease.

Once 2C32-R606A & C have drifted UP to the High RWL setpoint of 54.5 inches and bothRFPTs have tripped, as long as the Condensate System remains in service, the RFPTs will lowerin speed to approximately 100 rpm and be windmilling. If the Condensate System is not inservice or the minimum flow valve does not open, then RFPTs will lower in speed and go onTurning gear operation.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to predict the INITIAL RFPTspeed response to a reference leg break associated with R606A (R606C uses this same referenceleg) and the final RFPT speed.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sincethe RFPTs are designed to automatically go on Turning Gear operation after the RFPT isshutdown.

The "C" distractor is plausible since there is a time when an instrument leg failure will causeR606A & R606C to decrease (Variable leg) which will cause RFPTs to increase in speed. Thesecond part is plausible since it is correct.

The "D" distractor is plausible since there is a time when an instrument leg failure will causeR606A & R606C to decrease (Variable leg) which will cause RFPTs to increase in speed. Thesecond part is plausible since the RFPTs are designed to automatically go on Turning Gearoperation after the RFPT is shutdown.

Tuesday, April 19, 2016 3:14:40 PM 78


K/A:

259001 Reactor Feedwater System

A1. Ability to predict and/or monitor changes in parameters associated with operating theREACTOR FEEDWATER SYSTEM controls including: (CFR: 41.5 / 45.5)

A1.04 RFP turbine speed: Turbine-Driven-Only . . . . . . . . . . . . . . . . . 2.8 2.7


C32-RWLC-LP-00202, Reactor Water Level Control, Ver. 6.2, EO 002.020.A.06


34SO-N21-007-2, Condensate And Feedwater System, Ver. 54.2

Tuesday, April 19, 2016 3:14:40 PM 79

ILT-10 SRO NRC EXAM25. 259002K4.14 001

Unit 2 is at 75% RTP with the following RWLC System indications:

ILLUMINATED EXTINGUISHED

NOTE: 2C32-R600, FW Master Controller, is operating in AUTOMATIC.


2C32-R600, FW Master Controller, is currently using the __________ as its value for RWL.

The FW Master Controller is currently operating in __________ Control.

instrument selected from the Reactor Water Level Select switch; Single Element

instrument selected from the Reactor Water Level Select switch; Three Element

Median Level Signal Processor, 2C32-K648 output; Single Element

Median Level Signal Processor, 2C32-K648 output; Three Element

A.

B.

C.

D.

Description:

Reactor Water Level Mode Select;

AUTO positionPrimary: Utilizes the output of the Median Level Signal Processor.Alternate: Utilizes the signal selected by the Reactor Water Level Select switch as

the input signal for the RFPT Master Controller (C32-R600) & theSULCV Controller (C32-R619)

Tuesday, April 19, 2016 3:14:41 PM 80

ILT-10 SRO NRC EXAM

MAN positionPrimary: Utilizes the signal selected by the Reactor Water Level Select switch.Alternate: Utilizes the Median level Signal Processor as the input for the RFPT

Master Controller (C32-R600) & the SULCV Controller (C32-R619).

If the Master Controller determines that the primary input is over ranged high or low, it willautomatically swap to the alternate input REGARDLESS of the Reactor Water Mode Selectswitch position. The GREEN lamp above the Reactor Water Mode Select switch can be used todetermine which input the Master Controller is using. If the GREEN lamp above the ReactorWater Mode Select switch is ILLUMINATED then the Master Controller is using the MedianLevel Signal Output (K648). If the GREEN lamp above the Reactor Water Mode Select switchis EXTINGUISHED then the Master Controller is using the instrument selected by the ReactorWater Level Select switch.

With the green light above the "Reactor Level Mode Select" illuminated, the Master Controller isutilizing the MAN - Alternate which is the signal from K648 to control RFPTs.

The indicating light above the Feedwater Control Mode Select Switch on panel 2H11-P603ONLY ILLUMINATES WHEN the Master Controller is controlling using the 3 Element signal.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know the design feature ofthe RWLC System (Master Controller) in determining the RWL input for controlling RWL.

The "A" distractor is plausible is plausible since there is a time when "Manual" with "A" selectedwould be the input to the FW Master Controller and since the RWL Select Switch is selected to"A" position and the Reactor Level Mode Select is in "Manual" that the FW Master Controller isusing the currently selected RWL instrument. The second part is plausible since it is correct.

The "B" distractor is plausible is plausible since there is a time when "Manual" with "A" selectedwould be the input to the FW Master Controller and since the RWL Select Switch is selected to"A" position and the Reactor Level Mode Select is in "Manual" that the FW Master Controller isusing the currently selected RWL instrument. The second part is plausible since one of theswitches only functions in the selected position (RWL Select Switch position equals selectedinstrument) and since the Feedwater Control Mode Select Switch is in the 3 ELEM position,selects three element.

"C" is correct.

Tuesday, April 19, 2016 3:14:41 PM 81

ILT-10 SRO NRC EXAMone of the switches only functions in the selected position (RWL Select Switch position equalsselected instrument) and since the Feedwater Control Mode Select Switch is in the 3 ELEMposition, selects three element.

References:NONE

K/A:

259002 Reactor Water Level Control System

K4. Knowledge of REACTOR WATER LEVEL CONTROL SYSTEM design feature(s)and/or interlocks which provide for the following: (CFR: 41.7)

K4.14 Selection of various instruments to provide reactor water level input . . . 3.4 3.4


C32-RWLC-LP-00202, Reactor Water Level Control, Ver. 6.2, EO 002.019.A.01


34SO-N21-007-2, Condensate And Feedwater System, Ver. 54.2 Bank question from HLT Database Q#LT-002020-020

Tuesday, April 19, 2016 3:14:41 PM 82

ILT-10 SRO NRC EXAM26. 261000A4.01 001

BOTH Units are operating at 100% RTP with the Unit 2 refueling hatch installed.

Subsequently, the following occurs at the listed times:

At 10:00: A RWCU System break in the Unit 2 Reactor Building occurs At 10:05: 2D11-K609A-D, Rx. Bldg. Contaminated Area Radiation, monitors

indicate 20 mR/hr

Based on the above conditions and with NO operator actions,

At 10:08, the Unit 2 Reactor Building Stack release rate will be _________ than at 10:04.

At 10:15, the TOTAL number of SBGT fans that will be operating is __________ .

lower; four (4)

lower; three (3)

higher; four (4)

higher; three (3)

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:41 PM 83


A primary system leak in the Reactor Bldg results in Secondary Containment isolation signal at10:05. The Unit 1 & Unit 2 Reactor Building & Refuel Floor ventilation systems will isolate andall Unit 1 and Unit 2 SBGT trains will auto start.

SBGT will be discharging into the Main Stack. SBGT flow is less than the combined Rx. Bldg.& R/F flow, therefore at 10:08, the Rx. Bldg. Stack release rate will be lower than before theisolation at 10:04, since the Rx. Bldg. flow has stopped discharging into the Rx. Bldg. Stack.

Each Unit 1 SBGT train discharges into an 18 inch diameter duct. These two ducts tie togethernear the fan discharge into a single, common 18 inch duct which discharges to the main stack. Thus, it has been determined that one fan most likely operates near full flow while the secondfan operates with low flow.

At 10:15, one of the U1 SBGT fans (3 total discharging into Main Stack) has automaticallyshutdown due to low flow, therefore, the Main Stack release rate will be lower than the value at10:08 (4 total discharging into Main Stack).

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to monitor off-site release ratesthrough the Rx Bldg. Stack based on SBGT system operation.

The "A" distractor is plausible since the first part is correct. The second part is plausible if theapplicant remembers that all four SBGT fans are in service and does not consider that one fan onU1 will auto shutdown.

"B" is correct.

The "C" distractor is plausible if the applicant does not realize the Secondary Containmentisolation setpoint has been exceeded therefore, Normal Ventilation will still be in servicedischarging the untreated Rx. Bldg. atmosphere to the Rx. Bldg. Stack, causing the release rateto be higher. The second part is plausible if the applicant remembers that all four SBGT fans arein service and does not consider that one fan on U1 will auto shutdown.

The "D" distractor is plausible if the applicant does not realize the Secondary Containmentisolation setpoint has been exceeded therefore, Normal Ventilation will still be in servicedischarging the untreated Rx. Bldg. atmosphere to the Rx. Bldg. Stack, causing the release rateto be higher. The second part is plausible since it is correct.

Tuesday, April 19, 2016 3:14:41 PM 84


K/A:

261000 Standby Gas Treatment System


A4.01 †Off-site release levels: Plant-Specific . . . . . . . . . . 3.2* 4.0*


EOP-SCRR-LP-20325, Secondary Containment / Radioactivity Release Control, Ver. 3.0, EO 201.080.A.02


34AB-T22-003-2, Secondary Containment Control (SBGT Auto Starts)Unit Two FSAR 6.2.4.1.1 SBGT Safety Design Bases (Limits release)Modified question from HLT Database which was used in 2012 HLT-7 NRC Exam 2012-301Q#29

Original Question

Unit 2 is operating at 100% RTP with the following conditions:

o Unit 2 Refueling Hatch installed o 2A Standby Gas Treatment (SBGT) Fan is Danger Tagged out for maintenance

Subsequently, the following occurs at the listed times:

10:00 A RWCU System break in the Unit 2 Reactor Building 10:05 2D11-K609A-D, "Rx. Bldg. Contaminated Area Radiation increase to 20 mr/hr 10:10 The Supply breaker for 2R24-S012, 600V MCC, trips OPEN

With NO operator action, which ONE of the choices below predicts how the Off-siteRadioactive Release Rates will be affected?

At 10:08, the Reactor Building Stack release rate will be _________ than at 10:04.

Tuesday, April 19, 2016 3:14:41 PM 85

ILT-10 SRO NRC EXAMAt 10:15, the Main Stack release rate will be __________ at 10:08.

A. higher; approximately the same as

B. higher; lower than

C. lower; approximately the same as

D. lower; lower than

Tuesday, April 19, 2016 3:14:41 PM 86

ILT-10 SRO NRC EXAM27. 262001A1.04 001

Unit 2 is at 70% RTP with an operator transferring 4160 VAC Bus 2D to its NORMAL supply. The following conditions currently exist:

o Voltages are matched o Sync switch for the 4160 VAC 2D Normal breaker is in the ON position o Sync light is at its dimmest (12 O'Clock position and steady) o Station SVC Interlock Cutout switch for ACB 135484-135534 is in the OFF-(DOWN)

position

Subsequently, the operator places the control switch for 135484 (Normal supply breaker) in theclose position for one (1) second and then releases the switch.


The ampere indication will be as shown on __________ .

Figure 1 Figure 2 Figure 3 Figure 4

Figure 1

Figure 2

Figure 3

Figure 4

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:41 PM 87


The normal and alternate supply breakers are interlocked to prevent paralleling the 4160 VACbuses. Interlock cutout switches are provided to bypass this interlock between the normal andalternate supply breakers on 4160 VAC busses 2A, 2B, 2C, & 2D for a period of time whenswapping the power supplies is desired. This allows the buses to be paralleled only during thetime the house loads are being swapped from the startup source to the normal generator on lineoperation. Use of these switches is also allowed when shutting down to swap the house loadsover to the startup source. If both the alternate and supply breakers are closed at the same timewith the Interlock Service cutout switch in the NORMAL (UP) position, both breakers will tripopen.

Under heavy loads with both the alternate and normal supply breakers on the station service busclosed at the same time. A note in the 34SO-R22-001-2 states that "the current readings mayincrease on both the normal AND alternate supplies, WHEN both the normal AND alternatesupply breakers are closed" per engineering evaluations.

K/A JUSTIFICATION:

This question satisfies the K/A statement by requiring the applicant to know the expectedresponse concerning load currents to the electrical plant based on breaker interlocks.

The "A" distractor is plausible because on the emergency buses, once you place the normalsupply breaker to close, the alternate supply breaker will automatically open.

The "B" distractor is plausible because if the Sync Switch for 135434 is OFF, the normal supplybreaker will NOT close. The alternate supply breaker will remain closed.

"C" is correct.

The "D" distractor is plausible because if the Interlock Cutout switch is in NORMAL (UP) andnormal supply breaker is closed, both breakers will receive a trip signal.

References:NONE

K/A:

262001 A.C. Electrical Distribution

A1. Ability to predict and/or monitor changes in parameters associated with operating theA.C. ELECTRICAL DISTRIBUTION controls including: (CFR: 41.5 / 45.5)

A1.04 Load currents . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.7 2.9

Tuesday, April 19, 2016 3:14:41 PM 88

ILT-10 SRO NRC EXAM


R22-ELECT-LP-02702, 4160 VAC, Ver 7.0, EO 027.010.A.02


34SO-R22-001-2, 4160 VAC System, Ver 21.2


Original Question

Unit 2 is at 60% RTP with an operator transferring 4160 VAC Bus 2A to its Alternate supply. The following conditions currently exist:

o Voltages are matched o Sync switch for the 4160 VAC 2A Alternate breaker is in the ON positiono Sync light is at its dimmest (12 O'Clock position and steady)o Station SVC Interlock Cutout switch for ACB 135434-135454 is in the NORMAL (UP)

position

Subsequently, the operator places the control switch for ACB 135454 (Alternate supply breaker)in the close position and IMMEDIATELY releases the switch.


Tuesday, April 19, 2016 3:14:41 PM 89

ILT-10 SRO NRC EXAMTen (10) seconds after the operator releases the control switch, the ampere indication will be as shown on __________ .

Figure 1 Figure 2 Figure 3 Figure 4

A. Figure 1

B. Figure 2

C. Figure 3

D. Figure 4

Tuesday, April 19, 2016 3:14:41 PM 90

ILT-10 SRO NRC EXAM28. 262002K4.01 001

Unit 2 was operating at 100% RTP when the following occured:

o Loss of Off-Site Power (LOSP) o EDG 2A fails to start

Based on the above conditions and without any operator actions,

After the LOSP, the Unit 2 Vital AC Bus will INITIALLY receive power from __________ .

Subsequently, if this power supply is lost, the Vital AC Bus __________ automaticallytransfer to ANOTHER power source.

the Vital AC batteries; will

the Vital AC batteries; will NOT

600 V Bus 2D via Vital AC battery charger; will

600 V Bus 2D via Vital AC battery charger; will NOT

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:41 PM 91


The Vital AC bus has three different power supplies:

The normal power supply is from 600VAC Bus D (R23-S004) through the Vital AC BatteryCharger. The backup DC power supply is from the 240 VDC Vital AC Batteries. The batterywill supply Vital AC loads for approximately two hours without recharging. The alternatepower supply is from 600VAC Bus (R23-S003), through Vital AC 600 to 120 VAC EssentialTransformer A.

The LOSP and failure of EDG 2A causes 2R23-S003, 600VAC Bus 2C, to be de-energized. This bus supplies the Alternate power for Vital AC. Vital AC will not transfer to its alernatepower supply, 600VAC Bus 2C, until the non-essential loads are restored. If alternate power isnot available from 600VAC bus 2C and there was a drop in the inverter input voltage, theinverter would not attempt to transfer due to a sensing circuit which senses a loss of alternatepower and blocks the transfer.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know the designfeature/interlock for transferring Vital AC power from the preferred source to an alternatesource. The design feature/interlock is that Vital AC power will not shift to a second alternatesource unless the second alternate source is energized eventhough DC voltage is below thetransfer setpoint.

The "A" distractor is plausiblesince the first part is correct. The second part is plausible if theapplicant remembers that Vital AC will automatically swap to Alternate (600VAC Bus 2C)without any operator actions and does not consider the status of 600VAC Bus 2C. Alternate,600VAC Bus 2C, is de-energized due to EDG 2A failure and Vital AC will not transfer to ade-energized alternate (another) supply.

"B" is correct.

The "C" distractor is plausible if the applicant remembers that Vital AC can be powered by 600VBus 2D but does not remember which supply is Normal/Alternate/Backup, therefore selecting600VAC Bus D as the initial power source. The second part is plausible if the applicantremembers that Vital AC will automatically swap to Alternate (600VAC Bus 2C) without anyoperator actions and does not consider the status of 600VAC Bus 2C. Alternate, 600VAC Bus2C, is de-energized due to EDG 2A failure and Vital AC will not transfer to a de-energizedalternate (another) supply.

The "D" distractor is plausible if the applicant remembers that Vital AC can be powered by 600VBus 2D but does not remember which supply is Normal/Alternate/Backup, therefore selecting600VAC Bus 2D as the initial power source. The second part is correct.

Tuesday, April 19, 2016 3:14:41 PM 92


K/A:

262002 Uninterruptable Power Supply (A.C./D.C.)

K4. Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) design feature(s)and/or interlocks which provide for the following: (CFR: 41.7)

K4.01 Transfer from preferred power to alternate power supplies . . . . 3.1 3.4


R25-ELECT-LP-02705, Vital AC Electrical System, Ver. 3.1, EO 200.020.A.05


34SO-R25-002-2, 120/240 Volt Vital AC System, Ver. 5.3 34AR-651-150-2, Vital AC Sys On Alt Supply, Ver. 1.0 Bank question from HLT Database which was used on the 2012 HLT-7 NRC Exam 2012-301Q#31

Tuesday, April 19, 2016 3:14:41 PM 93

ILT-10 SRO NRC EXAM29. 263000A1.01 001

Unit 2 Division 1 125VDC Station Service Battery Chargers are being operated in theEQUALIZE Mode.

Based on the above conditions and IAW 34SO-R42-001-2, 125 VDC And 125/250 VDCSystem,

In EQUALIZE Mode, the charger output voltage to the battery will be __________ thanwhen the charger is operating in the FLOAT Mode.

Without re-charging, the 125 VDC Station Service batteries are sized to have adequatestorage capacity to carry the required load for approximately __________ .

equal; 2 hours

equal; 8 hours

higher; 2 hours

higher; 8 hours

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:41 PM 94


The normal lineup for all DC systems is energized with the battery chargers in service. Batterychargers supply all DC system loads with the batteries automatically picking up the load if poweris lost to the battery chargers. Battery chargers can be supplying a float or equalizing charge forthe batteries during their normal lineup.

If a battery is on a float charge, the battery chargers output voltage is equal to the battery voltageand the battery is "floating" on the system. The battery charger is supplying enough output toequal the load on the bus with current input and output from the battery equal.

If a battery is on an equalizing charge, the battery charger output voltage is elevated slightlyover battery voltage, with voltage input to the battery greater than the output. In this case, thebattery acts as a load on the battery charger. An equalizing charge brings the battery up to a fullycharged condition.

IAW 34SO-R42-001-2, 125 VDC & 125/250 VDC System, NOTE in Section to remove allchargers from service: "The batteries are rated for 2 hours service without the chargers inservice."

The 24/48VDC Division I battery is rated for 8 hours of continuous discharge from full tominimum charge.

The 125 VDC Substation Battery System contains two batteries with each battery rated for an8-hour continuous discharge.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to predict the charger outputvoltage (Battery charging) in the "Equalize" Mode and determine how long the 125VDC StationService Battery System will supply loads (discharging) without chargers available.

The "A" distractor is plausible since the first part would be correct if the batteries were on floatcharge. The second part is plausible since it is correct.

The "B" distractor is plausible since the first part would be correct if the batteries were on floatcharge. The second part is plausible since there are battery systems with an 8 hour limit(24/48VDC System or the 125VDC Switchyard) and would be correct if askin one of thosesystems.

"C" is correct.

The "D" distractor is plausible since the first part is correct. The second part is plausible sincethere are battery systems with an 8 hour limit (24/48VDC System or the 125VDC Switchyard)and would be correct if askin one of those systems. References:

Tuesday, April 19, 2016 3:14:41 PM 95


K/A:

263000 D.C. Electrical Distribution

A1. Ability to predict and/or monitor changes in parameters associated with operating theD.C. ELECTRICAL DISTRIBUTION controls including: (CFR: 41.5 / 45.5)

A1.01 Battery charging/discharging rate . . . . . . 2.5 2.8


R42-ELECT-LP-02704, DC Electrical Distribution, Ver. 7.1 EO 027.044.A.03


34SO-R42-001-2, 125 VDC And 125/250 VDC System, Ver. 10.0

Bank question from 2009 Duane Arnold Energy Center NRC Exam Q#14 for use at Plant Hatch.Modified question from 2015 Hatch NRC Exam Q#31

Original Question from DAEC 2009 Q#14

The Div 1 125 VDC battery charger is being operated in the equalize mode.

Which one of the following describes:

(1) the voltage relationship between the charger and the batteries

AND

(2) the design rating of the batteries if a loss of AC power occurred?

a. (1) In equalize, the charger output to the battery will be a higher voltage than when in thefloat mode

(2) The 125 VDC batteries are sized to supply emergency power for a 4-hour timeperiod.

b. (1) In equalize, the charger output to the battery will be a lower voltage than when in thefloat mode

(2) The 125 VDC batteries are sized to supply emergency power for a 4-hour time

Tuesday, April 19, 2016 3:14:41 PM 96

ILT-10 SRO NRC EXAM

c. (1) In equalize, the charger output to the battery will be a higher voltage than when in thefloat mode

(2) The 125 VDC batteries are sized to supply emergency power for an 8-hour timeperiod.

d. (1) In equalize, the charger output to the battery will be a lower voltage than when in thefloat mode

(2) The 125 VDC batteries are sized to supply emergency power for an 8-hour timeperiod.

Original Question from Hatch 2015 Q#31

Unit 2 is operating at 100% RTP with the "2A" and "2B" 125/250VDC Station Service Batterieson "Equalize" charge.

Subsequently,

o All Control Building Chillers and fans trip and can NOT be restoredo Control Building temperatures start increasing

With the "2A" and "2B" 125/250VDC Station Service Batteries on "Equalize" charge, the125/250VDC Station Service Battery Chargers output voltage will be __________ thebattery voltage.

With Control Building temperatures increasing and NO operator actions, the 125/250VDCStation Service Battery __________ .

A. equal to;Room Hydrogen concentrations will rise in each of the battery rooms

B. equal to;Chargers will trip when their high temperature trip setpoint is reached

C. greater than;Room Hydrogen concentrations will rise in each of the battery rooms

D. greater than;Chargers will trip when their high temperature trip setpoint is reached

Tuesday, April 19, 2016 3:14:41 PM 97

ILT-10 SRO NRC EXAM30. 264000K1.05 001

Which ONE of the following knobs can be used to completely shut off the fuel supply to arunning Plant Hatch Emergency Diesel Generator?

LOAD LIMIT

SYN INDICATOR

SPEED DROOP

SYNCHRONIZER

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:41 PM 98


Load Limit – controls the amount of fuel the DG can have. A setting of 0 is equal to 0% fuel.

Syn Indicator – indicates the number of complete turns of the Synchronizer knob.

Speed Droop – changes how quickly the DG recovers to 900 rpm after load is changed.

Synchronizer – adjusts speed/load of the DG. locally and a motor, attached to the governor,allows remote control from the control room.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know the cause/effect ofturning the load limit knob resulting in shutting off the fuel oil supply to a running EDG.

"A" is correct.

The "B" distractor is plausible since this is a knob on the EDG governor and would be correct ifasking the indications for the number of complete turns of the Synchronizer knob.

The "C" distractor is plausible since this is a knob on the EDG governor and would be correct ifasking how quickly the DG recovers to 900 rpm after load is changed.

The "D" distractor is plausible since this is a knob on the EDG governor and would be correct ifasking how to change the engine speed when running alone or to change the load (kW) when thediesel generator is paralleled to the grid.

Tuesday, April 19, 2016 3:14:41 PM 99


K/A:

264000 Emergency Generators (Diesel/Jet)

K1. Knowledge of the physical connections and/or cause effect relationships betweenEMERGENCY GENERATORS (DIESEL/JET) and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.05 Emergency generator fuel oil supply system . . . . . . . . . 3.2 3.3


R43-EDG-LP-02801, Emergency Diesel Generators, Ver. 8.0, 028.022.a.03


34SO-R43-001-2, Diesel Generator Standby AC System, Ver. 28.2 Bank question from HLT Database which was used on the 2009 Hatch NRC Exam Q#33

Tuesday, April 19, 2016 3:14:42 PM 100

ILT-10 SRO NRC EXAM31. 264000K5.05 001

IAW 34SO-R43-001-2, Diesel Generator Standby AC System, Section 7.2.1.2, SynchronizingDiesel Generator 2A to an Energized Bus,

When manually synchronizing EDG 2A to an energized bus, the synchroscope isREQUIRED to be rotating in a direction which will REDUCE the probability of causing _________ .

After EDG 2A output breaker is closed, if Crankcase pressure exceeds its trip setpoint, EDG 2A __________ automatically trip.

excessive VARs; will

excessive VARs; will NOT

a reverse power trip; will

a reverse power trip; will NOT

A.

B.

C.

D.

Description:

IAW 34SO-R43-001-2, Diesel Generator Standby AC System, steps;

Tuesday, April 19, 2016 3:14:42 PM 101

ILT-10 SRO NRC EXAM

5.1 P&Ls

5.1.2 Operating the diesel at near zero load OR zero output amperage may result in adiesel trip due to reverse power.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know the synchroscopedirection when paralleling AC sources to prevent reverse power (operational implication).

The "A" distractor is plausible since there is a time when excessive VARS will be encountered(large difference between EDG voltage and Emergency Bus voltage) and applying this to thiscondition. The second part is plausible since it is correct.

The "B" distractor is plausible since there is a time when excessive VARS will be encountered(large difference between EDG voltage and Emergency Bus voltage) and applying this to thiscondition. The second part is plausible since there is a time when crankcase pressure will notautomatically trip the EDG (LOSP) and applies this to this condition.

"C" is correct.

The "D" distractor is plausible since the first part is correct. The second part is plausible sincethere is a time when crankcase pressure will not automatically trip the EDG (LOSP) and appliesthis to this condition.

Tuesday, April 19, 2016 3:14:42 PM 102


K/A:

264000 Emergency Generators (Diesel/Jet)

K5. Knowledge of the operational implications of the following concepts as they apply toEMERGENCY GENERATORS (DIESEL/JET) : (CFR: 41.5 / 45.3)

K5.05 Paralleling A.C. power sources . . . . . . . . . . . 3.4 3.4


R43-EDG-LP-02801, Emergency Diesel Generators, Ver. 8.0, EO 028.023.A.11 & EO 028.023.A.02


34SO-R43-001-2, Diesel Generator Standby AC System, Ver. 28.2

Tuesday, April 19, 2016 3:14:42 PM 103

ILT-10 SRO NRC EXAM32. 272000K5.01 001

Unit 2 is at 50% RTP with the following conditions:

o The Hydrogen Injection System controller is in AUTOMATIC/EXTERNALo The Hydrogen flow setpoint is set at 5 SCFM

Subsequently, Reactor Power is increased to 100% RTP.

As reactor power is increased, the Hydrogen flow rate will __________ .

If, after reaching 100% RTP, Hydrogen flow is ISOLATED, the Main Steam Line radiationlevels will __________ .

INCREASE; DECREASE

INCREASE; REMAIN the SAME

REMAIN the SAME; DECREASE

REMAIN the SAME; REMAIN the SAME

A.

B.

C.

D.

Description:

Hydrogen injection involves radiological concerns. Nitrogen-16 is formed from theneutron-proton reaction of Oxygen-16. Nitrogen-16 has a short half-life (about 7 seconds) andemits a highly radioactive gamma when it decays. Without Hydrogen injection the bulk of theN-16 is converted to non-volatile nitrites and nitrates (NO2 and NO3).

With Hydrogen injection the chemically stable form of Nitrogen in the coolant changes fromnon-volatile Nitrites to volatile Ammonia (NH3). Due to its volatility, Nitrogen is carried out ofthe Reactor with steam flow. The increase in N-16 causes an increase in radiation levelsthroughout the plant.

When the Hydrogen Flow Controller is operating in AUTOMATIC/INTERNAL the desired flowrate is set with the Pulser Knob. In AUTOMATIC/EXTERNAL (Load Following), theHydrogen Controller will receive input to the controller from Total Feedwater Flow and basedon this signal, Hydrogen injection flow rate will be adjusted to a value dependent upon percentof total Feedwater flow.

Tuesday, April 19, 2016 3:14:42 PM 104

ILT-10 SRO NRC EXAM

If the HWC System is in "AUTOMATIC/INTERNAL", then as power level changes, theintroduction of hydrogen to the vessel will not change (remains at its current SCFM value). As aresult there will be more hydrogen than is needed for a lower Rx power or less hydrogen than isneeded for a higher Rx power. For a lower Rx power, this increased concentration of hydrogencombines with more nitrogen since its production is constant and oxygen production haslowered. The combination of hydrogen with nitrogen produces ammonia which is volatile. Since there is less oxygen for the nitrogen to combine with, the nitrogen is available forcombination with the hydrogen. So there will be less NO2 and NO3 and more NH3.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know how the MSLradiation levels (operational implication) will respond to the reduction in Hydrogen flow(decrease from less violatile ammonia).

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sincethe increase in reactor power will normally increase the MSL radiation levels compensating forthe reduction in hydrogen concentration.

The "C" distractor is plausible since there is a time when Hydrogen flowrate will remainapproximately the same (Internal operation of the controller instead of External operation) andwould be correct is asking Internal operation. The second part is plausible since it is correct.

The "D" distractor is plausible since there is a time when Hydrogen flowrate will remainapproximately the same (Internal operation of the controller instead of External operation) andwould be correct is asking Internal operation. The second part is plausible since the increase inreactor power will normally increase the MSL radiation levels compensating for the reduction inhydrogen concentration.

Tuesday, April 19, 2016 3:14:42 PM 105


272000 Radiation Monitoring System

K5. Knowledge of the operational implications of the following concepts as they apply toRADIATION MONITORING SYSTEM : (CFR: 41.7 / 45.4)

K5.01 Hydrogen injection operation's effect on process radiation indications: Plant-Specific . . . . . . . . . . 3.2 3.5


P73-HWCI-LP-07301, Reactor Water Chemical Injection Systems, Ver. 9.3, EO 026.035.A.01 & EO 026.034.A 08


Bank question from HLT Database Original question was from 2008 Browns Ferry NRC Exam Q#37

Tuesday, April 19, 2016 3:14:42 PM 106

ILT-10 SRO NRC EXAM33. 286000K2.02 001

IAW 34SO-X43-001-1, Fire Pumps Operating Procedure,

The power supply to 1X43-C001, Electric Fire Pump, is 4160VAC Bus __________ .

1E, 1R22-S005

1F, 1R22-S006

2F, 2R22-S006

2G, 2R22-S007

A.

B.

C.

D.

Description:IAW 34SO-X43-001-1, Fire Pumps Operating Procedure,

1R22-S005 4160V Bus 1E, Diesel Building, Switchgear Room 1EFR 6 Electric Fire Pump, 1X43-C001 RACKED

IN

K/A

This question satisfies the K/A statement by asking the applicant to know the electrical powersupply to the Electric Fire Pump.

"A" is correct.

The "B" distractor is plausible because this is an emergency bus that would receive power froman EDG during a LOSP and also has Division 1 & Division 2 4160 VAC pumps on it.

The "C" distractor is plausible because this is an emergency bus that would receive power froman EDG during a LOSP and has 4160 VAC pumps on it.

The "D" distractor is plausible because this is an emergency bus that would receive power froman EDG during a LOSP and has 4160 VAC pumps on it.

Tuesday, April 19, 2016 3:14:42 PM 107


K/A:

286000 Fire Protection System

K2. Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.02 Pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.9 3.1


X43-FPS-LP-03601, Fire Protection, Ver. 5.0


34SO-X43-001-1, Fire Pumps Operating Procedure, Ver. 6.1

Tuesday, April 19, 2016 3:14:42 PM 108

ILT-10 SRO NRC EXAM34. 290002K5.01 001

Unit 2 is operating at 100% RTP. The following information was obtained from a 3D MONICORE Monitor Case printout:

MFLCPR LOC MFLPD LOC MAPRAT LOC1.011 11-26 0.974 29-42-20 0.859 29-42-20


At this time, __________ of the thermal limits have been EXCEEDED.

IF a transient or accident were to occur, sufficient margin __________ exist to ensure at least99.9% of the fuel rods will avoid transition boiling.

Two (2);does NOT

Two (2);does

Only one (1);does NOT

Only one (1);does

A.

B.

C.

D.

Description:

Minimum Critical Power Ratio (MCPR) establishes a thermal limit for normal operation ANDtransient events. This is done to maintain the integrity of the fuel cladding by limiting the fuelrod power density to maintain nucleate boiling around the fuel rods thus avoiding a transition tofilm boiling.

For the MCPR limit, a ratio termed MFLCPR (Maximum Fraction of Limiting CriticalPower Ratio) is used by the Process Computer

Average Planar Linear Heat Generation Rate (APLHGR) is established as the thermal limit forpostulated accidents to maintain core geometry by minimizing gross cladding failure due to rapidfuel rewetting following a loss-of-coolant accident (LOCA).

Tuesday, April 19, 2016 3:14:42 PM 109

ILT-10 SRO NRC EXAM

For APLHGR, the Process Computer compares the APLHGR in a node to the (Flow orpower adjusted) APLHGR Limit for that node in a ratio termed MAPRAT

Linear Heat Generation Rate (LHGR) establishes a thermal limit for normal operation ANDtransient events. This is to maintain integrity of the fuel cladding by limiting the fuel rod powergeneration per linear foot to avoid overstressing the fuel cladding due to pellet-clad differentialexpansion.

The LHGR in a node is compared to the limit for that node in a ratio termed MFLPD(Maximum Fraction of Limiting Power Density).

All MAPRAT, MFLPD, AND MFLCPR values must be maintained < 1.0. IF any of theseparameters exceed 1.0, actions must be initiated such that the parameter is returned to < 1 within2 hours.

If any thermal limit ratio (MFLCPR, MFLPD, MAPRAT) increases to 0.980 or greater theSS/STA shall direct the Reactor Engineer on call to run predictions for the projected transientresponse. The Reactor Engineer will recommend appropriate reactivity adjustments if needed toensure all thermal limit ratios remain less than 0.990. If any thermal limit ratio increases to0.990 or greater the SS will take action to reduce the ratio to less than 0.990, as recommended bythe Reactor Engineer.

MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition tothe actual fuel assembly power. The MCPR Safety Limit is set such that 99.9% of the fuel rodsare expected to avoid boiling transition if the limit is not violated. The operating limit MCPRis established to ensure that no fuel damage results during anticipated operational occurrences(AOOs).

The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at anyaxial location. Limits on the APLHGR are specified to ensure that the peak cladding temperature(PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed2,200°F.

The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axiallocation. A value of 1% plastic strain of the fuel cladding has been defined as the limit belowwhich fuel damage caused by overstraining of the fuel cladding is not expected to occur

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine which thermallimits have exceeded the allowable values, based on a printout from the Process Computer. Andbased on the conditions, determine if an assumption regarding an initial condition of transientanalyses is satisfied (operational implication).

Tuesday, April 19, 2016 3:14:42 PM 110

ILT-10 SRO NRC EXAMThe "A" distractor is plausible because while thermal limit ratios must be maintained < 1, thecritical power ratio (CPR) thermal limit must be maintained > 1. The second part is correct.

The "A" distractor is plausible because while thermal limit ratios must be maintained < 1, thecritical power ratio (CPR) thermal limit must be maintained > 1. The second part is plausible ifthe applicant believes the MAPRAT thermal limit ratio has been violated, or if the applicantapplies the other thermal limit failure mechanisms (2,200°F or % plastic strain) to MFLCPR.

"C" is CORRECT.

The "D" distractor is plausible because the first part is correct. The second part is plausible if theapplicant believes the MAPRAT thermal limit ratio has been violated, or if the applicant appliesthe other thermal limit failure mechanisms (2,200°F or % plastic strain) to MFLCPR.

Tuesday, April 19, 2016 3:14:42 PM 111


K/A:

290002 Reactor Vessel Internals

K5. Knowledge of the operational implications of the following concepts as they apply toREACTOR VESSEL INTERNALS : (CFR: 41.5 / 45.3)

K5.01 †Thermal limits . . . . . . . . . . . . . . . . 3.5 3.9

ABOVE K/A REPLACES THE BELOW K/A AFTER PHONE CONVERSATION WITHCHIEF EXAMINER JOE VIERA ON 1/20/2016.

290002 Reactor Vessel Internals

K5. Knowledge of the operational implications of the following concepts as they apply toREACTOR VESSEL INTERNALS : (CFR: 41.5 / 45.3)

K5.04 †PCIOMR Plant-Specific . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 3.7


C95-PC-LP-40001, Process Computer, Ver. 3.0 / EO 400.060.E.01


34GO-OPS-005-2, Power Changes, Ver. 28.634SV-SUV-020-0, Core Parameter Surveillance, Ver. 17.19Unit 2 Tech Spec Bases, LCO 3.2.2, MCPRUnit 2 Tech Spec Bases, Safety Limits, 2.1.1.2

Tuesday, April 19, 2016 3:14:42 PM 112

ILT-10 SRO NRC EXAM35. 290003K3.02 001

Unit 1 is in Mode 3. Unit 2 is in Mode 4 currently, moving irradiated fuel in the Unit 2 Fuel Pool.

Subsequently, the following occurs;

o 600VAC Bus 1C DE-ENERGIZES o 600VAC Bus 1D DE-ENERGIZES


LCO 3.7.5, Control Room Air Conditioning (AC) System, is APPLICABLE to __________ .

Main Control Room instrumentation temperatures will __________ .

BOTH Units; start slowly increasing

BOTH Units; remain approximately the same

Unit 1 ONLY; start slowly increasing

Unit 1 ONLY; remain approximately the same

A.

B.

C.

D.

Description:

For Unit 1 and Unit 2, MCREC is required to be operable IAW LCO 3.7.4:

LCO 3.7.4 Two MCREC subsystems shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3,During movement of irradiated fuel assemblies in the secondary containment,During CORE ALTERATIONS,During operations with a potential for draining the reactor vessel (OPDRVs).

Tuesday, April 19, 2016 3:14:42 PM 113

ILT-10 SRO NRC EXAMPower Supplies:

600VAC Bus 1C:

C012A, 1R24-S002 (powered from 600V 1C) AHU 1Z41-B003A/8A is powered by 600 VAC MCC, 1R24-S002

600VAC Bus 1D:

C012B, 1R24-S003 (powered from 600V 1D) AHU 1Z41-B003B/8B is powered by 600 VAC MCC, 1R24-S003

AHU 1Z41-B003C/8C is powered by 600 VAC MCC, 1R24-S029. 1R24-S029 is powered from1R24-S003 with alternate power form 1R24-S002 (either 600VAC 1C or 1D).

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine Main ControlRoom computer and instrumentation temperatures with a loss of MCREC ventilation.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible if theapplicant remembers that some U1 components are powered by U2 600VAC Buses (RHRInjection valves) and some U2 components are powered by U1 600VAC Buses (RHR Injectionvalves & 2P41-C002 - Standby PSW pump ) and then applies this to the conditions given toselect temperatures will remain approximately the same.

The "C" distractor is plausible since it would be correct if U2 was not moving fuel in SecondaryContainment. The second part is plausible since it is correct.

The "D" distractor is plausible since it would be correct if U2 was not moving fuel in SecondaryContainment. The second part is plausible if the applicant remembers that some U1 componentsare powered by U2 600VAC Buses (RHR Injection valves) and some U2 components arepowered by U1 600VAC Buses (RHR Injection valves & 2P41-C002 - Standby PSW pump ) andthen applies this to the conditions given to select temperatures will remain approximately thesame.

Tuesday, April 19, 2016 3:14:42 PM 114


K/A:

290003 Control Room HVAC

K3. Knowledge of the effect that a loss or malfunction of the CONTROL ROOM HVACwill have on following: (CFR: 41.7 /45.6)

K3.02 Computer/instrumentation: Plant-Specific . . . . . . 3.3 3.6


Z41-MCREC-LP-03701, Main Control Room Environmental Control, Ver. 6.2, EO H-OP-90000.001


34SO-Z41-001-1, Control Room Ventilation System, Ver. 22.4

Tuesday, April 19, 2016 3:14:42 PM 115

ILT-10 SRO NRC EXAM36. 295001AK3.04 001


Subsequently, ONE Recirc Pump trips.

Unit 2 stabilizes inside the Region of Potential Instabilities on the power to flow map.

APRM bandwidth oscillations are observed and are INCREASING.


The LOWEST listed APRM peak to peak oscillation that REQUIRES a reactor scram is __________ .

The reason for the reactor scram is to avoid EXCEEDING the Limit for __________ .

3%; MCPR

3%; MAPRAT

7%; MCPR

7%; MAPRAT

A.

B.

C.

D.

Description:


IAW 34AB-C51-001-2, Step 4.1 contains the following:

IF at any time while operating in the Region of Potential Instabilities (RPI), any of the followingconditions exist, enter 34AB-C71-001-2, Scram Procedure, AND SCRAM the Reactor:

Any APRM is observed to have bandwidth oscillations > 5% peak to peak oscillationsAND INCREASING.

Tuesday, April 19, 2016 3:14:42 PM 116

ILT-10 SRO NRC EXAM

Whenever taking APRM OR LPRM readings, confirm the neutron flux noise level isnormal. (Normal flux noise in the range of 3% - 5% at 100% power AND 1% - 3% at 50%power).

Basis: In plants with Thermal Trip Monitors, automatic scram protection is not provided untilAPRM oscillations reach a peak magnitude of 120% of rated. During regional oscillations,because of the partial cancellation of out-of-phase LPRM signals, local neutron flux can besignificantly higher than indicated by the APRM signal and the margin to the Safety LimitMCPR may not be assured under these conditions. The analyses indicate that the Safety LimitMCPR may be violated when APRM oscillations are approximately 15% to 45% peak-to-peakwhich is the practical threshold for operator detection of oscillations. Since Region A representsthe least stable region of the power/flow operating domain, the potential to rapidly encounterlarge magnitude core thermal hydraulic oscillations is increased. During transients, the operatormay not have sufficient time to manually insert control rods or increase core flow to mitigate theoscillations before they reach an unacceptable magnitude. Therefore, the prompt action ofmanually scramming the plant is required to ensure protection of the Safety Limit MCPR.

U2 Bases TS 3.4.1 states "With only one recirculation loop in operation, modifications to therequired APLHGR limits [LCO 3.2.1, “Average Planar Linear Heat Generation Rate(APLHGR)"], MCPR limits [LCO 3.2.2, "Minimum Critical Power Ratio (MCPR)"], LHGRlimits, [LCO 3.2.3, "Linear Heat Generation Rate (LHGR)"], and APRM Simulated ThermalPower - High setpoint (LCO 3.3.1.1) must be applied to allow continued operation.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant the reason for a reactor scramwhen core flow is reduced (inside RPI) with APRM oscillations due to reduced flow.

The "A" distractor is plausible since this is the normal value for APRM neutron flux noise(oscillations). The second part is plausible since it is correct.

The "B" distractor is plausible since this is the normal value for APRM neutron flux noise(oscillations). The second part is plausible since there is a time when APLHPR/MAPRAT mustbe adjusted based on a lower core flow value (Single Loop operation) and applies this as thereason for scramming the reactor.

"C" is correct.

The "D" distractor is plausible since the first part is correct. The second part is plausible sincethere is a time when APLHPR/MAPRAT must be adjusted based on a lower core flow value(Single Loop operation) and applies this as the reason for scramming the reactor.

Tuesday, April 19, 2016 3:14:42 PM 117


K/A:

APE: 295001 Partial or Complete Loss of Forced Core Flow Circulation

AK3. Knowledge of the reasons for the following responses as they apply to PARTIAL ORCOMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : (CFR: 41.5 / 45.6)

AK3.04 Reactor SCRAM . . . . . . . . . . . . . . 3.4 3.6


LT-SG-50462, Region of Potential Instabilities/Instabilities and Oscillations, Ver. 10.1, Obj. 200.078.F


34GO-OPS-005-2, Power Changes, Ver. 28.6 34AB-C51-001-2, Entry Into The Region Of Potential Instabilities (RPI) Or Reactor OperationsWith Inoperable OPRM System, Ver. 9.0

Modified for Hatch from a question used on the 2008 River Bend Station NRC Exam Q#1.

Original Question

What is the reason for an automatic scram upon entry into the Exclusion Region of the Power toFlow map?

A. To avoid exceeding the Reactor Pressure Safety Limits during flux oscillations.

B. To avoid exceeding the MCPR Safety Limit during flux oscillations.

C. To avoid exceeding the MAPRAT operating limit due to low coolant flow.

D. To avoid exceeding the LHGR operating limit due to low coolant flow.

Tuesday, April 19, 2016 3:14:42 PM 118


Unit 1 is operating at 100% RTP when 1R25-S036, Essential Cabinet 1A, de-energizes andCANNOT be re-energized.

The SRO directs a NPO to cross connect Instrument Buses IAW 34AB-R25-002-1, Loss Of Instrument Buses.

Based on the above conditions and IAW 34AB-R25-002-1,

To energize 1R25-S064, Instrument Bus 1A, from 1R25-S065, Instrument Bus 1B, the NPO must utilize __________ .

Once power is restored to 1R25-S064, 1N62-F527, Off Gas Stack Inlet Valve, __________ .

breakers ONLY; will automatically open

breakers ONLY; must be manually opened

breakers and disconnects; will automatically open

breakers and disconnects; must be manually opened

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:42 PM 119


Unit 1 does not have the keylock switch arrangement; circuit breaker manipulation is utilized tocrosstie instrument buses 1R25-S064 and S065.

Unit 2 utilizes a keylock disconnect switch arrangement for cross-tying instrument buses2R25-S064 and S065.

Per 34AB-R25-002-1, when power is restored to instrument Bus 1A, place the 1N62-F527, Off Gas Stack Inlet Valve,to open and confirm the the valve opens on 1N62-P600.

4.1.2 Automatic Actions

5. 1N62-F527, Off Gas Stack Inlet Valve, CLOSES.7. The following temperature controllers lose power and their associated TCV fails in the

FULL OPEN position. These controllers return to AUTO upon power restoration.o 1P41-R658, Main Generator H2 Temp Controllero 1P41-R657, Main Turbine Oil Temp Controllero 1N32-R602, EHC Oil Temp Controllero 1P41-R901A, "A" RFPT Oil Temp Controllero 1P41-R901B, "B" RFPT Oil Temp Controller

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know how A.C. distrubtionis affected with a partial loss of the Essential buses (1R25-S036, Essential Cabinet). .

The "A" distractor is plausible since the first part is correct. The second part is plausible becausethere several temperature control valves (TCV) that automatically return to to service and controltemperatures (Main Generator H2 TCV, Main Turbine Oil TCV, EHC Oil TCV, and both RFPTOil TCVs).

"B" is correct.

The "C" distractor is plausible because the first part it is correct for UNIT 2. The second part isplausible because there several temperature control valves (TCV) that automatically return to toservice and control temperatures (Main Generator H2 TCV, Main Turbine Oil TCV, EHC OilTCV, and both RFPT Oil TCVs).

The "D" distractor is plausible because the first part it is correct for UNIT 2. the second part iscorrect.

Tuesday, April 19, 2016 3:14:42 PM 120


K/A:

APE: 295003 Partial or Complete Loss of A.C. Power

AK2. Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C.POWER and the following: (CFR: 41.7 / 45.8)

AK2.03 A.C. electrical distribution system . . . . . . . . . . . . . . . 3.7 3.9


R23-ELECT-LP-02703, 600/480/208 VAC, Ver. 5.1, EO 200.019.M.04 / 200.019.M.03 /027.026.O.01


34AB-R25-002-1, Loss Of Instrument Buses, Ver. 7.0

Tuesday, April 19, 2016 3:14:42 PM 121

ILT-10 SRO NRC EXAM38. 295004AA1.03 001


o A loss of 2R22-S016, 125/250VDC Switchgear 2A, occurs o Subsequently, a Main Turbine trip occurs

Based on the above conditions and IAW 34AB-R22-001-2, Loss Of DC Buses,

The Main Generator output breakers __________ .

The 4160 VAC Station Service Buses __________ automatically transfer to their alternate supply.

will automatically open; will

will automatically open; will NOT

must be manually opened; will

must be manually opened; will NOT

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:42 PM 122


IAW 34AB-R22-001-2, "Caution" at step 4.1.4 states "If a turbine trip signal is received (notimmediately expected due to loss of 2R22 S016) the turbine will trip but generator outputbreakers will not auto open. 4160V buses will not transfer to startup supply. To preventexcessive motoring of the generator and overheating of the last stage low pressure turbinebuckets, the main generator output breakers must be opened within 4 minutes of the turbine trip,and exhaust hood spray available and functioning properly."

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to monitor the AC distribution(Station Service Buses will not transfer) with the loss of the DC Bus (2R22-S016).

The "A" distractor is plausible since normally this is how the Main Generator PCBs function andwould also be correct if 2R25-S002 had been lost instead of 2R25-S003 via 2R22-S016. Thesecond part is plausible if the applicant remembers that normally, this is the correct operation ofthese breakers. Also plausible since it would be the correct answer if 2R22-S017 had been lostrather than 2R22-S016.

The "B" distractor is plausible since normally this is how the Main Generator PCBs function andwould also be correct if 2R25-S002 had been lost instead of 2R25-S003 via 2R22-S016. Thesecond part is plausible since it is correct.

The "C" distractor is plausible since the first part is correct. The second part is plausible if theapplicant remembers that normally, this is the correct operation of these breakers. Also plausiblesince it would be the correct answer if 2R22-S017 had been lost rather than 2R22-S016.

"D" is correct.

Tuesday, April 19, 2016 3:14:42 PM 123


K/A:

APE: 295004 Partial or Complete Loss of D.C. Power

AA1. Ability to operate and/or monitor the following as they apply to PARTIAL ORCOMPLETE LOSS OF D.C. POWER : (CFR: 41.7 / 45.6)

AA1.03 A.C. electrical distribution . . . . . . . . . . . . . . . . . . 3.4 3.6


R42-ELECT-LP-02704, DC Electrical Distribution, Ver. 7.1, EO 200.018.a.01


34AB-R22-001-2, Loss of DC Buses, Ver. 6.0 Bank question from HLT Database which was used on 2009 HLT-4 NRC Exam Q#39

Tuesday, April 19, 2016 3:14:42 PM 124

ILT-10 SRO NRC EXAM39. 295005G2.2.37 001

Unit 2 was operating at 75% RTP with 4160 VAC 2G Emergency Bus on Alternate Supply dueto the Normal Supply breaker DANGER tagged out of service.

Subsequently, the Unit 2 Main Turbine trips on Generator Differential Overcurrent.

Based on the above conditions and one (1) minute after the Main Turbine trips,

The Recirc pumps __________ be available to prevent Thermal Stratification in the RPV.

4160 VAC 2C __________ to its Startup Auxiliary Transformer.

will; must be manually transferred

will; will automatically transfer

will NOT; must be manually transferred

will NOT; will automatically transfer

A.

B.

C.

D.

Description:

NOTE: Startup Auxiliary Transformer (SAT) Unit Auxiliary Transformer (UAT)

Upon a loss of the Normal Power source to 4160 VAC busses A, B, C and D, the normal supplybreakers will automatically open and the alternate supply breakers will automatically close whenboth generators output PCBs are opened. This is known as a FAST TRANSFER. Thisoperation normally occurs following a manual or auto turbine trip. It ensures that all loadsremain energized. There are several conditions in the electrical distribution system which willlockout (prevent) a fast transfer. Those conditions are as follows:

o If the fast transfer does not occur within 0.2 seconds, the automatic fast transfer is lockedout, requiring a manual transfer to re-energize the 4160 VAC Station Service Busses 2A,B, C, & D.

o If any of the 4160 VAC Emergency Busses are tied to the SAT 2C (Alternate), a fasttransfer of house loads is prohibited and the SAT supply breakers to 4160 2A and Breceive a trip signal. However 2C AND 2D can be manually transferred to AlternateSupply (SAT 2D).

Tuesday, April 19, 2016 3:14:42 PM 125

ILT-10 SRO NRC EXAMSince SAT 2C remains energized, the emergency buses will be energized. When the MainTurbine trips, the Fast transfer will not occur for 4160 VAC Busses 2A, 2B, 2C and 2D.

4160 VAC Station Service Buses A, B, C and D, located on the 130' elevation Turb Bldg,receives power from two different sources depending upon plant conditions.

1. Unit Shutdown A & B from Startup Auxiliary Transformer (SAT) C C & D from SAT D

2. Unit in Operation o A & B from Unit Auxiliary Transformer (UAT) B o C & D from UAT A

4160 VAC Emergency Buses E, F, and G, located in the Diesel Generator Building, receivepower from one of three (3) sources.

o Normally supplied by SAT "D" o Alternate supplied by SAT "C" o Emergency supplied by - Emergency Diesel Generator

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine the availability(energized) of the 4160VAC Buses (3 Safety Related equipment).

The "A" distractor is plausible because the first part is correct when 4160 VAC 2G EmergencyBus is aligned to its NORMAL supply, SAT 2D, when the Main Turbine/Generator tripped. Thesecond part is corretc.

The "B" distractor is correct when 4160 VAC 2G Emergency Bus is aligned to its NORMALsupply, SAT 2D, when the Main Turbine/Generator tripped.

"C" is correct.

The "D" distractor is plausible since the first part is correct. The second part is correct when4160 VAC 2G Emergency Bus is aligned to its NORMAL supply, SAT 2D, when the MainTurbine/Generator tripped.

References:NONE

Tuesday, April 19, 2016 3:14:42 PM 126

ILT-10 SRO NRC EXAM

APE: 295005 Main Turbine Generator Trip

G2.2.37 Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12) . . . . . . . . . . . . . . 3.6 4.6


R22-ELECT-LP-02702, 4160 VAC, Ver. 7.0, EO 027.009.A.03


34AB-R22-004-2, Loss of 4160V Bus 2A, 2B, 2C or 2D, Ver. 1.11 34SO-R22-001-2, 4160V AC System Operation, Ver. 21.2

HLT Database Q#262001A2.01-001 which was used on the 2012 HLT-06 NRC Exam Q#28

Original Question

Unit 2 is operating at 50% RTP with 4160 VAC 2E, 2R22-S005, powered from StartupAuxiliary Transformer (SAT) 2C.

Subsequently, the Unit 2 Main Turbine trips.

After the Main Generator trips, the MAXIMUM number of Station Service Buses that will beenergized is __________ .

At this time, 34SO-R22-001-2, 4160V AC System Operation, can be used to MANUALLYre-energize 4160V Buses __________ .

A. zero (0); 2C and 2D

B. zero (0); 2A and 2B

C. two (2); 2C and 2D

D. two (2); 2A and 2B

Tuesday, April 19, 2016 3:14:42 PM 127


Unit 2 was at 100% RTP when a spurious reactor scram occurred.

ALL Control Rods fully inserted with the following EXCEPTIONS:

Control Rod Position Control Rod Position26-31 02 10-27 02 26-23 02 30-11 02 22-35 02 30-47 02 22-27 02 46-19 02 50-27 02 02-31 02


Without injecting boron or an evaluation from Reactor Engineering, the reactor __________ remain shutdown under all conditions.

IAW Plant Procedures, ALL automatic scram signals __________ ALLOWED to be BYPASSED.

will; are

will; are NOT

will NOT; are

will NOT; are NOT

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:42 PM 128


RWM Shutdown Confirmation criteria - if all rods are at or inserted past position 02 thenshutdown is confirmed. Allthough 10 rods are at position 02, the reactor will remain shutdownunder all conditions. If one of the rods listed was at a position other than 02, then shutdownconfirmation would not be confirmed. 34AB-C11-005-2 does not allow bypassing the autoscram signals to insert rods. This is only found in 31EO-EOP-103-2.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine if the reactor willremain shutdown (shutdown margin) based on control rod position only and then if all automaticscram signals can be bypassed (operational implications).

The "A" distractor is plausible since the first part is correct. The second part is plausible becauseif control rods are inserted IAW EOP-103, all scram signals are bypassed. Since all control rodsare inserted to at least position 02, EOP-103 is NOT entered.

"B" is correct.

The "C" distractor is plausible because not all control rods are inserted to position 00. Thesecond part is plausible because if control rods are inserted IAW EOP-103, all scram signals arebypassed. Since all control rods are inserted to at least position 02, EOP-103 is NOT entered.

The "D" distractor is plausible because not all control rods are inserted to position 00. Thesecond part is correct.

Tuesday, April 19, 2016 3:14:42 PM 129


K/A:

APE: 295006 SCRAM

AK1. Knowledge of the operational implications of the following concepts as they apply toSCRAM : (CFR: 41.8 to 41.10)

AK1.02 Shutdown margin . . . . . . . . . . . . . . . . . . . 3.4 3.7


C11-RWM-LP-05403, Rod Worth Minimizer Ver. 6.2, EO 001.013.B.02


34GO-OPS-065-0, Control Rod Movement, Ver. 12.6 Bank question from HLT Database Q#295006AK1.02-001

Tuesday, April 19, 2016 3:14:42 PM 130


Unit 2 has experienced an ATWS with the MSIVs closed.

The following conditions exist at the given time:

At 10:05, RPV pressure is 1056 psig At 10:08, RPV pressure is 1075 psig

At 10:11, LLS is controlling RPV pressure with the following:

o 2B LLS valve continuously OPEN o 2G LLS valve cycling on LLS setpoint


IAW 34AB-C71-001-2, Scram Procedure, the EARLIEST listed time that LLS isALLOWED to be manually ARMED is __________ .

At 10:11, Reactor Power is __________ .

10:05; below 5% RTP

10:05; above 5% RTP

10:08; below 5% RTP

10:08above 5% RTP

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:42 PM 131


1056 psig RPV pressure is one psig above the alarm setpoint (1055 psig) for Reactor PressureHigh, 34AR-603-114-2.

1075 psig RPV pressure is one psig above the scram setpoint (1074 psig) for Reactor VesselHigh Pressure Trip, 34AR-603-105-2. The setpoint of 1074 psig RPV Pressure is also used toarm the LLS logic.

To manually arm the LLS logic, perform the following:Confirm RPV pressure is > 1074 psigOpen any SRV by placing its control switch to OPEN

Each SRV is capable of passing approximately 8% of rated steam flow. The number of SRVsopen can give an estimate of power during an ATWS with failure of nuclear instrumentation. With one SRV continuously open, reactor power is approximately 8%.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine reactor powerbased on the SRV response to high RPV pressure.

The "A" distractor is plausible because 1055 psig is the alarm setpoint for Reactor VesselPressure High, 603-114. The second part is plausible because only one SRV cycling on LLS isnormal at decay heat levels with MSIVs closed.

The "B" distractor is plausible because 1055 psig is the alarm setpoint for Reactor VesselPressure High, 603-114. The second part is correct.

The "C" distractor is plausible because the first part is correct. The second part is plausiblebecause only one SRV cycling on LLS is normal at decay heat levels with MSIVs closed.

"D" is correct.

Tuesday, April 19, 2016 3:14:42 PM 132


K/A:

APE: 295007 High Reactor Pressure

AA2. Ability to determine and/or interpret the following as they apply to HIGHREACTOR PRESSURE : (CFR: 41.10 / 43.5 / 45.13)

AA2.02 Reactor power . . . . . . . . . . . . . . . . . . . . . . . . . 4.1* 4.1*


B21-SLLS-LP-01401, Main Steam & LLS, Ver. 10.0, EO 014.003.A.05 / 200.009.A.03


34AB-C71-001-2, Scram Procedure, Ver. 12.634SO-B21-001-2, ADS and LLS Systems, Ver. 13.1534AR-603-105-2, Reactor Vessel High Pressure Trip, Ver. 4.1 34AR-603-114-2, Reactor Pressure High, Ver. 6.1

Tuesday, April 19, 2016 3:14:42 PM 133


Unit 1 is operating at 100% RTP when Drywell pressure starts slowly increasing.

Current Drywell parameters:

o Drywell presssure 0.8 psig o Drywell temperature 151°F

IAW 34SO-T48-002-1, Containment Atmospheric Control and Dilution Systems, Normal Drywell venting is placed in service.


The reason Drywell venting is placed in service is to prevent a __________ .

IAW 34SO-T48-002-1, if Drywell pressure increases to 1.9 psig, the NPO will confirmDrywell venting __________ .

loss of both Drywell Chillers; is still in progress

loss of both Drywell Chillers; has isolated

high Drywell pressure Reactor Scram; is still in progress

high Drywell pressure Reactor Scram; has isolated

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:42 PM 134


The initial action taken to control primary containment pressure employs the same methodstypically used during normal plant operations. Actions include monitoring containment pressureand using containment and drywell pressure control systems (including Standby Gas TreatmentSystem) as required to maintain containment pressure below the high drywell pressure scramsetpoint.

With Drywell pressure 0.8 psig, ARP 603-115 will direct entry into 34AB-T23-002-1, SmallPipe Break Inside Primary Containment to monitor Drywell parameters and 34SO-T48-002-1,Containment Atmospheric Control and Dilution Systems, to vent the Drywell.

34AB-T23-002-1 states "IF a high Drywell pressure (1.85 PSIG) OR low RWL (+3 inches)condition exists, a Reactor Scram will occur."

34AB-C71-001-1, Scram Procedure, Attachment 1 step 2.1 EITHER of the following conditionscause isolation:

Reactor Water Level Low (Level 3) (+3") Drywell Pressure High (1.85 psig)

Step 2.2 Confirm the following valves have CLOSED:1T48-F334A Cad A Drywell Vent Isol Vlv 1T48-F335A Cad A Drywell Vent Isol Vlv 1T48-F334B Cad B Drywell Vent Isol Vlv 1T48-F335B Cad B Drywell Vent Isol Vlv

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant the reason Drywell venting isplaced into service with a high Drywell pressure condition.

The "A" distractor is plausible because on Unit 2 both Drywell Chillers will shutdown if Drywellpressure exceeds the high Drywell pressure setpoint. The second part is plausible because 1.92psig is the TS value for High Drywell Pressure.

The "B" distractor is plausible because on Unit 2 both Drywell Chillers will shutdown if Drywellpressure exceeds the high Drywell pressure setpoint. The second part is correct.

The "C" distractor is plausible since the first part is correct. The second part is plausible because1.92 psig is the TS value for High Drywell Pressure.

"D" is correct.

Tuesday, April 19, 2016 3:14:42 PM 135


K/A:

APE: 295010 High Drywell Pressure

AK3. Knowledge of the reasons for the following responses as they apply to HIGHDRYWELL PRESSURE : (CFR: 41.5 / 45.6)

AK3.01 Drywell venting . . . . . . . . . . . . . . . . . . . . . 3.8 4.0*


EOP-PC-LP-20310, Primary Containment Control (PC), Ver. 4.0, EO 201.076.A.01


34AR-603-115-1, PRIMARY CNMT PRESSURE HIGH, Ver. 7.0 34AB-T23-002-1, Small Pipe Break Inside Primary Containment, Ver. 4.16 34SO-T48-002-1, Containment Atmospheric Control and Dilution Systems, Ver. 23.4

Tuesday, April 19, 2016 3:14:42 PM 136

ILT-10 SRO NRC EXAM43. 295014G2.1.28 001

Unit 2 is operating at 75% RTP when HPCI inadvertently starts and injects at rated flow into theRPV.

The following parameters exist:

o RWL maintained in NORMAL band o RPV pressure 980 psig and stable o Drywell pressure 0.3 psig and stable o HPCI Initiation signal WHITE light will NOT reset

IAW 34SO-E41-001-2, High Pressure Coolant Injection (HPCI) System, a NPO has justdepressed the HPCI Trip Pushbutton.

Based on the above conditions and IAW 34SO-E41-001-2,

The HPCI Trip Pushbutton will be RELEASED __________ .

The purpose for this action is to __________ on a subsequent initiation signal.

WHEN the 2E41-F001, HPCI Turb Steam Supply valve, indicates full CLOSED; prevent HPCI from automatically restarting

WHEN the 2E41-F001, HPCI Turb Steam Supply valve, indicates full CLOSED; allow HPCI to automatically restart

ONLY after HPCI TURBINE BRG OIL PRESS LOW, 601-112, alarm is received; prevent HPCI from automatically restarting

ONLY after HPCI TURBINE BRG OIL PRESS LOW, 601-112, alarm is received; allow HPCI to automatically restart

A.

B.

C.

D.

Description:

34SO-E41-001-2 contains the following:

7.3.1 Control Room Shutdown

7.3.1.1 IF the HPCI System must be shutdown with an initiation signal present,THEN use the Preventing HPCI Injection to the RPV subsection of thisprocedure.

7.3.1.2 Confirm the HPCI system is no longer required to be in operation for reactorvessel level OR pressure control.

Tuesday, April 19, 2016 3:14:43 PM 137

ILT-10 SRO NRC EXAM

7.3.1.3 Confirm OR place in START control switch for 2E41-C002-2, Barom CndsrVacuum Pump.

7.3.1.4 Confirm a HPCI Initiation Signal is NOT present, by depressing the HPCIInitiation Signal Reset pushbutton on panel 2H11-P601 AND confirming thewhite light EXTINGUISHES.

7.3.1.4.1 IF HPCI Initiation signal "White Light" will NOT reset, THEN shutdownHPCI per the 'Preventing HPCI Injection to the RPV', subsection 7.4.4.

7.3.1.5 Using 2E41-R612, HPCI Flow Control, reduce turbine speed to about 2000 rpmAND depress AND hold the Turbine Trip pushbutton.

7.3.1.6 Confirm 2E41-C002-3, Aux Oil Pump, AUTO starts, PRIOR to the turbinespeed decreasing below 1500 rpm.

7.3.1.7 Close 2E41-F001, Turb Steam Supply Valve.7.3.1.8 WHEN 2E41-F001, Turb Steam Supply Valve, is fully CLOSED, release the

Turbine Trip pushbutton.

7.4.4 Preventing HPCI Injection To The RPV

7.4.4.2 IF HPCI is operating, perform the following: 7.4.4.2.1 Depress AND hold the HPCI Turbine Trip pushbutton. 7.4.4.2.2 WHEN the HPCI turbine has stopped, place 2E41-C002-3, HPCI Aux Oil

Pump, in PULL-TO-LOCK. 7.4.4.2.3 WHEN HPCI TURBINE BRG OIL PRESS LOW alarm is received,

release the HPCI Turbine Trip push-button.

With step 7.4.4.2.2 complete, HPCI will not automatically restart on an initiation signal since theAux Oil pump has been placed in the PULL-TO-LOCK (off) position.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to have knowledge of howHPCI will respond once it has been securred from an inadvertent initiation with the Aux Oilpump placed in the off-pull-to-lock position (function of components).

The "A" distractor is plausible because it is the normal method of securing HPCI with theinitiation signal reset. The second part is correct.

The "B" distractor is plausible because it is the normal method of securing HPCI with theinitiation signal reset. The second part is plausible because there are other ECCS that willautomatically restart (if secured) on another initiation signal (CS).

"C" is correct.

Tuesday, April 19, 2016 3:14:43 PM 138

ILT-10 SRO NRC EXAM

The "D" distractor is plausible since the first part is correct. The second part is plausible becausethere are other ECCS that will automatically restart (if secured) on another initiation signal (CS).

References:NONE

K/A:

APE: 295014 Inadvertent Reactivity Addition

G2.1.28 Knowledge of the purpose and function of major system components and controls.(CFR: 41.7) . . . . . . . 4.1 4.1


E41-HPCI-LP-00501, High Pressure Coolant Injection (HPCI), Ver. 6.0, EO 005.004.A.02


34SO-E41-001-2, High Pressure Coolant Injection (HPCI) System, Ver. 30.0 34AB-E10-001-2, Inadvertent Initiation Of ECCS/RCIC, Ver. 0.5

Tuesday, April 19, 2016 3:14:43 PM 139


The Main Control Room has been evacuated.

The Unit 2 reactor was NOT shutdown prior to leaving the Control Room.

o Local actions have been taken to scram the reactor o ALL RSDP transfer switches have been placed in-the EMERGENCY position

Based on the above conditions and IAW 31RS-OPS-001-2, Shutdown From Outside ControlRoom,

From the Remote Shutdown Panel, __________ can be started.

The reactor __________ be confirmed shutdown from the TSC.

ONLY one CRD pump; can NOT

ONLY one CRD pump; can

BOTH CRD pumps; can NOT

BOTH CRD pumps; can

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:43 PM 140


31RS-OPS-001-2 contains the following guidance for starting CRD pumps;

To run the 'B' CRD pump, at panel 2H21-P173 (inside 2C82-P001 at 130RBR15), PLACE2C82-S71, CRD Pump B Transfer Switch, in EMERGENCY. THEN start the pump byplacing the control switch to START.

To run the "A" CRD pump AND/OR maximize system flow, utilize procedure31RS-C11-001-2, which will start the pump at the breaker.

If SPDS is available, (step 4.6) the procedure will confirm the reactor is shutdown by usingSPDS in the TSC.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine reactor power (byconfirming the reactor is shutdown on SPDS) from outside the Main Control Room.

The "A" distractor is plausible since the first part is correct. The second part is plausible if theapplicant remembers that the RSDP has various RPV/Containment indications (pressure, RWL,temperatures, etc) and applies this for confirmation of reactor shutdown.

"B" is correct.

The "C" distractor is plausible if the applicant remembers both CRD pumps can be started fromoutside the Main Control Room and then applies this to starting both pumps. The second part isplausible if the applicant remembers that the RSDP has various RPV/Containment indications(pressure, RWL, temperatures, etc) and applies this for confirmation of reactor shutdown.

The "D" distractor is plausible if the applicant remembers both CRD pumps can be started fromoutside the Main Control Room and then applies this to starting both pumps. The second part isplausible since it is correct.

References:NONE

Tuesday, April 19, 2016 3:14:43 PM 141

ILT-10 SRO NRC EXAMK/A:

APE: 295016 Control Room Abandonment

AA2. Ability to determine and/or interpret the following as they apply to CONTROLROOM ABANDONMENT : (CFR: 41.10 / 43.5 / 45.13)

AA2.01 Reactor power . . . . . . . . . . . . . . . . . . . . 4.1* 4.1*


C82-RSDP-LP-05201, Remote Shutdown Panel (RSDP), Ver. 4.0, EO 001.023.A.01


31RS-OPS-001-2, Shutdown From Outside Control Room, Ver. 5.24 Modified question from HLT Database which was used on 2009 HLT-4 NRC Exam Q#45

Original Question

An evacuation of the Main Control room has occurred.

o The Unit 1 reactor was not scrammed prior to leaving the Control Room.o SPDS is NOT available.

Which ONE of the following choices completes both statements IAW 31RS-OPS-001-1,"Shutdown From Outside Control Room,"

Guidance is given to locally scram the reactor by tripping the Scram Discharge Volume(SDV) (1) .

This procedure provides guidance to confirm Reactor shutdown by visually verifying thateach (2) .

A. (1) Thermal Level Switches (2) SDV Vent and Drain valve is CLOSED

B. (1) Thermal Level Switches(2) HCU Scram Inlet and Outlet valve is OPEN

C. (1) Float Level Switches(2) SDV Vent and Drain valve is CLOSED

Tuesday, April 19, 2016 3:14:43 PM 142

ILT-10 SRO NRC EXAMD. (1) Float Level Switches

(2) HCU Scram Inlet and Outlet valve is OPEN

Tuesday, April 19, 2016 3:14:43 PM 143


A Site Area Emergency has just been declared on Unit 2 due to a leak into secondarycontainment. A Prompt Offsite Dose Assessment is in progress but results are unavailable atthis time.

SPDS "MIDAS" Screen is attached on the next page.

The following Table is from NMP-EP-111-002, Emergency Notification NetworkCommunicator Instructions - Hatch:

Wind Direction From

Rally Point

Site Exit Route

Evacuation Route/State Reception Center

340º - 60º Gate 17 Main Access Road

U.S. Hwy. 1 - North to Toombs Co. High School/ Lyons

61º - 110º PESB

Road behind Low Level Radwaste Building

U.S. Hwy. 1 - South to Appling Co. High School/ Baxley

111º - 225º PESB Main Access Road

U.S. Hwy. 1 - South to Appling Co. High School/ Baxley

226º - 339º PESB Main Access Road

EITHER direction on U.S. Hwy. 1 to Toombs Co. High School/ Lyons

OR Appling Co. High School/ Baxley


Radiation levels in the Main Stack __________ exceeded the HI-HI setpoint.

IAW NMP-EP-111-002, the site evacuation route is __________ .

have;US #1 North

have;US #1 South

have NOT;US #1 North

have NOT;US #1 South

A.

B.

C.

D.

Description:

Tuesday, April 19, 2016 3:14:43 PM 144

ILT-10 SRO NRC EXAMThere are two instruments for the Main Stack, 1D11-K600A & B, for radiation monitoring. Ifeither detector has a Hi-Hi alarm, the Main Stack effluent accident range monitor activates andthe normal Main Stack monitor isolates.

On the SPDS Midas Information page, the Main Stack normal range monitors will display avalue and the Main Stack accident range monitor will be blank, until the HI-HI setpoint isexceeded. Once exceeded, only the Main Stack accident range monitor will display a value.

OFF GAS VENT RADIATION HIGH-HIGH, 601-412Radiation level at the main stack has exceeded the High-High SetpointConfirm the accident range effluent monitor on panel 1D11-P006 has auto started

IAW NMP-EP-104-F07, Offsite Dose Assessment Hatch Prompt Offsite Dose Assessment:

IF a radiological release is in progress, THEN, based on wind direction, determine the rallypoint, site exit route, evacuation route, and State Reception Center with the chart below...

The 15 minute average wind direction information must be read using the meteorologicalinstrumentation that corresponds to the primary release point:The 100M instrument must be used for Main Stack (elevated) release.The 10M instrument must be used for RB Vent (ground) release.

K/A JUSTIFICATION:

This question satisfies the K/A statement by evaluating the applicants' ability to monitorinformation on SPDS and determine the evacuation route with a release in progress.

"A" is correct.

The "B" distractor is plausible because the first part is correct. The applicant might determine aground released is in progress because the Unit 2 Rx Bldg vent stacks are in service and select awind direction of 65° (10M instrument) . Even if a Secondary Isolation had occurred in the Unit2 Rx Bldg, other buildings (Turb. Bldg, Radwaste..) would continue to discharge to the Rx Bldgstack. If a wind direction of 65° were used, the evacuation route would be South on US hwy 1.

The "C" distractor is plausible because the applicant might determine that the reading given onthe Main Stack accident range monitor is below the HI-HI setpoint (The HI-HI setpoint is 1E6cpm on the normal range). The second part is correct.

The "D" distractor is plausible because the applicant might determine that the reading given onthe Main Stack accident range monitor is below the HI-HI setpoint (The HI-HI setpoint is 1E6cpm on the normal range). The applicant might determine a ground released is in progressbecause the Unit 2 Rx Bldg vent stacks are in service and select a wind direction of 65° (10Minstrument) . Even if a Secondary Isolation had occurred in the Unit 2 Rx Bldg, other buildings(Turb. Bldg, Radwaste..) would continue to discharge to the Rx Bldg stack. If a wind directionof 65° were used, the evacuation route would be South on US hwy 1.

Tuesday, April 19, 2016 3:14:43 PM 145


K/A:

APE: 295017 High Off-Site Release Rate

AA1. Ability to operate and/or monitor the following as they apply to HIGH OFF-SITERELEASE RATE : (CFR: 41.7 / 45.6)

AA1.05 SPDS/ERIS/CRIDS/GDS: Plant-Specific . . . . . . . . . . . . 2.7 3.2


LR-LP-20017, Offsite Dose Assessment, Ver. 6.1 / EO 200.059.A.01LT-LP-20017, Meteorological Data Collecting System, Ver. 5.1D11-PRM-LP-10007, Process Radiation Monitors, Ver. 5.1


NMP-EP-111-002, Emergency Notification Network Communicator Instructions - Hatch, Ver. 1.3NMP-EP-104, Dose Assessment, Ver. 8.0NMP-EP-104-F07, Offsite Dose Assessment Hatch Prompt Offsite Dose Assessment, Ver. 1.0

Tuesday, April 19, 2016 3:14:43 PM 146

ILT-10 SRO NRC EXAM295017AA1.05 002

Tuesday, April 19, 2016 3:14:43 PM 147

ILT-10 SRO NRC EXAM

Tuesday, April 19, 2016 3:14:43 PM 148



Subsequently, one (1) RBCCW pump trips.

RBCCW Header pressure stabilizes at 93 psig.

Based on the above conditions and IAW 34AB-P42-001-2, Loss Of Reactor Building ClosedCooling Water, the NPO will __________ the Standby RBCCW pump.

A Standby RBCCW pump automatic start signal __________ when the condition clears.

confirm the automatic start of; will automatically reset

confirm the automatic start of; must be manually reset

manually start; will automatically reset

manually start; must be manually reset

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:43 PM 149


Standby RBCCW Pump auto-starts if system pressure decreases to 90 psig with the pumpcontrol switch in AUTO.

Standby PSW pump will auto start if system header pressure decreases to 95 psig.

Leakage out of the system is monitored by a Timer/Counter system. The counter-timer willcause an alarm if there has been more than one fill cycle in 13 hours, which indicates at least 134gallons leakage (.17 gpm). An annunciator sounds in the Main Control Room indicating excessleakage is occurring in the RBCCW System. The timer/counter must be locally reset (resetpushbutton ) before it will resume automatic operation. If after only one fill cycle, the 13 hourtimer times out, the system automatically resets to 13 hours and a counter of 0002.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to monitor when the RBCCWstandby pump will auto start and then determine how this backup system (pump) will be reset.

The "A" distractor is plausible if the applicant remembers a cooling water system pump willautomatically start at 95 psig (Plant Service Water - PSW) and applies this to the pressure for theRBCCW Standby pump. The second part is plausible since it is correct.

The "B" distractor is plausible if the applicant remembers a cooling water system pump willautomatically start at 95 psig (Plant Service Water - PSW) and applies this to the pressure for theRBCCW Standby pump. The second part is plausible if the applicant remembers there is asignal that must be reset locally for the RBCCW System (RBCCW excessive fill cycles) andapplies this for the condition of low RBCCW Header pressure.

"C" is correct.

The "D" distractor is plausible since the first part is correct. The second part is plausible if theapplicant remembers there is a signal that must be reset locally for the RBCCW System(RBCCW excessive fill cycles) and applies this for the condition of low RBCCW Headerpressure.

Tuesday, April 19, 2016 3:14:43 PM 150


K/A:

APE: 295018 Partial or Complete Loss of Component Cooling Water

AA1. Ability to operate and/or monitor the following as they apply to PARTIAL ORCOMPLETE LOSS OF COMPONENT COOLING WATER : (CFR: 41.7 / 45.6)

AA1.01 Backup systems . . . . . . . . . . . . . . . . . . . . . . 3.3 3.4


P42-RBCCW-LP-00901, Reactor Building Closed Cooling Water, Ver. 3.0, EO 200.014.A.06


34AB-P42-001-2, Loss Of Reactor Building Closed Cooling Water, Ver. 2.6 34SO-P42-001-2, Reactor Building Closed Cooling Water System, Ver. 18.0 34AR-650-248-2, RBCCW SURGE TK LEVEL LOW OR EXCESS LEAKAGE, Ver. 2.1

Tuesday, April 19, 2016 3:14:43 PM 151

ILT-10 SRO NRC EXAM47. 295019G2.1.32 001

On Unit 2, IAW 34AB-P51-001-2, Loss of Instrument And Service Air System Or WaterInstrusion Into the Service Air System,

2P52-F565, Rx Bldg Inst N2 To Non-Int Air El. 185 Isol Vlv, will FIRST open whenNon-Interruptible Essential Air Pressure DECREASES to __________ .

IF 2P52-F565 is continuously cycling, THEN 2P52-F565 is OPENED and the supply breakeris turned OFF to __________ .

90 psig;stabilize nitrogen system pressure

90 psig;prevent failure of the valve motor

80 psig;stabilize nitrogen system pressure

80 psig;prevent failure of the valve motor

A.

B.

C.

D.

Description:

The Essential Instrument Air Header is divided up into the Interruptible and Non-InterruptibleAir Headers. A Nitrogen backup system is provided to the Non-Interruptible Essential Airheader through isolation valve 2P52-F565 (On Unit 1 1P52-F874, F875, F876, F877 & F878).

On Unit 2, 2P52-F565 will automatically open when Non-Interruptible Essential Air headerpressure decreases to 80 psig (On Unit 1, all valves open at 90 psig and decreasing exceptF875, which opens at 80 psig and decreasing).

IAW 34AB-P51-001-2, Loss of Instrument And Service Air System Or Water Instrusion Into theService Air System:

ALLOWING 2P52-F565 TO CONTINUOUSLY CYCLE FULL OPEN TO FULL CLOSEDCOULD CAUSE EXCESSIVE DUTY ON THE VALVE MOTOR AND MAY CAUSETHE EVENTUAL FAILURE OF THE MOTOR.

IF 2P52-F565 is continuously cycling, THEN OPEN 2P52-F565 AND turn OFF the supplybreaker (MCC 2R24-S011, FR 22E, 130RB) to the valve.

IAW 34SO-T48-002-2, Containment Atmospheric Control and Dilution Systems,

Tuesday, April 19, 2016 3:14:43 PM 152

ILT-10 SRO NRC EXAM

Normal Nitrogen Storage Tank pressure is between 130-150 psig

IF NITROGEN INERTING SYSTEM IS SUPPLYING BACKUP NITROGEN TO UNITONE OR UNIT TWO DRYWELL PNEUMATIC SYSTEM OR REACTOR BUILDINGINSTRUMENT AIR SYSTEM, NITROGEN STORAGE TANK PRESSURE MUST BEMAINTAINED GREATER THAN 125 PSIG.

DO NOT RAISE TANK PRESSURE HIGHER THAN 150 PSIG, TO AVOID LIFTING 2T48-F461, TANK SAFETY RELIEF VALVE, (SET AT 155 PSIG),AND/OR 2P70-F102, DRYWELL PNEUMATIC RECEIVER TANK RELIEF VALVE.

IF it is desired to DECREASE Nitrogen Storage Tank pressure, perform the following steps:

Slowly Open 2T48-F456 Nitrogen Storage Tank vent OR 2T48-F448, Full Trycock vent

WHEN Nitrogen Storage Tank pressure is 130-150 psig, on 2T48-R001, Close 2T48-F456 Nitrogen Storage Tank Vent, AND/OR 2T48-F448 Full Trycock vent

IF it is desired to INCREASE Nitrogen Storage Tank pressure, perform the following steps: Open 2T48-F437, 2T48-PCV-F466 Bypass Valve

WHEN Nitrogen Storage Tank pressure is 130-150 psig on 2T48-R001,

Close 2T48-F437, 2T48-PCV-F466 Bypass Valve

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to expalin the reason why anaction is taken, based on a CAUTION in the system abnormal procedure.

The "A" distractor is plausible because on Unit 1, 4 of the 5 nitrogen backup valvesautomatically open when pressure decreases to 90 psig. If the nitrogen system is supplying theInstrument Air system, 34SO-T48-002-2 contains actions to manually adjust/maintain nitrogensystem pressure.

The "B" distractor is plausible because on Unit 1, 4 of the 5 nitrogen backup valvesautomatically open when pressure decreases to 90 psig. The second part is correct.

The "C" distractor is plausible because the first part is correct. If the nitrogen system issupplying the Instrument Air system, 34SO-T48-002-2 contains actions to manuallyadjust/maintain nitrogen system pressure.

"D" is correct.

Tuesday, April 19, 2016 3:14:43 PM 153


K/A:

APE: 295019 Partial or Complete Loss of Instrument Air

G2.1.32 Ability to explain and apply system limits and precautions. (CFR: 41.10 / 43.2 / 45.12) . . . . . . . . . . . . . . 3.8 4.0


LT-SG-50303, Loss Of Instrument And Service Air, Ver. 11.1 Obj #3 200.025.D


34AB-P51-001-2, Loss of Instrument And Service Air System Or Water Instrusion Into theService Air System, Ver. 5.0

Tuesday, April 19, 2016 3:14:43 PM 154



o RPS B DE-ENERGIZES

Subsequently, 2R22-S017, 125/250 VDC Batt Swgr 2B, DE-ENERGIZES.

Based on the above electrical losses,

PCIS Group 2 __________ valves will automatically CLOSE.

Ten (10) seconds later, Main Steam from the RPV to the Main Turbine __________ haveautomatically isolated.

Inboard; will

Inboard; will NOT

Outboard; will

Outboard; will NOT

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:43 PM 155


Valve operation is controlled by a 4-way spool valve which is positioned by two solenoids, oneAC powered and one DC powered. With both solenoids de-energized the spool valve suppliespneumatics to close the MSIV. With either of the solenoids energized the spool is positioned toopen the MSIV. Having both AC and DC Solenoid Valves provides redundancy to prevent anisolation on the loss of one power supply and also provides a means of opening the MSIVs in anemergency, if either AC or DC power is available. The inboard MSIVs solenoids are poweredfrom RPS A and R25-S001 (via 2R22-S016). The outboard MSIV solenoids are powered fromRPS B and R25-S002 (via 2R22-S017). These solenoids can also be verified energized bychecking the RED LEDs are illuminated on the vertical section of H11-P602 for the InboardMSIVs or H11-P601 for the Outboard MSIVs. These LEDs were added to allow the Operator tomonitor the status of each of the MSIVs solenoid valves to reduce the possibility of aninadvertent MSIV closure during testing.

RPS A power supply to Group 2 PCIS inboard valvesRPS B power supply to Group 2 PCIS outboard valves

On loss of an RPS bus half the Group 2 valves go close. The inboard valves will isolate on a lossof RPS A, and the outboard valves isolate on loss of RPS B.

K/A JUSTIFICATION:

This question satisfies the K/A statement by requiring the applicant to know the effect on theMSIVs closing and isolating the Main Steam Lines (inadvertent isolation) due to a loss of power to the pilot valve solenoid with the given plant conditions.

"A' is plausible because it is correct for a loss of RPS A. The second part is correct.

The "B" is plausible because it is correct for a loss of RPS A. The second part is plausiblebecause of ECCS Logic which will cross over to ensure an isolation would occur upon a powerfailure.

The "C" is correct.

The "D" distractor is plausible since the first part is correct. The second part is plausible becauseof ECCS Logic which will cross over to ensure an isolation would occur upon a power failure.

References:NONE

Tuesday, April 19, 2016 3:14:43 PM 156

ILT-10 SRO NRC EXAM

K/A:

APE: 295020 Inadvertent Containment Isolation

AK2. Knowledge of the interrelations between INADVERTENT CONTAINMENTISOLATION and the following: (CFR: 41.7 / 45.8)

AK2.01 Main steam system . . . . . . . . . . . . . . . . . . . 3.6 3.7


C71-RPS-LP-01001, Reactor Protection System, Ver. 8.2 EO 010.002.A.01


34AB-C71-002-2, Loss of RPS, Ver. 5.4 Modified from a bank question from HLT Database Q#LT-010002-01

Original Question


o RPS A DE-ENERGIZES

Two (2) minutes later, a DC Switchgear DE-ENERGIZES as indicated below on Panel, 2H11-P651;

Tuesday, April 19, 2016 3:14:43 PM 157

ILT-10 SRO NRC EXAM

Based on the above electrical losses,

The 125/250 VDC Battery Switchgear that de-energized is __________ .

Ten (10) seconds later, __________ of the Main Steam Isolation Valves (MSIVs) will haveautomatically isolated.

A. 125/250 VDC Batt Swgr 2B, 2R22-S017; four (4)

B. 125/250 VDC Batt Swgr 2B, 2R22-S017; none

C. 125/250 VDC Batt Swgr 2A, 2R22-S016; four (4)

D. 125/250 VDC Batt Swgr 2A, 2R22-S016; none

Tuesday, April 19, 2016 3:14:43 PM 158


Unit 2 is maintaining Mode 4 with RHR Loop A aligned for Shutdown Cooling with thefollowing conditions:

o Reactor Shutdown Two (2) days ago o Reactor Coolant temperature 180°F, steady o Corrected RWL 40 inches, steady o 2B21-F003, Reactor Head Vent Valve OPEN o 2B21-F004, Reactor Head Vent Valve OPEN

Subsequently, a loss of SDC occurs resulting in the following conditions:

o Corrected RWL 40 inches (highest achievable) o Reactor Coolant temperature Increasing o 2B21-F003, Reactor Head Vent Valve Failed OPEN o 2B21-F004, Reactor Head Vent Valve Failed OPEN


IAW 34AB-E11-001-2, Loss of Shutdown Cooling, Reactor bulk coolant temperature isexpected to reach 212°F in approximately __________ .

If Reactor bulk coolant temperature reaches 250°F, Drywell temperature indications will __________ .

REFERENCE PROVIDED

1 hour 5 minutes; be slowly increasing

1 hour 5 minutes; still be approximately the same

3 hours 20 minutes; be slowly increasing

3 hours 20 minutes; still be approximately the same

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:43 PM 159


34AB-E11-001-2 Att. 1 lists the time to reach 212°F is 1 hour and 4 minutes.

The RPV Head Vents tap off the Main Steam Lines prior to the MSIVs. One vent path is fromthe RPV head through normally open manual valve (F005) to MSL "A". This allows acontinuous venting of non-condensable gasses during power operations. The second path isfrom the RPV head to the Drywell Equipment Sump Drain through Air Operated Valves (AOV)F003 and F004. This size pipe is insufficient to prevent the reactor vessel pressurization fromdecay heat 36 hours after shutdown. This vent path is used during startup or when filling theRPV.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to monitor the status ofDrywell temperature (increase/remain the same) with a loss of Shutdown Cooling.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible becausethere is a head vent valve that is routed to MSL 2A which provides a continuous vent path.

The "C" distractor is plausible because the boil-off time for Case 2 Attachment for the Cavityflooded with the gates installed is 3 hr 20 min. The second part is plausible since it is correct.

The "D" distractor is plausible because the boil-off time for Case 2 Attachment for the Cavityflooded with the gates installed is 3 hr 20 min. The second part is plausible because there is ahead vent valve that is routed to MSL 2A which provides a continuous vent path.

Tuesday, April 19, 2016 3:14:43 PM 160

ILT-10 SRO NRC EXAMReferences:34AB-E11-001-2, Attachments 1, 2 & 3 ONLY

K/A:

APE: 295021 Loss of Shutdown Cooling

AA1. Ability to operate and/or monitor the following as they apply to LOSS OFSHUTDOWN COOLING : (CFR: 41.7 / 45.6)

AA1.06 Containment/drywell temperature . . . . . . . . . . . . . 2.8 3.0




34AB-E11-001-2, Loss Of Shutdown Cooling, Ver. 6.15

Tuesday, April 19, 2016 3:14:43 PM 161


Both Units are operating at 100% RTP with the following activities in progress:

o Irradiated fuel movement in progress between the Fuel Pools o Unit 1 Normal Drywell venting in progress o Unit 2 Normal Drywell venting in progress

A malfunction occurs resulting in a dropped irradiated fuel bundle on the Refueling Floor.

The following indications are observed in the Main Control Room:

o ALL Refueling Floor (RF) ARMs 30 mR/hr o 1D11-K611 A-D, RF Vent Exhaust Radiation 20 mR/hr o 2D11-K611 A-D, RF Vent Exhaust Radiation 15 mR/hr o 2D11-K634 A-D, RF Vent Exhaust Radiation 7 mR/hr o 2D11-K635 A-D, RF Vent Exhaust Radiation 7 mR/hr


Normal Drywell venting __________ isolated on BOTH Units.

The resultant release of fission products from the Refuel Floor will be through __________release point.

has; an Elevated

has; a Ground Level

has NOT; an Elevated

has NOT; a Ground Level

A.

B.

C.

D.

Description:

Secondary Containment will isolate on the following signals:

Unit 1 or Unit 2 LOCA Signal (DW pressure > 1.85 psig or Low Rx Water Level of -35") Unit 1 or Unit 2: 18 mrem/hr on 1/2D11-K609 A-D Unit 1 or Unit 2: 18 mrem/hr on 1/2D11-K611 A-D Unit 2: 6.9 mrem/hr on 2D11-K634 A-D,

Tuesday, April 19, 2016 3:14:43 PM 162

ILT-10 SRO NRC EXAMUnit 2: 5.7 mrem/hr on 2D11-K635 A-D

An isolation signal is received on U2 2D11-K635A-D.

When a Unit 1 or Unit 2 Reactor or Refueling zone automatic isolation signal is received, all ofthe Unit 1 and Unit 2 Reactor/Refueling ventilation systems zones are shut down, the ducts areisolated, and both Units Standby Gas Treatment Systems are started and begins exhausting fromthe isolated zones to the main stack. The Unit 1 Secondary Containment isolation signal willisolate both units Secondary Containment but NOT Unit 2 Primary Containment. The Unit 2Secondary Containment isolation signal will isolate both units Secondary Containment but NOTUnit 1 Primary Containment.

Based on the conditions, each units Sec. Containment signal will isolate the respective PrimaryContainment resulting in Normal Drywell venting to isoalte on Both Units.

The Main Stack is an Elevated release point where the Normal release point is through the Rx.Bldg. stack which is a Ground Level release point.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant the implications of a SecondaryContainment Isolation on the ability to vent the associated Primary Containment and theresultant release due to the Secondary Containment isolation (dropped fuel bundle).

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible becausethe normal release point is considered "Ground Level" if flow is being exhausted through the Rx.Bldg. Stack.

The "C" distractor is plausible if the applicant because a Secondary Containment isolation signalwill isolate the associated Units DW vent valves and the U2 K611 A-D are NOT above the tripsetpoint. The second part is correct.

The "D" distractor is plausible if the applicant because a Secondary Containment isolation signalwill isolate the associated Units DW vent valves and the U2 K611 A-D are NOT above the tripsetpoint. The second part is plausible because the normal release point is considered "GroundLevel" if flow is being exhausted through the Rx. Bldg. Stack.

References:NONE

K/A:

APE: 295023 Refueling Accidents Tuesday, April 19, 2016 3:14:44 PM 163

ILT-10 SRO NRC EXAM

AK1. Knowledge of the operational implications of the following concepts as they apply toREFUELING ACCIDENTS : (CFR: 41.8 to 41.10)

AK1.01 Radiation exposure hazards . . . . . . . . . . . . . . . . . . 3.6 4.1


T41-SC HVAC-LP-01303, Secondary Containment HVAC Systems, Ver. 5.1, EO 037.011.A.12


34AB-T22-003-1, Secondary Containment Control, Attachment 7, Ver. 5.14 34AB-T22-003-2, Secondary Containment Control, Attachment 7, Ver. 4.2 Modified from HLT Database Q#295023AK1.01-002

ORIGINAL QUESTION

Both Units are operating at 100% RTP with the following activities in progress:

o Unit 1 Irradiated fuel movement in progress in Fuel Pool o Unit 2 Normal Drywell venting in progress

A malfunction occurs resulting in a dropped irradiated fuel bundle on the Refueling Floor.

The following indications are observed in the Main Control Room:

o ALL Refueling Floor ARMs 50 mr/hr

o 1D11-K611 A-D, Refueling Floor Vent Exhaust Radiation 20 mr/hr

o 2D11-K611 A-D, Refueling Floor Vent Exhaust Radiation 10 mr/hr o 2D11-K634 A-D, Refueling Floor Vent Exhaust Radiation 6 mr/hr o 2D11-K635 A-D, Refueling Floor Vent Exhaust Radiation 5 mr/hr


Unit 2 Normal Drywell venting __________ .

The resultant release of fission products from the Refuel Floor will be through __________release point.

Tuesday, April 19, 2016 3:14:44 PM 164

ILT-10 SRO NRC EXAM

A. has isolated; an Elevated

B. has isolated; a Ground Level

C. is still in progress; an Elevated

D. is still in progress; a Ground Level

Tuesday, April 19, 2016 3:14:44 PM 165

ILT-10 SRO NRC EXAM51. 295024EK2.13 001

Unit 2 is operating at 75% RTP. Reactor power is being reduced due to a leaking SRV.

The following conditions exist:

o Drywell pressure 0.4 psig o Torus pressure 0.2 psig o RHR pump 2A operating o Torus Cooling in sevice o Torus Spray in service

At 13:00, Drywell pressure starts increasing at 0.3 psig/minute.

At 13:03, DRYWELL PRESS HIGH, 602-210, ILLUMINATED.


The EARLIEST listed time that RHR Loop A in Torus Spray will automatically isolate is __________ .

At 13:10, to return Torus Spray to service, the Containment Spray Vlv Control switch __________ REQUIRED to be placed in the MANUAL position.

13:05; is

13:05; is NOT

13:03; is

13:03; is NOT

A.

B.

C.

D.

Description:


F028A/B valves are interlocked closed for the duration of the LOCA (RWL <-101 inches orDrywell pressure >1.85 psig) signal. However, during a LOCA condition, it may be necessary tospray the containment or place Torus cooling in service. Therefore, the valves are interlockedclosed for the duration of the LOCA signal UNLESS:

Tuesday, April 19, 2016 3:14:44 PM 166

ILT-10 SRO NRC EXAM1) With a LOCA signal present the containment spray valve control switch (pistol grip) is

positioned to MANUAL or: 2) With a LOCA signal present and water level <-193", the Containment Spray Valve

Manual Override keylock permissive switch (2/3 core height) is placed to MANUALOVERRIDE and then the containment spray valve control switch (pistol grip) is placedto MANUAL.

NOTE: If in Torus cooling and a LOCA signal occurs, the F028 and F024 will close. Afterplacing the containment spray valve control switch to MANUAL, F028 will reopen if itskeylock switch is still in OPEN, however, F024 must be reopened manually with itscontrol switch.

3) F006A/C (B/D) valves must be closed to open F028A (B) or F016A (B) which preventsdraining the RPV to the Torus.

E11-F027A/B has the same interlocks as F028A/B. RHR Test Line Discharge Isolation,F024A/B have the same interlock as F028A/B.

At 13:03, Drywell pressure will be 1.3 psig resulting in alarm 602-210, being ILLUMINATED(actual setpoint is 1.2 psig).

At 13:05, Drywell pressure will be 1.9 psig resulting in Torus Spray automatically isolating.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine when Torus Spraywill automatically isolate based on Drywell pressure.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible becausethe CNMT SPRAY VLV CNTL 2/3 CORE HT PERMIS switch required to be manipulated tospray the containment when RWL is < -192".

The "C" distractor is plausible because at 13:03 Drywell pressure has increased to 1.2 psig,which is the setpoint for 602-210. The second part is plausible since it is correct.

The "D" distractor is plausible because at 13:03 Drywell pressure has increased to 1.2 psig,which is the setpoint for 602-210. The second part is plausible because the CNMT SPRAYVLV CNTL 2/3 CORE HT PERMIS switch required to be manipulated to spray the containmentwhen RWL is < -192".

Tuesday, April 19, 2016 3:14:44 PM 167


K/A:

EPE: 295024 High Drywell Pressure

EK2. Knowledge of the interrelations between HIGH DRYWELL PRESSURE and thefollowing: (CFR: 41.7 / 45.8)

EK2.13 Suppression pool spray: Plant-Specific . . . . . . . . . . . . . . . 3.8 3.8




34SO-E11-010-2, Residual Heat Removal System, Ver. 42.2

Tuesday, April 19, 2016 3:14:44 PM 168

ILT-10 SRO NRC EXAM52. 295025EA2.01 001

REVISE DESCRIPTION31EO-EOP-107-2, Altenate RPV Pressure Control, is in progress.

o RPV pressure 1060 psig and slowly rising o HPCI system is aligned in Pressure Control Mode o 2E41-R612, HPCI flow controller is in AUTOMATIC with the setpoint at 2500 gpm

Based on the above conditions and IAW 31EO-EOP-107-2,

To stabilize RPV pressure, the operator will __________ .

RAISE the setpoint on 2E41-R612, HPCI flow controller

LOWER the setpoint on 2E41-R612, HPCI flow controller

throttle 2E41-F011, Test to CST VLV, in the CLOSE direction

throttle 2E41-F011, Test to CST VLV, in the OPEN direction

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:44 PM 169


With RPV pressure rising and HPCI aligned for pressure controller, raising the 2E41-R612,HPCI flow controller, setpoint will lower RPV pressure (new flow setpoint will increase thesteam demand to the HPCI turbine to achieve the new flow demand by the controller).

2E41-F011 isolates HPCI/RCIC combined discharge line to the CST. Isolation only on throttlecapability.

2E41-F008, Test to CST VLV (used to isolate HPCI when 2E41-F011 is aligned for RCIC Testto CST) is up stream of 2E41-F011 and has throttle capability unlike 2E41-F011. Throttlingclosed will cause HPCI flow controller to increase steam demand to HPCI turbine to maintainthe desired flow rate.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to interpret RPV pressure andthen take actions per Emergency Operating Procedures.

The "A" is correct.

"B" is plausible because the first part is correct.. The second part is correct when rasing RPVpressure with the HPCI flow controller.

The "C" distractor first part is also correct if the 2E41-F008, Test to CST VLV, is used. Thesecond part is correct when using 2E41-F008 to lower RPV pressure.

The "D" distractor first part is also correct if the 2E41-F008, Test to CST VLV, is used. Thesecond part is correct when using 2E41-F008 to raise RPV pressure.

Tuesday, April 19, 2016 3:14:44 PM 170


K/A:

EPE: 295025 High Reactor Pressure

EA2. Ability to determine and/or interpret the following as they apply to HIGHREACTOR PRESSURE: (CFR: 41.10 / 43.5 / 45.13)

EA2.01 Reactor pressure . . . . . . . . . . . . . . . . . . . . 4.3* 4.3*


B21-SLLS-LP-01401, Main Steam and Low Low Set, Ver. 10.0, EO 014.003.A.06


34SO-B21-001-2, Automatic Depressurization (ADS) and Low-Low Set (LLS) Systems, Ver. 13.15

Tuesday, April 19, 2016 3:14:44 PM 171


Unit 2 is operating at 100% RTP with SRV 2L experiencing elevated leakage.

31EO-EOP-012-2, PC Primary Containment Control, is in progress.

Based on the above conditions and IAW 31EO-EOP-012-2,

The LOWEST listed Torus temperature requiring entry into RC flowchart at point "A" is __________ .

Entering RC flowchart at Point "A" assures that, if possible, the reactor is shutdown before Torus temperature reaches __________ limit.

111°F; the BIIT Curve

111°F; any ECCS NPSH

121°F; the BIIT Curve

121°F; any ECCS NPSH

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:44 PM 172


IAW 34AB-T23-003-2, Torus Temperature Above 95° F, IF Suppression Pool bulk averagetemperature exceeds 110ºF, enter 34AB-C71-001-2, Scram Procedure, AND SCRAM the reactorAND verify temperature < 120ºF once per 30 minutes.

Entering either RC or RCA flowchart at point "A" assures that, if possible, the reactor isscrammed and shutdown by control rod insertion before the requirement for boron injection(BIIT) is reached.

The NPSH (Net Positive Suction Head) Limit is defined to be the highest Torus temperaturewhich provides adequate net positive suction head for HPCI/RCIC or an ECCS pump takingsuction on the Pool. This temperature is a function of individual pump flow and SuppressionChamber overpressure (airspace pressure plus hydrostatic head over the HPCI/RCIC or ECCSsuction), and the limit is utilized to preclude pump damage due to cavitation.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant why the reactor is shutdownbased on high Torus water temperature.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible becauseall NPSH graphs have Torus temperature.

The "C" distractor is plausible because 120°F is the temperature at which entry into34GO-OPS-013-2, Normal Plant Shutdown, is required in order to reduce RPV pressure to < 200PSIG within 12 hours. The second part is plausible since it is correct.

The "D" distractor is plausible because 120°F is the temperature at which entry into34GO-OPS-013-2, Normal Plant Shutdown, is required in order to reduce RPV pressure to < 200PSIG within 12 hours. The second part is plausible because all NPSH graphs have Torustemperature.

References:NONE

K/A:

EPE: 295026 Suppression Pool High Water Temperature

EK3. Knowledge of the reasons for the following responses as they apply toSUPPRESSION POOL HIGH WATER TEMPERATURE: (CFR: 41.5 / 45.6)

Tuesday, April 19, 2016 3:14:44 PM 173

ILT-10 SRO NRC EXAM

EK3.05 Reactor SCRAM . . . . . . . . . . . . . . . . . . . 3.9 4.1


EOP-PC-LP-20310, Primary Containment Control , Ver. 4.0, EO 201.074.A.08


31EO-EOP-012-2, PC Primary Containment Control, Ver. 7.0 Modified question from HLT Database Q#295026EK3.05-003

Original Question

Unit 2 is operating at 60% when SRV "A" inadvertently opens causing Torus temperature to risequickly.

Which one of the below choices correctly completes the following statements?

Based on Unit 2 Technical Specifications, the LOWEST torus temperature requiringIMMEDIATE placement of the mode switch in shutdown once exceeded is ________.

Based on 31EO-EOP-012-2 (Primary Containment flow chart), the unit is required to bescrammed before torus temperature reaches the limit in the ___________.

A. 110°F; BIIT curve (graph 5)

B. 110°F; DWSIL curve (graph 8)

C. 120°F; BIIT curve (graph 5)

D. 120°F; DWSIL curve (graph 8)

Tuesday, April 19, 2016 3:14:44 PM 174

ILT-10 SRO NRC EXAM54. 295028G2.4.31 001

Unit 2 has scrammed from 100% RTP which results in the following conditions;

o RWL indication is -10 inches and steady o DRYWELL TEMP HIGH, 657-043, ILLUMINATED o DRYWELL TEMP HIGH, 654-066, ILLUMINATEDo Bulk Average Drywell temperature is 152°F

Based on the above conditions and IAW the above ARPs,

34AB-B21-002-2, RPV Water Level Corrections, __________ REQUIRED to be entered.

31EO-EOP-012-2, PC Primary Containment Control, __________ REQUIRED to be entered.

is; is NOT

is; is

is NOT; is NOT

is NOT; is

A.

B.

C.

D.

Description:

654-066 step 5.5 states "IF drywell average air temperature exceeds 150°F, perform water levelcorrections, if necessary, due to high drywell temperature per 34AB-B21-002-2, RPV WaterLevel Corrections."

657-043 step 5.4 states "IF drywell average air temperature exceeds 150°F, enter31EO-EOP-012-2, Primary Containment Control."

IAW 34AB-B21-002, step 1.0 states "RPV water level indication corrections are requiredWHEN:

1.1.1 Fuel Zone Uncompensated Instrumentation is required for level monitoring.

OR

Tuesday, April 19, 2016 3:14:44 PM 175

ILT-10 SRO NRC EXAMfor level monitoring AND Bulk Average Drywell temperature is greater than 150°F.

OR

1.1.3 Floodup Range Instrumentation is required for level monitoring AND the followingconditions exist:

1.1.3.1 Reactor pressure is < 100 PSIGAND/OR

1.1.3.2 ANY Drywell group 1 RTD above 212°F"

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know the annunciatorresponse procedures for high Drywell temperature in determining entry into other procedures.

The "A" distractor is plausible since the first part is correct. The second part is plausible becauseDrywell temperature of 212°F is a condition in 34AB-B21-002-2 to correct RWL on the FloodupRange instruments.

"B" is correct.

The "C" distractor is plausible because RWL corrections on the Fuel Zone Range instruments isNOT required since the Fuel Zone Range instruments are upscale (>-17 inches). The second partis plausible because Drywell temperature of 212°F is a condition in 34AB-B21-002-2 to correctRWL on the Floodup Range instruments.

The "D" distractor is plausible because RWL corrections on the Fuel Zone Range instruments isNOT required since the Fuel Zone Range instruments are upscale (>-17 inches). The second partis plausible since it is correct.

Tuesday, April 19, 2016 3:14:44 PM 176


K/A:

EPE: 295028 High Drywell Temperature

G2.4.31 Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3) . . . . . . . . . . . . . 4.2 4.1




34AR-657-043-2, Drywell Temp High, Ver. 5.2 34AR-654-066-2, Drywell Temp High, Ver. 5.1 34AB-B21-002-2, RPV Water Level Corrections, Ver. 6.16 31EO-EOP-012-2, PC Primary Containment Control, Ver. 7.0

Tuesday, April 19, 2016 3:14:44 PM 177


Unit 1 has experienced a leak resulting in Torus water level decreasing.

Based on the above conditions and IAW 31EO-EOP-012-1, Primary Containment Control,

If the leak is isolated, to restore Torus water level, the __________ System suction pipingfrom the CST can be used.

If the leak can NOT be isolated, the reason for performing an Emergency Depressurizationon low Torus water level is to prevent __________ .

Core Spray; direct pressurization of the Torus air space

Core Spray; excessive loading of the SRV tail pipes and supports

HPCI; direct pressurization of the Torus air space

HPCI; excessive loading of the SRV tail pipes and supports

A.

B.

C.

D.

Description:

The RPV is not permitted to remain at pressure (ED) if suppression of steam discharged fromthe RPV into the drywell cannot be assured. When the downcomer vent openings are notadequately submerged, any steam discharged from the RPV into the drywell may not condense inthe torus before torus pressure reaches unacceptable levels. Emergency RPV depressurizationwill be required at or before the point at which this low water level condition occurs (98" Unit2) (102" Unit 1). The determination that torus level cannot be maintained above 98 inches shouldbe made BEFORE reaching the actual limit based on level trend or other plant conditions.

Maximum Pressure Suppression Primary Containment Water Level (MPSPCWL) needs to beaddressed, if in the unsafe region of this SRVTPLL due to exceeding this limit (193"). The RPVis not permitted to remain at pressure if operation of any SRV may cause the SRV dischargeline(s) or associated components to fail. The consequences of operating SRVs when the statusof torus water level and RPV pressure exceeds the SRV Tail Pipe Level Limit may include directpressurization of the containment if a break in any SRV discharge line occurs. This could resultin primary containment pressurization and possible containment failure. To preclude thissituation from occurring, the RPV is depressurized before conditions significantly worsen.

Tuesday, April 19, 2016 3:14:44 PM 178

ILT-10 SRO NRC EXAM

IAW 31EO-EOP-012-1, PC, contains the following step to maintain Torus water level:

Both of the above procedures require the use of the Core Spray suction piping to restore Toruslevel.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant the reason for an EmergencyDepressurization with low Torus water level.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible becauseexcessive loading of the SRV tailpipes and supports is the reason for an EmergencyDepressurization (exceeding SRVTPLL/MPSPCWL) for high torus water level.

The "C" distractor is plausible if the applicant since the HPCI System can be aligned to the Torusor the CST and the applicant realizing Torus level will increase if the Torus and CST valves aremanipulated open at the same time. The second part is correct.

The "D" distractor is plausible if the applicant since the HPCI System can be aligned to theTorus or the CST and the applicant realizing Torus level will increase if the Torus and CSTvalves are manipulated open at the same time. The second part is plausible because excessiveloading of the SRV tailpipes and supports is the reason for an Emergency Depressurization(exceeding SRVTPLL/MPSPCWL) for high torus water level.

Tuesday, April 19, 2016 3:14:44 PM 179


K/A:

EPE: 295030 Low Suppression Pool Water Level

EK3. Knowledge of the reasons for the following responses as they apply to LOWSUPPRESSION POOL WATER LEVEL: (CFR: 41.5 / 45.6)

EK3.01 Emergency depressurization . . . . . . . . . . . . . . . . . . . . . 3.8 4.1




31EO-EOP-012-1, PC Primary Containment Control, Ver. 7.0 34SO-E21-001-1, Core Spray System, Ver. 24.2 34GO-OPS-087-1, Torus Fill And Drain, Ver. 8.4

Tuesday, April 19, 2016 3:14:44 PM 180



An event occurs causing RWL to decrease and results in the following 1C32-R606A-C GEMAC, indications:

Time RWL

10:00 2 inches (LOWEST REACHED) 10:01 9 inches (slowly increasing) 10:02 18 inches (slowly increasing)


The NPO __________ REQUIRED to perform RWL control actions per Placard RC-2.

At 10:01, alarm, REACTOR VESSEL WATER LEVEL HIGH/LOW, 603-141, will be __________ .

is; ILLUMINATED

is; EXTINGUISHED

is NOT; ILLUMINATED

is NOT; EXTINGUISHED

A.

B.

C.

D.

Description:

Reactor Vessel Low Water Level, Plant setpoint: 3 inches above Instrument Zero. RWL ismonitored to ensure that the core remains covered with water, thus ensuring adequate corecooling. The Reactor scram protects the fuel by reducing the fission heat generation within thecore. The Reactor low water level scram setpoint is chosen far enough below normal operationallevels to avoid spurious scrams, but high enough above the top of active fuel to assure thatenough water is available to keep the core adequately cooled following the most severe accidentinvolving a level transient. Four independent level transmitters, two on each side of the ReactorBuilding, provide levels signals to ATTS. Each side of the Reactor Building has transmitter thatfeeds RPS "A" and a transmitter that feeds RPS "B".

Tuesday, April 19, 2016 3:14:44 PM 181

ILT-10 SRO NRC EXAM0 inches is Instrument Zero and is also the point where the Steam Dryer skirt is uncovered.

Narrow Range RWL instruments, 1C32-R606A-C GEMAC Range 0-60 inches

Reactor Vessel Water Level High/Low, 603-141, Low +32 inches, High +42 inches

Alarms 603-117, Reactor Auto Scram System A Trip & 603-118, Reactor Auto Scram System BTrip, must be manually reset once the RPS scram signal is in a reset condition. 603-108 willautomatically reset once the condition has cleared.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know when the ReactorProtection System will generate an automatic scram signal based on low RWL.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible if theapplicant remembers that alarm 603-117/118 must be manually reset with the Scram ResetSwitch and applies this for alarm 603-108.

The "C" is plausible because zero (0) inches is the bottom of the Narrow Range RWLinstruments 1C32-R606A-C GEMAC, the preferred instrumentation during normal operation. Additionally the Reactor Vessel Instrument zero is the bottom lip of the steam dryer skirt. SinceRWL only reached two (2) inches a reactor scram did not occur, therefore, not requiring RC-2 tobe performed. The second part is plausible since it is correct.

The "D" is plausible because zero (0) inches is the bottom of the Narrow Range RWLinstruments 1C32-R606A-C GEMAC, the preferred instrumentation during normal operation. Additionally the Reactor Vessel Instrument zero is the bottom lip of the steam dryer skirt. SinceRWL only reached two (2) inches a reactor scram did not occur, therefore, not requiring RC-2 tobe performed. The second part is plausible if the applicant remembers that alarm 603-117/118must be manually reset with the Scram Reset Switch and applies this for alarm 603-108.

Tuesday, April 19, 2016 3:14:44 PM 182


K/A:

EPE: 295031 Reactor Low Water Level

EK2. Knowledge of the interrelations between REACTOR LOW WATER LEVEL and thefollowing: (CFR: 41.7 / 45.8)

EK2.11 Reactor protection system . . . . . . . . . . . . . . . . . . 4.4* 4.4*


C71-RPS-LP-01001, Reactor Protection System, Ver. 8.2, EO 300.008.A.02 B11-RPV-LP-04402, Reactor Vessel Internals, Ver. 3.0, LT-04402.001 & LT-04402.005B11-RXINS -LP-04404, Reactor Vessel Instrumentation, Ver. 7.2, EO 200.002.A.07


34AB-C71-001-1, Scram Procedure, Ver. 12.8 34AR-603-108-1, Reactor Vessel Low Level Trip, Ver. 4.1 34AR-603-117-1, Reactor Auto Scram System A Trip, Ver. 6.0 34AR-603-118-1, Reactor Auto Scram System B Trip, Ver. 6.0 Unit 1 TS 3.3.1.1 RPS Instrumentation, Table 3.3.1.1-1 (page 2 of 3), Item 4-Reactor VesselWater Level - Low, Level 3, Amendment 232

Tuesday, April 19, 2016 3:14:44 PM 183


Unit 2 is operating at 100% RTP with the following HPCI Pump Room Cooler alignment:

o HPCI Pump Room Cooler "A" (2T41-B005A) is operating with its control switch in the "RUN" position

o HPCI Pump Room Cooler "B" (2T41-B005B) is NOT running with its control switch in the "AUTO" position

Subsequently, a steam leak develops in the HPCI Room causing HPCI Area temperatures toincrease.

At 01:00, the following exits:

o HPCI Pump Room temperature is 105°F o HPCI is in a Standby lineup

At 01:10, the following exits:

o HPCI Pump Room temperature is 170°F o HPCI is in a Standby lineup


At 01:00, HPCI Pump Room Cooler B (2T41-B005B), __________ be running.

At 01:10, IAW 31EO-EOP-014-2, SC EOP flowchart, the HPCI System __________REQUIRED to be manually isolated.

will; is NOT

will; is

will NOT; is NOT

will NOT; is

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:44 PM 184

ILT-10 SRO NRC EXAM

Description:

In AUTO the southeast diagonal Core Spray and RHR coolers, B002A/B, will bothautomatically start on any one of the following:

a. Core spray pump C001B (Unit 1 C001A) running b. RHR pump C002B (Unit 1 - C002A) running c. RHR pump C002D (Unit 1 - C002C) running d. High SE diagonal temperature (A: 135°F, B: 140°F)

In AUTO the HPCI room coolers B005A/B will automatically start on any one of the following:

a. HPCI initiation (F001 open) b. High HPCI room temperature

(A: 95°F, B: 100°F)

IAW 34SO-E41-001-2, P/L;

Step 5.2.3 The HPCI System will automatically isolate upon receipt of any of the followingsignals:

5.2.3.1 HPCI Steam Line High Differential Pressure (Steam line break): InstrumentSetpoint, 202" H2O increasing OR -100" H2O decreasing. (TechnicalSpecification Limit <303%).

5.2.3.2 HPCI Steam Supply Low Pressure: Instrument Setpoint 134 psig (2E41-N058A,2E41-N058B, 2E41-N058C, & 2E41-N058D). (Technical Specification Limit >100 psig).

5.2.3.3 Turbine Exhaust Diaphragm High Pressure: Instrument Setpoint 10 psig. (Technical Specification Limit < 20 psig).

5.2.3.4 HPCI Emergency Area Cooler Temperature High of 165°F (TechnicalSpecification Limit < 169°F).

5.2.3.5 HPCI Pipe Penetration Room Temperature High of 165°F (Technical SpecificationLimit < 169°F).

5.2.3.6 Suppression Pool Area Ambient Temperature - High of 165°F for more than 14minutes. (Technical Specification Limit < 169°F).

5.2.3.7 Suppression Pool Area Differential Temp. - High 36°F for more than 14 minutes. (Technical Specification Limit < 42°F).

Tuesday, April 19, 2016 3:14:44 PM 185

ILT-10 SRO NRC EXAM5.2.3.8 Manual - the push-button on panel 2H11-P601 affects Logic B only AND

CLOSES 2E41-F003 AND 2E41-F041. Manual isolation will only functionWHEN HPCI Initiation Signal is present, depressing the pushbutton without theinitiation signal (white light) illuminated will not affect system operation.

5.2.3.9 Drywell Pressure - High 1.85 psig with low steam line pressure 134 psig(2E41-N058A, 2E41-N058B, 2E41-N058C & 2E41-N058D) isolates only:2E41-F1112E41-F104

Main Steam Tunnel Temp. High (190 degrees F)

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to interpret SecondaryContainment area temperature (HPCI Room) and determine if the Standby HPCI Cooler has autostarted and if the HPCI System should be manually isoalted.

The "A" distractor is plausible since the first part is correct. The second part is plausible becauseit is correct when HPCI is required for damage control and required to be operated by proceduresdeveloped from the EOPs.

"B" is correct.

The "C" distractor is plausible because 135°F is a correct temperature that will auto start a cooler(RHR/CS SE diagonal). The second part is plausible because it is correct when HPCI is requiredfor damage control and required to be operated by procedures developed from the EOPs.

The "D" distractor is because 135°F is a correct temperature that will auto start a cooler(RHR/CS SE diagonal). The second part is plausible since it is correct.

Tuesday, April 19, 2016 3:14:44 PM 186


K/A:

EPE: 295032 High Secondary Containment Area Temperature

EA2. Ability to determine and/or interpret the following as they apply to HIGHSECONDARY CONTAINMENT AREA TEMPERATURE : (CFR: 41.10 / 43.5 / 45.13)

EA2.01 Area temperature . . . . . . . . . . . . . . . . . . . . . . 3.8* 3.8


T41-SCHVAC-LP-01303, Secondary Containment HVAC Systems, Ver. 5.1, EO 037.003.A.06E41-HPCI-LP-00501, High Pressure Coolant Injection (HPCI), Ver. 6.0, EO 005.012.a.04


34SO-E41-001-2, High Pressure Coolant Injection (HPCI) System, Ver. 30.0

Tuesday, April 19, 2016 3:14:44 PM 187


Unit 2 was operating at 100% RTP when an ATWS results in the following:

o Reactor power is 18% RTP o SBLC System fails to initiate from the Control Room o An Operator attempts to locally initiate SBLC, but SBLC fails to initiate

Based on the above conditions and IAW 31EO-EOP-109-2, Alternate Boron Injection,

To inject Boron into the RPV using the CRD System , the operator will DIRECTLY connect hoses to __________ .

When boron is being injected, 2C11-F015A & F015B, CRD Pump Minimum Flows, are REQUIRED to be in the __________ position.

2C41-F034, SBLC Combined Drain Valve; located on the 185 ft. elevation, Rx. Bldg; closed

2C41-F034, SBLC Combined Drain Valve; located on the 185 ft. elevation, Rx. Bldg; open

2C41-F015, SBLC Suction Line Drain Valve, located on the 203 ft. elevation, Rx. Bldg; closed

2C41-F015, SBLC Suction Line Drain Valve, located on the 203 ft. elevation, Rx. Bldg; open

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:44 PM 188


RCIC - When this line up is complete the following flow path will exist:

o SBLC storage tank into the enclosure area, through 2C41-F034 to the RCIC suctionpiping test connections, through the RCIC pump discharge line to the RPV.

HPCI - When this lineup is complete the following flow path will exist.

o SBLC storage tank into the enclosure area, through 2C41-F034 to the HPCI suctionpiping test connections, through the HPCI pump discharge line to the RPV.

CRD - When this line up is complete the following flow path will exist.

o SBLC storage tank into the enclosure area, through 2C41-F034 to the CRD pump suctionfilter boron injection adapter flange and valve assembly, through the CRD Pump TestBypass Valve 2C11-F018 to the RPV.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know how SBLC can beinjected into the RPV using an alternate method of boron injection.

"A" is correct.

The "B" distractor is plausible because the first part is correct. The second part is correct duringnormal system operation when the valve is locked opened.

The "C" distractor is plausible because this is the correct valve to drain the SBLC storage tank tothe SBLC enclose. The second part is plausible since it is correct.

The "D" distractor is plausible because this is the correct valve to drain the SBLC storage tank tothe SBLC enclose. The second part is correct during normal system operation when the valve islocked opened.

Tuesday, April 19, 2016 3:14:44 PM 189


K/A:

EPE: 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale orUnknown

EA1. Ability to operate and/or monitor the following as they apply to SCRAMCONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE ORUNKNOWN : (CFR: 41.7 / 45.6)

EA1.10 Alternate boron injection methods: Plant-Specific . . . . 3.7 3.9


EOP-109-LP-20320, EOP: 109 Alternate Boron Injection, Ver. 2.0, EO 039.020.A.03


31EO-EOP-109-2, Alternate Boron Injection, Ver. 5.4

Tuesday, April 19, 2016 3:14:44 PM 190


Unit 2 was operating at 100% RTP when an radiological event occurred that resulted in SBGTautomatically starting.

o Reactor Bldg. to Outside Air dP is -0.6 inches water column (WC)

An emergency is declared based on a radiological release at the following times:

At 09:00, Alert (RA1) At 10:00, Site Area (RS1)


The EARLIEST listed time that an entry condition into the RR portion of 31EO-EOP-014-2,SC-Rad Release Control, EOP Flowchart, will exist is __________ .

IAW 34SO-T46-001-2, SBGT System Operation, to MINIMIZE Offsite release rates, operate _________ on Unit 2.

09:01; ONLY one (1) SBGT Train

09:01; BOTH SBGT Trains

10:01; ONLY one (1) SBGT Train

10:01; BOTH SBGT Trains

A.

B.

C.

D.

Description:

34SO-T46-001-2, Stand-by Gas Treatment System, Section 4.2.1 AUTOMATIC STARTUP,step 4.0 contains a NOTE which states;

o Normally only one train of SBGT is required for design operation. o Operation with both trains of SBGT will increase offsite release rates AND deplete both

charcoal filter trains. o Operate both trains of SBGT only WHEN required per another procedure OR IF required

to maintain adequate negative pressure.

Tuesday, April 19, 2016 3:14:44 PM 191

ILT-10 SRO NRC EXAM

Step 4.2.1.4 states "IF both trains of SBGT have started AND are operating properly ANDoperation of both trains is NOT required for maintaining greater than or equal to 0.20 inch watervacuum in Unit 1 AND/OR Unit 2 Secondary Containment, THEN place one train of SBGT inSTANDBY as follows: Place either SBGT A or SBGT B Filter 2T46-D001A or 2T46-D001B,control switch, in the OFF position. "

With Reactor Bldg. to Outside Air dP is -0.6 inches water column (WC), one train SBGT fan isallowed to be secured.

RR - RADIOACTIVITY RELEASE CONTROL portion of 31EO-EOP-014-2Entry condition: Alert Theshold value RA1

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know how the release rateswill be affected if both SBGT fans remain in service (protection of general public) and whetherentry condition in RR portion of 31EO-EOP-014-2, SC-Rad Release Control, EOP Flowchart, exist.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible becausewith both SBGT fans in service that almost twice the RB/RF atmosphere is being treated beforebeing released.

The "C" distractor is plausible because it is mitigating strategy prior to exceeding PAGs (1000mR TEDE or 5000 mR thyroid CDE, ie also RG1) to SCRAM & Emergency Depress. Thepreceeding EAL action level prior to RG1 is RS1. The second part is plausible since it is correct.

The "D" distractor is plausible because it is mitigating strategy prior to exceeding PAGs (1000mR TEDE or 5000 mR thyroid CDE, ie also RG1) to SCRAM & Emergency Depress. Thepreceeding EAL action level prior to RG1 is RS1. The second part is plausible because withboth SBGT fans in service that almost twice the RB/RF atmosphere is being treated before beingreleased.

Tuesday, April 19, 2016 3:14:44 PM 192

ILT-10 SRO NRC EXAM References:NONE

K/A:

EPE: 295038 High Off-Site Release Rate

EK1. Knowledge of the operational implications of the following concepts as they apply toHIGH OFF-SITE RELEASE RATE : (CFR: 41.8 to 41.10)

EK1.02 †Protection of the general public . . . . . . . . . . . . . . . . . . . 4.2* 4.4*


T46-SBGT-LP-03001, Standby Gas Treatment System, Ver. 6.0, EO 030.002.A.07


34SO-T46-001-2, SBGT System Operation, Ver. 15.0 Bank question from HLT Database Q#295038EK3.02-001

Tuesday, April 19, 2016 3:14:44 PM 193

ILT-10 SRO NRC EXAM60. 300000K2.01 001

The power supply for Unit 2 Station Service Air Compressor (SSAC) 2A is __________ .

The power supply for Unit 2 SSAC 2C is __________ .

600VAC Bus 2C; 600VAC Bus 2A

600VAC Bus 2C; 600VAC Bus 2BB

600VAC Bus 2D; 600VAC Bus 2A

600VAC Bus 2D; 600VAC Bus 2BB

A.

B.

C.

D.

Description:

Electrical Power Supplies;

600VAC Bus 2C (1C) 2R23-S003 (1R23-S003) supplies Service Air Compressor “A” 600VAC Bus 2D (1D) 2R23-S004 (1R23-S004) supplies Service Air Compressor “B” 600VAC Bus 2A (1BB) 2R23-S001 (1R23-S012) supplies Service Air Compressor “C”

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know the power supplies tothe air compressors on Unit 2.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible andwould be correct if asking the power supply for Unit 1 SSAC "1C".

The "C" distractor is plausible since 600VAC 2D does power one of the SSACs (SSAC "2B")and would be correct if asking for SSAC 2B. The second part is plausible since it is correct.

The "D" distractor is plausible since 600VAC 2D does power one of the SSACs (SSAC "2B")and would be correct if asking for SSAC 2B. The second part is plausible and would be correctif asking the power supply for Unit 1 SSAC "1C".

Tuesday, April 19, 2016 3:14:44 PM 194


K/A:

300000 Instrument Air System (IAS)

K2 Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.01 Instrument air compressor . . . . . . . . . . . . 2.8 2.8


P51-P52-P70-PLANT AIR-LP-03501, Plant Air System, Ver. 5.0, EO 035.001.A.03


OPS-1935, Electrical Lineup, Ver. 1.0 Bank question from HLT Database Q#LT-035001-005

Tuesday, April 19, 2016 3:14:44 PM 195

ILT-10 SRO NRC EXAM61. 400000A2.04 001

The following annunciator on Unit 1 is in the alarm condition:

o SERVICE WATER EFFLUENT RADIATION HIGH, 601-407

Based on the above condition,

The flowpath containing this alarm __________ AUTOMATICALLY isolate.

IAW 601-407, the NPO will notify __________ .

will; Chemistry to sample Circulating Water to determine if effluent limits have been exceeded

will; the Radwaste Operator to confirm closed / close this discharge flowpath

will NOT; Chemistry to sample Circulating Water to determine if effluent limits have been exceeded

will NOT; the Radwaste Operator to confirm closed / close this discharge flowpath

A.

B.

C.

D.

Description:


The discharge from the Reactor Building and Radwaste Building tie into the discharge from theTurbine Building before exiting the Turbine Building. At this point, the PSW is monitored forradioactivity release by a scintillation detector which is monitored in the Control Room. ThePlant Service Water detector is located in the overhead of the 112' elevation in the PSWcombined discharge header. The detector is located prior to the discharge line splitting into twolines with one going to the river and the other discharging to the Circulating Water flume.

Some Process Radiation Monitors cause systems to isolate such as Main Steam Line (MSL)Radiation Monitors. A MSL High-High radiation signal results in the following valves to close:

o 2B31-F019 AND 2B31-F020, Reactor Water Sample Valves o 2T48-F209, 2T48-F210, 2T48-F211, AND 2T48-F212, Drywell To Torus Differential

Pressure System Isolation Valves

Tuesday, April 19, 2016 3:14:44 PM 196

ILT-10 SRO NRC EXAMsystems include Off Gas Pre-Treatment, PSW and RBCCW monitors, Carbon Vault, and thevarious other monitors in the plant.

IAW 601-407 Step 5.4 requires Chemistry Lab to sample Circulating Water for high radiationand determine if effluent limits are exceeded.

IAW 601-401 Step 5.2 reguires the Radwaste Discharge valves to be confirmed isolated.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to predict if the system beingmonitored will automatically isolate based on the radiation monitor and then use the ARP(procedure) to mitigate the alarming radiation monitor.

The "A" distractor is plausible because there are radiation monitors which automatically isolatesystems (Main Steam Line Rad Monitors). The second part is plausible since it is correct.

The "B" distractor is plausible because there are radiation monitors which automatically isolatesystems (Main Steam Line Rad Monitors). The second part is plausible because there is a timewhen confirming the discharge path is isolated (Radwaste Monitor Radiation High).

"C" is correct.

The "D" distractor is plausible since the first part is correct. The second part is plausible becausethere is a time when confirming the discharge path is isolated (Radwaste Monitor RadiationHigh).

References:NONE

K/A:

400000 Component Cooling Water System (CCWS)

A2. Ability to (a) predict the impacts of the following on the CCWS and (b) based on thosepredictions, use procedures to correct, control, or mitigate the consequences of thoseabnormal operation: (CFR: 41.5 / 45.6)

A2.04 Radiation monitoring system alarm . . . . . . . . 2.9 3.0

Tuesday, April 19, 2016 3:14:44 PM 197

ILT-10 SRO NRC EXAM

D11-PRM-LP-10007, Process Radiation Monitors, Ver. 5.1, EO 200.030.A.13


34AR-601-401-2, RADWASTE EFFL RADIATION HIGH, Ver. 0.3 34AR-601-407-2, SERVICE WATER EFFLUENT RADIATION HIGH, Ver. 3.6 34AB-B21-001-2, MAIN STEAM LINE HIGH RADIATION OR SUSPECTED FUELELEMENT FAILURE, Ver. 4.13 Modified from HLT Database Q# 295038EA1.03-001 which was used on 2011 HLT-6 NRCExam Q#61

Original Question

The following annunciator is in the alarm condition for 2D11-K605, Service Water LiquidRadiation Monitor:

o SERVICE WATER EFFLUENT RADIATION HIGH, (601-407)

Based on the above plant conditions.

The system being monitored by this detector is ________ .

This flowpath ________ AUTOMATICALLY isolate due to this High radiation signal.

A. RHR Service Water (RHRSW); will NOT

B. Plant Service Water (PSW); will NOT

C. RHR Service Water (RHRSW); will

D. Plant Service Water (PSW); will

Tuesday, April 19, 2016 3:14:44 PM 198

ILT-10 SRO NRC EXAM62. 400000K3.01 001

Unit 1 was operating at 100% RTP when a total loss of Reactor Building Closed Cooling Water(RBCCW) occurred.


The LOWEST RWCU Non-Regenerative Heat Exchanger outlet temperature which willresult in a RWCU system isolation is __________ .

This isolation signal will automatically close __________ .

130°F; ONLY 1G31-F001, Rx Water Cleanup Valve

130°F;ONLY 1G31-F004, Rx Water Cleanup Valve



A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:44 PM 199


RBCCW cools the RWCU Non-Regen Heat Exchanger. When temperature reaches 140 deg F,then ONLY the F004 valve closes (normally on a Group 5 isolation, both F001 and F004 close).

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know how the loss ofRBCCW (CCWS) will effect RWCU (loads).

The "A" distractor is plausible because the 130°F is the setpoint for alarm 602-427. The secondpart is plausible because 1G31-F001 is a RWCU isolation valve.

The "B" distractor is plausible because the 130°F is the setpoint for alarm 602-427. The secondpart is plausible since it is correct.

The "C" distractor is plausible since the first part is correct. The second part is plausible because1G31-F001 is a RWCU isolation valve.

"D" is correct.

References:NONE

K/A:

400000 Component Cooling Water System (CCWS)

K3. Knowledge of the effect that a loss or malfunction of the CCWS will have on thefollowing: (CFR: 41.7 / 45.6)

K3.01 Loads cooled by CCWS . . . . . . . . . . . . . . . . 2.9 3.3


G31-RWCU-LP-00301 Rx Water Cleanup, Ver. 5.2, EO 003.013.A.03


34SO-G31-003-2, Reactor Water Cleanup System, Ver. 43.5 Bank question from HLT Database which was used on 2009 HLT-4 NRC Exam Q#62

Tuesday, April 19, 2016 3:14:45 PM 200


The LOWEST listed Hydrogen concentration that will REQUIRE entry into31EO-EOP-012-1, PC Primary Containment Control, due to Primary Containment Hydrogenconcentration, is __________ .

IAW 31EO-EOP-104-1, Primary Containment Venting for Hydrogen/Oxygen Control, the PREFERRED method to vent containment is to align the CAD System vent valves to the __________ .

1.6%; Torus

1.6%; Drywell

4.1%; Torus

4.1%; Drywell

A.

B.

C.

D.

Description:

If any of the following entry conditions exist, ALL paths on the PC flowchart are entered andsteps performed concurrently until an appropriate Wait Until is reached:

Torus temperature above 100°F (most limiting torus temperature LCO)Drywell temperature above 150°F (drywell temperature LCO)Drywell pressure above 1.85 psig (high drywell pressure scram setpoint)Torus water level above 150" (maximum torus water level LCO)Torus water level below 146" (minimum torus water level LCO)Primary containment hydrogen concentration above 1.5% (High Hydrogen Alarm Setpoint)

Procedure 31EO-EOP-104-2, Primary Containment Venting for Hydrogen Control providesexplicit instructions for venting and purging the Torus, Drywell, and Primary Containment. Thepreferred vent path is through the Torus to take advantage of scrubbing. Vent and purge areperformed to reduce the Hydrogen concentration to less than the explosive level.

The containment is vented first via the Torus to gain the advantage of the radionuclidescrubbing action of the Torus water. The procedure allows the operator to use either or both ofthe CAD loops to control containment pressure.

Tuesday, April 19, 2016 3:14:45 PM 201

ILT-10 SRO NRC EXAMAlthough the primary containment is preferentially vented from the torus to take advantage oftorus scrubbing for minimizing the amount of radioactivity released, this is appropriate only iftorus venting alone is sufficient to control primary containment pressure. Concurrent ventingfrom the drywell and torus may be appropriate for evolutions that require rapid depressurizationof the containment.

Normal containment venting will be via the Drywell (Drywell pressure increasing) using theCAD valves.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know the preferred methodto vent containment (operational implication) and whether a high Hydrogen concentration entrycondition to the EOP flowchart exists.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sinceContainment will be vented via the Drywell if Torus venting is unsuccessfull but is not thepreferred method. Also plausible since normal Containment venting is via the Drywell.

The "C" distractor is plausible is plausible since 4% concentration is the value from the34SV-SUV-019-1, Control Room Rounds, and selecting 4.1% as the entry condition for the PCEOP flowchart. The second part is plausible since it is correct.

The "D" distractor is plausible is plausible since 4% concentration is the value from the34SV-SUV-019-1, Control Room Rounds, and selecting 4.1% as the entry condition for the PCEOP flowchart. The second part is plausible since Containment will be vented via the Drywell ifTorus venting is unsuccessfull but is not the preferred method. Also plausible since normalContainment venting is via the Drywell.

References:NONE

K/A:

EPE: 500000 High Containment Hydrogen Concentration

EK1. Knowledge of the operational implications of the following concepts as they apply toHIGH CONTAINMENT HYDROGEN CONCENTRATIONS: (CFR: 41.8 to 41.10)

EK1.01 Containment integrity . . . . . . . . . . . . . . . . . . 3.3 3.9


Tuesday, April 19, 2016 3:14:45 PM 202

ILT-10 SRO NRC EXAMEOP-PC-LP-20310, Primary Containment Control (PC), Ver. 4.0, EO 201.093.A.01


31EO-EOP-012-1, PC Primary Containment Control, Ver. 7.0 Modified from 2011 Browns Ferry NRC Exam Q#27

Original Question

Which ONE of the following completes the statements below?

Primary Containment hydrogen concentration above __(1)__ is an entry condition for 1-EOI-2,Primary Containment Control.

The method used to control hydrogen and oxygen concentration in accordance with 1-EOI-2, isto inject the __(2)__ in to containment.

A. (1) 4%(2) “A” Nitrogen Storage Tank

B. (1) 4%(2) “B” Nitrogen Storage Tank

C. (1) 2.4%(2) “A” Nitrogen Storage Tank

D. (1) 2.4%(2) “B” Nitrogen Storage Tank

Tuesday, April 19, 2016 3:14:45 PM 203


Unit 1 is shutdown with RPV pressure at 100 psig.

Subsequently, a fire has started in Fire Zone 1408, Switchgear Room 1F.

The fire brigade has requested all electrical switchgear located in the room to be de-energized toaid in suppressing the fire.


Shutdown Cooling can be placed into service using __________ to achieve and maintain aCold Shutdown condition.

RHR pump 1B and RHRSW pump 1C

RHR pump 1B and RHRSW pump 1D

RHR pump 1C and RHRSW pump 1C

RHR pump 1C and RHRSW pump 1D

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:45 PM 204


FEEDERBREAKER

DESCRIPTION

1R22-S005 4160V STA SERV SWGR 1E DIESEL BUILDING (1E SWGR RM)Frame 2 RHR Service Water Pump 1A, 1E11-C001AFrame 7 RHR Pump 1A, 1E11-C002A

FEEDERBREAKER

DESCRIPTIONLOAD

1R22-S006 4160V STA SERV SWGR 1F DIESEL BUILDING (1F SWGR RM)Frame 3 RHR Service Water Pump 1C, 1E11-C001CFrame 7 RHR Pump 1D, 1E11-C002DFrame 9 RHR Pump 1C, 1E11-C002C

FEEDERBREAKER

DESCRIPTION

1R22-S007 4160V STA SERV SWGR 1G DIESEL BUILDING (1G SWGR RM)Frame 2 RHR Service Water Pump 1B, 1E11-C001BFrame 3 RHR Service Water Pump 1D, 1E11-C001DFrame 7 RHR Pump 1B, 1E11-C002B

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine which equipmentwill be affected (available for Shutdown Cooling operation) by the Fire Brigades fire fightingactivities.

The "A" distractor is plausible because 2 RHRSW pumps are powered from a different4160VAC Emergency Bus. This combination would require RHRSW Loops to be crosstiedwhich is available.

"B" is correct.

The "C" distractor is plausible because 2 RHR pumps are powered from one 4160VACEmergency Bus and 2 RHRSW pumps are powered from a different 4160VAC Emergency Bus.

The "D" distractor is plausible because 2 RHR pumps are powered from one 4160VACEmergency Bus.

Tuesday, April 19, 2016 3:14:45 PM 205


K/A:

APE: 600000 Plant Fire On Site

AA2 Ability to determine and interpret the following as they apply to PLANT FIRE ONSITE:

AA2.14 Equipment that will be affected by fire suppression activities in each zone . . 3.0 3.6


R22-ELECT-LP-02702-7.0, 4160 VAC System, Ver. 7.0, EO 200.017.A.01


34SO-R22-001-1, 4160 VAC System, Ver. 22.0 Modified from 2010 Grand Gulf NRC Exam Q#77

Original Question

Use your provided references to answer this question.

A fire has started in fire zone 1A208, ESF Switchgear Room.

The fire brigade has requested all electrical switchgear located in the room be de–energized toaid in suppressing the fire.

For which of the following systems will this action result in the loss of equipment required toachieve and maintain cold shutdown?

A. RHR ‘A’

B. RCIC

C. LPCS

D. RWCU

Tuesday, April 19, 2016 3:14:45 PM 206

ILT-10 SRO NRC EXAM65. 700000G2.4.45 001

Unit 1 is operating at 100% RTP. Unit 2 is in a Refueling outage.

Subsequently, a grid disturbance results in the following conditions/alarms:

o 4160 BUS 1E VOLTAGE LOW, (652-122), ILLUMINATED o 4160 BUS 1F VOLTAGE LOW, (652-222), ILLUMINATED o 4160 BUS 1G VOLTAGE LOW, (652-322), ILLUMINATED

All 4160 VAC Emergency Buses indicate 3700 VAC.

The load dispatcher reports that these conditions will exist for 4 hours.

The following parameters currently exist:

o Generator H2 pressure 58 psig o Generator Megawatts 930 MWe o Generator Megavars +400 MVARs


IAW 34AB-S11-001-0, Operation With Degraded Voltage, one (1) hour later, __________REQUIRED to be supplying power to an Emergency Bus on Unit 1.

The Unit 1 Main Generator is operating __________ the limits of the Generator CapabilityCurve.

REFERENCE PROVIDED

NO EDGs are; within

NO EDGs are; outside

ONLY one EDG is; within

ONLY one EDG is; outside

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:45 PM 207


34AB-S11-001-0 states that if voltages are <3825VAC for 30 minutes, then the 1A and 2AEDGs will be supplying power to their respective emergency buses.

Step 4.2 states; Monitor the main generator for compliance with the capability curve per34SO-N40-001-1 or 34SO-N40-001-2. To ensure proper operation of the Main Generator withthe different types of loading, the Generator capabilities curve is used. The curve is based onHydrogen pressure, megawatts, and megavars. Using the value for machine gas pressure, thegenerator pressure operating line should be the next lower pressure line. With Hydrogenpressure, MWe, & MVARs at their present value, the generator is operating at a point that isoutside the limits of the Generator Capability curve (Att. 1).

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to interpret alarms (just aboveTS value) and in conjunction with the voltages on the Emergency Buses determines that4160VAC Emergency Buses are below their TS limit requiring one (1) EDG on Unit 1 besupplying its respective bus.

The "A" distractor is plausible because 3675 VAC is the setpoint for ARP 652-102-1. Thesecond part is plausible if the applicant uses the wrong Hydrogen pressure line, 60 psig vs. 52psig line.

The "B" distractor is plausible because 3675 VAC is the setpoint for ARP 652-102-1. Thesecond part is plausible since it is correct.

The "C" distractor is plausible since the first part is correct. The second part is plausible if theapplicant uses the wrong Hydrogen pressure line, 60 psig vs. 52 psig line.

"D" is correct.

Tuesday, April 19, 2016 3:14:45 PM 208

ILT-10 SRO NRC EXAMReferences:34SO-N40-001-1, Main Generator Operation, Att. 1, Generator Capability Curve

K/A:

APE: 700000 Generator Voltage and Electric Grid Disturbances

G2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12) . . . . . . . . . . . 4.1 4.3


S11-LP-02706-02, Basic Grid Operating Concepts, Ver. 2.0, EO 200.116.A.04


34AB-S11-001-0, Operation With Degraded System Voltage, Ver. 4.0 34SO-N40-001-1, Main Generator Operation, Attachment 1 - Generator Capability Curve, Ver. 18.0

Tuesday, April 19, 2016 3:14:45 PM 209

ILT-10 SRO NRC EXAM66. G2.1.5 001

Both Units are operating at 100% RTP.

Based on the above conditions and IAW NMP-AD-016-003, Scheduling and Calculating WorkHours,

The MAXIMUM number of hours that a Nuclear Plant Operator may work in any 24 hour period is _________ .

The MAXIMUM number of hours that a Nuclear Plant Operator may work in any 7-day/168-hour period is _________ .

14 hours; 84 hours

14 hours; 72 hours

16 hours; 84 hours

16 hours; 72 hours

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:45 PM 210


NMP-AD-016-003 step 4.1.2 states "The following work hour ceiling limits apply to coveredindividuals regardless of unit status unless the unit status is related to declared plant emergenciesor an unannounced emergency preparedness exercise:

No more than 16 work hours in any 24-hour period No more than 26 work hours in any 48-hour periodNo more than 72 work hours in any 7-day/168-hour period"

30AC-OPS-003-0 step 4.1.2.b states "Shift crew composition may be less than the minimumrequirement for less than 2 hours in order to accommodate unexpected absences providedimmediate actions are taken to restore minimum crew composition to WITHIN the minimumrequirements."

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to use NMP-AD-016-003,Scheduling and Calculating Work Hours, (procedures) related to work hour and overtimelimitations.

The "A" distractor is plausible if the applicant remembers 2 hours is allowed to call in a relieftherefore adding 2 hours to the normal 12 hour shift results in 14 hours as the maximum. Thesecond part is plausible if the applicant doubles the normal 12 hour shift resulting in 24 hours asthe maximum.

The "B" distractor is plausible if the applicant remembers 2 hours is allowed to call in a relieftherefore adding 2 hours to the normal 12 hour shift results in 14 hours as the maximum. Thesecond part is plausible since it is correct.

The "C" distractor is plausible since the first part is correct. The second part is plausible if theapplicant doubles the normal 12 hour shift resulting in 24 hours as the maximum.

"D" is correct.

Tuesday, April 19, 2016 3:14:45 PM 211


K/A:

2.1 Conduct of Operations

G2.1.5 Ability to use procedures related to shift staffing, such as minimum crewcomplement, overtime limitations, etc. (CFR: 41.10 / 43.5 / 45.12) . . . . . . . 2.9* 3.9


LT-LP-30003, Tech Specs/Administrative Controls, Ver. 8.0, EO 300.001.B.02


30AC-OPS-003-0, Plant Operations, Ver. 29.0 NMP-AD-016-003, Scheduling and Calculating Work Hours, Ver. 7.1

Original Question

Tuesday, April 19, 2016 3:14:45 PM 212


Unit 1 is in REFUEL with a core shuffle currently in progress.

Based on the above conditions and IAW 34FH-OPS-001-0, Fuel Movement Operation,

The individual in the Main Control Room on the headset with the Refueling SRO __________ REQUIRED to have a NRC License.

The Reactor Mode Switch __________ be LOCKED in the REFUEL position.

is; will

is; will NOT

is NOT; will

is NOT; will NOT

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:45 PM 213


34FH-OPS-001-0 contains the following:

Step 4.3.4 Fuel movements involving core alterations will be followed in the Control Roomon an appropriate core representation by a licensed individual. This individualwill monitor core parameters and communicate trends to the SRO on therefueling platform.

Step 5.2.2 Fuel movements in the reactor vessel may be performed only WHEN the ReactorMode switch is LOCKED in the REFUEL position.

Step 5.2.9 IF, during fuel movement, it is found that conditions have changed such that anyof the requirements of this procedure are no longer satisfied, any member of therefueling bridge team has the authority to halt fuel movement.

Step 5.2.15.4 The SRO must ensure that the control room is aware of conditions on therefueling floor. Constant communications will be maintained with a licensedindividual in the control room when core alterations are in progress.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know who can communicatewith the Refueling SRO and the status of equipment (Reactor Mode Switch) supporting fuelmovement.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible becausethere are times when the Reactor Mode switch can be in the REFUEL position and unlocked(NO Core Alterations in progress).

The "C" distractor is plausible because there are times when a NPOIT can be involved with theControl Room (Local ASD operation). The second part is plausible since it is correct.

The "D" distractor is plausible because there are times when a NPOIT can be involved with theControl Room (Local ASD operation). The second part is plausible because there are timeswhen the Reactor Mode switch can be in the REFUEL position and unlocked (NO CoreAlterations in progress).

References:NONE

K/A:

Tuesday, April 19, 2016 3:14:45 PM 214

ILT-10 SRO NRC EXAM2.1 Conduct of Operations

G2.1.44 Knowledge of RO duties in the control room during fuel handling such asresponding to alarms from the fuel handling area, communication with the fuel storagefacility, systems operated from the control room in support of fueling operations, andsupporting instrumentation. (CFR: 41.10 / 43.7 / 45.12) . . . . . . . . . . . . 3.9 3.8


F15-RF-LP-04502, Refueling, Ver. 5.0, EO 300.048.A.01 & EO 400.013.O.01


34FH-OPS-001-0, Fuel Movement Operation, Ver. 25.5 Modified from HLT Database which was used on 2011 HLT-6 NRC Exam Q#91

Original Question

Unit 1 is in REFUEL with core reload in progress.

The Control Room informs the Refueling SRO that the individual on the headset with them hasto be relieved.

IAW 34FH-OPS-001-0, Fuel Movement Operation.

The individual who relieves the person in the Main Control Room __________ REQUIREDto have a NRC License.

The fuel movement prerequisites must be completed __________ .

A. is; ONLY once during the refueling outage (prior to the initial fuel movement)

B. is NOT;ONLY once during the refueling outage (prior to the initial fuel movement)

C. is; at EACH shift change (12 hour shift) during fuel movement

D. is NOT;at EACH shift change (12 hour shift) during fuel movement

Tuesday, April 19, 2016 3:14:45 PM 215


IAW the Unit 1 Facility Operating License, Southern Nuclear is authorized to operate Plant Hatch at a MAXIMUM steady-state reactor core power level of __________ .

2777 MWth

2804 MWth

2818 MWth

2832 MWth

A.

B.

C.

D.

Description:

IAW the Unit 1 Facility Operating License, Renewed License No. DPR-57, 2.C states: Page 4 of 8, C. This renewed license shall be deemed to contain, and is subject to, the conditions specified in

the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; allapplicable provisions of the Act and the rules, regulations, and orders of the Commissionnow or hereafter in effect; and the additional conditions specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady-state reactor core powerlevels not in excess of 2,804 megawatts thermal.

IAW 34GO-OPS-022-0, Maintaining Rated Thermal Power, Attachment 1 CTP Program CoreThermal Power Data, Acceptance Criteria has the following table:

SUBSECTION 7.1 DATA 0630-1830 1830-0630 ACCEPTANCECRITERIA

AVG CORE THERMAL PWR(CMWt)

< 2804 CMWT

MAX TIME OVER 100.0% (MIN) < 15 minMAX TIME OVER 100.5% (MIN) = 0 minMAX TIME OVER 101.0% (MIN) = 0 minMAX TIME OVER 102.0% (MIN) = 0 min

INITIAL

Cross Flow System out of service is 2777 MWTh. The 100.5% RTP value is 2818 MWTh. The 101.0% RTP value is 2832 MWTh.

Tuesday, April 19, 2016 3:14:45 PM 216

ILT-10 SRO NRC EXAMK/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know the maximum corethermal power as stated in the facility license.

The "A" distractor is plausible because 2777 Megawatts Thermal value for when Cross FlowSystem is out of service per 34GO-OPS-022-0 "Limitations".

"B" is correct.

The "C" distractor is plausible because 2818 Megawatts Thermal is the value for 100.5% CTP of34GO-OPS-022-0 acceptance criteria.

The "D" distractor is plausible because 2832 Megawatts Thermal is the value for 101.0% CTP of34GO-OPS-022-0 acceptance criteria.

References:NONE

K/A:

2.2 Equipment Control

G2.2.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13) . . . . . . . . . . . . . . 3.6 4.5


C95-PC-LP-40001, Process Computer, Ver. 3.0


Edwin I. Hatch Nuclear Plant, Unit No. 1 Renewed Facility Operating License DPR-57,Renewed License No. DPR-57 34GO-OPS-022-0, Maintaining Rated Thermal Power, Ver. 21.1

Tuesday, April 19, 2016 3:14:45 PM 217


Unit 2 is shutdown with the following conditons:

o RPV pressure 134 psig o Recirculation pump 2A OFF o Recirculation pump 2B Running

Based on the above conditions and IAW Tech Spec 3.4.7, Residual Heat Removal (RHR)Shutdown Cooling System - Hot Shutdown,

The MINMUM number of RHR Shutdown Cooling (SDC) subsystems required to beoperable, without requiring entry into a Required Action Statement (RAS), is __________ .

A RHR SDC subsystem __________ REQUIRED to be in operation.

one (1); is

one (1); is NOT

two (2); is

two (2); is NOT

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:45 PM 218


U2 TS 3.4.7, Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown.

LCO 3.4.7 Two RHR shutdown cooling subsystems shall be OPERABLE and, with norecirculation pump in operation, at least one RHR shutdown cooling subsystem shallbe in operation.

APPLICABILITY: MODE 3 with reactor steam dome pressure less than the RHR low pressurepermissive pressure.

Given plant conditions of <135 psig RPV pressure and 1 Recirc Pump operating, 2 subsystemsare required, but neither are required to be in operation.

With NO Recirc Pumps running one RHR SDC subsystem would be required to be in service.

In Mode 5, only one RHR subsystem is required IAW TS 3.9.7.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to apply Tech Specs for asystem (Number of required RHR SDC subsystems & if required to place in service).

The "A" distractor is plausible if the applicant remembers there is a time when only one RHRSDC subsystem is required to be operable (Mode 5 Refueling) and applies it to this condition. The second part is plausible if the applicant remembers there is a time when the RHR SDCsubsytem is required to be in operation (NO Recirc pump in service) and applies it to thiscondition.

The "B" distractor is plausible if the applicant remembers there is a time when only one RHRSDC subsystem is required to be operable (Mode 5 Refueling) and applies it to this condition. The second part is plausible since it is correct.

The "C" distractor is plausible since the first part is correct. The second part is plausible if theapplicant remembers there is a time when the RHR SDC subsytem is required to be in operation(NO Recirc pump in service) and applies it to this condition.

"D" is correct.

References:NONE

Tuesday, April 19, 2016 3:14:45 PM 219

ILT-10 SRO NRC EXAMK/A:


G2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 / 43.5 / 45.3) . . . . . . . . . . 3.4 4.7


E11-RHR-LP-00701, Residual Heat Removal System, Ver. 9.2, EO 300.006.a.34


U2 Tech Specs 3.4.7, Residual Heat Removal (RHR) Shutdown Cooling System - HotShutdown, Amendment 210

U2 Tech Specs 3.9.7 Residual Heat Removal (RHR) - High Water Level, Amendment 210 Modified from HLT Database which was used on HLT-5 NRC Exam Q#49

Original Question

Unit 2 is shutdown with the following conditons

o Rx pressure ............................... 134 psigo "2A" Rx Recirculation pump ...... Runningo "2B" Rx Recirculation pump ...... Off

Which ONE of the choices below completes BOTH the following statement?

IAW Tech Spec 3.4.7, "Residual Heat Removal (RHR) Shutdown Cooling System - HotShutdown" the MINMUM number of RHR Shutdown Cooling (SDC) subsystems required tobe operable, (without requiring entry into a Required Action Statement) (RAS), is _______ .

Also, IAW with Tech Spec 3.4.7 and with current plant conditions, ________ RHR SDCsubsystem is required to be in operation.

A. one;one

B. one;neither

C. two;

Tuesday, April 19, 2016 3:14:45 PM 220

ILT-10 SRO NRC EXAMone

D. two;neither

Tuesday, April 19, 2016 3:14:45 PM 221


On Unit 2 the following annunciator has been determined to be INOPERABLE:

o TURBINE TROUBLE, 650-105

IAW 31GO-OPS-014-0, Annuciator and Plant Component Control, alarm window 650-105 isDE-ACTIVATED.

Based on the above condition and IAW 31GO-OPS-014-0,

With 650-105 de-activated, a __________ will be installed beside the annunciator numberlabel under the annunciator window.

yellow magnetic "P" tile

yellow magnetic "R" tile

white magnetic "P" dot

white magnetic "R" dot

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:45 PM 222


IAW 31GO-OPS-014-0: Whenever an APC Log entry has been made and an APC Sheet written, THEN INSTALL anAPC blank YELLOW magnetic tile beside the annunciator number label under the Annunciatorwindow. If an alarm is deactivated, the YELLOW magnetic blank tile will be replaced with aYELLOW magnetic "P" tile beside the Annunciator number label under the Annunciatorwindow. The YELLOW "P" tile will signify that the annunciator has been "pulled".

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine how an inoperablealarm, that has been deactiviated, is labeled (track).

"A" is correct.

The "B" distractor is plausible because to de-activate an alarm either the annunciator card has tobe removed (R) or wires have to removed (lifted). Yellow is the color of the proper tile.

The "C" distractor is plausible because to de-activate an alarm either the annunciator card has tobe removed (R) or wires have to removed (lifted). There is a time when a "White" item is placedon equipment/indications/recorders that are not function properly in the Main Control Room("White DOT).

The "D" distractor is plausible because to de-activate an alarm either the annunciator card has tobe removed (R) or wires have to removed (lifted). There is a time when a "White" item is placedon equipment/indications/recorders that are not function properly in the Main Control Room("White DOT). References:NONE

K/A:


G2.2.43 Knowledge of the process used to track inoperable alarms. (CFR: 41.10 / 43.5 / 45.13) . . . . . . . . . . . . . . . . . . 3.0 3.3


LT-LP-30004, Admin Procedures, Ver. 14.0,


31GO-OPS-014-0, Annuciator and Plant Component Control, Ver. 3.0

Tuesday, April 19, 2016 3:14:45 PM 223


Unit 2 is operating at 100% RTP with a steam leak in the 2A SJAE room. A SO has beenassigned to investigate. RP reports that the general area radiation levels are 120 mR/hr.


A __________ RWP will be used to enter the SJAE room and __________ is the type ofinformation you would expect to find on this type of RWP.

Specific; Dose rates at Hot Spots

Specific; Digital Alarming Dosimeter (DAD) settings

General; Dose rates at Hot Spots

General; Digital Alarming Dosimeter (DAD) settings

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:45 PM 224


IAW 60AC-HPX-004, Radiation and Contamination Control, step 4.2.2.1 states "General RWPsare used for inspection, plant tours and general area walk-downs in areas other than HighlyContaminated, High Radiation, Very High Radiation or Airborne Radioactivity Areas."

Step 4.2.3 states "Specific RWPs are used for controlling work that is not controlled by aGeneral RWP and occurs in a High Radiation Area, High Contamination Area or areas withchanging radiological conditions."

A Specific RWP will be considered if any of the following conditions exist:

o General area radiation levels >100 mrem/hr o Airborne radioactivity concentration greater than 30% of Derived Air Concentrations

(DACs) specified in 10CFR20 o Loose surface contamination levels greater than 200,000 dpm/100 cm2 o Breach of a highly contaminated system, or in a High or Very High Radiation Area

Since the general dose rate is 120 mr/hr, a "Specific" RWP will be used. Digital AlarmingDosimeter (DAD) settings will be listed on the RWP.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine which RWP tocomply with during entry into a high radiation area.

The "A" distractor is plausible since the first part is correct. The second part is plausible becauseHot Spots are identified by radiological surverys and maintained using the RadIS databasesystem.

"B" is correct.

The "C" distractor is plausible because a "General" RWP is used for inspections of areas if thearea dose rate was less than the high radiation area value (<100 mr/hr). The second part isplausible because Hot Spots are identified by radiological surverys and maintained using theRadIS database system.

The "D" distractor is plausible because a "General" RWP is used for inspections of areas if thearea dose rate was less than the high radiation area value (<100 mr/hr). The second part iscorrrect.

Tuesday, April 19, 2016 3:14:45 PM 225


K/A:

2.3 Radiation Control

G2.3.7 Ability to comply with radiation work permit requirements during normal orabnormal conditions. (CFR: 41.12 / 45.10) . . . . . . . . 3.5 3.6


LT-LP-30008, Radiation Control Adminstration Procedures And Instrumentation, , Ver. 9.0, LO LT-30008.003


2011 Hatch HLT-6 Audit ExamQ#72 River Bend 2008 NRC Exam Q#72 60AC-HPX-004-0, Radiation And Contamination Control, Ver. 22.0

Tuesday, April 19, 2016 3:14:45 PM 226


Unit 2 Radwaste is discharging Waste Sample Tank A to the canal.

Subsequently, the following indication is received:

o 2G11-R045, Total Plant Dilution Flow, recorder indicates 9500 gpm


The Radwaste discharge to the canal __________ .

With the existing Specific Release Permit, __________ permitted to restart the discharge ofWaste Sample Tank A to the canal.

will automatically terminate; ONLY one (1) restart is

will automatically terminate; NO restarts are

must be manually terminated;ONLY one (1) restart is

must be manually terminated;NO restarts are

A.

B.

C.

D.

Description:

The discharge must be stopped and the discharge line isolated upon occurrence of any of thefollowing:

a. Rad Monitor Reading drops more than 5% below the background reading indicated on theDischarge Permit.

b. Dilution flow decreases below 10,000 gpm.c. Rad Monitor Reading exceeds the High Radiation trip setpoint. d. Downscale/INOP trip is set at 3 cps

NOTE: a or c above will cause the Radwaste Effluent High Radiation Alarm to Annunciate.

NOTE: b,c or d above will auto close 2G11-F184 and 2G11-F185, Radwaste Canal DischargeLine Auto-Isolation valves.

Tuesday, April 19, 2016 3:14:45 PM 227

ILT-10 SRO NRC EXAM34SO-G11-021-2 contains the following CAUTIONS:

Cauton before step 7.2.6: The required minimum dilution flowrate for all discharges will always be 10,000 gpm. Therewill be no deviation from this specified value. If the dilution flowrate during a discharge tocanal as measured on 2G11-R045 falls below 10,000 gpm, then the discharge will be suspendedor terminated. Only one restart of a specific discharge permit is permissible after a loss ofdilution flow and the restart must be accomplished within 6 hours of the time the discharge wassuspended.

Cauton before step 7.2.8: If the Radwaste Effluent monitor reading in CPS, approximately 10 minutes after the start of thedischarge, is > 5% below the radwaste effluent monitor background as indicated on the dischargepermit. If the monitor reading is not acceptable, the discharge must be stopped and it isrecommended that the monitor be backwashed. A new background must be determined and anew trip setpoint calculated.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know how radiation releasesare terminated (controlled) when Radwaste is discharging radioactive water to the canal and ifthe release can be restarted with the current release permit.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sincethere is a time when a restart is not permitted with the existing release permit (Radwaste Effluentmonitor reading is > 5% below the radwaste effluent monitor background as indicated on thedischarge permit that a new background must be determined and a new trip setpoint calculated,indicative of a new release permit).

The "C" distractor is plausible since the valve which controls dilution flow (2N71-F009, CircWater Blowdown) must be manually positioned, therefore, applying that the discharge to thecanal must be manually positioned. The second part is plausible since it is correct.

The "D" distractor is plausible since the valve which controls dilution flow (2N71-F009, CircWater Blowdown) must be manually positioned, therefore, applying that the discharge to thecanal must be manually positioned. The second part is plausible since there is a time when arestart is not permitted with the existing release permit (Radwaste Effluent monitor reading is >5% below the radwaste effluent monitor background as indicated on the discharge permit that anew background must be determined and a new trip setpoint calculated, indicative of a newrelease permit).

Tuesday, April 19, 2016 3:14:45 PM 228


K/A:


G2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10) . . . . . 3.8 4.3


G11-RW-LP-02901, Radwaste System, Ver. 3.1, EO 029.179.B.07


34SO-G11-021-2, Radwaste Sample Tank Operating Procedure, Ver. 11.7 Bank question from HLT Database Q#LT-029179-001

Tuesday, April 19, 2016 3:14:45 PM 229


The power supplies for the Unit 1 Post-Treatment Radiation monitors, 1D11-K615A & K615B, is __________ .

The location of the Unit 1 Post-Treatment Radiation detectors NORMAL sample point is just prior to __________ .

RPS Bus B & Instrument Bus 1A; EXITING the Main Stack

RPS Bus B & Instrument Bus 1A; ENTERING the Main Stack

24/48 VDC Cabinet 1A & 1B EXITING the Main Stack

24/48 VDC Cabinet 1A & 1B ENTERING the Main Stack

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:45 PM 230


The Offgas Post Treatment Radiation Monitoring has two gamma sensitive Scintillationdetectors. The detectors have three sampling points. The normal sample point is in the Offgassystem just prior to release into the stack. The sample can also be taken from two differentpoints between the adsorber vessels, one from between vessels D008 and D009, and the otherfrom between vessels D014 and D015. The detectors are powered from 24 VDC Cabinet "A"and "B".

There are two gamma sensitive Scintillation detectors in the Main Stack Radiation MonitoringSystem. A gas sample is drawn through an Isokinetic Probe which is located high enough inthe main stack vent stream to assure representative sampling. The sample passes through twoshielded chambers where the radiation level of the vent gas is measured.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to have knowledge of thepower supplies and normal sample point for fixed readiation monitor (Post-Treatment).

The "A" distractor is plausible since this is the power supplies to the Pre-Treatment RadMonitors and applying this to the Post-Treatment monitors. The second part is plausible sincethis is the location for a radiation monitor sample point (Main Stack Radiation monitor) andapplying this to the Post-Treatment monitors.

The "B" distractor is plausible since this is the power supplies to the Pre-Treatment RadMonitors and applying this to the Post-Treatment monitors. The second part is plausible since itis correct.

The "C" distractor is plausible since the first part is correct. The second part is plausible sincethis is the location for a radiation monitor sample point (Main Stack Radiation monitor) andapplying this to the Post-Treatment monitors.

"D" is correct.

Tuesday, April 19, 2016 3:14:45 PM 231


K/A:


G2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors andalarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.12 / 43.4 / 45.9) . . . . . . 2.9 3.1


D11-PRM-LP-10007, Process Radiation Monitors, Ver. 5.1, EO 200.030.A.05


34AB-R22-001-1, Loss Of DC Buses, Ver. 6.0 A-10182, Load List For Distribution Panel 1R25-S015, Ver. 4.0 A-10183, Load List For Distribution Panel 1R25-S016, Ver. 3.0 A-10228, Load List For Distribution Panel 1R25-S064, Ver. 52.0

Tuesday, April 19, 2016 3:14:45 PM 232


An ATWS exists on Unit 1.

o Reactor Power is slowly DECREASING o Boron has NOT been injected into the RPV o All SRMs and IRMs have been fully inserted

The following conditions occur:

Time Reactor Power01:00 APRMs indicate 4% RTP 01:10 IRMs on Range 7 01:20 IRMs on Range 5 01:30 SRMs indicate 6.0 x 104 CPS

Based on the above conditions and IAW the EOPs,

The EARLIEST time the reactor can be declared SHUTDOWN with the reactorSUBCRITICAL is at __________ .

01:00

01:10

01:20

01:30

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:45 PM 233


IAW the EOP, the reactor is considered shutdown if it is subcritical with reactor power belowthe heating range. The definition of subcritical (IRMs below range 6 with no boron injected)is the point where the EOPs directs the operators to exit the EOPs and continue shutdownactivities with normal plant procedures (Scram Procedure).

This definition of shutdown is not the same as "shutdown under all conditions", which is a termalso used in the EOPs. The reactor is considered shutdown under all conditions if either, a coldshutdown rod configuration exists, or cold shutdown boron weight (CSBW) has been injected.

For conditions where a cold shutdown rod configuration does NOT exist, if the reactor isshutdown and no boron has been injected, an exit to the scram procedure is appropriate eventhough the margin to criticality is small. A return to criticality is manageable because guidancein the RPV pressure control will terminate any cooldown, stopping the reactor power increase.

K/A JUSTIFICATION:

This question satisfies the K/A statement because the applicant must demonstrate knowledge ofwhen the reactor is considered shutdown, based on the EOPs.

The "A" distractor is plausible because if reactor power is still above the APRM downscalesetpoint (5%) following a scram, then it is a clear indication of a failure to scram and alternatemethods of reactivity control must be used. However, with reactor power below 5%, the actionsto control reactivity are very similiar to, if not exactly the same as, the actions taken during ashutdown with normal plant procedures. IAW the EOPs and scram procedure, if an AWTS exitsand reactor is below 5%: the EOP entry condition is NOT met and the scram procedure isentered, Recirc pumps are NOT tripped and boron is NOT injected.

The "B" distractor because reactor power is below the power range and in the intermediate range(neglecting overlap). A caution statement in 34GO-OPS-013-1, Normal Plant Shutdown, statesthat reactor power below IRM range 7 is below the heating range.

"C" is correct.

The "D"distractor because at this plant condition the IRMs are now fully inserted and wouldindicate down scale and the SRMs are the normal method for monitoring reactor power whileshutdown. Additionally both criteria for calling the reactor critical IAW 34GO-OPS-001-1, PlantStartup, 1) SRM period meters indicate a stable positive period, 2) SRM levels are increasingwithout requiring additional control rod withdrawl. This answer is correct but not the earliestshutdown condition listed.

Tuesday, April 19, 2016 3:14:45 PM 234


K/A:

2.4 Emergency Procedures / Plan

G2.4.17 Knowledge of EOP terms and definitions. (CFR: 41.10 / 45.13) . . . . . . . . . . . . . . . 3.9 4.3


EOP-TERMS-LP-20303, EOP Terms & Definitions, Ver. 3.2, EO 201.093.A.08


EOP/SAG Guidelines, Technical Basis Vol 1, Rev. 3.031GO-OPS-013-1, Normal Plant Shutdown, Ver. 29.1

Tuesday, April 19, 2016 3:14:45 PM 235


An emergency has been declared at Plant Hatch.


The LOWEST listed emergency classification that will REQUIRE activation of the Technical Support Center (TSC) is __________ .

The ALTERNATE location for the TSC is __________ .

an Alert Emergency; Classroom 172, Simulator Building

an Alert Emergency; the Main Control Room

a Notification of Unusual Event; Classroom 172, Simulator Building

a Notification of Unusual Event; the Main Control Room

A.

B.

C.

D.

Tuesday, April 19, 2016 3:14:45 PM 236


Emergency Response Facilities, in addition to the Control Room, include a Technical SupportCenter (TSC), an Operations Support Center (OSC), and an Emergency Operations Facility(EOF). The TSC, OSC, and EOF must be activated within 60 minutes of declared emergenciesrequiring their support. Although the Control Room is not normally considered an emergency response facility, it servesthis purpose in the truest sense of the term. If an abnormal or emergency condition arises, itmerely changes from a "normal" mode to an "emergency" mode. Control Room personnel areresponsible for identifying and classifying any emergency event in accordance with emergencyimplementing procedures. The Control Room is of course used for all emergency classifications,and is the only facility "activated" for NUEs.

The TSC handles the overall management of the plant's emergency response activities. Itprovides plant management and technical support personnel with adequate space to assist plantoperating personnel located in the Control Room during an emergency. The TSC is equippedwith technical data displays and has ready access to plant records to allow TSC personnel toperform detailed analysis and diagnosis of abnormal plant conditions, including assessment ofany release or radioactivity to the environment. The TSC is activated for Alert, Site-Area andGeneral Emergencies. The TSC is located in the Unit 2 Service Building annex (see Figure 02). The Control Roomserves as the alternate location for the TSC management and supervision should the TSCbecome uninhabitable. All other staff are directed to other facilities.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to know (emergency responsefacilities) the TSC activation emergency and the alternate location of the TSC.

The "A" distractor is plausible since the first part is correct. The second part is plausible sincethis is an alternate location for one of the emergency response facilities (OSC).

"B" is correct.

The "C" distractor is plausible since during a NOUE the emergency response facility that isactivated is the Control Room and since the NOUE is the lowest emergency and applies this tothe TSC. The second part is plausible since this is an alternate location for one of the emergencyresponse facilities (OSC).

The "D" distractor is plausible since during a NOUE the emergency response facility that isactivated is the Control Room and since the NOUE is the lowest emergency and applies this tothe TSC. The second part is plausible since it is correct.

Tuesday, April 19, 2016 3:14:45 PM 237


K/A:


G2.4.42 Knowledge of emergency response facilities. (CFR: 41.10 / 45.11) . . . . . . . . 2.6 3.8


H-EP-SS-20001, Initial EP Response, Ver. 11.1, EO H-EP-002.041.A.02 & EO H-EP-002.041.A.03


73EP-EIP-021-0, Alternate Operations Support Center (OSC) Activation, Ver. 1.3 73EP-EIP-063-0, Technical Support Center Activation, Ver. 11.6 NMP-EP-111, Emergency Notifications, Ver. 11.0

Tuesday, April 19, 2016 3:14:45 PM 238

NRC SRO REFERENCES SRO EXAM

1. 34AB-E11-001-2, Loss Of Shutdown Cooling, Attachments 1, 2 & 3

2. 34SO-N40-001-1, Main Generator Operation, Attachment 1 Capability Curve

3. U1 TS 3.3.1.1 RPS Instrumentation, Pages 3.3-1 – 3.3-3

4. LFD-1-RPS-09, TS 3.3.1.1-1, Item 4 Reactor Protection System Instrumentation Reactor Vessel Water Level – Low, Level 3

5. U1 TS Table 3.3.1.1-1, RPS Instrumentation

6. U1 TS 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring

SNC PLANT E. I. HATCH Pg 15 of 34 DOCUMENT TITLE:

LOSS OF SHUTDOWN COOLING DOCUMENT NUMBER:

34AB-E11-001-2 Ver No:

6.15 ATTACHMENT 1 Att. Pg.

TITLE: CORE FUEL IN UNFLOODED RPV 1 of 2

Case 1 : CORE FUEL in UNFLOODED RPV (Normal RPV water level)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime

1.0 57.275 52 min 4 hr 50 min 5 hr 42 min 2.0 47.051 1 hr 4 min 5 hr 53 min 6 hr 57 min 3.0 40.644 1 hr 14 min 6 hr 48 min 8 hr 3 min 4.0 36.112 1 hr 23 min 7 hr 40 min 9 hr 3 min 5.0 32.822 1 hr 32 min 8 hr 26 min 9 hr 58 min 6.0 30.372 1 hr 39 min 9 hr 6 min 10 hr 46 min 7.0 28.494 1 hr 45 min 9 hr 42 min 11 hr 29 min 8.0 27.011 1 hr 51 min 10 hr 14 min 12 hr 6 min 9.0 25.806 1 hr 57 min 10 hr 43 min 12 hr 40 min 10.0 24.799 2 hr 1 min 11 hr 9 min 13 hr 11 min 11.0 23.935 2 hr 6 min 11 hr 34 min 13 hr 40 min 12.0 23.177 2 hr 10 min 11 hr 56 min 14 hr 7 min 13.0 22.501 2 hr 14 min 12 hr 18 min 14 hr 32 min 14.0 21.887 2 hr 18 min 12 hr 38 min 14 hr 57 min 15.0 21.323 2 hr 21 min 12 hr 59 min 15 hr 20 min 16.0 20.800 2 hr 25 min 13 hr 18 min 15 hr 44 min 17.0 20.310 2 hr 28 min 13 hr 37 min 16 hr 6 min 18.0 19.850 2 hr 32 min 13 hr 56 min 16 hr 29 min 19.0 19.415 2 hr 35 min 14 hr 15 min 16 hr 51 min 20.0 19.002 2 hr 39 min 14 hr 34 min 17 hr 13 min 21.0 18.610 2 hr 42 min 14 hr 52 min 17 hr 35 min 22.0 18.235 2 hr 45 min 15 hr 10 min 17 hr 56 min 23.0 17.877 2 hr 49 min 15 hr 29 min 18 hr 18 min 24.0 17.535 2 hr 52 min 15 hr 47 min 18 hr 39min 25.0 17.207 2 hr 55 min 16 hr 5 min 19 hr 1 min

NOTE: • Saturation Time is the time differential between the time of complete loss of Shutdown cooling and the

onset of boiling.

• Evaporation Time is the time differential between Saturation Time and the Boil-off Time.

• Boil-off Time is the time differential between the time of complete loss of Shutdown cooling and the time water level reaches the top of fuel.



34AB-E11-001-2 Ver No:



Case 1 : CORE FUEL in UNFLOODED RPV (cont'd) (Normal RPV water level)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime

26.0 16.893 2 hr 58 min 16 hr 23 min 19 hr 22 min 27.0 16.592 3 hr 2 min 16 hr 41 min 19 hr 43 min 28.0 16.302 3 hr 5 min 16 hr 59 min 20 hr 4 min 29.0 16.024 3 hr 8 min 17 hr 16 min 20 hr 25 min 30.0 15.756 3 hr 11 min 17 hr 34 min 20 hr 46 min 31.0 15.499 3 hr 15 min 17 hr 51 min 21 hr 6 min 32.0 15.251 3 hr 18 min 18 hr 9 min 21 hr 27 min 33.0 15.013 3 hr 21 min 18 hr 26 min 21 hr 47 min 34.0 14.783 3 hr 24 min 18 hr 43 min 22 hr 8 min 35.0 14.562 3 hr 27 min 19 hr 0 min 22 hr 28 min 36.0 14.348 3 hr 30 min 19 hr 17 min 22 hr 48 min 37.0 14.142 3 hr 33 min 19 hr 34 min 23 hr 8 min 38.0 13.944 3 hr 36 min 19 hr 51 min 23 hr 28 min 39.0 13.752 3 hr 39 min 20 hr 7 min 23 hr 47 min 40.0 13.567 3 hr 42 min 20 hr 24 min 24 hr 7 min 41.0 13.388 3 hr 45 min 20 hr 40 min 24 hr 26 min 42.0 13.215 3 hr 48 min 20 hr 57 min 24 hr 45 min 43.0 13.048 3 hr 51 min 21 hr 13 min 25 hr 4 min 44.0 12.886 3 hr 54 min 21 hr 29 min 25 hr 23 min 45.0 12.730 3 hr 57 min 21 hr 44 min 25 hr 42 min 46.0 12.579 4 hr 0 min 22 hr 0 min 26 hr 1 min 47.0 12.432 4 hr 3 min 22 hr 16 min 26 hr 19 min 48.0 12.290 4 hr 6 min 22 hr 31 min 26 hr 37 min 49.0 12.152 4 hr 8 min 22 hr 47 min 26 hr 55min 50.0 12.018 4 hr 11 min 23 hr 2 min 27 hr 13 min

NOTE: • Saturation Time is the time differential between the time of complete loss of Shutdown cooling and

the onset of boiling. • Evaporation Time is the time differential between Saturation Time and the Boil-off Time. • Boil-off Time is the time differential between the time of complete loss of Shutdown cooling and the

time water level reaches the top of fuel.



34AB-E11-001-2 Ver No:


TITLE: CORE FUEL IN FLOODED RPV 1 of 2

Case 2 : CORE FUEL in FLOODED RPV(Cavity flooded and gates installed)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime

1.0 57.275 2 hr 24 min 33 hr 10 min 35 hr 55 min 2.0 47.051 3 hr 20 min 40 hr 23 min 43 hr 43 min 3.0 40.644 3 hr 51 min 46 hr 55 min 50 hr 37 min 4.0 36.112 4 hr 20 min 52 hr 37 min 56 hr 58 min 5.0 32.822 4 hr 46 min 57 hr 53 min 62 hr 40 min 6.0 30.372 5 hr 9 min 62 hr 34 min 67 hr 44 min 7.0 28.494 5 hr 30 min 66 hr 41 min 72 hr 12 min 8.0 27.011 5 hr 48 min 70 hr 21 min 76 hr 9 min 9.0 25.806 6 hr 4 min 73 hr 38 min 79 hr 43 min 10.0 24.799 6 hr 19 min 76 hr 38 min 82 hr 57 min 11.0 23.935 6 hr 33 min 79 hr 23 min 85 hr 57 min 12.0 23.177 6 hr 46 min 81 hr 59 min 88 hr 45 min 13.0 22.501 6 hr 58 min 84 hr 27 min 91 hr 25 min 14.0 21.887 7 hr 10 min 86 hr 49 min 93 hr 59 min 15.0 21.323 7 hr 21 min 89 hr 7 min 96 hr 28 min 16.0 20.800 7 hr 32 min 91 hr 22 min 98 hr 54 min 17.0 20.310 7 hr 43 min 93 hr 34 min 101 hr 17 min 18.0 19.850 7 hr 54 min 95 hr 44 min 103 hr 38 min 19.0 19.415 8 hr 4 min 97 hr 53 min 105 hr 57 min 20.0 19.002 8 hr 15 min 100 hr 0 min 108 hr 15 min 21.0 18.610 8 hr 25 min 102 hr 7 min 110 hr 32 min 22.0 18.235 8 hr 36 min 104 hr 13 min 112 hr 49 min 23.0 17.877 8 hr 46 min 106 hr 18 min 115 hr 4 min 24.0 17.535 8 hr 56 min 108 hr 22 min 117 hr 19 min 25.0 17.207 9 hr 5 min 110 hr 26 min 119 hr 33 min

NOTE:

• Saturation Time is the time differential between the time of complete loss of Shutdown cooling and the onset of boiling.





34AB-E11-001-2 Ver No:



Case 2 : CORE FUEL in FLOODED RPV (cont'd) (Cavity flooded and gates installed)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime


NOTE:






34AB-E11-001-2 Ver No:


TITLE: CORE AND POOL FUEL IN FLOODED RPV AND POOL 1 of 2

Case 3 : CORE and POOL FUEL in FLOODED RPV and POOL(Cavity flooded and gates removed)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime


NOTE:

• Saturation Time is the time differential between the time of complete loss of cooling (Shutdown & Fuel Pool) and the onset of boiling.





34AB-E11-001-2 Ver No:



Case 3 : CORE and POOL FUEL in FLOODED RPV and POOL (cont'd) (Cavity flooded and gates removed)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime


NOTE: • Saturation Time is the time differential between the time of complete loss of cooling (Shutdown &

Fuel Pool) and the onset of boiling. • Evaporation Time is the time differential between Saturation Time and the Boil-off Time. • Boil-off Time is the time differential between the time of complete loss of Shutdown cooling and the



MAIN GENERATOR OPERATION DOCUMENT NUMBER:

34SO-N40-001-1 Ver No:


TITLE: OPERATING LIMITS FOR GENERATOR VOLTAGE REGULATOR IN AUTOMATIC 1 of 1

-700

-600

-500

-400

-300

-200

-100

0

100

200

300

400

500

600

700

800

900

0 100 200 300 400 500 600 700 800 900 1000 1100 1200

MEGAWATTS

(LE

AD

) M

EG

AV

AR

S (

LA

15psig 30psig45psig

52psig60psig

(SNC27067)

Name: ________________________________ ILT 10 SRO NRC EXAM PRE DRAFTSUBMITTAL

Form: 0 Version: 0

1. 206000A2.11 001

Unit 2 has experienced a non-isolable RCIC pipe break in the Main Steam Tunnel (Chase)resulting in the following conditions:

o MSIVs CLOSED o RWL -25 inches and lowering o HPCI ISOL TIMER INITIATED, 601-105, ILLUMINATED


The HIGHEST listed RWL value that HPCI will receive an automatic start signal is __________ .

To maintain HPCI for RWL control, the Shift Supervisor will direct bypassing the HPCI High Area Temperature Isolation using __________ .

-36 inches; 31EO-EOP-100-2, Miscellaneous Emergency Overrides

-36 inches; 31EO-EOP-110-2, Alternate RPV Water Level Control

-48 inches; 31EO-EOP-100-2, Miscellaneous Emergency Overrides

-48 inches; 31EO-EOP-110-2, Alternate RPV Water Level Control

A.

B.

C.

D.

Wednesday, April 13, 2016 1:48:51 PM 1

ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTAL

Description:

The HPCI initiation signal for RWL is -35 inches. The Tech Spec allowable value is -47 inches.

On the RC EOP flowchart, if RWL cannot be restored and maintained above -155 in., then entryto CP-1, Alternate Level Control, is REQUIRED.

On the CP-1 flowchart, there is an override that states: IF RWL can be restored and maintainedabove -155 in., THEN perform RC/L (non-ATWS).

With HPCI injecting and in either case above (CP-1 or RC/L) to maintain RWL, level controlis to be directed from the TABLE L-1 Preferred Injection Systems which allows defeating ofhigh area temoeratures IAW 31EO-EOP-100-2, Miscellaneous Emergency Overrides.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of continuous re-check statements on the EOPflowcharts. The continuous re-check statements will determine which EOP flowcharts areentered, and what actions are taken, for RWL control. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by evaluating the applicants' ability to determine theHPCI system response to Low RWL, based on the given conditions. Also, the SRO has todetermine which EOP flowchart is REQUIRED to be used for RWL control.

The "A" is CORRECT.

The "B" distractor is plausible because the first part is correct. In the second part,31EO-EOP-110-2, Alternate RPV Water Level Control, is correct when necessary to operateHPCI IAW 31RS-E41-001-2, HPCI Operation from Outside Control Room.

The "C" distractor is plausible because this would be correct for an initiation setpoint at the TechSpec allowable value. The second part is correct.

The "D" distractor is plausible because this would be correct for an initiation setpoint at the TechSpec allowable value. second part is correct. In the second part, 31EO-EOP-110-2, AlternateRPV Water Level Control, is correct when necessary to operate HPCI IAW 31RS-E41-001-2,HPCI Operation from Outside Control Room.


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALReferences:NONE

K/A:

206000 High Pressure Coolant Injection System

A2. Ability to (a) predict the impacts of the following on the HIGH PRESSURECOOLANT INJECTION SYSTEM ; and (b) based on those predictions, use procedures tocorrect, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.11 Low reactor water level: BWR-2,3,4 . . . . . . . . . . . . 4.1 4.2

SRO only because of link to 10CFR55.43(b)(5): Assessment of facility conditions andselection of appropriate procedures during normal, abnormal, and emergency situations.


E41-HPCI-LP-00501, Ver. 6.0 / EO 005.005.A.01EOP-CP1-LP-20309, Contingency Procedures (CP-1), Ver. 3.0 / EO 201.083.A.04


SRO ONLY Guideline31EO-EOP-010-2, RPV Control (Non-ATWS), Ver. 1031EO-EOP-015-2, Alternate Level Control, Steam Cooling & Emergency RPV Depress, Ver. 10


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTAL2. 212000A2.05 001

Unit 1 is in Mode 2 making preparations to enter Mode 1.

At 08:00, I & C reports the following instruments have failed high and will not provide a tripsignal:

o 1B21-N080A, Narrow Range RWLo 1B21-N080C, Narrow Range RWL

At 08:30, the Channel associated 1B21-N080A has been placed in the TRIPPED condition.


At 08:01, if RWL decreases to zero (0) inches, a FULL Reactor Scram __________ automatically occur.

At 09:01, IAW TS 3.0.4, WITHOUT performing a Risk Evaluation, Tech Specs __________ALLOW the Reactor Mode switch to be placed in the RUN position.

REFERENCE PROVIDED

will;does

will;does NOT

will NOT;does

will NOT;does NOT

A.

B.

C.

D.


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALDescription:

The RPS is comprised of two independent trip systems (A & B) with two logic channels in eachtrip system (A1 & A2, B1 & B2). The outputs of the logic channels in a trip system arecombined in a one-out-of-two logic so that either channel can trip the associated trip system. Thetripping of both trip systems will produce a reactor scram. This arrangement is referred to as aone-out-of-two taken twice logic.

To scram the reactor on low RWL, channels in one of the following combinations must be eitheroperable or maintained in the tripped condition: (A1 or A2) AND (B1 or B2)

With the instrumention in channels A1 (1B21-N080A) and A2(1B21-N080C) failed high andunable to provide a trip signal on low RWL, trip capability is NOT maintained.

TS LCO 3.3.1.1, RPS Instrumentation

Condition A. One or more required channels inoperable.

Place channel in trip, OR place associated trip system in trip, within 12 hours

Condition B. One or more Functions with one or more required channels inoperable in both tripsystems.

Place channel in one trip system in trip, OR place one trip system in trip, within 6 hours

Condition C. One of more Functions with RPS trip capability not maintained.

Restore RPS trip capability within 1 hour

If RPS trip capability is not resored within 1 hours, Condition G is entered.

Be in MODE 3 within 12 hours

3.0 Limiting Condition for Operation Applicability

LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified Conditionin the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continuedoperation in the MODE or other specified condition in theApplicability for an unlimited period of time,

b. After performance of a risk assessment addressing inoperablesystems and components, consideration of the results,determination of the acceptability of entering the MODE or otherspecified condition in the Applicability, and establishment of riskmanagement actions, if appropriate; exceptions to this


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALspecification are stated in the individual Specifications, or

c. When an allowance is stated in the individual value, parameter, orother Specification.

This Specification shall not prevent changes in MODES or other specifiedconditions in the Applicability that are required

SRO JUSTIFICATION:

The SRO must have detailed knowledge of the application of generic LCO requirement 3.0.4. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine how a nuclearboiler instrumentation failure will effect the RPS trip capability; and, use Tech Specs to control consequences for the instrumentation failure.

The "A" distractor is plausible because ECCS logic is arranged into two independent tripsystems (A & B) with two logic channels in each trip system (A1 & A2, B1 & B2). However,this arrangement is NOT a one-out-of-two taken twice logic. To maintain initiation capability,channels in one of the following combinations must be either operable or maintained in thetripped condition: (A1 or B1) & (A2 or B2). If the A1 and A2 channels were failed high andunable to provide an initiaiton signal, initiation capability would be maintained. The second partwould be correct if trip capability were maintained. IAW Condition A, the inoperable channel isrequired to be placed in trip within 12 hours. Operation in this condition is allowed for anunlimited period of time. IAW LCO 3.0.4.a, a mode change to MODE 1 would be allowed.

The "B" distractor is plausible because ECCS logic is arranged into two independent trip systems(A & B) with two logic channels in each trip system (A1 & A2, B1 & B2). However, thisarrangement is NOT a one-out-of-two taken twice logic. To maintain initiation capability,channels in one of the following combinations must be either operable or maintained in thetripped condition: (A1 or B1) & (A2 or B2). If the A1 and A2 channels were failed high andunable to provide an initiaiton signal, initiation capability would be maintained. The second partis correct.

The "C" distractor is plausible because the first part is correct. The second part would be correctif trip capability were maintained. IAW Condition A, the inoperable channel is required to beplaced in trip within 12 hours. Operation in this condition is allowed for an unlimited period oftime. IAW LCO 3.0.4.a, a mode change to MODE 1 would be allowed.

"D" is CORRECT.


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALReferences:Unit 1 TS 3.3.1.1 Reactor Protection System (RPS) Instrumentation, page 3.3-1 thru 3.3-3, Table 3.3.1.1-1 (page 2 of 3 WITHOUT VALUES) ONLY, & T3.3.1.1-1 Item 4LFD-1-RPS-09.

K/A:

212000 Reactor Protection System

A2. Ability to (a) predict the impacts of the following on the REACTOR PROTECTIONSYSTEM ; and (b) based on those predictions, use procedures to correct, control, ormitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.05 Nuclear boiler instrument system failure . . . . . . . . . . . . . . . . . . . . . 3.4 3.7

SRO only because of link to 10CFR55.43(b)(2): Facility operating limitations in thetechnical specifications and their bases.


B11-RXINS-LP-04404, Ver. 7.2, EO 300.010.A.23


SRO ONLY GuidelineUnit 1 Tech Specs, LCO 3.3.1.1, RPS Instrumentation



Unit 2 is conducting a reactor startup.

o Control rod 26-27 is inop and fully inserted (TS required actions complete) o Control rod 18-19 has just been withdrawn to position 48

A coupling check on control rod 18-19 results in the following:

o 18-19 four rod display position BLANK o 18-19 Full-Out light EXTINGUISHED o ROD OVERTRAVEL, 603-248, ILLUMINATED


With NO operator action, one (1) minute later, control rod 18-19 four rod display positionindication will __________ .

IAW TS 3.1.3, Control Rod Operability, if control rod 18-19 can NOT be re-coupled, the EARLIEST REQUIRED action is to __________ .

remain BLANK; fully insert and disarm the control rod

remain BLANK; place the reactor in hot shutdown

have returned to position 48; fully insert and disarm the control rod

have returned to position 48; place the reactor in hot shutdown

A.

B.

C.

D.



Description:

IAW 34GO-OPS-065-0, Control Rod Movement:

If a control rod is moved either intentionally OR unintentionally to position 48, a couplingcheck will be performed as required by TS.

4.7 Control Rod Coupling

Perform a coupling integrity check of a control rod, by performance of the following steps:

1) Perform a withdraw signal, by PLACING the Rod Movement Control switch to thewithdraw position 2) Confirm the withdraw logic actuates, by ILLUMINATION of the red withdraw light. 3) Confirm the rod attempts to move, by a momentary loss of the "48" position indication. 4) Confirm the rod does NOT move beyond position 48. 5) Confirm annunciator ROD OVERTRAVEL (603-248) is NOT received

If the rod drive piston moves to an overtravel position, an alarm is sounded in the control room. The overtravel alarm provides a means to verify that the drive-to-rod coupling is intact. With thecoupling in its normal condition, the drive cannot be physically withdrawn to the overtravelposition. Coupling integrity can be checked by attempting to withdraw the drive to theovertravel position and observe that no overtravel alarm occurs.

ROD OVERTRAVEL, 603-248, contains the following note:

IAW TS 3.1.3, Control Rod Operability:

Condition C. One or more control rods inoperable for reasons other than Condition A or B.

Fully insert inoperable control rod within 3 hours AND disarm within 4 hours.



Condition E. Required Action and associated Completion Time of Condition A, C, or D not met.

Be in MODE 3 in 12 hours.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of Tech Spec required actions for an INOP (uncoupled)control rod. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to predict the response of RPISto a failed control rod coupling check, and use a procedure/TS to mitigate the condition.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sincethis would be correct if the TS required action completion times for Condition C are not met.

The "C" distractor is plausible because only a momentary loss of the 48 position indicationoccurs when a coupling check is performed successfully. The position indication will go blank,then return to 48.

The "D" distractor is plausible because only a momentary loss of the 48 position indicationoccurs when a coupling check is performed successfully. The position indication will go blank,then return to 48. The second part is plausible since this would be correct if the TS requiredaction completion times for Condition C are not met.



K/A:

214000 Rod Position Information System

A2. Ability to (a) predict the impacts of the following on the ROD POSITIONINFORMATION SYSTEM ; and (b) based on those predictions, use procedures to correct,control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.03 Overtravel/in-out . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6 3.9



C11-RMCS-LP-05401, Ver. 6.0, EO 300.006.C.02


SRO ONLY Guideline

34AR-603-248-2, Rod Overtravel, Ver. 2.134GO-OPS-065-0, Control Rod Movement, Ver. 12.7Unit 2 TS 3.1.3, Control Rod OperabilityBank question from HLT Database which was used on 2012 HLT-7 Audit Exam


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTAL4. 233000G2.2.22 001

Unit 2 is in MODE 5.

o The Reactor has been shutdown for 5 dayso Fuel movement from the RPV to the Spent Fuel Pool is in progress

Subsequently, all Unit 2 Fuel Pool Cooling is lost and can NOT be restored.

o Spent Fuel Pool temperature has increased to 150°F


IAW TS Bases 3.7.8, Spent Fuel Pool Water Level, the LOWEST listed Spent Fuel Poolwater level that will still meet the LCO 3.7.8 requirements WITHOUT entering a RequiredAction Statement (RAS) is __________ .

IAW TS Bases 3.7.8, Spent Fuel Pool Water Level, the MINIMUM water level in the Spent Fuel Pool is REQUIRED to __________ .

21.1 feet; remove decay heat from irradiated fuel assemblies

21.1 feet; provide for absorption and transport delay of fission product gases

22.1 feet; remove decay heat from irradiated fuel assemblies

22.1 feet; provide for absorption and transport delay of fission product gases

A.

B.

C.

D.



Description:

Joe, this was a Pre-submittal SRO question. Changes were incorporated based on yourES-401-9 comments.

TS LCO 3.7.8, Spent Fuel Storage Pool Water Level: The spent fuel storage pool water levelshall be > 21 feet over the top of irradiated fuel assemblies seated in the spent fuel storage poolracks. APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuelstorage pool.

TS BASES LCO 3.7.8: The water level above the irradiated fuel assemblies is an explicitassumption of the fuel handling accident. The water level in the spent fuel storage pool providesfor absorption of water soluble fission product gases and transport delays of soluble andinsoluble gases that must pass through the water before being released to the secondarycontainment atmosphere. This absorption and transport delay reduces the potential radioactivityof the release during a fuel handling accident.

TS 3.9.7 Residual Heat Removal (RHR) - High Water Level

LCO 3.9.7 One RHR shutdown cooling subsystem shall be OPERABLE and inoperation. APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV)and the water level > 22 ft 1/8 inches above the top of the RPV flange.

TS 3.9.8 Residual Heat Removal (RHR) - Low Water Level

LCO 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and one RHRshutdown cooling subsystem shall be in operation. APPLICABILITY: MODE 5 with irradiatedfuel in the reactor pressure vessel (RPV) and the water level < 22 ft 1/8 inches above the top ofthe RPV flange.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of Tech Spec Bases for Spent Fuel Pool water level. This is above the RO knowledge level.



K/A JUSTIFICATION:

This question satisfies the K/A statement by evaluating the applicants' knowledge of limitingconditions for operation associated with the abnormal procedure and licensing document, duringa loss of fuel pool cooling.

The "A" distractor is plausible because the first part is correct. The second part is plausiblebecause IAW TS Bases, RWL is required to be above the top of the active fuel to provide corecooling capability. With fuel in the reactor vessel during periods when the reactor is shut down,consideration must be given to water level requirements due to the effect of decay heat. If thewater level should drop below the top of the active irradiated fuel during this period, the abilityto remove decay heat is reduced.

"B" is CORRECT.

The "C" distractor is plausible since RHR operability requirements change above/below 22 feet1/8 inch (TS 3.9.7 & 3.9.8). The second part is plausible because IAW TS Bases, RWL isrequired to be above the top of the active fuel to provide core cooling capability. With fuel inthe reactor vessel during periods when the reactor is shut down, consideration must be given towater level requirements due to the effect of decay heat. If the water level should drop below thetop of the active irradiated fuel during this period, the ability to remove decay heat is reduced.

The "D" distractor is plausible since RHR operability requirements change above/below 22 feet1/8 inch (TS 3.9.7 & 3.9.8). The second part is correct.



K/A:

233000 Fuel Pool Cooling and Clean-up

G2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2 / 45.2) . . . . . . . . . . . . 4.0 4.7



G41-FPC-LP-04501, Fuel Pool Cooling, Ver. 5.1, EO 045.025.A.01


SRO ONLY Guideline Unit 2 Tech Spec LCO 3.7.8, Spent Fuel Pool Water Level, Amendment 210 Unit 2 Tech Spec Bases, LCO 3.7.8, Spent Fuel Pool Water Level, Revision 74.0 Unit 2 Tech Spec LCO TS 3.9.7 RHR - High Water Level, Amendment 210 Unit 2 Tech Spec LCO TS 3.9.8 RHR - Low Water Level, Amendment 210



Unit 2 is operating at 90% RTP when SRV 2B21-F013B inadvertently OPENS.


o SAFETY/BLOWDOWN VLV PILOT/SEAT LEAKING, 603-122, ILLUMINATED o SAFETY BLOWDOWN PRESSURE HIGH, 602-311, ILLUMINATED o Torus water temperature INCREASES to 85°F

Five (5) seconds after opening and with NO operator action, the SRV goes CLOSED:

Based on the above conditions and IAW 34AB-B21-003-2, Failure of Safety/Relief Valves,

Tech Spec surveillance requirement __________ is REQUIRED to be performed.

SR 3.6.2.1, verify Torus average temperature is within the applicable limits

SR 3.4.4.1, verify RCS unidentified & total leakage and unidentified leakage

SR 3.6.1.6.2, verify the LLS system actuates on an actual or simulated automatic initiation signal

SR 3.6.1.8.2, perform a functional test of each required Torus-to-Drywell vacuum breakers

A.

B.

C.

D.




IAW 34AB-B21-003-2, Failure of Safety / Relief Valves,

1. If SRVs are OPEN electrically due to an inadvertent initiation, then cycle the controlswitch several times.2. After cycling the switch, attempt to reset LLS / ADS.3. If SRVs are OPEN mechanically, then cycle the control switch several times.4. If the SRV did not close, remove all the fuses to the SRV.5. Confirm the SRV is CLOSED o Confirm Rx power & generator output return to normal6. If the SRV is stuck OPEN, then SCRAM the reactor. o If RPV Pressure decreases to 500 psig, then close the outboard MSIVs7. If Torus water temperature exceeds 95°F, enter 34AB-T23-003-2.8. Notify the NRC.9. Refer to Tech Specs.10. Notify the Shift Manager to determine if an EAL has been exceeded.11. Write a Condition Report and comply with IEB 80-25.12. Demonstrate operability of the Suppression Chamber Drywell Vacuum Breakers WITHIN 12 hours after ANY discharge of steam to the Suppression Chamber from theSafety/Relief Valves per 34SV-T48-002-2, Suppression Chamber To Drywell VacuumBreaker System Operability.13. Notify Chemistry

TS SR 3.4.4.1, Verify RCS unidentified and total LEAKAGE and unidentified LEAKAGEincrease are within limits IAW the Surveillance Frequency Control Program.

SR 3.6.1.6.2, verify the LLS system actuates on an automatic initiation signal IAW theSurveillance Frequency Control Program (every 24 months).

SR 3.6.2.1, verify Torus average temperature is within limits IAW the Surveillance FrequencyControl Program (every 24 hours) AND every 5 minutes when performing testing that adds heatto the Torus.

Testing is NOT being performed (inadvertent SRV lift), therefore SR 3.6.2.1 is not requiredto be performed every 5 minutes.

34AB-T23-003-2, Torus Temperature Above 95°F, together with 34SV-B21-004-2,34SV-E41-002-2, 34SV-E41-005-2, 34SV E51 002 2, 34SV-E51-004-2, 34IT-E51-003-2, and34IT-E41-003-2, meets the requirements of SR 3.6.2.1.1 during testing that adds heat to thesuppression pool.

Torus Temp doesn't increase above 85°F, therefore 34AB-T23-003-2 is not required to beentered.


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALSRO JUSTIFICATION:

The SRO must have detailed knowledge of 34AB-B21-003-2 to determine which TS SR isrequired to be performed when a SRV lifts and discharges steam to the Torus. This is above theRO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant detailed procedure actions inresponse to an inadvertent SRV actuation.

The "A" distractor is plausible because SR 3.6.2.1 is required to be performed during TESTINGthat adds heat to the Torus.

The "B" distractor is plausible, IAW TS, when Primary system leakage is occuring TS requires aquantitative determination of the total leakage rate to demonstrate if TS LCO requirements arebeing met.

The "C" distractor is plausible because, IAW TS, when certain system are declared INOP, asurveillance is REQUIRED to be performed to demonstrate the operability of other componentsin the system (examples: SR 3.8.1.1, SR 3.8.1.2)

"D" is CORRECT.



K/A:

239002 Relief/Safety Valves

G2.4.31 Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3) . . . . . . . . . . . . . . . . . . 4.2 4.1



B21-LP-01401, Ver. 10.0, EO 200.009.A.02


SRO ONLY Guideline

34AB-B21-003-2, Failure of Safety / Relief Valves, Ver. 8.034SO-B21-001-2, ADS and LLS Systems, Ver. 13.1534IT-B21-002-2, Relief Valve Testing, Ver. 0.5



Unit 1 is in Mode 5 with all control rods fully inserted and Shutdown Cooling In Service.

Maintenance replaced the below relays on RPS MG Set 1A.

57SV-C71-005-1, RPS Power Monitors FT&C, has just been completed.

At 12:00 on 7/1, the “AS LEFT” trip setpoints for two Equipment Protection Circuit Breakersare as follows:

1C71-P003A 1C71-P003C Breaker 52-3A Breaker 52-3C

Overvoltage 131.5 volts 132.4 volts Undervoltage 107.6 volts 108.1 volts Underfrequency 56.8 Hz 57.7 Hz

Based on the above conditions and IAW U1 TS 3.3.8.2, Reactor Protection System (RPS)Electric Power Monitoring,

The EARLIEST listed TS Required Action is to __________ .

REFERENCE PROVIDED

remove RPS MG Set 1A from service no later than 12:00 on 7/4

remove RPS MG Set 1A from service no later than 13:00 on 7/1

isolate RHR in SDC no later than 13:00 on 7/1

isolate RHR in SDC no later than 12:00 on 7/1

A.

B.

C.

D.



RPS Electric Power Monitoring System is provided to isolate the RPS bus from the motorgenerator (MG) set or an alternate power supply in the event of overvoltage, undervoltage, orunderfrequency. This system protects the loads connected to the RPS bus against unacceptablevoltage and frequency conditions and forms an important part of the primary success path of theessential safety circuits. Some of the essential equipment powered from the RPS buses includesthe RPS logic, scram solenoids, and various valve isolation logic (e.g., residual heat removalshutdown cooling). RPS electric power monitoring assembly will detect any abnormal high or low voltage or lowfrequency condition in the outputs of the two MG sets or the alternate power supply and willde-energize its respective RPS bus, thereby causing all safety functions normally powered by thisbus to de-energize. Together, a circuit breaker and its sensing logic constitute an electric power monitoringassembly. If the output of the MG set or the alternate power supply exceeds predeterminedlimits of overvoltage, undervoltage, or underfrequency, a trip coil driven by this logic circuitryopens the circuit breaker, which removes the associated power supply from service.

IAW TS 3.3.8.2, - Overvoltage, < 132 V, with time delay set to < 4 seconds Undervoltage, > 108 V, with time delay set to < 4 seconds Underfrequency, > 57 Hz, with time delay set to < 4 seconds

RPS Electric Power Monitoring, Section 3.3.8.2 1. Two RPS electric power monitoring assemblies shall be OPERABLE for each in-service RPS

motor generator set or alternate power supply in:

Modes 1, 2, or 3, and

Modes 4 or 5, with any control rod withdrawn from a core cell containing one or morefuel assemblies, or with both RHR SDC isolation valves open.

Required Action A.1 states: One or both in-service power supplies with one electric powermonitoring assembly inoperable:

Remove associated in-service power supply(s) from service within 72 hours

Since each power supply has two assemblies (breakers) in series, if one assembly for each powersupply remains OPERABLE, 72 hours is an acceptable time to rely on only one assembly whilerepairs are made to the INOPERABLE assembly.

Required Action B.1 states: One or both in-service power supplies with both electric powermonitoring assemblies inoperable:

Remove associated in-service power supply(s) from service within 1 hour.

If BOTH assemblies for any power supply are INOPERABLE, RPS components are in jeopardyof severe damage from voltage or frequency fluctuations. Therefore, the unprotected powersupply must be removed from service. One hour is allowed only to provide sufficient time tosafely transfer to an OPERABLE power supply.


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALWith 52-3A Undervoltage & Underfrequency and 52-3B Overvoltage values out of spec,BOTH RPS power monitoring breakers are inop requiring the 1 hour TS Required Action.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of TS Bases 3.3.8.2, RPS Electric Power Monitoring, indetermining how many components are required to ensure operability and then the requiredaction associated with the RPS Electric Power Monitoring. (1 hour distractor is not RO levelbecause the data in the table must be analyzed and compared against the surveillancerequirements.) This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine the maintenanceeffect (degraded power devices) on the operability (LCO) of the power monitoring breakers,following replacement. The RPS power monitoring breakers are considered UninterruptablePower Supply (A.C./D.C.).

The "A" distractor is plausible if the applicant remembers there are other TS RPS componentsthat require 2 of 3 or 3 of 4 per function and thinks that only 2 of the three power monitoringrelays are required to ensure RPS protection. Since only one relay on 52-3C has exceeded its tripvalue, 52-3C Equipment Protection Circuit Breaker is operable. With one power monitoringbreaker inop (52-3A) the required action is to remove the inservice power supply from servicewithin 72 hours. Also would be correct if only one power monitoring breaker was inoperable.

"B" is correct.

The "C" distractor is plausible if the applicant realizes that both power monitoring breakers areinoperable but does not take into consideration that the inservice power supply must be removedfrom service within the first hour. After an hour has elasped, if the inservice power supply isstill providing power to RPS components, then RHR in SDC must be isolated at 13:00 on 7/1. Also would be correct if the inservice power supply was not removed from service within thefirst hour.

The "D" distractor is plausible if the applicant realizes that both power monitoring breakers areinoperable but does not take into consideration that the inservice power supply must be removedfrom service within the first hour. Required Action D 2.2 would be entered immediately whichwould result in RHR in SDC isolated immediately at 12:00 on 7/1. Also would be correct if theinservice power supply was not removed from service within the first hour.


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALReferences:Unit 1 TS 3.3.8.2, Reactor Protection System (RPS) Electric Power Monitoring,Amendment 210

K/A:

262002 Uninterruptable Power Supply (A.C./D.C.)

G2.2.36 Ability to analyze the effect of maintenance activities, such as degraded powersources, on the status of limiting conditions for operations. (CFR: 41.10 / 43.2 / 45.13) . . . . . . . . . . . . . 3.1 4.2



C71-RPS-LP-01001, Reactor Protection System, Ver. 8.2, EO 300.010.A.14


SRO ONLY Guideline U1 TS 3.3.8.2, Reactor Protection System (RPS) Electric Power Monitoring, Amendment 210 U1 Bases 3.3.8.2, Reactor Protection System (RPS) Electric Power Monitoring, Rev. 1.0

Modified from HLT Database Q#212002G2.2.36-001 which was used on 2009 NRC ExamQ#76.

Original Question

Unit 1 is in Mode 5 with all control rods fully inserted and Shutdown Cooling In Service. BothRPS MG Sets are in service following the completion of 57SV-C71-005-1, RPS Power MonitorsFT&C.

While reviewing the completed surveillance package, the SS notes the following “AS LEFT” tripsetpoints for two Equipment Protection Circuit Breakers:

1C71-P003A 1C71-P003CBreaker 52-3A Breaker 52-3COvervoltage 131.5 volts 131.4 voltsUndervoltage 107.6 volts 108.1 voltsUnderfrequency 56.8 Hz 57.7 Hz



Which ONE of the following identifies the status of the Equipment Protection Circuit Breakersand the required action, if any?

[Reference provided]

A. ONLY one breaker is INOPERABLE; no required Tech Spec actions at this time

B. BOTH breakers are INOPERABLE; no required Tech Spec actions at this time

C. ONLY one breaker is INOPERABLE; place the associated RPS Bus on the alternatepower supply in 72 hours

D. BOTH breakers are INOPERABLE; place the associated RPS Bus on the alternate powersupply in 1 hour



Unit 1 was operating at 100% RTP when an unisolable steam line break occurred in the Main Steam Tunnel (Chase) area.

The following Reactor Building radiation conditions exist:

o 1D11-K609 A-D, Reactor Bldg Pot Cont Area Rad Monitors 21 mR/hro 1D21-K601F, Tip Area Radiation Monitor 1100 mR/hro The next highest area radiation monitor 200 mR/hr


Secondary Containment on __________ will have received an isolation signal.

IAW 31EO-EOP-014-1, SC EOP flowchart, a reactor shutdown is REQUIRED by entering __________ .

Unit 1 ONLY;34GO-OPS-014-1, Fast Reactor Shutdown

Unit 1 ONLY;31EO-EOP-010-1, RPV Control (Non-ATWS)

Unit 1 AND Unit 2;34GO-OPS-014-1, Fast Reactor Shutdown

Unit 1 AND Unit 2;31EO-EOP-010-1, RPV Control (Non-ATWS)

A.

B.

C.

D.



Description:

IAW 34AB-T22-003-1, Secondary Containment Control:

IF secondary containment HVAC exhaust radiation level exceeds the secondary containmentHVAC isolation setpoint:

o Confirm OR manually initiate isolation of secondary containment HVACo Confirm initiation of OR manually initiate SBGT

High radiation in the Refueling Floor or Reactor Building Exhaust (either unit) causesvarious Group 2 (Primary Containment) isolation valves to close along with isolation of theUnit 1 and Unit 2 Secondary Containment.

Any ONE of the following will generate an isolation signal for Unit 1 and Unit 2 SecondaryContainment:

o Unit 1 or 2 Reactor Bldg vent exhaust high radiation:

Unit 1: 18 mrem/hr on 1D11-K609 A-D, ORUnit 2: 18 mrem/hr on 2D11-K609 A-D

o Unit 1 or 2 Refueling Floor vent exhaust high radiation:

Unit 1: 18 mrem/hr on 1D11-K611-A-D, ORUnit 2: 18 mrem/hr on 2D11-K611 A-D 6.9 mrem/hr on 2D11-K634 A-D 5.7 mrem/hr on 2D11-K635 A-D

o High drywell pressure (Either Unit): 1.85 psig

o Low reactor water level (Either Unit): -35 inches

IAW SC EOP Flowchart:

With a primary system (Main Steam) discharging into the secondary containment, BEFOREarea radiation level reaches Max Safe, enter the RC EOP flowchart and scram the reactor. IfMax Safe is reached prior to the reactor scram, it is still required to scram the reactor.

The Tip Area, North CRD HCU & Tip Probe Drives Area radiation monitors are all in ONEarea (130' elevation Rx Bldg North Area). The Max Safe radiation level for thesemonitors is 1000 mr/hr.



If radiation levels in more than one area exceed Max Safe levels and a primary system isNOT discharging into the secondary containment, a reactor shutdown is REQUIRED per34GO-OPS-013-1 or 34GO-OPS-014-1.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of actions required by the SC EOP flowchart, based ongiven conditions, including which procedure is required to be entered to shutdown the reactor. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to predict the response ofSecondary Containment to a high radiation condition, and select the correct procedure formitigating the condition.

The "A" distractor is plausible because high radiation in the Reactor Building Exhaust causesvarious Group 2 (PCIVs) isolation valves to close, but only on the effect unit. For the givenconditions, only Group 2 isolation valves on Unit 1 will close. Direction is given on the SC EOPflowchart to perform a Fast Rx Shutdown if Max Safe radiation levels are exceeded in more thanone area.

The "B" distractor is plausible because high radiation in the Reactor Building Exhaust causesvarious Group 2 (PCIVs) isolation valves to close, but only on the effect unit. For the givenconditions, only Group 2 isolation valves on Unit 1 will close. The second part is correct.

The "C" distractor is plausible the first part is correct. The second part is plausible becausedirection is given on the SC EOP flowchart to perform a Fast Rx Shutdown if Max Saferadiation levels are exceeded in more than one area.

"D" is CORRECT.

References:NONE



K/A:

290001 Secondary Containment

A2. Ability to (a) predict the impacts of the following on the SECONDARYCONTAINMENT ; and (b) based on those predictions, use procedures to correct, control,or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.04 High airborne radiation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4 3.7



T41-SC HVAC-LP-01303, Ver. 5.0, EO 037.011.A.10


SRO ONLY Guideline31EO-EOP-014-1, Secondary Containment Control / Radioactivity Release Control, Ver. 1234AB-T22-003-1, Secondary Containment Control, Ver. 5.14

Modified from 2009 HLT-4 NRC Exam Q#82

Original Question

Unit 2 was operating at 100% power when an unisolable steam line break occurred in the MainSteam Line Tunnel area.

The following Reactor Building radiation conditions currently exist:

o 2D11-K609A, RB POT CONTAM AREA RADIATION ..................... 22 mR/hro 2D11-K609B, RB POT CONTAM AREA RADIATION ..................... 20 mR/hro 2D11-K609C, RB POT CONTAM AREA RADIATION ..................... 20 mR/hro 2D11-K609D, RB POT CONTAM AREA RADIATION ..................... 19 mR/hro 2D21-K601F, 130' ELEVATION AREA (NORTHWEST) Tip Area .... 1100 mR/hr



Which ONE of the following choices completes the following statements?

The Unit 2 Standby Gas Treatment system __________ received an auto start signal.

IAW 31EO-EOP-014-2, "Secondary Containment Flowchart," the SS is required to enter__________.

A. has; 34GO-OPS-013-2, Plant Shutdown

B. has; 31EO-EOP-010-2, "RC RPV Control" point A

C. has NOT; 34GO-OPS-013-2, Plant Shutdown

D. has NOT; 31EO-EOP-010-2, "RC RPV Control" point A


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTAL8. 295001AA2.02 001

Unit 2 is operating at 90% RTP when Reactor Engineering notifies the Main Control Room ofthe following:

o ALL OPRM Period Based algorithms will NOT provide an OPRM Upscale Trip o ALL OPRM Amplitude Based and Growth Rate algorithms are operating normally

Subsequently, the following occurs;

o Recirc Pump 2A trips o Reactor power stabilizes at 64% RTP o Core flow stabilizes at 48%

Based on the above conditions and IAW 34AB-B31-001-2, Reactor Recirculation Pump(s) Trip,Or Recirc Loops Flow Mismatch, Or ASD Power Cell Failure,

The APRM __________ setpoint is REQUIRED to be adjusted for Single Loop operationIAW LCO 3.4.1.

The Power-To-Flow Map in __________ will be used to plot the Unit 2 operating point.

Simulated Thermal Power - High; 34GO-OPS-005-2, Power Changes

Simulated Thermal Power - High; 34AB-C51-001-2, Entry Into the Region Of Potential Instabilities, (RPI) or Reactor Operations With Inoperable OPRMs

Neutron Flux - High; 34GO-OPS-005-2, Power Changes

Neutron Flux - High; 34AB-C51-001-2, Entry Into the Region Of Potential Instabilities, (RPI) or Reactor Operations With Inoperable OPRMs

A.

B.

C.

D.



Description:

IAW TS Bases 3.3.1.1 Function 2f, there are three algorithms for detecting thermal-hydraulicinstability related neutron flux oscillations: the period based detection algorithm, the amplitudebased algorithm, and the growth rate algorithm. All three are implemented in the OPRMUpscale Function, but the safety analysis takes credit only for the period based detectionalgorithm. The remaining algorithms provide defense in depth and additional protection againstunanticipated oscillations. OPRM Upscale Function OPERABILITY for TechnicalSpecifications purposes is based only on the period based detection algorithm.

34AB-B31-001-2, Reactor Recirculation Pump(s) Trip, Or Recirc Loops Flow Mismatch, OrASD Power Cell Failure contains steps to:

enter 34AB-C51-001-2, Entry into RPI or Reactor Operation with Inoperable OPRM System,if OPRM system inoperable. Where 34AB-C51-001-2 Attachment 1 must be used whenOPRMs are INOP. With reactor power at 64% RTP & core flow at 48% the RPI has beenentered.

confirm proper operation on the Power-To-Flow Map in 34GO-OPS-005-2, Power Changes,Attachment 1. With reactor power at 64% RTP & core flow at 48%, the OPRM ArmedRegion will be entered (>25% RTP with core flow <60%). However the RPI has not beenentered per 34GO-OPS-005-2, Attachment 1.

TS 3.4.1 Recirculation Loops Operating

LCO 3.4.1 Two recirculation loops with matched flows shall be in operation,OR

One recirculation loop shall be in operation with the following limitsapplied when the associated LCO is applicable:

d. LCO 3.3.1.1, “Reactor Protection System (RPS) Instrumentation,”Function 2.b (Average Power Range Monitor Simulated ThermalPower - High), Allowable Value of Table 3.3.1.1-1 is reset forsingle loop operation.



The SRO must have detailed knowledge of procedures to determine which procedure has thecorrect Power to Flow Map for the given conditions, This level of procedure knowledge isabove the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine which procedureto use to if the OPRM (Neutron monitoring) ARMED Region has been entered on the partialloss of core flow circulation (SRO part) and the TS Actions for single loop (RO part).

The "A" distractor is plausible since the first part is correct. The second part is correct if theOPRM system was not INOP.

"B" is correct.

The "C" distractor is plausible because the Average Neutron Flux is used to create the SimulatedThermal Power signal. The second part is correct if the OPRM system was not INOP

The "D " distractor is plausible because the Average Neutron Flux is used to create theSimulated Thermal Power signal. The second part is plausible since it is correct.



K/A:

APE: 295001 Partial or Complete Loss of Forced Core Flow Circulation

AA2. Ability to determine and/or interpret the following as they apply to PARTIAL ORCOMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : (CFR: 41.10 / 43.5 / 45.13)

AA2.02 Neutron monitoring . . . . . . . . . . . . . . . . . 3.1 3.2




C51-PRNM-LP-01203, Power Range Neutron Monitoring System, Ver. 9.0, EO 300.006.C.02 LT-LP-30005, Technical Specifications, Ver. 10.3, EO 300.006.A.27


SRO ONLY Guideline 34AB-B31-001-2, Reactor Recirculation Pump(s) Trip, Or Recirc Loops Flow Mismatch, OrASD Power Cell Failure, Ver. 10.10 34GO-OPS-005-2, Power Changes, Attachment 1, Power to Flow Map, Ver. 28.6 U2 TS Bases 3.3.1.1, RPS Instrumentation, Function 2F, Rev. 16.0




o Supply breaker to Essential Cabinet 2A trips OPEN o Subsequently, the Instrument Buses are cross-tied


IAW TS Bases 3.8.7, Instrument Bus 2A __________ OPERABLE.

After power is restored to Instrument Bus 2A, 34AB-R25-002-2, Loss Of Instrument Buses,directs the operator to transfer RFPT __________ to the M/A Station.

is; 2A

is; 2B

is NOT; 2A

is NOT; 2B

A.

B.

C.

D.



When the normal supply breaker (from 600 VAC 2C) to Essential Cabinet 2A trips OPEN, theconditions are met to enter 34AB-R24-001-2, Loss of Essential AC Distribution Buses. WhenEssential Cabinet 2A is deenergized, Instrument Bus 2A is also deenergized. The conditions toenter 34AB-R25-002-2, Loss of Instrument Buses, are also met.

Also, 34AB-R24-001-2 contains explicit direction to enter 34AB-R25-002-2:

Step 4.1) Enter 34AB-R25-002-2, Loss of Instrument Buses and perform sections 4.1 and4.3 concurrently with this procedure.

IAW 34AB-R25-002-2, Loss of Instrument Buses:

IF fuses are NOT BLOWN, AND Essential Cabinet 2A cannot be energized, Then ENERGIZE Instrument Bus 2A, from Instrument Bus 2B.

TS Bases 3.8.7:

The AC electrical power distribution subsystem requires the associated buses and electricalcircuits to be energized to their proper voltages.

Tie breakers between redundant safety related AC and DC power distribution subsystems, ifthey exist, must be open. If any tie breakers are closed, the electrical power distributionsubsystem which is not being powered from its normal source (i.e., it is being poweredfrom its redundant electrical power distribution subsystem) is considered inoperable.

Therefore, when Instrument Bus 2A is cross-tied to Instrument Bus 2B, it is considered INOPeven though it is energized.

34AB-R25-002-2, Loss Of Instrument Buses,4.1 LOSS OF INST BUS 2A, 2R25 S064

4. SUBSEQUENT OPERATOR ACTIONSi. WHEN power is restored to Instrument Bus 2A, perform the following:

(5) Recover RFP 'A' operating from the TMR to M/A Station per the "Recovery From Loss of Normal Speed Control Signal", subsection, in34SO-N21-007-2.



The SRO must have detailed knowledge of the Bases for TS LCO 3.8.7 to determine thatInstrument Bus 2A is INOP, even though it is energized, when the tie breakers are closed. Detailed knowledge of procedural step to select correct procedure to recover RFPT 2A. This isabove the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine what TS actionsare required, due to the loss of an AC distribution cabinet and the resultant electrical distributionlineup.

The "A" distractor is plausible because the distribution bus is energized (cross-tied to the otherInstrument Bus). IAW LCO 3.8.7 Bases, if any tie breakers are closed, the electrical powerdistribution subsystem which is not being powered from its normal source is consideredinoperable. The second part is correct.

The "B" distractor is plausible because the distribution bus is energized (cross-tied to the otherInstrument Bus). IAW LCO 3.8.7 Bases, if any tie breakers are closed, the electrical powerdistribution subsystem which is not being powered from its normal source is consideredinoperable. The second part is plausible because the RFPT 2B has several components poweredby Instrument Bus A (speed indicator 2N32-R603, lube oil temp controller 2N34-N306, highpress stop valve temp indicator, minimun flow valve 2N21-F117B).

"C" is CORRECT.

The "D" distractor is plausible because the first part is correct. The second part is plausiblebecause the RFPT 2B has several components powered by Instrument Bus A (speed indicator2N32-R603, lube oil temp controller 2N34-N306, high press stop valve temp indicator, minimunflow valve 2N21-F117B).



K/A:

APE: 295003 Partial or Complete Loss of A.C. Power

AA2. Ability to determine and/or interpret the following as they apply to PARTIAL ORCOMPLETE LOSS OF A.C. POWER : (CFR: 41.10 / 43.5 / 45.13)

AA2.04 System lineups . . . . . . . . . . . . . . . . . . . . 3.5 3.7



R23-ELECT-LP-02703, 600/400/208 VAC Electrical, Ver. 5.1, EO 300.006.A.29


SRO ONLY Guideline 34AB-R24-001-2, Loss of Essential AC Distribution Buses, Ver. 1.1034AB-R25-002-2, Loss of Instrument Buses, Ver. 8.0Tech Spec Bases, 3.8.7 Distribution Systems - Operating




o 2R22-S016, 125/250VDC Switchgear 2A, de-energizes and can NOT be restored


IAW 34AB-R22-001-2, Loss Of DC Buses, a reactor shutdown per __________ isREQUIRED.

To trip the RFPTs, the trip pushbuttons are REQUIRED to be depressed __________ .

34GO-OPS-013-2, Normal Plant Shutdown;locally at the RFP front standard

34GO-OPS-013-2, Normal Plant Shutdown;on panel 2H11-P650

34GO-OPS-014-2, Fast Reactor Shutdown;locally at the RFP front standard

34GO-OPS-014-2, Fast Reactor Shutdown;on panel 2H11-P650

A.

B.

C.

D.

Description:



IAW 34AB-R22-001-2, Loss of DC Buses:

IF power CANNOT be restored to Bus 2A AND a Reactor Scram has NOT occurred, initiate 34GO-OPS-014-2, Fast Reactor Shutdown

RFPT 2A & 2B will not trip manually or auto due to loss of power to the trip circuits

TRIP both RFPTs by depressing both trip pushbuttons for each RFPT, locally at RFP frontstandard

IAW 34GO-OPS-014-2, when Recirc Pumps are at minimum speed, enter the Scram Procedureand scram the reactor. The immediate operator actions for the Scram Procedure is to performRC-1, RC-2 & RC-3. A subsequent action is to perform Main Turbine Trip Actions (TC-1).

TC-1 actions include manually tripping the Main Turbine and confirming or tripping open theMain Generator output breakers. With 2R22-S016 de-energized, when the Main GeneratorOutput breakers are open, 4160 VAC station service buses will NOT auto fast transfer. 4160VAC C & D will de-energize, resulting in a loss of all Condensate pumps & Condensate Boosterpumps and RFP main oil pumps. The RFPTs are then required to be tripped.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of procedure content, with respect to procedureselection. Specifically, the requirement in 34AB-R22-001-2 to shutdown the reactor per34GO-OPS-014-2 if power can NOT be restored to the electrical bus. This is above the ROknowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to demonstrate the ability toexecute required actions in 34AB-R22-001-2 for shutting down the reactor and tripping theRFPTs during a partial loss of DC power.

The "A" distractor is plausible because a reactor shutdown per 34GO-OPS-013-2 is the normalmethod of the performing a shutdown. For some DC cabinets listed in 34AB-R22-001-2, ifpower is lost and can NOT be restored, a reactor shutdown is required. However, for thesecabinets, 34GO-OPS-014-2 is NOT explicitly required. The second part is correct.



The "B" distractor is plausible because a reactor shutdown per 34GO-OPS-013-2 is the normalmethod of the performing a shutdown. For some DC cabinets listed in 34AB-R22-001-2, ifpower is lost and can NOT be restored, a reactor shutdown is required. However, for thesecabinets, 34GO-OPS-014-2 is NOT explicitly required. The second part is plausible becauseIAW 34SO-N21-007-2, the normal method for tripping the RFPTs is the trip pushbuttons onpanel 2H11-P650. Also, if power is lost to 2R22-S017, depressing the trip pushbuttons on panel2H11-P650 will trip the RFPTs.

"C" is correct.

The "D" distractor is plausible because the first part is correct. The second part is plausiblebecause IAW 34SO-N21-007-2, the normal method for tripping the RFPTs is the trippushbuttons on panel 2H11-P650. Also, if power is lost to 2R22-S017, depressing the trippushbuttons on panel 2H11-P650 will trip the RFPTs. References:NONE

K/A:

APE: 295004 Partial or Complete Loss of D.C. Power

G2.1.20 Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12) . . . . . . . . . . . . 4.6 4.6



R42-ELECT-LP-02704, DC Electrical Distribution, Ver. 7.1, EO 200.018.A.05


SRO ONLY Guideline34AB-R22-001-2, Loss of DC Buses, Ver. 6.0 34GO-OPS-014-2, Fast Reactor Shutdown, Ver. 12.0



Unit 2 has experienced an ATWS.

o Reactor Power 17% RTPo MSIVs Openo Main Generator 135 MWe


Time Event07:45 Main Generator load is 115 MWe08:00 Main Generator load is 75 MWe


The EARLIEST listed time that the Main Turbine is procedurally REQUIRED to be tripped is __________ .

IAW 31EO-EOP-011-2, RCA EOP flow chart, following the Main Turbine trip, RPV Pressure is REQUIRED to be stablized below a MAXIMUM value of __________ .

07:45;845 psig

07:45;1074 psig

08:00;845 psig

08:00;1074 psig

A.

B.

C.

D.



Description:

The RCA EOP flowchart is entered if any of the following conditions exist:

o RWL below +3 inches o Reactor pressure above 1074 psig o Drywell Pressure above 1.85 psig o A condition which requires a reactor scram and reactor power is above 5% o A condition which requires a reactor scram and reactor power CANNOT be determined

34AB-C71-001-2, Scram Procedure, is entered following any automatic OR manual Reactorscram. IAW 30AC-OPS-013-0, Use of Emergency Operating Procedures:

A scram may require entry into the EOPs depending on whether one of the entry conditionsis reached. IF EOP entry is necessary, THEN the Scram Procedure AND the EOP willbe performed concurrently. IF conflicts arise between the Scram Procedure AND the EOP,the EOP actions take precedence.

IAW 34AB-C71-001-2:

WHEN Main Generator output is below 80 GMWe, perform Main Turbine Trip Actions perPlacard TC-1.

To control RPV Pressure during an ATWS, the following steps are given in the RCA EOPflowchart, pressure path:

1) Stablize reactor pressure below 1074 psig using the main turbine bypass valves.

2) If the main turbine is off-line and any MSL is open, stabilize pressure below 845 psigwith the bypass valves.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of the RCA EOP flowchart and the Scram Procedure. The SRO is required to assess plant conditions during an emergency (ATWS) and, based onknowledge of the content of the above procedures, know that the Main Turbine is required to betripped and, following the Main Turbine trip, RPV Pressure is required to be stablized below 845psig (EOP Override). This is above the RO knowledge level.



K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to interpret indicationsavailable in the control room and know the required operation of the Main Turbine, and howoperation of the Main Turbine will affect the RPV Pressure limit.

The "A" distractor is plausible because IAW 34GO-OPS-013-2, Normal Plant Shutdown, whenMain Generator load decrease to 120 MWe, the Main Turbine and Main Generator are requiredto be shutdown. Based on the conditions given, 34GO-OPS-013-2 will not be used to controlplant conditions. In fact, during an ATWS, only a specific section of 34GO-OPS-013-2 isentered to begin a plant cooldown when cold shutdown boron weight is inject or the reactor issubcritical based on control rod position. The second part is correct.

The "B" distractor is plausible because IAW 34GO-OPS-013-2, Normal Plant Shutdown, whenMain Generator load decrease to 120 MWe, the Main Turbine and Main Generator are requiredto be shutdown. Based on the conditions given, 34GO-OPS-013-2 will not be used to controlplant conditions. In fact, during an ATWS, only a specific section of 34GO-OPS-013-2 isentered to begin a plant cooldown when cold shutdown boron weight is inject or the reactor issubcritical based on control rod position. The second part is plausible because the RCA EOPflowchart contains a step in the pressure path to "stablize reactor pressure below 1074 psig usingmain turbine bypass valves...". If the Main Turbine is ON-LINE, this step is accomplished byallowing the turbine pressure regulator to control RPV pressure at the normal regulator setting(960 psig).

"C" is correct.

The "D" distractor is plausible because the first part is correct. The second part is plausiblebecause the RCA EOP flowchart contains a step in the pressure path to "stablize reactor pressurebelow 1074 psig using main turbine bypass valves...". If the Main Turbine is ON-LINE, this stepis accomplished by allowing the turbine pressure regulator to control RPV pressure at the normalregulator setting (960 psig).



K/A:

APE: 295005 Main Turbine Generator Trip

G2.2.44 Ability to interpret control room indications to verify the status and operation of asystem, and understand how operator actions and directives affect plant and systemconditions. (CFR: 41.5 / 43.5 / 45.12) . . . . . . . . . . . . . 4.2 4.4



EOP-RCA-LP-20328, Ver. 4.0, EO 201.069.A.13


SRO ONLY Guideline34AB-C71-001-2, Scram Procedure, Ver. 12.531EO-EOP-011-2, RPV Control (ATWS), Ver. 12.0



Unit 2 was operating at 100% RTP when a scram occurred. The following conditions exist:

o Reactor Power 3% RTP, slowly decreasingo RPV Pressure 950 psig, slowly decreasingo RWL 72 inches, slowly increasingo SPDS Unavailableo HPCI injecting into the RPV 4000 gpm


IAW 34AB-B21-002-2, RPV Water Level Corrections, RWL indication __________REQUIRED to be corrected for RPV Pressure.

The SS is REQUIRED to terminate injection into the RPV IAW __________ .

is;31EO-EOP-113-2, Terminating and Preventing Injection into the RPV

is;34AB-C32-001-2, Reactor Water Level Above +60 Inches

is NOT;31EO-EOP-113-2, Terminating and Preventing Injection into the RPV

is NOT;34AB-C32-001-2, Reactor Water Level Above +60 Inches

A.

B.

C.

D.



Description:

Joe, the original question was a Pre-submittal SRO question. The K/A was replaced due tobeing hard to hit at the SRO level.

With RWL > 60", only Floodup Range Instrumention can be used to monitor RWL. NormalRange & Wide Range Instrumentation do NOT indiciate RWL > 60".

IAW 34AB-B21-002-2, RWL indication corrections are REQUIRED when:

Floodup Range Instrumentation is required for level monitoring AND the followingconditions exist: Reactor pressure is < 100 PSIG AND/OR ANY Drywell group 1 RTDabove 212°F

RPV Pressure is > 100 psig, therefore the Floodup Range Instrumentation is NOT required to becorrected for RPV Pressure.

IAW 34AB-C32-001-2:

IF RFPTs, HPCI, OR RCIC fail to trip on High level, take Manual Control per the systemOperating Procedure to prevent uncontrolled injection.

PRIOR to RWL exceeding +100 inches, TERMINATE ALL injection to the Reactorvessel, except CRD, unless directed by the EOPs.

SRO JUSTIFICATION:

The SRO must have knowledge of procedure content to determine which procedure is used toterminate injection into the RPV, based on the conditions given. This is above the ROknowledge level.

K/A JUSTIFICATION:

This question satisifies the K/A statement by requiring the applicant to determine if RWLcorrections are required (determine RWL) during a high RWL condition (With RWL > 60", theFloodup range instrumentation is used to monitor RWL).


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALThe "A" distractor is plausible since RWL is in the range of the Floodup Range Instrumentswhich are required to be used for level monitoring but does not consider the requirement forDrywell temperature and RPV pressure. The second part is plausible since this is the correctprocedure to use in an ATWS if reactor power is greater than 5% RTP.

The "B" distractor is plausible since RWL is in the range of the Floodup Range Instrumentswhich are required to be used for level monitoring but does not consider the requirement forDrywell temperature and RPV pressure. The second part is correct.

The "C" distractor is plausible since the first part is correct. The second part is plausible sincethis is the correct procedure to use in an ATWS if reactor power is greater than 5% RTP.

"D" is correct.



K/A:

APE: 295008 High Reactor Water Level

AA2. Ability to determine and/or interpret the following as they apply to HIGHREACTOR WATER LEVEL : (CFR: 41.10 / 43.5 / 45.13)

AA2.01 Reactor water level................................... 3.9 3.9



295008 High Reactor Water Level

AA2. Ability to determine and/or interpret the following as they apply to HIGHREACTOR WATER LEVEL : (CFR: 41.10 / 43.5 / 45.13)

AA2.02 Steam flow/feedflow mismatch . . . . . . . 3.4 3.4

Difficulty in writing a discriminatory question to this K/A.


LT-LP-20201, Introduction To Abnormal Procedures, Ver. 12.0, EO LT-20201.019


SRO ONLY Guideline 34AB-C32-001-2, Reactor Water Level Above +60 Inches, Ver. 1.131EO-EOP-017-2, CP3 ATWS Level Control, Ver. 16



Unit 2 was operating at 100% RTP when a transient resulted in the following RWL values at thegiven times:

Time RWL

11:30 -36 inches 12:00 -102 inches 12:30 -156 inches 13:00 -194 inches

Based on the above conditions and IAW NMP-EP-111, Emergency Notifications,

The LATEST listed time that State and Local Governments must be notified based on low RWL is __________ .

11:58

12:28

12:58

13:28

A.

B.

C.

D.



Description:

Based on the Fission Product Barrier Chart when RWL reaches -155 inches a "Potential" Losswill exist in the Fuel Clad Barrier and an "Actual" Loss will exist in the RCS Barrier. A loss orpotential loss of two barriers constitutes a Site Area Emergency (FS1). All RWL values are 1inch below the required value.

IAW NMP-EP-111, Emergency Notifications, Section 4.1, PRECAUTIONS ANDLIMITATIONS, step 1 states "Notifications of applicable State and Local Agencies shall beaccomplished as soon as practicable and within 15 minutes of the declaration of anemergency, an upgrade to a higher emergency classification level, or the approval of protectiveactions recommendations."

The earliest time that RWL is below EAL threshold value is 12:30. The ED SHALL assess,classify, AND declare an emergency condition within 15 minutes after the availability ofindications to plant operators that an EAL has been exceeded. Once declared, notifications ofapplicable State and Local Agencies shall be accomplished as soon as practicable and within 15minutes of the declaration of an emergency. 14 minutes to declare plus 14 minutes to notify theState & Local Agencies equals a total time of 28 minutes which is added to each listed time.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of EAL threshold valves for RWL, based on the FissionProduct Barrier Chart. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to interpret RWL for anEmergency declaration based on low RWL and then the latest time for notifying externalagencies (State & Local).

The "A" distractor is plausible because IAW NMP-AD-031, when RWL decreases to -35 inches,a notification to the NRC is required due to a valid actuation of PCIS, ECCS and RCIC.

The "B" distractor is plausible because IAW NMP-AD-031, when RWL decreases to -101inches, a notification to the NRC is required due to a valid actuation of PCIS, ECCS and EDGs.

"C" is correct.

The "D" distractor is plausible because IAW NMP-EP-110-GL02, a RWL of -193 inches is the"Loss" threshold value for the Fuel Clad barrier. In this situation, an emergency is required to bedeclared (if not already declared), and notification of State & Local authorities is required within15 minutes after the declaration.



K/A:

APE: 295009 Low Reactor Water Level

G2.4.30 Knowledge of events related to system operation/status that must be reported tointernal organizations or external agencies, such as the State, the NRC, or the transmissionsystem operator. (CFR: 41.10 / 43.5 / 45.11) . . . . . . . . . . . . 2.7 4.1

SRO only because of link to 10CFR55.43(b)(7): Fuel handling facilities and procedures(Reporting requirements and Emergency classifications).


EP-LP-20101, Initial and Terminating Activities, Ver. 1.1, EO 001.017.A.01 & EO 001.065.A.04


SRO ONLY Guideline NMP-EP-110-GL02, Figure 1, Ver. 4.0 NMP-EP-111, Emergency Notifications, Ver. 11.0



Unit 2 is operating at 100% RTP when a loss of Drywell Cooling occurs.

Drywell pressure is being manually controlled between 0.5 psig and 1.2 psig.

34AB-T47-001-2, Complete Loss of Drywell Cooling, is in progress.

At 10:00, Drywell temperatures reach the temperature limits on Attachment 1, Peak DrywellTemperature.

Drywell temperatures are continuing to slowly increase.

The decision has been made to shutdown Unit 2.

Based on the above conditions and IAW 34AB-T47-001-2,

The shutdown procedure that is REQUIRED to be entered is __________ .

The EARLIEST time that this shutdown procedure is REQUIRED to be entered is __________ .

34GO-OPS-014-2, Fast Reactor Shutdown; 10:30

34GO-OPS-014-2, Fast Reactor Shutdown; 11:00

34GO-OPS-013-2, Normal Plant Shutdown; 10:30 34GO-OPS-013-2, Normal Plant Shutdown; 11:00

A.

B.

C.

D.



Description:

34AB-T47-001-2, Complete Loss of DW Cooling, contains a subsequent action that if any of thetemperatures are exceeded in Attachment 1, then a 30 minute clock starts for restoringtemperatures. If this time limit is exceeded, then a Fast Reactor Shutdown will be initiated per34GO-OPS-014-2.

TS LCO 3.6.1.5, Drywell Temperature, Condition A.1 states with Drywell average airtemperature not within limit to restore drywell average air temperature to within limit in 8 hours.

TS LCO 3.6.1.4, Drywell Pressure, Condition A.1 states with Drywell pressure not within limitto restore drywell pressure to within limit in 1 hour.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of the abnormal procedure and the necessaryconditions/parameters required for a fast reactor shutdown. Also SRO only because this is adetailed knowledge of when to implement attachments and appendices, including how tocoordinate these items with procedure steps. The amount of knowledge required is above theRO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by requiring the applicant to know the abnormalDrywell temperature conditions for requiring a fast reactor shutdown and then the appropriatetime constraint for when to enter the fast reactor shutdown procedure based upon high Drywelltemperatures.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sinceTS LCO 3.6.1.4, Drywell Pressure, allows a one (1) hour completion time to restore Drywellpressure if exceeded, therefore, adding one (1) hour to the time that Attachment 1 temperatureswere exceeded and selecting 11:00.

The "C" distractor is plausible since TS LCO 3.6.1.5, Drywell Air Temperature, allows an eight(8) hour completion time to restore Drywell temperatures and 12 more hours to reach Mode 3 &36 hours to reach Mode 4, therefore, selecting the normal Shutdown procedure(34GO-OPS-013-2). The second part is plausible since it is correct.


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALThe "D" distractor is plausible since TS LCO 3.6.1.5, Drywell Air Temperature, allows an eight(8) hour completion time to restore Drywell temperatures and 12 more hours to reach Mode 3 &36 hours to reach Mode 4, therefore, selecting the normal Shutdown procedure(34GO-OPS-013-2). The second part is plausible since TS LCO 3.6.1, Drywell Pressure, allowsa one (1) hour completion time to restore Drywell pressure if exceeded, therefore, adding one (1)hour to the time that Attachment 1 temperatures were exceeded and selecting 11:00. References:NONE

K/A:

APE: 295012 High Drywell Temperature

AA2. Ability to determine and/or interpret the following as they apply to HIGHDRYWELL TEMPERATURE : (CFR: 41.10 / 43.5 / 45.13)

AA2.01 Drywell temperature . . . . . . . . . . . . . . . . . . . 3.8 3.9



T23-PC-LP-01301, Primary Containment, Ver 7.1, EO 200.032.A.01


SRO ONLY Guideline 34AB-T47-001-2, Complete Loss of Drywell Cooling, Ver. 1.10 34GO-OPS-014-2, Fast Reactor Shutdown, Ver. 12.0 Modified from 2015 ILT NRC Exam Q#88

Original Question

Unit 1 is operating at 100% RTP when a loss of Drywell Cooling occurs.

34AB-T47-001-1, Complete Loss of Drywell Cooling, is in progress.

The Drywell temperatures on Attachment 1, Peak Drywell Temperature, are exceeded.

The following Drywell conditions exist:


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALo Average Drywell temperature is 190°F


IAW 34AB-T47-001-1, the MAXIMUM amount of time that the existing Drywell conditionscan exist before requiring a fast reactor shutdown is __________ .

IAW Unit 1 TS Bases 3.6.1.5, Drywell Air Temperature, if a DBA LOCA were to occur, theresultant peak accident temperature __________ EXCEED the Drywell design temperature.

A. 30 minutes; will

B. 30 minutes; will NOT

C. 60 minutes; will

D. 60 minutes; will NOT


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTAL15. 295031EA2.01 001

Unit 1 was operating at 100% RTP when a transient resulted in the following conditions:

o Reactor power 18% RTP o RWL 9 inches o MSIVs CLOSED o Torus temperature is above BIIT o RPV Pressure controlled by LLS

Based on the above conditions and IAW 31EO-EOP-017-1, CP-3, ATWS Level Control,

The SS will direct the operator to __________ .

Once a RWL band is established, if reactor power remained above 10% RTP throughout theRWL decrease, the SS will order an UPPER RWL band of __________ .

lower RWL to maintain between -60 inches and -90 inches with Table L-3 Systems; -60 inches

lower RWL to maintain between -60 inches and -90 inches with Table L-3 Systems; -155 inches

perform 31EO-EOP-113-1, Terminating and Preventing Injection Into The RPV; -60 inches

perform 31EO-EOP-113-1, Terminating and Preventing Injection Into The RPV; -155 inches

A.

B.

C.

D.



Description:

IAW the CP-3 EOP flowchart:

IF all of the following exist:If reactor power is above 5%RWL is above -155"Torus water temp. is above BIITAny SRV is open or opens

THEN Terminate and prevent using 31EO-EOP-113-1

Terminate and prevent is performed until any of the following conditions are satisfied:Reactor power remains < 5% and RWL is < -60"RWL is at -155"Drywell pressure remains < 1.85 psig and all SRVs remain closed and RWL is < -60"

When the terminate and prevent is complete, RWL will be maintained between -180" and theRWL value where one of the above conditions was satisfied.

Assuming reactor power remains > 10%, the RWL value which will stop the lowering will be-155 ", therefore the SS will order RWL to be maintained between -180" and -155".

SRO JUSTIFICATION:

The SRO must have detailed knowledge of decision points in the EOPs that involve transitionsto event specific sub-procedures. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine the required actionfor RWL and then, based on the taken action and plant conditions, determine the correct band formaintaining RWL.

The "A" distractor is plausible because IAW CP-3, if the conditions for performing Terminateand Prevent do NOT exist, then the following step is executed: IF Reactor power is > 5% ANDRWL is > -60 ", THEN lower RWL to maintain between -60" and -90" with Table L-3 Systems. The second part is plausible because IAW CP-3, If RWL is lowered to -60" to -90" and a furtherreduction in RWL is needed, then maintain RWL -60" to -180". Also, if Terminate and Preventis performed, EOP-113 is exited if reactor power remains < 5% and RWL is < -60". If EOP-113is exited based on this condition, -60" becomes the upper limit on the RWL band.


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALThe "B" distractor is plausible because IAW CP-3, if the conditions for performing Terminateand Prevent do NOT exist, then the following step is executed: IF Reactor power is > 5% ANDRWL is > -60 ", THEN lower RWL to maintain between -60" and -90" with Table L-3 Systems.The second part is correct.

The "C" distractor is plausible since the first part is correct. The second part is plausible becauseIAW CP-3, If RWL is lowered to -60" to -90" and a further reduction in RWL is needed, thenmaintain RWL -60" to -180". Also, if Terminate and Prevent is performed, EOP-113 is exited ifreactor power remains < 5% and RWL is < -60". If EOP-113 is exited based on this condition,-60" becomes the upper limit on the RWL band.

"D" is correct.

References:U1 Graph 5 (BIIT)

K/A:

EPE: 295031 Reactor Low Water Level

EA2. Ability to determine and/or interpret the following as they apply to REACTOR LOWWATER LEVEL : (CFR: 41.10 / 43.5 / 45.13)

EA2.01 Reactor water level . . . . . . . . . . . . . . . . 4.6* 4.6*



EOP-CP3-LP-20327, Level / Power Control (CP-3 ), Ver. 5.0, EO 201.089.A.01


SRO ONLY Guideline 31EO-EOP-017-1, CP-3, ATWS Level Control, Ver. 13.0



Unit 1 is operating at 30% RTP when the Main Turbine trips.

NO RPS Scram signal is received. Reactor power is 33% RTP.

The OATC performs RC-1 actions and manually initiates Alternate Rod Insertion (ARI).

ARI initiation results in an ALL RODS IN condition.

Based on the above conditions and IAW NMP-EP-110-GL02 - HNP EALs - ICs, ThresholdValues and Basis,

Initiating ARI from the Main Control Room __________ considered a Manual Scram.

An EAL Threshold value __________ been EXCEEDED.

is; has

is; has NOT

is NOT; has is NOT; has NOT

A.

B.

C.

D.



Description:

Basis for: ALERT

SUBCRITICAL: IRMs below Range 6 and Period is negative

This condition indicates failure of the automatic protection system to SCRAM the reactor. Thiscondition is more than a potential degradation of a safety system in that a front line automaticprotection system did not function in response to a plant transient and thus the plant safety hasbeen compromised, and design limits of the fuel may have been exceeded. An Alert is indicatedbecause conditions exist that lead to potential loss of fuel clad or RCS. Reactor protectionsystem setpoint being exceeded, rather than limiting safety system setpoint being exceeded, isspecified here because failure of the automatic protection system is the issue. A manual scramis any set of actions by the reactor operator(s) at the reactor control console (1/2H11-P603 Panel)which causes control rods to be rapidly inserted into the core and brings the reactorSUBCRITICAL (e.g., reactor trip button, Alternate Rod Insertion). Failure of manual scramwould escalate the event to a Site Area Emergency.

Since ARI inserted all of the control rods, an Alert emergency will be the highest emergencydeclared.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of the threshold value basis and EAL knowledge inNMP-EP-110-GL02 to answer this question. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant the basis for the threshold valuedefinition of a manual scram (ARI considered a manual scram Y/N) and then if emergencyclassification exists based on given conditions.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sincethe initiation of ARI caused all control rods to fully insert, which is equivalent to a manualscram, therefore, no EAL Threshold value has been exceeded.



The "C" distractor is plausible if the applicant remembers that ARI takes approximately 35seconds to insert all control rods and compares this to a manual scram which inserts all controlrods in approximately 7 seconds, therefore with 35 seconds compared to 7 seconds, selects ARIis NOT a manual scram. The second part is plausible since it is correct.

The "D" distractor is plausible if the applicant remembers that ARI takes approximately 35seconds to insert all control rods and compares this to a manual scram which inserts all controlrods in approximately 7 seconds, therefore with 35 seconds compared to 7 seconds, selects ARIis NOT a manual scram. The second part is plausible since the initiation of ARI caused allcontrol rods to fully insert, and if reactor power was 26% RTP, no EAL Threshold value wouldhave been exceeded.


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTAL References:NONE

K/A:

APE: 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale orUnknown

G2.4.41 Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10 / 43.5 / 45.11) . . . . . . 2.9 4.6



295019 Partial or Complete Loss of Instrument Air

G2.4.41 Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10 / 43.5 / 45.11) . . . . . . 2.9 4.6


EP-LP-20101, Initial and Terminating Activities, Ver. 1.1, EO 001.017.A.01


SRO ONLY Guideline NMP-EP-110-GL02, HNP EALs - ICs, Threshold Values and Basis, Ver. 4.2


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTAL17. 295030EA2.03 001

Unit 1 was operating at 100% RTP when an event occurred causing Torus Water level to begindecreasing.

At 10:00, the following conditions exist:

o Torus water level 130 inches decreasing 2 inches/minute o MSIVs Closedo RPV Pressure controlled between 800 - 1074 psig o RWL -65 inches increasing 1 inch/minute o RCIC Danger Tagged out of service o HPCI System injecting

ALL Low Pressure Injection Systems are available.

At 10:15, Maintenance isolates the leak and Torus level stabilizes at 100 inches.


At 10:08, HPCI is REQUIRED to be __________ .

At 10:15, RPV pressure is REQUIRED to be controlled using EOP flowchart __________ .

injecting to maintain Reactor water level; RC/P

injecting to maintain Reactor water level; CP-1 Point G

shutdown and prevented from operating; RC/P

shutdown and prevented from operating; CP-1 Point G

A.

B.

C.

D.



Description:

With the current trend on Torus water level, Torus level will be 114 inches (130 - (8x2) ) at1008. IAW the PC flowchart, since adequate core cooling is assured, the HPCI System will besecured when Torus level cannot be maintained >115 (U2 110) inches due to uncovering theHPCI Exhaust line and pressurizing Primary Containment. The EOPs allow RCIC to beoperated below this value. No instructions regarding RCIC operation is included in this step (orin an equivalent step) for two reasons:

1. The exhaust flow rate of RCIC is approximately equal to that of decay heat, and is thusconsistent with the basis used for determining the Primary Containment Pressure Limit.

2. Elevated torus pressure will cause the RCIC turbine to trip much sooner than the HPCIturbine.

At 1015, Torus level is 100 inches which is below the level at which an Emergency Depress willbe required, therefore, allowing CP-1 Point G to be used to control reactor pressure.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of both units PC EOP Flowcharts (SP/L) Paths alongwith EOP Overrides to answer this question. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to determine how RPV pressurewill be controlled with a low suppression pool water level.

The "A" distractor is plausible since the first part would be correct if on U2. The second part isplausible if the applicant does not realize that an Emergency Depress is required on U1. Also isplausible since it would be correct on U2.

The "B" distractor is plausible since the first part would be correct on U2. The second part iscorrect.

The "C" distractor is plausible since the first part is correct. The second part is plausible if theapplicant does not realize that an Emergency Depress is required on U1. Also is plausible sinceit would be correct on U2.

"D" is correct.



K/A:

EPE: 295030 Low Suppression Pool Water Level

EA2. Ability to determine and/or interpret the following as they apply to LOWSUPPRESSION POOL WATER LEVEL : (CFR: 41.10 / 43.5 / 45.13)

EA2.03 Reactor pressure . . . . . . . . . . . . . . . . . . . . 3.7 3.9

SRO only because of link to 10CFR55.43(b)(5): Assessment of facility conditions andselection of appropriate procedure, recalling the action in the body of procedure and whento take the action.


EOP-PC-LP-20310, Primary Containment Control PC, Ver. 4.0, EO 201.075.A.12 & EO 201.075.A.13


SRO ONLY Guideline 31EO-EOP-012-1 & 2, PC Primary Containment Control, Ver. 7.0 Bank question from HLT Database which was used on 2012 NRC Exam Q#89




o INSTR AIR PREFLTR D103A DIFF PRESS HIGH, 700-225, ILLUMINATED o INSTR AIR DRYERS SYS PRESS LOW, 700-219, ILLUMINATEDo Prefilter D103A Differential pressure is 15 PSID o Non-Essential Instrument Air Header pressure is 55 psig

The following procedures are entered:

o 34AR-700-225-2, INSTR AIR PREFLTR D103A DIFF PRESS HIGH o 34AB-P51-001-2, Loss Of Instrument And Service Air System Or Water Intrusion Into

The Service Air System


The Non-Essential Instrument Air Header Isolation Valve, 2P52-F015, will __________ .

The procedure that contains detailed guidance on how to place Prefilter D103B in serviceand remove Prefilter D103A from service is __________ .

remain open; 34AB-P51-001-2

remain open; 34AR-700-225-2

close and remain closed; 34AB-P51-001-2

close and remain closed; 34AR-700-225-2

A.

B.

C.

D.



Description:


The prefilter D103A malfunction has resulted in the air pressure downstream decreasing. Whenair pressure reaches 50 psig, the Non-Essential Instrument Air Header Isolation Valve,2P52-F015, will close and remain closed. With decreasing air pressure, 34AB-P51-001-2, willalso be entered. Since the differential pressure is above the required value, the SRO will directthe NPO to swap prefilters IAW 34AR-700-225-2.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of 34AB-P51-001-2 & 34AR-700-225-2 to correctlyassess plant conditions and then select the correct procedure to mitigate the event. This is abovethe RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to predict (RO part) theresponse to an air filter malfunction and determine which procedure (SRO part) will provide thedetatiled guidance to correct the malfunction.

The "A" distractor is plausible since the first part is correct. The second part is plausible if theapplicant remembers that these alarms are mentioned in this procedure, therefore thinking theabnormal procedure will direct the activity.

"B" is correct.

The "C" distractor is plausible if the applicant remembers there is a valve in the Air system(F017) which will close and remain closed when the associated pressure is less than 70 psig. The second part is plausible if the applicant remembers that these alarms are mentioned in thisprocedure, therefore thinking the abnormal procedure will direct the activity.

The "D" distractor is plausible if the applicant remembers there is a valve in the Air system(F017) which will close and remain closed when the associated pressure is less than 70 psig. The second part is plausible since it is correct.


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALNONE

K/A:

300000 Instrument Air System (IAS)

A2. Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEMand (b) based on those predictions, use procedures to correct, control, or mitigate theconsequences of those abnormal operation: (CFR: 41.5 / 45.6)

A2.01 Air dryer and filter malfunctions . . . . . . . . . . . . . . . . . . . . . . . 2.9 2.8



P51-P52-P70-PLANT AIR-LP-03501, plant Air System, Ver. 5.0, EO 035.001.A.02


SRO ONLY Guideline 34AR-700-225-2, INSTR AIR PREFLTR D103A DIFF PRESS HIGH, Ver. 1.2 34AB-P51-001-2, Loss Of Instrument And Service Air System Or Water Intrusion Into TheService Air System, Ver. 5.0

Modified for Hatch from 2010 April Brunswick NRC Exam Q#92

Original Question

The following plant conditions exist on Unit Two due to a malfunction of the Air Dryer:

- SERVICE AIR PRESS-LOW is in alarm- RB INSTR AIR RECEIVER 2A PRESS LOW is in alarm- RB INSTR AIR RECEIVER 2B PRESS LOW is in alarm- Instrument Air pressure is 93 psig and recovering

Based on the above indications, which one of the following identifies:

(1) the status of the Service Air Dryer Bypass Valve, SA-PV-5067, and (2) the procedure that contains the steps to close the Reactor Building Inboard and


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALOutboard Isolation Valves (BFIVs)?

A. (1) open (2) OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures

B. (1) open(2) 2APP-UA-O1, Se,vice Air Press-Low

C. (1) closed (2) OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures

D. (1) closed (2) 2APP-UA-O1, Se,vice Air Press-Low


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTAL19. G2.1.3 001

The following Modes exist for both units at the given time:

Time Unit 1 Unit 2

10:00 Mode 4 Mode 2 11:00 Mode 4 Mode 3 17:00 Mode 5 Mode 3 20:00 Mode 5 Mode 4

Based on the above conditions and IAW Tech Specs,

The EARLIEST listed time that an extra (Licensed) Reactor Operator (RO) can assume the Control Room Command Function is __________ .

10:05

11:05

17:05

20:05

A.

B.

C.

D.



Description:

IAW TS 5.1 Responsibility, 5.1.5, A Senior Reactor Operator (SRO) shall be responsible for thecontrol room command function. During any absence of the responsible SRO from the controlroom while either unit is in MODE 1, 2, or 3, an individual with an active SRO license shall bedesignated to assume the control room command function.

During any absence of the responsible SRO from the control room while both units are inMODE 4 or 5, an individual with an active SRO license or Reactor Operator license shall bedesignated to assume the control room command function.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of TS 5.1 "Responsibility" of the Control RoomCommand Function as defined by TS 5.1.5. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant the Shift Manager's short-termrelief of the Control Room Command Function.

The "A" distractor is plausible if the applicant remembers that Mode 4 is a condition where aLicensed RO can assume this function but does not consider Unit 2 is in Mode 2 which requiresa SRO. This would be correct if Unit 2 was in Mode 4 or 5.

The "B" distractor is plausible if the applicant remembers that Mode 4 is a condition where aLicensed RO can assume this function and since Unit 2 is shutdown in Mode 3 thinks conditionsare met. This would be correct if Unit 2 was in Mode 4 or 5.

The "C" distractor is plausible if the applicant remembers that Mode 5 is a condition where aLicensed RO can assume this function and since Unit 2 is still shutdown in Mode 3 thinksconditions are met. This would be correct if Unit 2 was in Mode 4 or 5.

"D" is correct.



K/A:

2.1 Conduct of Operations

G2.1.3 Knowledge of shift or short-term relief turnover practices. (CFR: 41.10 / 45.13) . . . . . . . . . . . . . . . 3.7 3.9

SRO only because of link to 10CFR55.43(b)(1): Conditions and limitations in the facilitylicense.


LT-LP-30003, Tech Specs / Administrative Controls, Ver. 8.0, LT-30003.002


SRO ONLY Guideline U1 TS 5.0, Administrative Controls, 5.1.5, Responsibility, Amendment 252 U2 TS 5.0, Administrative Controls, 5.1.5, Responsibility, Amendment 196



A proposed plant modification must ALWAYS have prior approval from the NRC if itinvolves any __________ .

system that requires a 50.59 screening

change to any system included in Tech Specs

design basis limit for Primary Containment being altered

change to the Technical Requirements Manual (TRM) Bases

A.

B.

C.

D.



Description:

IAW NMP-AD-010, 10CFR 50.59 Screenings and Evaluations, section 6.1.1 states "(c)(2) Alicensee shall obtain a license amendment pursuant to 50.90 prior to implementing a proposedchange, test, or experiment if the change, test, or experiment would:

(i) Result in more than a minimal increase in the frequency of occurrence of an accidentpreviously evaluated in the final safety analysis report (as updated);

(ii) Result in more than a minimal increase in the likelihood of occurrence of amalfunction of a structure, system, or component (SSC) important to safety previouslyevaluated in the final safety analysis report (as updated);

(iii) Result in more than a minimal increase in the consequences of an accident previouslyevaluated in the final safety analysis report (as updated);

(iv) Result in more than a minimal increase in the consequences of a malfunction of anSSC important to safety previously evaluated in the final safety analysis report (asupdated);

(v) Create a possibility for an accident of a different type than any previously evaluated inthe final safety analysis report (as updated);

(vi) Create a possibility for a malfunction of an SSC important to safety with a differentresult than any previously evaluated in the final safety analysis report (as updated);

(vii) Result in a design basis limit for a fission product barrier as described in theFSAR (as updated) being exceeded or altered; or

(viii) Result in a departure from a method of evaluation described in the FSAR (as updated)used in establishing the design bases or in the safety analyses.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of NMP-AD-010 to determine which one of theconditions requires prior NRC approval.



K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant when prior NRC approval isrequired prior to making a design change.

The "A" distractor is plausible since this type of evaluation will determine if NRC approval isrequired.

The "B" distractor is plausible since this must have a 50.59 evaluation but not necessarily priorNRC approval.

"C" is correct.

The "D" distractor is plausible since this must have a 50.59 evaluation but not necessarily priorNRC approval if the applicant remembers TRM and applies this to the TRM Bases.



K/A:


G2.2.5 Knowledge of the process for making design or operating changes to the facility. (CFR: 41.10 / 43.3 / 45.13) . . . . . . . . 2.2 3.2

SRO only because of link to 10CFR55.43(b)(3): Facility licensee procedures required toobtain authority for design and operating changes in the facility.


LT-LP-30004, Administrative Procedures, Ver. 14.0, EO 300.017.A.03


SRO ONLY Guideline NMP-AD-010, 10CFR 50.59 Screenings and Evaluations, Ver. 13.0 Bank question used on 2009 Perry NRC Exam Q#SRO 21 Bank question from HLT Databse which was used on 2011 NRC Exam Q#96



Operations has identified an issue that needs troubleshooting activities to be performed in anarea posted as "PROTECTED EQUIPMENT AREA".

Troubleshooting will be performed near "PROTECTED" equipment.

Based on the above conditions and IAW NMP-OS-010, Protected Train/Division and ProtectedEquipment Program,

The MINIMUM level of authorization REQUIRED to enter the area near the"PROTECTED" equipment to perform troubleshooting activities is the __________ .

Shift Manager

Shift Supervisor

Shift Support Supervisor

Shift Operations Manager

A.

B.

C.

D.



Description:

IAW NMP-OS-010:

The Shift Manager has the responsibility of authorizing work on or near a ProtectedTrain/Division or Protected Equipment.

The Shift Manager can grant permission to Work Near Protected Train/Division or ProtectedEquipment after all of the following have occurred:

The SM has evaluated the need for access and has determined that access permission shouldbe granted.The area has been evaluated for any compensatory actions (e.g., oversight, detailed briefing,bump protection, protective barriers, etc.) needed to provide reasonable assurance that noadverse effects could occur.The SM has verified all compensatory actions have been put in place.All persons needing to enter the area have been identified and briefed by the SM.The Work Group Supervisor SHALL provide periodic oversight of the work activity.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of administrative procedures to coordinate normal plantactivities. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant detailed knowledge foradministrative requirements in NMP-OS-010 for managing troubleshooting activities nearProtected Equipment.

"A" is correct.

The "B" distractor is plausible because IAW NMP-AD-002, the Shift Supervisor is required toapprove a troubleshooting plan (except for troubleshooting within a clearance boundary).



The "C" distractor is plausible because IAW NMP-OS-007, the Shift Support Supervisorauthorizes maintenance and/or testing activities to be performed, and ensures plant conditionsare suitable for performing such activities.

The "D" distractor is plausible because IAW NMP-OS-007, the Shift Operations Manager is thefinal authority for work prioritization in the Work Planning and Scheduling process for onlinework activities. Also, the Shift Operations Manage provides direction to the Shift Managers forroutine scheduling and coordination of Operations shift activities, including interfacing withother plant departments.

References:NONE

K/A:


G2.2.20 Knowledge of the process for managing troubleshooting activities. (CFR: 41.10 / 43.5 / 45.13) . . . . . . . . . . . . . 2.6 3.8



HLT-SIM-LP-00003, ILT Simulator Curriculm Modules C, D, E, Ver. 8.4,


SRO ONLY Guideline NMP-OS-010, Protected Train/Division and Protected Equipment Program, Ver. 7.2NMP-AD-002, Problem Solving and Troublehooting Guidelines, Ver. 11.0 NMP-OS-007, Conduct of Operations, Ver. 11.0



During a Radiological Emergency, an Operator is needed to enter an area to protect valuableequipment. The Operator will exceed their Federal exposure limit during this entry.


IAW 73EP-EIP-017-0, Emergency Exposure Control, the MAXIMUM Dose Limit allowedto protect valuable property is __________ .

IAW NMP-EP-110, Emergency Classification Determination and Initial Action, theindividual who is REQUIRED to authorize exceeding this exposure limit is the __________ .

10 Rem Emergency Director

10 Rem RP Supervisor

25 Rem Emergency Director

25 Rem RP Supervisor

A.

B.

C.

D.



Description:

IAW NMP-EP-110, Emergency Classification Determination and Initial Action, step 2.1.1, 5thbullet, requires the Shift Manager (Emergency Director) to have the responsibility to authorizeplant personnel to exceed 10CFR20 radiation exposure limits.

IAW 73EP-EIP-017-0, Emergency Exposure Control, step 7.4.1, the Exposure Limit is 25 REMfor life saving rescue or protection of large populations. The Exposure Limit is 10 REM forprotecting valuable property.

IAW 60AC-HPX-001-0, Radiation Exposure Limits, the RP Supervisor is in the chain ofprogression for authorizing exceeding Tier 1 limits. Occupational dose to individual adults shallbe limited to:

An annual limit which is the more limiting of: (1) The total effective dose equivalent (DDE + CEDE) being equal to 5 rem;

OR (2) The sum of the Deep Dose Equivalent (DDE) and the Committed Dose Equivalent

(CDE) to any individual organ or tissue, other than the lens of the eye, being equal to50 rem.

An annual limit to the lens of the eye (LDE) 15 rem. A Shallow Dose Equivalent (SDE) to the skin or any extremity of 50 rem/yr.

SRO JUSTIFICATION:

The SRO must realize what constitues an emergency exposure and then determine who providesthe authorization and the value based on a table inside of the procedure. ED responsibilites areabove the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to have knowledge of raditationexposure limits and the appropriate authorization to exceed the exposure limits in an emergency.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sincethe RP Supervisor is responsible for radiation protections at Plant Hatch and will be asssitingwith radiation decisions, but can NOT authorize exceeding 10CFR20 limits (NMP-EP-110 step5.1.1).


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALThe "C" distractor is plausible since it would be correct if asking for the value to save a life. Thesecond part is correct.

The "D" distractor is plausible since it would be correct if asking for the value to save a life. Thesecond part is plausible since the RP Supervisor is responsible for radiation protections at PlantHatch and will be asssiting with radiation decisions, but can NOT authorize exceeding 10CFR20limits (NMP-EP-110 step 5.1.1). References:NONE

K/A:


G2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10) . . . . . . . . . . . . . . 3.2 3.7

SRO only because of link to 10CFR55.43(b)(4): Radiation hazards that may arise duringnormal and abnormal situations, including mainteneance activities and variouscontamination conditions.


LT-LP-30008, Radiation Control Adminstration Procedures And Instrumentation Ver. 9.0, EO 100.012.A.03


SRO ONLY Guideline 60AC-HPX-001-0, Radiation Exposure Limits, Ver. 10.9 73EP-EIP-017-0, Emergency Exposure Control, Ver. 4.0 NMP-EP-110, Emergency Classification Determination and Initial Action, Ver. 8.1 Modified from HLT Database Q#G2.3.4-002 which was used on 2011 HLT-6 Audit

Original Question

During an Emergency an individual is injured and is unconscious in the Reactor Building. AnOperator is needed to rescue the individual and will receive 12 rem exposure during the rescue. The Operator has already received 1500 mrem Total Effective Dose Equivalent (TEDE) for theyear.

Which ONE of the choices below is the amount of exposure the Operator will have to receive


ILT 10 SRO NRC EXAM PRE DRAFT SUBMITTALexceeding this exposure limit?

A. 3 Rem Emergency Director ONLY

B. 3.5 Rem Emergency Director ONLY

C. 3 Rem Emergency Director or their designee

D. 3.5 Rem Emergency Director or their designee



IAW 73EP-RAD-001-0, Radiological Event,

The __________ is the MINIMUM level of qualification necessary to declare a Radiological Event AND will make immediate decisions concerning Emergency Call Listnotifications.

A Radiological Event will be declared when one Area Radiation Monitor UNEXPECTEDLYalarms and radiation levels have increased to a MINIMUM of __________ their normalvalues.

Shift Manager; three (3) times

Shift Manager; ten (10) times

Control Room Shift Supervisor; three (3) times

Control Room Shift Supervisor; ten (10) times

A.

B.

C.

D.



Description:

73EP-RAD-001-0, Section 6.0, states:

The Control Room Shift Supervisor, normally in consultation with RP Supervision, must havedetermined it to be prudent to alert plant personnel to an unusual radiological condition.

Such conditions include, but are not limited to the following:

An Area Radiation Monitor (ARM) UNEXPECTEDLY alarms indicating radiation levels inthe vicinity of the monitor of greater than 10 times normal.

An air sample taken in an occupied area which is normally not an Airborne RadioactivityArea UNEXPECTEDLY yields results of > 1.0 DAC.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of Radiation Event procedure including when to declarea Radiological Event and the proper authorization level. This is above the RO knowledge level.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant when a Radiological Event willbe declared and the minimum level of authorization based on radiation monitor alarms.

The "A" distractor is plausible since the Shift Manager can make the declaration if the ShiftSupervisor is incapacitated, but in this case is not the minimum level. The second part isplausible if the applicant remembers there are radiation instruments that will indicate fueldamage if they exceed a certain value (3 times normal for MSL rad monitors), therefore, therequirements for declaring a Radiological Event on radiation levels will be met.

The "B" distractor is plausible since the Shift Manager can make the declaration if the ShiftSupervisor is incapacitated, but in this case is not the minimum level. The second part is correct.

The "C" distractor is plausible since the first part is correct. The second part is plausible if theapplicant remembers there are radiation instruments that will indicate fuel damage if they exceeda certain value (3 times normal for MSL rad monitors), therefore, the requirements for declaringa Radiological Event on radiation levels will be met.

"D" is correct.



K/A:


G2.3.13 Knowledge of radiological safety procedures pertaining to licensed operatorduties, such as response to radiation monitor alarms, containment entry requirements, fuelhandling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 43.4 / 45.9 / 45.10) . . . . . . . . . . . . . . . 3.4 3.8

SRO only because of link to 10CFR55.43(b)(4): Radiation hazards that may arise duringnormal and abnormal situations, including maintenance activities and variouscontamination conditions.



SRO ONLY Guideline 73EP-RAD-001-0, Radiological Event, Ver. 2.1

Bank question from HLT Database Q#G2.3.13S-001 which was used on 2012 HLT-7 NRCExam Q#98



An event occurs on Unit 2 resulting in the need to Anticipate an Emergency Depress of the RPV.

Subsequently, the event degrades to where the Severe Accident Guidelines (SAGs) are entered.


IAW the EOPs, to Anticipate the Emergency Depress, the Main Turbine Bypass Valves willbe opened using __________ .

After the SAGs are entered, the EOP Flowcharts will be __________ .

34SO-N30-001-2, Main Turbine Operation; performed concurrently

34SO-N30-001-2, Main Turbine Operation; exited

31EO-EOP-107-2, Alternate RPV Pressure Control; performed concurrently

31EO-EOP-107-2, Alternate RPV Pressure Control; exited

A.

B.

C.

D.



Description:

On the RC EOP flowchart at C-1:

"If Emergency Depress is anticipated, except for low RWL, THEN rapidly depressurize with main turbine bypass valves..."

If based on the present trends, it is expected that Emergency Depressurization will be required,then it is expected that the crew will perform this step. Once the bypass valves have beenopened to begin rapidly depressurizing, the depressurization may be stopped only if the reasonfor anticipating emergency depressurization is corrected (e.g. a break is now isolated on SCchart).

34SO-N30-001-2, Main Turbine Operation, Attachment 12 contains the procedure direction foroperating the Main Turbine Bypass Valves for "Anticipating Emergency Depress."

The EOP flowcharts contain continuous recheck statements that provide direction when entryinto the SAGs is required. IF it is determined core damage is occurring, OR following an EmergDepress, RWL cannot be restored and maintained above -180" OR adequate CS cannot beestablished, THEN EXIT the EOPs and enter the SAGs. When the SAGs are entered, theEOP flowcharts are put down and not used. Entry into EOP supplemantal procedures is allowed,if directed by the SAGs.

IAW 30AC-OPS-013-0, Use of EOPs:

Exit from the EOP may be accomplished in four ways...The last method to exit from the EOP isIF [entry into the SAGs] is required. THEN the EOP will direct the flow chart user to exitthe EOP and enter the SAG.

SRO JUSTIFICATION:

The SRO must have detailed knowledge of administrative procedures that specify the EOP/SAGhierarchy; when SAGs are entered, the EOP flowcharts are exited and not re-entered. This isabove the required RO knowledge level.



K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant to have knowledge of whencontrol of emergency strategies is transfered from the EOPs to the SAGs.

The "A" distractor is plausible since the first part is correct. The second part is plausible becausewhen the SAGs are entered, the EOP flowcharts are exited but the EOP supplemental proceduresare still used. The EOP supplemental procedures are entered when directed by the SAGs and areperformed currently with the SAGs.

"B" is correct.

The "C" distractor is plausible because when the EOP flowcharts are entered, 31EO-EOP-107-2provides alternate methods to control RPV Pressure when normal pressure control methods arenot available. EOP-107 does not contain any procedural direction for operating the MainTurbine Bypass Valves. The second part is plausible because when the SAGs are entered, theEOP flowcharts are exited but the EOP supplemental procedures are still used. The EOPsupplemental procedures are entered when directed by the SAGs and are performed currentlywith the SAGs.

The "D" distractor is plausible because when the EOP flowcharts are entered, 31EO-EOP-107-2provides alternate methods to control RPV Pressure when normal pressure control methods arenot available. EOP-107 does not contain any procedural direction for operating the MainTurbine Bypass Valves. The second part is plausible since it is correct.



K/A:


G2.4.5 Knowledge of the organization of the operating procedures network for normal,abnormal, and emergency evolutions. (CFR: 41.10 / 43.5 / 45.13) . . . . . . 3.7 4.3

SRO only because of link to 10CFR55.43(b)(5): Assessment of facility conditions andselection of appropriate procedure during normal, abnormal, and emergency situations.


EOP-RULES-LP-20307, EOP Rules and Usage, Ver. 3.1, EO 201.093.A.13


SRO ONLY Guideline 30AC-OPS-013-0, Use Of Emergency Operating Procedures, Ver. 9.6



Unit 2 is operating at 100% RTP when a fire occurs on Startup Auxillary Transformer (SAT) 2Cresulting in a large amount of smoke.

34AB-X43-001-2, Fire Procedure, is in progress.

Based on the above conditions and IAW 34AB-X43-001-2,

While the fire is burning, the SS will direct an operator to align the Main Control RoomEnviromental Control (MCREC) System to the __________ .

Georgia Control Center (GCC) __________ REQUIRED to be notified prior tode-energizing SAT 2C.

Isolation Mode; is NOT

Isolation Mode; is

Pressurization Mode; is NOT

Pressurization Mode; is

A.

B.

C.

D.



Description:

Joe, this was a Pre-submittal SRO question. We rewrote the 2nd part. Changes wereincorporated based on your ES-401-9 comments.

SAT 2C is located on the West side of the Turbine Building. MCREC suction louvers arelocated on the West side of the Turbine Building. Smoke coming from SAT 2C can enter theMCREC suction louvers and be drawn into the Main Control Room, creating a smokeenvironment.

34AB-X43-001-2, Fire Procedure, contains the following steps:

4.6 IF an electrical fire exists: o De-energize the affected component. o IF the fire involves an electrical bus, THEN confirm open OR open the normal

and alternate supply breakers associated with the bus. o IF a transformer is supplying power to the component, THEN ensure the

transformer is DE-ENERGIZED.

4.12 IF the fire is outside (Transformer fire OR forest fire) AND is causing a smokeenvironment, THEN place Control Room ventilation system in Isolation mode per34SO-Z41-001-1.

5.15.2 IF CO2 discharge to the Cable Spreading Room OR Computer Room is required forfire suppression, perform the following:

5.15.2.1 Place the Main Control Room in the Pressurization mode per34SO-Z41-001-1.

IAW 34SO-S22-001-2, 500 KV SUBSTATION SWITCHING, step 4.3.6 which states: "Exceptduring emergency conditions, PRIOR to OPENING one of the following 230kv breakers, notifyGCC to contact Switchyard Maintenance to inspect the other breaker and associateddisconnect switches feeding the respective Startup Transformer (SUT) to ensure that noopen-phase conditions exist. (e.g.: If opening the 179520 PCB, then inspect the 179530 PCBand associated disconnects. Refer to Attachment 4.

SUT 1C & 2C SUT 1D SUT 2D179470 179440 179520179480 179450 179530

This inspection is not required if the purpose of opening the breakers is to de-energize / tag outthe respective Startup Transformer."



The SRO must have detailed procedure knowledge of 34AB-X43-001-2, Fire Procedure. Thefirst answer is located on page 5 of 114 and the second part is located in the Switchyardprocedure. This level of procedure knowledge is above the RO knowledge level of overallmitigating strategy.

K/A JUSTIFICATION:

This question satisfies the K/A statement by asking the applicant detailed procedure actionslocated inside the plant fire procedure.

"A" is correct.

The "B" distractor is plausible since the first part is correct. The second part is plausible sincethis would be correct if the PCBs are being opened normally to ensure that no open-phaseconditions exist.

The "C" distractor is plausible if the applicant remembers there is a time when this proceduredirects MCREC to be placed in the Pressurization Mode and would be correct if this was one ofthose times. The second part is plausible since it is correct.

The "D" distractor is plausible if the applicant remembers there is a time when this proceduredirects MCREC to be placed in the Pressurization Mode and would be correct if this was one ofthose times. The second part is plausible since this would be correct if the PCBs are beingopened normally to ensure that no open-phase conditions exist.



K/A:


G2.4.27 Knowledge of “fire in the plant” procedures. (CFR: 41.10 / 43.5 / 45.13) . . . . . . . 3.4 3.9



X43-FPS-LP-03601, Fire Protection, Ver. 5.0, EO 200.092.A.01


SRO ONLY Guideline 34AB-X43-001-2, Fire Procedure, Ver. 15.0 34SO-S22-001-2, 500 KV Substation Switching, Ver. 11.0

You have completed the test!Wednesday, April 13, 2016 1:48:54 PM 94

NRC SRO REFERENCES SRO EXAM

1. 34GO-OPS-005-2, Power Changes, Att. 1, Power Versus Flow Map

2. 34AB-E11-001-2, Loss Of Shutdown Cooling, Attachments 1, 2 & 3

3. 34SO-N40-001-1, Main Generator Operation, Attachment 1 Capability Curve

4. U1 TS 3.3.1.1 RPS Instrumentation, Pages 3.3-1 – 3.3-3

5. LFD-1-RPS-09, TS 3.3.1.1-1, Item 4 Reactor Protection System Instrumentation Reactor Vessel Water Level – Low, Level 3

6. U1 TS Table 3.3.1.1-1, RPS Instrumentation

7. U1 TS 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring


POWER CHANGES DOCUMENT NUMBER:

34GO-OPS-005-2 Ver No:


TITLE: POWER VERSUS FLOW MAP 1 of 1

Posted at 2H11-P603



34AB-E11-001-2 Ver No:



Case 1 : CORE FUEL in UNFLOODED RPV (Normal RPV water level)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime

1.0 57.275 52 min 4 hr 50 min 5 hr 42 min 2.0 47.051 1 hr 4 min 5 hr 53 min 6 hr 57 min 3.0 40.644 1 hr 14 min 6 hr 48 min 8 hr 3 min 4.0 36.112 1 hr 23 min 7 hr 40 min 9 hr 3 min 5.0 32.822 1 hr 32 min 8 hr 26 min 9 hr 58 min 6.0 30.372 1 hr 39 min 9 hr 6 min 10 hr 46 min 7.0 28.494 1 hr 45 min 9 hr 42 min 11 hr 29 min 8.0 27.011 1 hr 51 min 10 hr 14 min 12 hr 6 min 9.0 25.806 1 hr 57 min 10 hr 43 min 12 hr 40 min 10.0 24.799 2 hr 1 min 11 hr 9 min 13 hr 11 min 11.0 23.935 2 hr 6 min 11 hr 34 min 13 hr 40 min 12.0 23.177 2 hr 10 min 11 hr 56 min 14 hr 7 min 13.0 22.501 2 hr 14 min 12 hr 18 min 14 hr 32 min 14.0 21.887 2 hr 18 min 12 hr 38 min 14 hr 57 min 15.0 21.323 2 hr 21 min 12 hr 59 min 15 hr 20 min 16.0 20.800 2 hr 25 min 13 hr 18 min 15 hr 44 min 17.0 20.310 2 hr 28 min 13 hr 37 min 16 hr 6 min 18.0 19.850 2 hr 32 min 13 hr 56 min 16 hr 29 min 19.0 19.415 2 hr 35 min 14 hr 15 min 16 hr 51 min 20.0 19.002 2 hr 39 min 14 hr 34 min 17 hr 13 min 21.0 18.610 2 hr 42 min 14 hr 52 min 17 hr 35 min 22.0 18.235 2 hr 45 min 15 hr 10 min 17 hr 56 min 23.0 17.877 2 hr 49 min 15 hr 29 min 18 hr 18 min 24.0 17.535 2 hr 52 min 15 hr 47 min 18 hr 39min 25.0 17.207 2 hr 55 min 16 hr 5 min 19 hr 1 min

NOTE: • Saturation Time is the time differential between the time of complete loss of Shutdown cooling and the

onset of boiling.





34AB-E11-001-2 Ver No:



Case 1 : CORE FUEL in UNFLOODED RPV (cont'd) (Normal RPV water level)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime

26.0 16.893 2 hr 58 min 16 hr 23 min 19 hr 22 min 27.0 16.592 3 hr 2 min 16 hr 41 min 19 hr 43 min 28.0 16.302 3 hr 5 min 16 hr 59 min 20 hr 4 min 29.0 16.024 3 hr 8 min 17 hr 16 min 20 hr 25 min 30.0 15.756 3 hr 11 min 17 hr 34 min 20 hr 46 min 31.0 15.499 3 hr 15 min 17 hr 51 min 21 hr 6 min 32.0 15.251 3 hr 18 min 18 hr 9 min 21 hr 27 min 33.0 15.013 3 hr 21 min 18 hr 26 min 21 hr 47 min 34.0 14.783 3 hr 24 min 18 hr 43 min 22 hr 8 min 35.0 14.562 3 hr 27 min 19 hr 0 min 22 hr 28 min 36.0 14.348 3 hr 30 min 19 hr 17 min 22 hr 48 min 37.0 14.142 3 hr 33 min 19 hr 34 min 23 hr 8 min 38.0 13.944 3 hr 36 min 19 hr 51 min 23 hr 28 min 39.0 13.752 3 hr 39 min 20 hr 7 min 23 hr 47 min 40.0 13.567 3 hr 42 min 20 hr 24 min 24 hr 7 min 41.0 13.388 3 hr 45 min 20 hr 40 min 24 hr 26 min 42.0 13.215 3 hr 48 min 20 hr 57 min 24 hr 45 min 43.0 13.048 3 hr 51 min 21 hr 13 min 25 hr 4 min 44.0 12.886 3 hr 54 min 21 hr 29 min 25 hr 23 min 45.0 12.730 3 hr 57 min 21 hr 44 min 25 hr 42 min 46.0 12.579 4 hr 0 min 22 hr 0 min 26 hr 1 min 47.0 12.432 4 hr 3 min 22 hr 16 min 26 hr 19 min 48.0 12.290 4 hr 6 min 22 hr 31 min 26 hr 37 min 49.0 12.152 4 hr 8 min 22 hr 47 min 26 hr 55min 50.0 12.018 4 hr 11 min 23 hr 2 min 27 hr 13 min

NOTE: • Saturation Time is the time differential between the time of complete loss of Shutdown cooling and

the onset of boiling. • Evaporation Time is the time differential between Saturation Time and the Boil-off Time. • Boil-off Time is the time differential between the time of complete loss of Shutdown cooling and the




34AB-E11-001-2 Ver No:



Case 2 : CORE FUEL in FLOODED RPV(Cavity flooded and gates installed)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime


NOTE:






34AB-E11-001-2 Ver No:



Case 2 : CORE FUEL in FLOODED RPV (cont'd) (Cavity flooded and gates installed)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime


NOTE:






34AB-E11-001-2 Ver No:



Case 3 : CORE and POOL FUEL in FLOODED RPV and POOL(Cavity flooded and gates removed)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime


NOTE:

• Saturation Time is the time differential between the time of complete loss of cooling (Shutdown & Fuel Pool) and the onset of boiling.





34AB-E11-001-2 Ver No:



Case 3 : CORE and POOL FUEL in FLOODED RPV and POOL (cont'd) (Cavity flooded and gates removed)

Days AfterShutdown

Heat Load(MBTU/hr)

SaturationTime

EvaporationTime

Boil-OffTime


NOTE: • Saturation Time is the time differential between the time of complete loss of cooling (Shutdown &

Fuel Pool) and the onset of boiling. • Evaporation Time is the time differential between Saturation Time and the Boil-off Time. • Boil-off Time is the time differential between the time of complete loss of Shutdown cooling and the



MAIN GENERATOR OPERATION DOCUMENT NUMBER:

34SO-N40-001-1 Ver No:


TITLE: OPERATING LIMITS FOR GENERATOR VOLTAGE REGULATOR IN AUTOMATIC 1 of 1

-700

-600

-500

-400

-300

-200

-100

0

100

200

300

400

500

600

700

800

900

0 100 200 300 400 500 600 700 800 900 1000 1100 1200

MEGAWATTS

(LE

AD

) M

EG

AV

AR

S (

LA

15psig 30psig45psig

52psig60psig

(SNC27067)

Appendix D Scenario Outline Form ES-D-1

NRC DRAFT Facility: E. I Hatch Scenario No.: 10-1 Op-Test No.: 2016-301 Examiners: Operators: SRO RO BOP Initial Conditions. Unit 2 is operating at 90% RTP. 2B21-F013D, SRV 2D, is inoperable for LLS, RAS written. RCIC is inoperable, RAS written. “A” Type Secondary Containment with both RF plugs removed. Turnover: IAW 34SO-E51-001-2, unisolate RCIC and place in standby. 34SV-E51-001-2, RCIC Pump Operability, will be performed at the end of the next shift. After RCIC is placed in standby, increase Reactor power to 95% RTP using Recirc.

Event No.

Malf. No. Event Type*

Event Description

1 N/A N (BOP) IAW 34SO-E51-001-2, unisolate RCIC and place in standby.

2 N/A R (ATC) Raise Reactor power to 95% using Recirc.

3 mfB21_130F I (BOP) TS (SRO)

SRV 2F (LLS) cycles open/close until fuses are pulled. (Critical Task)

4 mfC11_31A C (ATC) CRD Flow Control Valve 2A Failure

5 mfG31_242 EGT46-11 EGT46-12

C (BOP) TS (SRO)

Leak in Drywell requiring DW venting. First SBGT started will fail to start, must use opposite train to vent the Drywell.

6 mfB31_41A C (ATC) TS (SRO)

Recirc Pump 2A High Vibration requiring reducing reactor power in an attempt to clear alarm. Alarm remains in until pump is tripped.

7 mfG31_242 M (ALL) Leak worsens in Drywell causes High Drywell pressure scram.

8 diE11-F016A diE11-F016B C (ATC) RHR 2E11-F016A/B stuck closed requiring swapping to other loop

of DW spray (Critical Task)

* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Required Operator Actions Form ES-D-2

Op-Test No.: 2016-301 Scenario No.: 10-1 Event No.: 1 Page 2 of 29

Event Description: IAW 34SO-E51-001-2, unisolate RCIC and place in standby.

Time Position Applicant's Actions or Behavior

10 Min SRO Directs BOP to unisolate RCIC and place in standby IAW 34SO-E51-001-2, RCIC System, starting at step 7.1.1.36.

NOTE: The RCIC Steam Line can be warmed and pressurized by performing either step 7.1.1.36.1 (2E51-F007) OR 7.1.1.36.2 (2E51-F008).

If the BOP chooses step 7.1.1.36.1, (2E51-F007), then the following steps are applicable.

BOP

Pressurizing RCIC steam supply with 2E51-F007:

• Confirm closed 2E51-F007, Steam Supply Isol Valve • Opens 2E51-F008 Steam Supply Line Isol Valve • Opens 2E51-F054 Steam Line Drain Valve • Slowly throttles open 2E51-F007 Steam Supply Isol Valve

• Fully Opens 2E51-F007 when turbine steam inlet pressure (2E51-R602) is within 50 psig of reactor pressure on 2B21-R623A or B (P601 panel)

• Confirms RCIC ISOLATION VLV F007/F008 NOT FULLY OPEN, (602-336) alarm clears

• Closes 2E51-F054, Steam Line Drain Valve, when alarm, RCIC TURBINE INLET DRAIN POT LEVEL HIGH, (602-308), clears

• Completes Attachment 1 and notifies SRO to be verified

BOP Notifies SSS to complete verification of Attachment 1.



Event Description: IAW 34SO-E51-001-2, unisolate RCIC and place in standby.


BOP

NOTE: The RCIC Steam Line can be warmed and pressurized by performing either step 7.1.1.36.1 (2E51-F007) OR 7.1.1.36.2 (2E51-F008).

If the BOP chooses step 7.1.1.36.2, (2E51-F008), then the following steps are applicable.

Pressurizing RCIC steam supply with 2E51-F008:

• Confirm closed 2E51-F008, Steam Supply Line Isol Valve • Opens 2E51-F007 Steam Supply Isol Valve • Opens 2E51-F054 Steam Line Drain Valve • Slowly throttles open 2E51-F008 Steam Supply Line Isol Valve

• Fully Opens 2E51-F008 when turbine steam inlet pressure (2E51-R602) is within 50 psig of reactor pressure on 2B21-R623A or B (P601 panel)

• Confirms RCIC ISOLATION VLV F007/F008 NOT FULLY OPEN, (602-336) alarm clears

• Closes 2E51-F054, Steam Line Drain Valve, when alarm, RCIC TURBINE INLET DRAIN POT LEVEL HIGH, (602-308), clears NOTE: Alarm 602-308 is not expected to be in alarm due to RCIC out of service time.

SRO May notify SSS to complete verification of Attachment 1.

Simulator Operator, at the Chief Examiners direction, PROCEEDS to the next event.



Event Description: Raise Reactor power to 95% using Recirc.


10 Min SRO • Directs ATC to increase reactor power to 95% by increasing Recirc flow. Power increases are NOT to exceed 10 MWe/min.

NOTE: May get the RBM UPSCALE, (603-202) and ROD OUT BLOCK, (603-238) alarm, if a peripheral control rod is NOT selected. This is expected and the operator MAY select a peripheral rod at this time. MAY also get Alarm HEATER TROUBLE, (650-135), alarm. This is expected at this power level.

ATC

• IAW 34SO-B31-001-2 (step 7.1.5) & 34GO-OPS-005-2, the ATC increases Recirc pump speed, NOT to exceed 10 MWE per minute by depressing the RAISE SLOW or RAISE MEDIUM pushbuttons on the Master (P603 panel) or Individual controls (P602 panel) until reactor power is 95%.

• If using Individual Controls, pump speed increases will alternate between

the “A” & “B” Recirc pumps to prevent excessive flow mismatches. • Monitors power increase by observing APRM and generator output

indications.

ATC

• Complies with 34SO-B31-001-2 , Limitation 5.2.15, which states:

WHEN changing Recirc pumps speed while in Two Loop operation maintain pump speeds to limit recirculation loop jet pump mismatch within the following limits:

• <10% of rated core flow (7.7 E6 lbm/hr) WHEN operating

< 70% of rated core flow; AND

• <5% of rated core flow (3.85 E6 lbm/hr) WHEN operating at > 70% of rated core flow.

ATC Notifies the SRO that reactor power has been increased to 95%.

Simulator Operator, at the Chief Examiners direction OR after power has been increased to 95%, PROCEEDS to the next event.



Event Description: SRV 2F (LLS) cycles open/close until fuses are pulled.


10 Min

Simulator operator: At the Chief Examiner’s direction, INSTRUCT the ATC operator by phone to stay on the line until told to hang up, then ENTER (RB-3) mfB21_130F, Main Steam Relief Valve Fails Open. ENSURE Event Trigger EGB21-12 ACTIVATES. This SRV will cycle such that it is open for 15 seconds, then stays closed for 45 seconds, then repeats this cycle for 5 minutes or until fuses for SRV 2F are removed.

ALL • Receive alarms:

• SAFETY/BLOWDOWN VALVE LEAKING, (603-122) • SAFETY BLOWDOWN PRESSURE HIGH, (602-311)

SRO

Directs operator to enter 34AB-B21-003-2, Failure of Safety/Relief Valves

• Enters a RAS for TS LCO 3.6.1.6, Condition A, for 2 or more LLS valves inop, which requires the unit to be in mode 3 in 12 hours and mode 4 in 36 hours.

BOP

• Enters 34AB-B21-003-2, Failure of Safety/Relief Valves • Determines SRV 2F is cycling open then close • Cycles the SRV 2F Control Switch several times • May depress the ADS Logic A Timer Reset pushbutton (2B21-S2A) • May depress the ADS Logic B Timer Reset pushbutton (2B21-S2B) • Depresses the LLS Channel A / C Reset pushbutton (2B21-S15A) • Depresses the LLS Channel B / D Reset pushbutton (2B21-S15B) • Informs SRO that SRV 2F is cycling and the fuses will have to be

removed for the valve • Notifies the SSS to remove the fuses for SRV 2F. (Critical Task)

BOP Receives DRYWELL/TORUS RCDR R627 TEMP HIGH, (650-204) alarm

• Determines Torus Water Temp (Point 2) is high at 98°F • When the SRV is closed Point 2 starts decreasing.





Simulator Operator, if requested to remove fuses for SRV 2F, wait 4 minutes, then DELETE mfB21_130F, 2F LLS Valve or Scenario EGB21-12, then ENTER (RB-1), rfB21_305, to simulate removing the fuses for SRV 2F. Then, NOTIFIES the crew that the fuses have been removed for SRV 2F.

BOP

• Confirms that SRV 2F is closed by monitoring one or more of the following: • SRV tailpipe temperature decreasing (Panel 2H11-P614) • Torus water level increase has stopped • Torus temperature increase has stopped • Reactor and Generator power return to the pre-event level • Resets the SRV leak detection by placing the Leak Detection Logic A

Reset keylock switch and Leak Detection Logic B Reset keylock switch to Reset position and back to Normal position

• Confirms that the Amber SRV indicating lights have Extinguished • SAFETY BLOWDOWN PRESSURE HIGH, (602-311), clears • Informs the SRO that SRV 2F is closed

SRO/ BOP

Informs the crew that operability of the suppression chamber-drywell vacuum breakers must be performed within 12 hours per 34SV-T48-002-2, Suppression Chamber To Drywell Vacuum Breaker System Operability. Notifies Chemistry and initiates a CR to initiate increased monitoring of vessel moisture content carryover per 64CH-SAM-025-0.

At this time, Torus temperature will still be below 95°F, therefore RHR is NOT required to be placed into Torus Cooling. The SRO may elect to place Torus Cooling in service, since steam was admitted to the Torus. The following steps are written IF the SRO elects to place Torus Cooling in service.





NOTE: The operator may place torus cooling in service by using the Placard that's available or using the appropriate section of the procedure. These steps assume the Placard is used. The A or B loop of RHR may be used. The following steps are written assuming "B" loop and "B" pump is used. If "A" loop is used, substitute "A" for "B" for valves and if "B" pump is not used substitute "A", "C", or "D" for "B" pump.

BOP • Enters 34SO-E11-010-2, Residual Heat Removal • Places RHRSW in service • Prelube RHRSW pump • Overrides 2E11-F068B Low Discharge Pressure Interlock • Positions 2E11-F068B to 45% OPEN • Starts RHRSW pump B • Places 2E11-F068B Low Discharge Pressure Interlock switch to

normal position. • Positions 2E11-F068B to obtain < 4400 gpm AND < 450 psig

BOP • IF desired to start a SECOND RHRSW pump, • Throttles 2E11-F068B to achieve max flow rate

(not to exceed 4400 GPM). • Opens 2E11-F068B an additional 5%. • Starts second RHRSW Pump. • Positions 2E11-F068B to obtain < 8800 gpm AND < 450 psig





BOP • Places RHR B Loop in Torus cooling per the placard by performing the following steps: • Opens 2E11-F048B • Closes 2E11-F047B. • Opens 2E11-F003B. • Starts RHR Loop B pump • Receives alarm, RHR LOW FLOW, (601-222) • Opens 2E11-F028B • Receives alarm, AUTO BLOWDOWN CS OR RHR PRESS

PERMISSIVE, (602-312) • Receives alarm, SEC SYSTEM AUTO INITIATION SIGNAL

PRESENT, (650-234) • Throttles OPEN 2E11-F024B • Alarm, RHR LOW FLOW, (601-222), clears • Opens 2E11-F047B • Ensures RHR flow is < 11,500 GPM, THEN Closes 2E11-F048B • Notifies the SRO that RHR "B" loop is in service • May place the second pump in service.




Event Description: CRD FCV 2A Slow Failure; Places standby FCV in service


12 Min Simulator operator: At the Chief Examiner’s direction, ENTER (RB-4) mfC11_31A, CRD Flow Control Valve Failure.

ALL • Receives the following annunciator:

• CRD HYDRAULIC TEMP HIGH, (603-140)

ATC

• Determines that the CRD Flow Control Valve A has failed closed • Notifies SRO that the CRD Flow Control Valve A has failed closed • Enters:

• CRD HYDRAULIC TEMP HIGH, (603-140) • 34AB-C11-001-2, Loss of CRD System

NOTE: With the loss of the CRD FCV, the ATC may FIRST address

34AB-C11-001-2, Loss of CRD System.

SRO

• Dispatches a SO & Maintenance to investigate the cause of the FCV failing closed

• Directs the operator to perform actions per the ARP and 34AB-C11-001-2, Loss of CRD System

• May direct operator to swap Flow Control Valves at this time, or may wait for report from SO, before ordering the swap

ATC • IAW CRD HYDRAULIC TEMP HIGH, (603-140):

• Dispatches a SO to monitor CRD temperatures locally at 2C11-R018 (130 Rx Bldg)



Event Description: CRD FCV 2A Slow Failure; Places standby FCV in service


Simulator Operator, wait 2 minutes and report, as SO at the FCV, that the FCV has failed closed due to an air leak on the supply line, and that you have isolated the air line and stopped the leak. Simulator Operator, IF/WHEN requested and if CRD HYDRAULIC TEMP HIGH is illuminated, as the SO checking CRD temps, report that 2 CRD drives are >250°F; 30-11 and 22-35 at 260°F and some others are slowly increasing.

SRO Directs operator to shift Flow Control Valves

ATC

• Enters 34SO-C11-005-2, CRD Hydraulic System, section 7.1.6, to shift Flow Control valves: • Confirms 2C11-R600 is in AUTO • Directs the SO to locally shift CRD FCV to B IAW step 7.1.6.1.2

through 7.1.6.1.9.

Simulator Operator, ENTER (RB-2) rfC11_24, CRD Flow Control F002 to “B”, then as the SO, report that the CRD FCV’s have been shifted locally.

Simulator Operator, If asked, as SO, INFORMS ATC that local controller is in Automatic.

ATC

• Returns CRD System to normal configuration: • Places 2C11-R600 to MANUAL • Adjusts 2C11-R600 to desired flow (~50 gpm) • Places 2C11-R600 to AUTO • Confirms system parameters are normal

ATC • CRD HYDRAULIC TEMP HIGH, (603-140) alarm clears • Reports to SRO that CRD flow control valves have been shifted and the

CRD System has been restored to normal configuration




Event Description: Leak in Drywell requiring DW venting. First SBGT started will fail to start, must use opposite train to vent the Drywell.


10 Min Simulator operator: At the Chief Examiner’s direction, ENTER (RB-5) mfG31_242, RWCU Non-Isol Leak in Drywell (final 0.02, ramp 100).

ALL Recognizes increasing Drywell Pressure from the following alarm:

• PRIMARY CNMT PRESSURE HIGH, (603-115)

SRO

• Directs an operator to enter 34AB-T23-002-2, Small Pipe Break Inside Primary Containment.

• When venting is required; Directs the BOP to vent the drywell with BOTH loops of CAD.

Simulator Operator, if asked to perform 34SV-SUV-019-2, Leakage Determination, INFORM that you will get another operator to perform the determination and will contact you back later for assistance.

ATC

• IAW 34AB-T23-002-2, performs the following using Placard: • Confirms automatic actions by:

• Trips RWCU Pump “2B” • Places control switch for 2G31-F001 to close • Places control switch 2G31-F004 to the close • Enters 34AB-G31-001-2, RWCU System Isolation • Notifies SRO of conductivity monitoring requirements of

TRM T3.4.1.

NOTE: EVENT TRIGGERS EGT46-11 & EGT46-12 will INSERT overrides to keep the first SBGT fan from starting and then remove the event trigger from the opposite SBGT train. The applicant will not be successful with the first SBGT train; however, the second SBGT train will work.





BOP

• Enters 34SO-T48-002-2, “Containment Atmosphere Dilution System” or uses placard to vent the Drywell.

• Enters 34SO-T46-001-2, “Standby Gas Treatment System” procedure or

uses placard to start SBGT 2A or 2B. • Opens 2T46-F001 (A or B) and/or 2T46-F003 (A or B). • Places SBGT Fan (A or B) control switch to “RUN • Receives alarm 2B (OR 2A) SBGT SWITCH NOT IN AUTO,

(654-076) or (657-091) • Determines that the first SBGT fan will not start • At this time, may inform SRO that SBGT (A or B) cannot be started or

may wait until other SBGT is in service (either is acceptable).

BOP

(Placard)

• Vents with the opposite train of SBGT per 34SO-T46-001-2 SBGT System procedure or uses placard. • Opens 2T46-F001 (A or B) and/or 2T46-F003 (A or B) for the

subsequent train. • Places SBGT Fan (A or B) control switch to RUN. • Receives alarm 2B (OR 2A) SBGT SWITCH NOT IN AUTO,

(654-076) or (657-091) • Confirms 2T46-F002 (A or B) OPENS • Confirms SBGT Heater red light illuminates.

BOP

(Placard)

• Opens 2T48-F334A or 2T48-F334B (both valves may be opened) • Receives alarm DW VENT EXH BYPASS VLV OPEN, (654-002) or

(657-008) • Receives alarm DRWL/TORUS N2 M/U 2 INCH ISOL VALVES OPEN,

(654-035) or (657-042) • Opens 2T48-F335A or 2T48-F335B (both valves may be opened) • Opens 2T48-F336A or 2T48-F336B. (both valves may be opened) • Monitors DW pressure • Notifies SRO DW venting is in progress and of the SBGT failure (if not

notified previously)





SRO

• Reviews TS 3.6.4.3, Standby Gas Treatment System • IAW TS 3.6.4.3 Condition B, • Declares SBGT fan A or B inoperable, • Must restore SBGT fan A or B to operable status within 7 days AND

30 days from discovery of failure to meet the LCO. • Direct that maintenance be contacted to determine problem with failed

SBGT fan. • IAW TRM T3.4.1 performs conductivity monitoring requirements.




Event Description: Recirc Pump 2A High Vibration requiring reducing reactor power in an attempt to clear alarm. Alarm remains in until pump is tripped.


18 Min

Simulator operator: At the Chief Examiner’s direction, contact BOP and INFORM BOP to remain on the phone, and then ENTER (RB-6) mfB31_41A, Recirc Pump A High Vibration.

ALL • Receives the following annunciator:

• RECIRC PUMP A HIGH VIBRATION, (602-104)

Simulator operator: If dispatched to the ASD locally, wait two minutes and report, as SO, that you have not found anything yet, but will continue monitoring.

ATC

• IAW RECIRC PUMP A HIGH VIBRATION, (602-104): • Depresses the High vibration reset pushbutton and determines that the

alarm does NOT clear. • Notifies the SRO that the vibration alarm did NOT clear. • Notifies the SRO that the ARP requires reducing Recirc flow UNTIL

the alarm will clear when the reset pushbutton is depressed.

SRO

• Directs the operator to reduce Reactor Power with Recirc flow per 34GO-OPS-005-2 and 34SO-B31-001-2, section 7.1.6, Two Loop Operation From Rated To Minimum Speed.

• Notifies Plant Management, Load Dispatcher, and Engineering that the power increase has been halted pending investigation of Recirc Pump 2A high vibration condition.

NOTE: IAW 34GO-OPS-005-2, Power Changes, Limitation 5.2.3, entry into the Immediate Exit Region is ALLOWED during a manual OR automatic power reduction initiated in response to an equipment problem.

NOTE: IAW 34GO-OPS-005-2, Power Changes, a rapid power reduction

may be performed at a rate > 10 MWE/min to put the plant in a safe condition.





ATC

• IAW 34GO-OPS-005-2, Power Changes: • Decreases Recirc Pump speed per 34SO-B31-001-2, Reactor Recirc

System.

• IAW 34SO-B31-001-2, Reactor Recirc System: • Decreases Recirc pump speed by depressing the desired LOWER

pushbuttons on the Master or Individual controls.

ATC

• Periodically stops the power reduction and depresses the High vibration reset pushbutton.

• Monitors power decrease by observing APRM and generator output indications.

• Notifies SRO when BOTH Recirc Pump speeds are at minimum. • Notifies the SRO that the vibration alarm DID NOT clear after reducing

BOTH Recirc pumps. • Notifies SRO of entering Region of Potential Instabilities/Immediate Exit

Region of Power to Flow Map.

NOTE: RECIRC PUMP A HIGH VIBRATION, (602-104) will NOT remain

clear UNTIL the 2A Recirc Pump is tripped.

BOP IAW 34GO-OPS-005-2, Power Changes, Step 5.2.14, notifies chemistry of 15% power change in 1 hour.

NOTE: HEATER TROUBLE ALARM, (650-135) may alarm due to plant conditions.

NOTE: 15% power change sample required IAW 34GO-OPS-005-2,

Limitation 5.2.14.





SRO

• Directs the ATC operator to shutdown ASD A IAW 34SO-B31-001-2, section 7.2.1.3, Abnormal Recirc Pump/ASD A Shutdown.

• As time allows, directs the BOP operator to secure one Condensate and one Condensate Booster Pump plus one RFPT.

ATC

• IAW 34SO-B31-001-2, Reactor Recirc System: • Places ASD A control switch 2B31-S002A to Pull to Lock

OR depresses the ASD A Shutdown pushbutton and places ASD A control switch to Pull to Lock on panel 2H11-P602.

• Enters 34AB-B31-001-2, Reactor Recirculation Pump(s) Trip, Recirc Loops Flow Mismatch, Or ASD Cell Bypass, for single Recirc pump trip. • Closes 2B31-F031A, Pump Discharge Valve.

Time: _________

• After 4-5 minutes, throttles 2B31-F031A, Pump Disch Valve OPEN

Time: _________

ALL

• Receives the following annunciators: • ASD A TRIP WARNING, (602-101) • ASD A FATAL FAULT, (601-102) • ASD A TROUBLE, (601-108) • RECIRC LOOP A OUT OF SERVICE, (601-127)

NOTE: IAW 34AB-B31-001-2, During single loop operation, WHEN the speed of the running pump decreases below approximately 35% speed, positive flow through the idle pump loop due to natural circulation overcomes the negative flow due to reverse flow. The total core flow summing circuitry will continue to subtract this positive idle loop flow from the running loop flow AND give a misleading LOW core flow indication. Total core flow can be calculated by adding the JET PUMP LOOP "A" AND the JET PUMP LOOP "B" flows.





ATC • Determines that the plant is in the Immediate Exit Region of the Power to Flow map.

NOTE: AFTER the trip of the 2A Recirc pump, at the Chief Examiners direction, PROCEED to the MAJOR EVENT. IF the crew makes the decision to SCRAM the Reactor, at the Chief Examiners direction, PROCEED to the MAJOR EVENT.

EXAMINER NOTE: Log time when Region of Instabilities/Immediate Exit

Region has been ENTERED. Time: _________

NOTE: IAW 34GO-OPS-005-2, Power Changes, Limitation 5.2.3.1:

If the ‘Region Of Potential Instabilities’ (RPI) is entered, IMMEDIATELY initiate actions to exit the (RPI), to return operation to the Analyzed Region of the Power/Flow Map outside of the RPI.

If the Immediate Exit Region is entered, within 5 minutes the operators are to initiate control rod movement to return to the Analyzed Region.

Operation within the Analyzed Region AND outside of the RPI must be restored within 1 hour.

SRO

• Directs ATC operator to insert rods to exit the RPI and Immediate Exit Region to return operation to the Analyzed Region of the Power/Flow Map.

• References Tech Spec 3.4.1.A.1 and has 24 hours to meet requirements for Single Loop Operation.

NOTE: RBM Downscale alarm may alarm during this movement due to the

significant rod worth of these rods. The RBM Downscale and Rod Out Block alarms may be flagged.





ATC

• Inserts control rods per 34GO-OPS-065-0, Control Rod Movement: • Inserts control rods per the Reactivity Briefing Sheet and rod pull

sheets. • Selects first Rod to be inserted • Places Control Rod movement switch to the IN position • Verifies Rod moves using Rod display information and Rx and

Generator power decreasing. • Releases Rod movement switch so that the control rod stops 1 position

before the insert limit, unless the insert limit is 00. • Initials Rod movement Sheet. • Verifier, if available, Initials Rod movement sheet. • Notifies the SRO when they are out of the RPI.

• If required, adjust 2C11-F003 to get 220 – 280 psid drive water dp

EXAMINER NOTE: Log time when Region of Instabilities/Immediate Exit

Region has been EXITED. Time: _________

BOP

• Removes condensate pumps (CP & CBP) from service prior to CBP discharge pressure < 525 psig OR power <70% OR as soon as practical. • For removal of CBP:

• If 2C is to be removed, removes the Hydrogen Injection System from service

• Sends SO to closed selected pumps discharge valve • Prior to discharge valve being full closed, trips pump and places

control switch in Auto or PTL • Has SO complete the procedure

• For removal of CP: • Sends SO to closed selected pumps discharge valve • Trips pump and places control switch in Auto or PTL • If pump left in Standby, has SO reopen discharge valve





NOTE: IAW 34GO-OPS-005-2, WHEN Feedwater flow is less than

7 mlbm/HR AND two Reactor Feed Pumps are running, one Reactor Feed Pump MAY be shutdown

BOP

Enters 34SO-N21-007-2, Condensate And Feedwater System, section 7.2.1, First Reactor Feed Pump Shutdown and Leaving in Standby.

• Confirms Feedwater Flow is less than 7 Mlbm/hr. • Confirms RFPT 2A AND RFPT 2B are in Automatic control on 2C32-

R600, Master Controller. • Places 2C32-R601A (2C32-R601B), RFP A (B) M/A Station, in

Manual, by depressing the 'M' pushbutton until it illuminates, panel 2H11-P603.

• Slowly decrease RFPT 2A (2B) speed with RFP A (B) M/A Station until the other RFP is controlling RWL.

NOTE: At this point the operator may stop here with the RFPT NOT

injecting and continue with this section as time allows.

BOP

• When the other RFP has control of water level, slowly decrease RFPT 2A (2B) speed with RFP A (B) M/A Station until no speed decrease is observed AND/OR place the RFPT A (B) TMR switch to SS AND confirm Speed Setter yellow light illuminates.

• Slowly lower RFPT 2A (2B) Speed Setter switch until RFPT speed is at 1000 rpm, at 2H11-P650.

• IF desired, reduce the RFPT 2A (2B) speed to minimum AND allow the RFPT to “windmill”, provided seal water, steam seals, AND lube oil systems remain in service.

Simulator Operator, at the Chief Examiners direction, PROCEEDS to the MAJOR EVENT.



Event Description: Leak worsens in Drywell causes High Drywell pressure scram


10 Min

Simulator Operator, at the Chief Examiners direction, MODIFY mfG31_242, RWCU Non-Isol Leak in Drywell:

• Final 1.0 • Ramp 1000

ALL

Recognizes increasing Containment Pressure from the following alarms: • PRIMARY CNMT HIGH PRESSURE TRIP, (603-106) • PRIMARY CNMT PRESSURE HIGH, (603-115) • DRYWELL PRESSURE HIGH, (602-210)

NOTE: The SRO may direct the BOP to vent the Drywell, but there will

NOT be time to complete the task before the LOCA signal is received.

SRO

• With Drywell pressure increasing and alarms PRIMARY CNMT PRESSURE HIGH, (603-115) and DRYWELL PRESSURE HIGH, (602-210), alarms being received, directs the ATC enter 34AB-C71-001-2, Scram Procedure, and Scram the reactor.

• Assigns the ATC to perform RC-1 • Assigns the BOP operator to perform RC-2 and RC-3 • Enters 31EO-EOP-010-2, RC (Non ATWS) &31EO-EOP-012-2, PC

Control, EOP flow charts





Simulator Operator, WHEN the reactor is scrammed, ENSURES EVENT TRIGGER EGC71-12 MODIFIES mfG31_242 Final to 1.5 with a ramp of 1000 AFTER 180 seconds.

ATC

• Performs RC-1 consisting of: • Inserts a manual scram. • Places the mode switch to SHUTDOWN. • Confirms all rods are inserted by observing full in lights, SPDS, or the

RWM display. • Notifies SRO of rod position check. • Places SDV isolation valve switch to "isolate" & confirms closed. • If NOT tripped, places the Recirc pumps at minimum speed. • Inserts SRMs and IRMs. • Shifts recorders to read IRMS, when required. • Ranges IRMS to bring reading on scale. • Notifies the SRO when the above actions are complete.

BOP

• Performs RC-2 actions consisting of: • Confirms proper Level Control response:

• Checks ECCS Injection Systems and secure as necessary. • Ensures FW Master Controller setpoint reduces to 9 inches and output

reduces to 25% of previous value. • IF set down does NOT auto function, then manually reduces FW

Master Controller setpoint to approximately 9 inches.





BOP

• Controls HPCI operation for level control by performing one or more of the following:

• Adjusts 2E41-R612, HPCI Flow Control, to the desired injection rate

• Transfers the flow controller to manual and adjust its speed demand output to obtain the desired pump flow

• Shutdown HPCI by: • Depresses and holds the HPCI Turbine Trip pushbutton • When HPCI turbine has stopped, places 2E41-C002-3, HPCI

Aux Oil Pump, in Pull-To-Lock • When HPCI TURBINE BRG OIL PRESS LOW alarm is

received, releases the HPCI Turbine Trip push-button.

BOP

• When feed flow is less than the capacity of the S/U level control valve (≈ 1.5 mlbm/hr), then: • Opens 2N21-F125. • Confirms 2C32-R619, FW S/U level control valve controller, in Auto,

set at approximately 9 inches. • Closes 2N21-F110. • May attempt to restart the CRD pumps • May attempt maximize CRD flow IAW 34SO-C11-005-2 • Controls RWL with the HPCI/FW System. • Notifies SRO if RWL gets outside assigned band.

NOTE to EXAMINER: SRVs actuate in LLS at 1120 psig and then control pressure between 850 - 990 psig.

BOP

• Performs RC-3 consisting of: • Monitors RPV pressure. • If necessary, allows RPV pressure to exceed 1074 psig then cycles any

SRV to initiate LLS. • If necessary, verifies LLS actuates at 1120 psig • Maintains RPV Pressure between 1074 and 800 psig. • Notifies SRO of pressure control system operation.





NOTE: The SRO may select a RPV pressure band which will lower the

driving head of the leak while maintaining <100°F/hr RPV cool down (typically between 500 psig & 920 psig).

SRO

As time allows, may; • Direct the ATC to decrease reactor pressure to reduce the driving head of

the leak using EHC pressure set. • Remove RWCU from service IAW 34AB-T23-002-2, Small Pipe Break

in Primary Containment.

ATC

Enters Attachment 11 of 34SO-N30-001-2, Main Turbine, or 34GO-OPS-013-2, Normal Plant Shutdown, and at the DEHC panel computer, performs ONE of the following:

1. Throttle Pressure Set

• Selects the Control → psi- load screen

• Selects the *Ramp Rate* button

• Enters a ramp rate

• Selects the *Pressure* button

• Enters desired target pressure 2. Bypass Valve Jack Positioning

• Control BPV position by intermittently using the *Raise* / *Lower* buttons until BYPASS VALVE JACK STATUS changes to *ACTIVE* AND desired cooldown rate is established

ATC Notifies the SRO that RPV Pressure is at the target psig.

Simulator Operator PROCEEDS to the next event at the Chief Examiner’s direction.



Event Description: RHR 2E11-F016A/B stuck closed requiring swapping to other loop of DW spray


SRO • Per the PC flowchart, verifies Torus level is <285 inches and directs an

operator to place Torus Sprays in service.

ATC

• Sprays the Torus per 34SO-E11-010-2 placard on the 2H11-P601 Panel as follows: • Places Cnmt Spray Vlv Cntl switch in the MANUAL position. • Starts RHR pump(s) in loop A/B, if NOT already running. • Opens 2E11-F028A/B. • Throttles Open 2E11-F027A/B. • Notifies SRO that RHR is in Torus Sprays.

NOTE: Only one loop of RHR will be placed in Torus Sprays. The flow is only 700 gpm.

Simulator Operator, ENSURE Event Triggers EGE11-4 & EGE11-5 is

ACTIVATED when the operator positions 2E11-F016A or B to open.

SRO

• When Torus pressure exceeds 11 psig, verifies that Torus Level is <215 inches, in the safe area of Graph 8 (DWSIL) and directs an operator to: • Place all DW cooling fans to OFF • Place both Recirc pumps to PTL OFF





ATC

• Places both Recirc pumps to PTL OFF on panel 2H11-P602. • Places the following DW cooling fans control switches to OFF on panels

2H11-P654 & P657: • 2T47-B007B, Drywell Cooling Top Head Area Unit • 2T47-B008B, Drywell Cooling Pedestal/Annular Area Unit • 2T47-B009B, Drywell Cooling Recirc Pump Area Unit • 2T47-C001B, Drywell Cooling Return Air Fan • 2T47-C002B, Drywell Cooling Return Air Fan • 2T47-B010B, Drywell Cooling EL 114 Unit

• 2T47-B007A, Drywell Cooling Top Head Area Unit • 2T47-B008A, Drywell Cooling Pedestal/Annular Area Unit • 2T47-B009A ,Drywell Cooling Recirc Pump Area Unit • 2T47-C001A, Drywell Cooling Return Air Fan • 2T27-C002A, Drywell Cooling Return Air Fan • 2T47-B010A, Drywell Cooling EL 114 Unit

SRO

• When all DW cooling fans are OFF AND both Recirc pumps are PTL OFF: • Directs an operator to spray the Drywell

ATC

• Sprays the Drywell per 34SO-E11-010-2 placard on the 2H11-P601 Panel as follows: • Places Cnmt Spray Vlv Cntl switch in the Manual position • Starts RHR pump(s) in loop A (B), if NOT already running • Opens 2E11-F021A or B • Opens 2E11-F016A or B (ONE WILL NOT OPEN AND OPERATOR

TRANSITIONS TO THE OTHER LOOP) • Informs SRO that the 2E11-F016A (or B) will NOT Open

• Opens 2E11-F021A or B • Throttles Open 2E11-F016A(B) to >5000 gpm (Critical Task) • Confirms Drywell pressure is reducing • Notifies SRO that RHR is in Drywell Sprays





BOP

Notifies the SSS to perform actions for RHR per 31EO-EOP-114-2. IAW 31EO-EOP-114-2, the operator performs the following:

• CLOSES RHR OUTBD INJ VLV, 2E11-F017A • CLOSES RHR OUTBD INJ VLV, 2E11-F017B • Notifies SSS to OPEN links & INSTALL jumpers for 2E11-F017A • Notifies SSS to OPEN links & INSTALL jumpers for 2E11-F017B • Confirms/CLOSES RHR OUTBD INJ VLV, 2E11-F017A • Confirms/CLOSES RHR OUTBD INJ VLV, 2E11-F017B • Confirms/CLOSES INBD DISCHARGE VLV, 2E21-F005A • Confirms/CLOSES INBD DISCHARGE VLV, 2E21-F005B • Trips Core Spray pump A, 2E21-C001A • Trips Core Spray pump A, 2E21-C001B

Notifies SRO 31EO-EOP-114-2 actions for RHR & CS are complete

SRO • As time allows, directs H2/O2 Analyzers placed in service IAW

34SO-P33-001-2.

ATC

• If directed, places H2/O2 Analyzers in service IAW 34SO-P33-001-2 or “Placard” by performing the following at 2H11-P700 panel: • Confirms closed 2P33-F605 • Places 2P33-S16, LOCA Override to ‘Bypass” • Places 2P33-S17, LOCA Override to ‘Bypass” • Confirms analyzers are running by either red analyzer lights illuminated or

values indicated on the Primary Display of SPDS • If analyzers red light is off, depresses Channel A and Channel B Reset

pushbuttons on 2H11-P700 panel. • Notifies SRO H2/O2 Analyzers are running.

With Chief Examiners Permission, the scenario should be terminated when the crew has sprayed the Drywell with the other Loop of RHR.

of 29 Appendix D Scenario Outline Form ES-D-1

NRC DRAFT Facility: E. I Hatch Scenario No.: 10-1 Op-Test No.: 2016-301 Examiners: Operators: SRO RO BOP

Initiating Conditions: Unit 2 is operating at 90% RTP. 2B21-F013D, SRV 2D, is inoperable for LLS,

RAS written. RCIC is inoperable, RAS written. “A” Type Secondary Containment with both RF plugs removed.

Turnover IAW 34SO-E51-001-2, unisolate RCIC and place in standby. 34SV-E51-001-2, RCIC Pump Operability, will be performed at the end of the next shift. After RCIC is placed in standby, increase Reactor power to 95% RTP using Recirc.

Summary: • Event 1: Normal; IAW 34SO-E51-001-2, unisolate RCIC and place in standby.

• Event 2: Reactivity; The ATC will raise Reactor power to 95% RTP.

• Event 3: Instrument; SRV 2F (LLS) cycles open and close until fuses are removed.

(Critical Task)

• Event 4: Component; The CRD FCV 2A will slowly fail causing a loss of CRD. The operator will take manual actions to swap FCV and return the CRD system to normal.

• Event 5: Component/TS; A leak in the Drywell causes DW pressure to increase. The team will receive a P603 alarm that will direct DW venting to be placed in service. BOP operator will attempt to start Standby Gas Treatment (SBGT). The first SBGT system the operator attempts to start will fail. The opposite SBGT train will then be started and DW venting will be aligned. The SRO will address TS for an inoperable train of SBGT.

• Event 6: Component/TS; Recirc Pump 2A will experience high vibration requiring reducing reactor power in an attempt to clear the alarm. Alarm remains in until pump is tripped. The SRO addresses TS for single loop operation. The plant will be operating in the Immediate Exit Region of the Power-to-Flow map. The ATC operator will insert control rods to exit the Immediate Exit Region of the P/F map.

• Event 7: Major; Pipe leak in Drywell worsens causing a High Drywell pressure scram.

• Event 8: Component; When Torus pressure exceeds 11 psig, the operator will attempt to spray the Drywell but one RHR DW spray valve will NOT open. The other loop of RHR will be used to spray the Drywell. The first DW spray valve attempted will NOT open but the other loop of RHR DW spray valve will work. (Critical Task)

of 29 Critical Tasks

NRC DRAFT Facility: E. I Hatch Scenario No.: 10-1 Op-Test No.: 2016-301 Critical Tasks

• SRV 2F (LLS) valve cycles open and close until the fuses are removed which will preclude a failure that will result in challenging Primary Containment integrity if manual action is not taken in a timely manner (prior to exceeding the BIIT curve) or challenge exceeding the Tech Spec cooldown rate limit. (Event 3)

• RHR 2E11-F016A (B) is stuck closed requiring the operator to swap to other loop of RHR to spray the Drywell while in the safe region of the Drywell Spray Initiation Limit (DSIL), prior to exceeding 340°F. (Event 8)

ES 301-4 Attributes Required Actual Items 1. Total Malfunctions 5-8 6 1. SRV 2F cycles open and closed (Event 3)

2. CRD FCV 2A failure (Event 4) 3. SBGT 2A(2B) failure (Event 5) 4. Recirc Pump 2A high vibration (Event 6) 5. RWCU pipe break LOCA (Event 7) 6. 2E11-F016A(B) fails to open (Event 8)

2. Malfunctions After EOP Entry

1-2 1 1. 2E11-F016A(B) fails to open (Event 8)

3. Abnormal Events 2-4 4 1. SRV 2F cycles open and closed (Event 3) 2. CRD FCV 2A failure (Event 4) 3. SBGT 2A(2B) failure (Event 5) 4. Recirc Pump 2A high vibration (Event 6)

4. Major Transients 1-2 1 1. RWCU pipe break LOCA (Event 7) 5. EOPs entered,

requiring substantive actions

1-2 2 1. 31EO-EOP-010-2, RC (Non ATWS) (Event 7) 2. 31EO-EOP-012-2, PC (Event 7)

6. EOPs contingencies requiring substantive actions

0-2 0 N/A

7. EOP Based Critical Tasks

2-3 2 1. Remove fuses for SRV 2F (Event 3) 2. Spray the Drywell (Event 8)

ILT 10 NRC DRAFT Scenario 1

SHIFT TURNOVER

Safety Focus

UNIT 1 STATUS

Plant Conditions: Unit 1 is operating at 100% RTP.

UNIT 2 STATUS

Plant Conditions:

• Unit 2 is operating at 90% RTP. • RCIC has been isolated for maintenance. Maintenance is

complete and RCIC is ready to be placed in Standby. • “A” Type Secondary Containment with both RF plugs removed.

Protected Train: EOOS:

Division I

Division II

Green Yellow

Orange

Red

Scheduled

evolutions :

� IAW 34SO-E51-001-2, unisolate RCIC and place in standby. � After RCIC is placed in standby, increase Reactor power to 95%

RTP using Recirc (No fuel preconditioning limitations).

Surveillances due

this shift :

� NONE; 34SV-E51-001-2, RCIC Pump Operability, will be performed at the end of the next shift.

Inop Equipment: � 2B21-F013D, SRV 2D, is inoperable for LLS, RAS written

� RCIC, RAS written

Active tagouts : � 2B21-F013D, SRV 2D

Rod Configuration : � See RWM


NRC DRAFT Facility: E. I Hatch Scenario No.: 10-2 Op-Test No.: 2016-301 Examiners: Operators: SRO RO BOP Initial Conditions. Unit 2 is operating at 100% RTP. 2B21-F013D, SRV 2D, is inoperable for LLS, RAS written. Turnover: IAW 34SO-N62-001-2, swap Cooler Condensers from 2N62-B003A to 2N62-B003B, starting at step 7.2.2.

Event No.


Event Description

1 N/A N (BOP) Swap Cooler Condensers from 2N62-B003A to 2N62-B003B IAW 34SO-N62-001-2, starting at step 7.2.2.

2/3 aoN40R600 mf65111604 mf65111605

C (BOP) R (ATC)

UAT 2B Hi temp/lower reactor power/swap house loads/remove from service

4 mf65021482 (ON) mfP41_292

C (ATC) TS (SRO)

2B PSW pump overload – manually trip 2B PSW pump & start 2D PSW pump.

5 mf65702215 C (BOP) RF Vent Filter 2T41-D007 high dP requiring swapping to D008

6 mfE51_114 diE51A-S17

C (ATC) TS (SRO) RCIC Inadvertent start with Trip pushbutton failure.

7

mfE51_250 svoE51074 svoE51075 diT41-B009 diT41-B026

M (ALL) Unisolable RCIC Steam leak in Reactor Building requiring a Reactor Manual scram. RCIC Group 4 signal failure.

8 N/A M (ALL) Emergency Depress when Max Safe exceeded in more than one area. (Critical Task).

9 mfE11_202B mfE21_202A mfN21_99

C (BOP) RHR & CS LOCA logic failure, SULCV closed – manual actions are required for proper RHR/CS & Cond/FW operation ensuring adequate core cooling. (Critical Task)




Event Description: IAW 34SO-N62-001-2, swap Cooler Condensers from 2N62-B003A to 2N62-B003B, step 7.2.2.


10 Min SRO Directs BOP to swap Cooler Condensers from 2N62-B003A to 2N62-B003B IAW 34SO-N62-001-2, step 7.2.2.

The BOP will perform the following at 2N62-P001 panel. Simulator Operator, at the Chief Examiner’s direction, when BOP is waiting for 15 minutes to elapse before opening 2N62-F025B, inform the BOP using Time Compression that 15 minutes has elapsed. NOTE: ALARM 600-020, Inlet Flow To Stack High, may alarm when 2N62-

F025B is opened. This is normal for this condition and the alarm will clear in approximately one minute.

BOP

• Monitors 2N62-R605, Glycol Pump Disch pressure indicator, and starts the idle 2N62-C001B, Glycol Sys Pump B, as necessary.

• Opens the following valves: • 2N62-F026B, Glycol Sys To Cndsr B • 2N62-F071B, Glycol Sys From Cndsr B • 2N62-F025B, Clr Cndsr B Inlet

BOP

• Closes the following valves: • 2N62-F026A, Glycol Sys To Cndsr A • 2N62-F071A, Glycol Sys From Cndsr A • 2N62-F025A, Clr Cndsr A Inlet • If the second Glycol pump was started, stops 2N62-C001B, Glycol Sys

Pump B • Confirms Glycol Pump discharge pressure remains at 10 to 20 psig, on

2N62-R605, Glycol Pump Disch pressure • Notifies SRO that the Cooler Condensers have been swapped

Simulator Operator – Continue with the next event at the Chief Examiners request.



Event Description: UAT 2B Hi temp requiring reactor power to be lowered and the removal from service.


15 Min

Simulator Operator At Lead Examiner’s direction, ACTIVATE: (RB-2)

• mf65111604, Unit Aux Xfmr “2B” Misc Alarm (Annunciator On) • aoN40R600, Gen & Transformer Temps, final of 110, ramp 100

THEN, 2 minutes later ACTIVATE: (RB-1)

• mf65111605, Unit Aux Xfmr “2B” Winding Temp High (Annunciator On)

ALL

UNIT AUX XMFR 2B MISC ALARM, (651-116), annunciates Two minutes later, UNIT AUX XMFR 2B WINDING TEMP HIGH, (651-117) annunciates.

BOP

• Responds to alarm UNIT AUX XMFR 2B MISC ALAR, (651-116) • Notifies GCC of the alarm. • Dispatches an operator to check the transformer local panel in the Low

Voltage Switchyard, 2H21-P214. • Responds to alarm UNIT AUX XMFR 2B WINDING TEMP HIGH,

(651-117) • Notifies the SRO that per the ARPs Rx Power will need to be reduced if

the winding temperature is increasing and to Transfer the loads to Startup Transformer (SUT) 2C, if the oil temperature is high.





SIMULATOR OPERATOR:

• 3 minutes from being dispatched, call the control room as the operator dispatched to the Unit 2B UAT and report that the 2B UAT oil temperature is 93°C and that the winding temperature is 106°C and slowly increasing. If asked, Transformer fans and oil pumps are running.

• DO NOT PROMPT TO TRANSFER LOADS

• If more updates of temperature are required, increase temps 1°C/min

EACH UPDATE, until load is transferred. • Temps will eventually stabilize above the setpoints.

When the UAT is unloaded, temps will slowly decrease, however the afore mentioned alarms will NOT clear until locally reset.

SRO • Directs BOP operator to enter 34SO-R22-001-2 for transferring 4160VAC

buses from Unit Aux Transformer (UAT) 2B to Start-Up transformer (SUT) 2C.

BOP • Enters 34SO-R22-001-2, 4160 VAC System

SRO • May direct the ATC to lower Rx power to LOWER the load on UAT 2B.

NOTE: If the SRO does not direct the lowering of reactor power, SIMULATOR OPERATOR as the Shift Manager in consultation with Switchyard Maintenance, NOTIFY the SRO to reduce reactor power to ~ 90% RTP due to increasing UAT 2B temperatures.

NOTE: The SRO may direct the operator to exceed 10MWe/min. It is allowable to enter the immediate exit region for protecting equipment, but in this particular event the power reduction should be completed prior to entering the immediate exit region.





SRO When reactor power is reduced Event #3, directs the BOP to swap 2A & 2B 4160V buses from UAT to the SAT.

BOP

Swapping of 2A 4160VAC from the UAT to the SAT • Confirms power is available to Startup Aux XFmr 2C as indicated by the

potential lights on panel 2H11-P651 • Confirms OPEN ACBs 135544, 135564 and 135584 (2H11-P652)

BOP

• Places 135434/135454 Station Svc Interlock Cutout switch in OFF-(DOWN)

• Places Sync Switch (SSW) ACB 135454 in ON • Confirms the sources of power to 4160V Bus 2A are synchronized and

voltage is normal on Start-Up Aux Transformer 2C

BOP

• Closes ACB 135454, 4160V Bus 2A Alternate Supply, AND confirms that current increases from Startup Auxiliary Transformer 2C

• Trips ACB 135434, 4160V Bus 2A Normal Supply • Places Sync Switch (SSW) ACB 135454 in OFF • Places 135434/135454 Station Svc Interlock Cutout switch in NORMAL-

(UP)

BOP

Swapping of 2B 4160VAC from the UAT to the SUT • Confirms power is available to Startup Aux XFmr 2C as indicated by the

potential lights on panel 2H11-P651 • Confirms OPEN ACBs 135544, 135564 and 135584 (2H11-P652)

• Places 135444/135464 Station Svc Interlock Cutout switch in OFF-(DOWN)

• Places Sync Switch (SSW) ACB 135464 in ON • Confirms the sources of power to 4160V Bus 2B are synchronized and

voltage is normal on Start-Up Aux Transformer 2C





BOP

• Closes ACB 135464, 4160V Bus 2B Alternate Supply, AND confirms that current increases from Startup Auxiliary Transformer 2C

• Trips ACB 135444, 4160V Bus 2B Normal Supply • Places Sync Switch (SSW) ACB 135464 in OFF • Places 135444/135464 Station Svc Interlock Cutout switch in NORMAL-

(UP) • Notifies the SRO that 4160 VAC 2A and 2B buses have been transferred

from the UAT to SUT 2C.

SIMULATOR OPERATOR, after UAT 2B is removed from service, MODIFIES Override aoN40-R600 to a Final of 50 with a 5 ramp rate.

SIMULATOR OPERATOR, at the Chief Examiners direction, PROCEEDS to the next event.



Event Description: Reduce Reactor power to ~ 90% RTP to remove UAT 2B from service.


NOTE: This reactivity change will occur as a result of the need to reduce

power to place the 2A and 2B 4160VAC buses on SAT 2C (Event 2).

NOTE: The SRO may direct the operator to exceed 10MWe/min. It is allowable to enter the immediate exit region for protecting equipment, but in this particular event the power reduction should be completed prior to entering the immediate exit region.

ATC

• Reduces Rx power using Recirc per SRO direction in Event #2 (approx 90% power).

• Enters the following procedures • 34GO-OPS-005-2, “Power Changes” • 34SO-B31-001-2, “Recirculation System”

ATC

• Depresses either the Master Recirc Flow Control Lower Fast/Medium/Lower Slow pushbuttons or the individual pump Speed Control Lower Fast/Medium/Lower Slow pushbuttons.

• Monitors power decrease by observing APRM and generator output indications.

• Monitors the Power to Flow map to determine if the Immediate Exit Region is entered.

• Notifies the SRO of the proximity to the Immediate Exit region.

NOTE: When Reactor power has been reduced to ~ 90% RTP return to Event

2 for removing loads from 2B UAT.

SRO

• If the plant is in the Immediate Exit Region, then: • Conducts a rod movement briefing. • Directs the ATC operator to insert control rods to exit the Immediate

Exit Region.

ATC • Attends the rod movement briefing. • Inserts control rods to exit the Immediate Exit Region

IAW 34GO-OPS-065-2.

Simulator Operator, WAIT until Event 2 (removing UAT 2B from service) is complete and at the Chief Examiner’s request, PROCEED to the next event.



Event Description: 2B PSW pump overload – manually trip2B PSW pump & start 2D PSW pump.


10 Min

At the Chief Examiner’s direction, Simulator Operator; Contact BOP and instruct to stay on the phone until told to hang, then ENTERS (RB-4) malfunction mf65021482 – Window 20 PSW PUMP 2B OVLD/LOCKOUT RELAY TRIP (ANNUNCIATOR ON).

ALL • Annunciator PSW PUMP 2B OVLD/LOCKOUT RELAY TRIP, (650-220) alarms

ATC

• Announces alarm to the SRO • Enters ARP 650-220 • Determines that the 2B PSW pump is still running • Informs the SRO that the 2B PSW pump failed to trip

SRO

• Directs the BOP to Trip the 2B PSW pump and place the 2D PSW pump is service.

• Notifies Maintenance (if BOP has NOT) to investigate 2B PSW pump. • Enters Tech Specs LCO 3.7.2 and determines that a 30 day RAS is

required IAW TS 3.7.2.A.

ATC

• Trips PSW Pump 2B And Verifies Green light illuminates • WHEN PSW Div. II pressure, as indicated on 2P41-R601B reaches 95

PSIG (decreasing), panel 2H11-P650, STARTS PSW Pump 2D. (Standby pump start failure may not be recognized if PSW 2D is started before low header pressure is received)

• As time allows, may enter 34AB-P41-001-2, Loss of Plant Service Water. • Notifies Maintenance (if SRO has NOT) to investigate 2B PSW pump.




Event Description: RF Vent Filter 2T41-D007 high dP requiring swapping to D008.


8 Min

Simulator Operator, at the direction of the Chief Examiner, ENTERS: (RB-5) malfunction mf65702215 RF D007 Filter Diff Pressure High alarm and override 2T41-R618P4. EGT41-2 will be used to clear alarms after filter swap.

ALL • On the 2H11-P650 panel alarm, PANEL 2H11-P657 SYSTEM TROUBLE, (650-224), is received.

NOTE: This failure will require entry into the SC EOP Flow Chart as well as

two Abnormal procedures. The following sequence may be slightly out of order because of this.

If the operator enters 657-036 first, then the following actions will be addressed, otherwise skip to the next alarm.

BOP

• Acknowledges and enters, RF VENT FLTR 2T41-D007 DIFF PRESS HIGH, (657-036) and performs the following: • Dispatches a SO/Maintenance to confirm the Refuel Floor Vent Filter

differential pressure is > 7” WC as indicated on 2T41-dpi-R004A. SIMULATOR OPERATOR, when dispatched to confirm Refuel Floor Vent Filter differential pressure is > 7” WC as indicated on 2T41-dpi-R004A, wait two minutes and REPORT, as SO, that differential pressure is > 10” WC and Maintenance requests that the filter be removed from service for replacement IAW the section for replacing the filter (4.3.7).

• Confirms Refueling Floor Exhaust Flow is < 30 KCFM as indicated by the purple pen on Rx Bldg/Refuel Floor HVAC Supply/Exhaust Flow Recorder, 2T41-R618. (Actual flow will be low, ~20,000)

• Confirms 2T41-F041B, Refuel Flr Vent Filter D008 Inlet Damper, is OPEN.

• Notifies OATC to record the following differential pressure in the Plant Operator’s log: • 2T41-D007 Pre-Filter as indicated on 2T41-dpi-R015A • 2T41-D007 Carbon Filter as indicated on 2T41-dpi-R015B • 2T41-D004 HEPA Filter as indicated on 2T41-dpi-R015C

• Notifies SRO of above alarm.





If the operator enters 657-001 first, then the following actions will be addressed.

BOP

• Acknowledges, RF INSIDE TO OUTSIDE AIR DIFF PRESS LOW, (657-001), alarm and enters (657-001) and performs the following actions:

• Confirms Refueling Floor to outside air differential pressure, is greater than -0.06" WC, as indicated on 2T46-R604B, Sec Cnmt Diff Press B, black pen, 2H11-P700.

• Determines that the 2T41-C005A, ‘RF Exhaust’ fan Filter dP is high (+0).

BOP

• IF an abnormal radioactive release is occurring, OR has occurred, enters 34AB-D11-001-2, Radioactivity Release Control. (Should not enter this one but okay if they do)

• Informs the SRO that the 2T41-C005A, ‘RF Exhaust’ fan Filter dP is high and that RF Bldg dP is >0”.

• Dispatches a SO/Maint to confirm 2R25-S064, BRKR 19 is closed.

• Enters 34AB-T22-002-2, ‘Loss of Secondary Containment Integrity’ and 34AB-T22-003-2, ‘Secondary Containment Control’, and notifies Management of the plant conditions.

BOP

• Enters 34AB-T22-003-2, ‘Secondary Containment Control’. • Monitors Secondary Containment parameters. • Notifies the SRO to enter 31EO-EOP-014-2, EOP Secondary

Containment flowchart due to low RF dP. • Dispatches a SO/Maint to investigate the low RF dP. • Enters 34AB-T22-002-2, ‘Loss of Secondary Containment Integrity’.





BOP • Enters 34SO-T41-006-2 step 4.3.7 to remove 2T41-D007, Refueling Floor

Vent Filter, from service for Maintenance and performs the following.

• Places the STBY 2T41-C002B, Refuel Flr Vent Supply Fan, control switch, in OFF

• Places the in service 2T41-C002A, Refuel Flr Vent Supply Fan, control switch, in OFF

• Places 2T41-C005B, Refuel Flr Vent Exh Fan, control switch, in OFF • Places 2T41-C005A, Refuel Flr Vent Exh Fan, control switch, in OFF • Dispatches SO to close 2T41-F115A, 2T41-D007 Inner Door • Dispatches SO to close 2T41-F115B, 2T41-D008 Inner Door

SIMULATOR OPERATOR, when dispatched to close 2T41-F115A & B, wait two minutes and REPORT, as SO, that 2T41-F115A & B are closed.

• Places 2T41-C005B, Refuel Flr Vent Exh Fan, control switch, in RUN. • Confirms 2T41-C005B, Refuel Flr Vent Exh Fan, starts. • Places 2T41-C002A or 2T41-C002B, Refuel Flr Vent Supply Fan,

control switch, in RUN. • Dispatches a SO to adjust 2T41-R032 and 2T41-R037, Flow

Controller, to obtain approximately 15 KCFM through 2T41-C002A Refueling Floor Supply Fan AND 2T41-C005B, Refueling Floor Exhaust Fan.

Simulator Operator, when dispatched, DEPRESS RB-3, which modifies RF Supply flow indication, THEN waits approximately 2 minutes and reports to the BOP, as a SO, that you have adjusted 2T41-R032 and 2T41-R037, to approximately 15 KCFM.

• Confirms 0.25 inches water negative pressure on the Refueling Floor as indicated on2T46-R604A AND 2T46-R604B, Sec Cnmt Diff Press A&B, on 2H11-P700.

• Closes 2T41-F041A, Refuel Flr Vent Filter D007 Inlet Damper. • Notifies SRO Vent Filter D007 is out of service and 657-036 has

cleared.





Simulator Operator waits 2 minutes and reports to the SRO, as a SO, that Breaker 19 at 2R25-S064 is closed.

SRO

• Dispatches a SO/Maint to determine why the RF Fan D007 experienced a high dP.

• Enters 31EO-EOP-014-2, EOP Secondary Containment flowchart.

Simulator Operator proceeds to the next event at the Chief Examiner’s direction.



Event Description: RCIC Inadvertent start with Trip pushbutton failure.


6 Mins Simulator Operator: At the direction of the Chief examiner, ACTIVATE: (RB-6) to ENTERS mfE51_114 and diE51A-S17 to “off”.

ALL • Recognizes that RCIC has started.

ATC • Determines RCIC has auto started and that RWL is normal.

SRO • May tell ATC that RWL is normal • Directs operator to trip RCIC

ATC

• Attempts to Trip RCIC by depressing the RCIC Trip pushbutton and recognizes that the Trip pushbutton is failed.

• Notifies the SRO that the RCIC trip pushbutton has failed and Trips RCIC by ONE of the following methods: • Closes the Trip and Throttle valve, 2E51-F524 OR • Places controller 2E51-R612 to Manual and reduces output to lower

RCIC discharge pressure to below reactor pressure. • Receives RCIC TURBINE BRG OIL PRESS LOW, (602-304) &

RCIC PUMP DISCHARGE FLOW LOW, (602-322)



Event Description: RCIC Inadvertent start with Trip pushbutton failure.


ATC

• Enters 34AB-E10-001-2, Inadvertent Initiation of ECCS/RCIC • Enters 34SO-E51-001-2, RCIC System • Dispatches RO/Maintenance to determine cause of initiation signal and the

cause of the Trip pushbutton failure. • May attempt to reset the Initiation signal • Will close 2E51-F524, Trip and Throttle Vlv, if not already closed. • Notifies SRO that RCIC is shutdown.

SRO

• May have the operator run the Trip and Throttle Valve down to in case RCIC is needed later.

• Enters TS RAS for RCIC 3.5.3 Condition A, which requires verifying HPCI is operable within 1 hour and restoring RCIC in 14 days.

NOTE: It is intended that RCIC is left in its’ current condition and not

returned to standby.




Event Description: Unisolable RCIC Steam leak in Reactor Building requiring a Reactor Manual scram. RCIC Group 4 signal failure.


15 Min

At the Chief Examiner’s direction, Simulator operator, ENTER (RB-7) mfE51_250, RCIC Steam Line, break 70/3.5, diT41-B009 & diT41-B026 to off. NOTE: SVOs svoE51074 (2E51-F007 Stuck Open) and svoE51075 (2E51-

F008 Stuck Open), are activated at the beginning. NOTE: It takes approximately 3 minutes for the first alarm, LEAK DET

DIFF TEMP HIGH, (601-321), to alarm.

ALL

• Receives the following: • LEAK DET DIFF TEMP HIGH, (601-321) in 3 minutes • RCIC ISOL TIMER INITIATED, (602-303) in 3.5 minutes • LEAK DET AMBIENT TEMP HIGH, (601-327) in 4 minutes • RCIC ISOLATION SIGNAL LOGIC A INITIATED, (602-307) after

timer times out • RCIC ISOLATION SIGNAL LOGIC B INITIATED, (602-313) after

timer times out

SRO

• Orders BOP to evaluate leak detection alarms on 2H11-P601. • Orders RCIC to be isolated. • Orders BOP to evacuate the Reactor Building. • May notify Maintenance for assistance in closing RCIC valves if

ATC/BOP does not.





ATC/BOP

• Responds to RCIC alarms • Observes RCIC Isolation valves have failed to close. • Places 2E51-F007, RCIC Isolation valve switch to Close. • Places 2E51-F008, RCIC Isolation valve switch to Close. • Notifies SRO of RCIC valve failures. • May notify Maintenance for assistance in closing RCIC valves if SRO

does NOT.

BOP

• Respond to annunciator LEAK DET DIFF TEMP HIGH, (601-321). • Addresses 2G31-R604 OR 2G31-R608 on 2H11-P614. • Identifies the following points on R604 increasing:

• 113, TORUS NW WALL is ~ 112°F • 114, TORUS SE WALL is ~ 97°F • 115, TORUS VENT AIR DIFF is ~ 31°F • 116, TORUS VENT AIR DIFF is ~ 18°F

• Identifies the following points on R608 increasing: • 115, TORUS WEST WALL is ~ 112°F • 116, TORUS NE WALL is ~ 98°F • 117, TORUS VENT AIR DIFF is ~ 31°F • 118, TORUS VENT AIR DIFF is ~ 18°F • 119, MAIN STEAM TNL ~ 140°F

• Reports temperatures to the SRO.





SRO

• Enters the SC flow chart. • Progresses down each path. • Proceeds down the SC/T path, directing:

• All available area coolers are operated • Operate the refueling floor ventilation • Operate the reactor building ventilation • Isolate all systems discharging into the area except those needed for

ACC, shutdown the reactor, suppress a fire, maintain primary containment

SRO

• Orders the reactor shutdown before any area exceeds Max Safe operating temperatures or delta temps since a primary system (RCIC) is discharging into Secondary Containment. (May direct this prior to Max Safe received).

SRO

• Assigns the ATC to perform RC-1. • Assigns the BOP operator to perform RC-2 and RC-3. • Enters 31EO-EOP-010-2, RC EOP flow chart if RWL decreases below 3

inches or if any area exceeds Max Safe. • Directs RWL Band of 3 to 50 inches.





ATC

• Performs RC-1 consisting of: • Inserts a manual scram. • Places the mode switch to shutdown. • Confirms all rods are inserted by observing full in lights, SPDS, or the

RWM display. • Notifies SRO of rod position check. • Places SDV isolation valve switch to "isolate" & confirms closed. • If not tripped, places the Recirc pumps at minimum speed. • Inserts SRMs and IRMs. • Shifts recorders to read IRMS, when required. • Ranges IRMS to bring reading on scale. • Notifies the SRO when the above actions are complete.

BOP


• Checks ECCS Injection Systems (will not be in service if RWL >-35 inches)

• Ensures FW Master Controller setpoint reduces to 9 inches and output reduces to 25% of previous value

• If Set down does not auto function, manually reduces FW Master Controller setpoint to approximately 9 inches.

• When feed flow is less than the capacity of the S/U level control valve (≈ 1.5 mlbm/hr), then: • Confirms/Opens 2N21-F125. • Confirms/places 2C32-R619, FW S/U level control valve

controller, in Auto, set at approximately 9 inches. • Closes 2N21-F110.

• If RFPTs are no longer available, will transition to HPCI to control RWL.





BOP

• If necessary, starts HPCI for level control by performing the following at 2H11-P602 panel: • If required, depresses High Water Level Reset P/B • Opens 2E41-F059 • Starts Barom Cndsr Vac Pump • Opens 2E41-F001 • Starts Aux Oil Pump • Opens 2E41-F006 • Confirms TCV and • Confirms/Closes 2E41-F012 at flow > 790 gpm • Adjusts controller for desired flow and with SRO permission will

raise RWL to 32 to 42 inches

BOP

• Performs RC-3 consisting of: • Monitoring RPV pressure. • Confirms proper operation of pressure control system (TBV, LLS,

etc.), at 2H11-P650 panel, by confirming TBVs are responding to control reactor pressure at the desired pressure setpoint.

• Maintains RPV pressure between 1074 and 800 psig until a different band is directed.

• Notifies SRO of pressure control system operation.

• If the MSIVs are closed, then the operator will perform the following: • Confirms closed Inboard MSIVs (2B21-F022A-D) and

places control switches to close • Confirms closed Outboard MSIVs (2B21-F028A-D) and

places control switches to close

• If the MSIVs are still open, SRO orders “Anticipate Emergency Depress, then the operator will perform the following: • At P650 HMI screen,

• Selects “Control” • Selects “Bypass Valve” • Inserts “Ramp Rate” of 100 then OK • Inserts BPV position of 100 then OK • Ensures Bypass Valve Jack Status is Active (controlling) • Notifies SRO that Bypass Valves are opening





SRO

• May order a lower Reactor pressure band to reduce the driving head. • May direct an operator to perform Rx Power, Level, and Pressure control,

so that the other operator can address Secondary Containment parameters.

Simulator operator PROCEEDS to the next event.



Event Description: Emergency Depress when Max Safe exceeded in more than one area.


ATC

• Provides periodic updates on temperature readings and delta temp readings to the SRO.

• Reports R604 points 115 is above Max Safe. • Reports R608 points 120 & 117 are above Max Safe. • When a second temperature or delta temp exceeds Max Safe levels, informs

the SRO.

SRO • May order a lower Reactor pressure band to reduce the driving head. • Transitions to CP-1 and orders 7 ADS valves open for Emergency

Depress.

ATC

• Provides periodic updates on temperature readings and delta temp readings to the SRO.

• When a second temperature or delta temp exceeds max safe levels, informs the SRO.

SRO • Transitions to CP-1 and orders 7 ADS valves open for emergency depress.



Event Description: Emergency Depress when Max Safe exceeded in more than one area.


ATC/BOP

• Places the switches for 7 ADS valves in the open position. (Critical Task – Open 7 SRVs when Max Safe exceeded in more than one area. Critical task is met when 5 SRVs have been opened).

• The amber lights for the SRVs will not illuminate if pressure has been

reduced to below approximately 300 psig. In this case the operator must use 2H11-P614 recorder indication to monitor tail pipe temperatures for the SRVs to verify the valves actually opened (Recorder 2B21-R614).

• Depending on Reactor Water Level prior to opening ADS valves, RWL

may swell to above 60 inches, requiring the operator to enter 34AB-C32-001-2, Reactor Water Level Above 60 inches. • Operator secures all injection other than CRD.

Simulator operator the next event was activated at the beginning of the scenario.



Event Description: RHR & CS LOCA logic failure, SULCV closed – manual actions are required for proper RHR/CS & Cond/FW operation ensuring adequate core cooling.


NOTE: mfE11_202B (LOCA Signal failure) & mfE21_202A (LOCA Signal

failure) was inserted at beginning.

BOP

• At panel 2H11-P601, Confirms the 4 RHR pumps did NOT auto start • If a LOCA signal is received, notifies SRO of RHR pump logic failure and

starts (Critical Task is met if one RHR pump is started) the 4 RHR pumps. • At panel 2H11-P601, Confirms 2 CS pumps did NOT auto start and starts

(Critical Task is met if one CS pump is started) the CS pumps





Simulator Operator, At the Chief Examiner’s direction AND AFTER the ADS valves have been opened (Event 8); when Reactor pressure is < 500 psig, ENSURE EGB21-5 INSERTS mfN21_99, SULCV fails closed. (Critical Task) – Throttle open 2N21-F110 for vessel injection to maintain RWL greater than -180 inches. If 2N21-F110 is NOT used, then a NEW Critical Task will be to align RHR/Core Spray systems for vessel injection to maintain RWL greater than -180 inches. EITHER IS ACCEPTABLE.

BOP

• The BOP recognizes the SULCV has failed closed.

• Performs ONE or MORE of the following to inject water into the reactor. • Throttles open Condensate System, 2N21-F110, and injects water into the

reactor (Critical Task).

NOTE: mfE21_202A (LOCA Signal failure) inserted at beginning.

OR • If a LOCA signal is received, notifies SRO of CS pump logic failure and

starts 2A (2B) Core Spray Pump. (Critical Task) • Confirm discharge pressure >265 psig. • Confirm 2E21-F031A(B) closes when flow >950 gpm. • Confirm a CS and RHR room cooler automatically starts. • When RPV pressure is <425 psig, throttles open 2E21-F005A(B).





NOTE: mfE11_202B (LOCA Signal failure) inserted at beginning.

OR • At panel 2H11-P601, Confirms the 4 RHR pumps did NOT auto start and

ONLY 2 Core Spray pumps are running. • If a LOCA signal is received, notifies SRO of RHR pump logic failure and

starts (Critical Task) is met if one RHR pump is started) the 4 RHR pumps.

• When RPV pressure is <425 psig, throttles open 2E11-F015A (B).

• Reports SULCV problem to the SRO.

With Chief Examiners Permission, the Scenario will be terminated when reactor water is being controlled following the Emergency Depressurization, with reactor pressure below 50 psig or as directed by the Chief Examiner.



Initiating Conditions: Unit 2 is operating at 100% RTP. 2B21-F013D, SRV 2D, is inoperable for LLS, RAS written.

Turnover IAW 34SO-N62-001-2, swap Cooler Condensers from 2N62-B003A to 2N62-B003B, starting at step 7.2.2.

Summary: • Event 1: Normal; Swap Cooler Condensers from 2N62-B003A to 2N62-B003B

IAW 34SO-N62-001-2.

• Event 2: Component; UAT 2B will experience a Hi temperature requiring the removal from service. The ATC will lower Reactor power prior to removing the UAT from service. Once power is ~ 90% RTP, the BOP will remove the UAT from service.

• Event 3: Reactivity; The ATC will lower Reactor power prior to removing the UAT from service. Once power is ~ 90% RTP, the BOP will remove the UAT from service.

• Event 4: Component/TS; 2B PSW pump overload requiring manual trip of 2B PSW pump & starting 2D PSW pump.

• Event 5: Component; RF Vent Filter 2T41-D007 high dP requiring swapping to D008.

• Event 6: Component/TS; RCIC will experience an inadvertent start with Trip pushbutton failing to trip RCIC. Operator will shutdown RCIC by either; closing T&TV or placing flow controller in manual and lowering speed to prevent injection.

• Event 7: Major; Unisolable RCIC Steam leak in Reactor Building requiring a Reactor Manual scram.

• Event 8: Major; Emergency Depress when > Max Safe in more than one area. (Critical Task)

• Event 9: RHR & CS LOCA logic failure, SULCV closed – manual actions are required for proper RHR/CS & Cond/FW operation ensuring adequate core cooling. (Critical Task)


NRC DRAFT Facility: E. I Hatch Scenario No.: 10-2 Op-Test No.: 2016-301 Critical Tasks • Emergency Depress when > Max Safe in more than one area. (Event 8)

• SULCV closed, RHR & CS LOCA logic failure – manual actions are required for proper RHR/CS

& Cond/FW operation ensuring adequate core cooling. (Event 9)

ES 301-4 Attributes Required Actual Items 1. Total Malfunctions 5-8 1. UAT 2B Hi temp (Event 2)

2. 2B PSW pump overload (Event 4) 3. RF Vent Filter 2T41-D007 high dP (Event 5) 4. RCIC inadvertent start (Event 6) 5. Unisolable RCIC Steam leak (Event 7) 6. RHR & CS LOCA logic failure, SULCV closed

(Event 9) 2. Malfunctions After

EOP Entry 1-2 1 1. RHR & CS LOCA logic failure, SULCV closed

(Event 9) 3. Abnormal Events 2-4 3 1. 2B PSW pump overload (Event 4)

2. RF Vent Filter 2T41-D007 high dP (Event 5) 3. RCIC inadvertent start (Event 6)

4. Major Transients 1-2 2 1. Unisolable RCIC Steam leak (Event 7) 2. Emergency Depress (Event 8)

5. EOPs entered, requiring substantive actions

1-2 2 1. RC (Non-ATWS) (Event 7) 2. SC/RR (Event 7)


0-2 1 1. CP-1 (Event 8)


2-3 2 1. Emergency Depress (Event 8) 2. Manual actions for adequate core cooling.

(Event 9)


SHIFT TURNOVER

Safety Focus

UNIT 1 STATUS


UNIT 2 STATUS



Division I

Division II

Green Yellow

Orange

Red

Scheduled evolutions : � IAW 34SO-N62-001-2, swap Cooler Condensers from 2N62-B003A to 2N62-B003B, starting at step 7.2.2.

Surveillances due this

shift :

� None


Active tagouts : � 2B21-F013D, SRV 2D



NRC DRAFT Facility: E. I Hatch Scenario No.: 10-3 Op-Test No.: 2016-301 Examiners: Operators: SRO RO BOP Initial Conditions. Unit 2 is operating at 60% RTP. 2B21-F013D, SRV 2D, is inoperable for LLS, RAS written. Turnover: Continue startup and place RFPT 2A in service, starting at step 7.1.11.2.08 of 34SO-N21-007-2. Once RFPT 2A is in service, notify Maintenance to perform walk down of RFPT 2A piping. Once Maintenance walk down is complete, increase Reactor power to 75% RTP.

Event No.


Event Description

1 N/A N (BOP) Continue startup and place RFPT 2A in service, starting at step 7.1.11.2.08 of 34SO-N21-007-2.

2

mfP51_222C loP51-C001BGl loP51- C001BR2 diP51- C001B

C (BOP) SSAC 2C Trips - 2B fails unloaded, 2A must be manually started.

3 mfE51_250 svoE51075

C (ATC) TS (SRO)

RCIC Steam Line breaks in the Rx bldg. RCIC isolation valves fail to auto close (Critical Task)

4 mf60131136 aoE21R600B

C (BOP) TS (SRO)

2B Loop of Core Spray experiences high discharge pressure (valve leakage). Perform ARP & restore.

5 svoB21036 mf60211154 mfC12_26_22-27

I (ATC) TS (SRO)

One (1) Reactor Pressure ATTS trip unit causes a half scram and Control Rod 22-27 scrams in due to a blown fuse. The control rod fuse is repaired (Time Compress) and the rod is withdrawn.

6 mf65702209 mf65702227 mf65321987

M (ALL) Earthquake requiring scram prior to 98” in Torus with a loss of Unit 1 SAT 1D.

7 N/A R (ATC) Reduce reactor power to between 40% & 50% due to earthquake.

8

svoT48140(70/.75) svoT48142(50/10) svoT48143(50/10) svoT48147(50/10) svoT48148(50/10)

M (ALL) Emergency Depress prior to Torus level reaches 98”. (Critical Task)

9

mfB21_129A mfB21_129E mfB21_129L mfB21_129M mfB21_129C mfB21_129K mfB21_129H

C (ATC) ALL ADS valves fail to open when Emergency Depress is required, EOP-108. (Critical Task)




Event Description: Place the RFPT ‘2A’ in service, starting at step 7.1.11.2.8 of 34SO-N21-007-2.


15 Mins

SRO Orders Operator to place RFPT ‘2A’ in service, starting at step 7.1.11.2.8 of 34SO-N21-007-2.

BOP • Enters 34SO-N21-007-2 at step 7.1.11.2.8.

• Confirms M/A station is tracking actual Speed Setter (RFPT) speed • Places the TMR Mode switch to M/A • Confirms the M/A Station green light illuminates

• Slowly changes the RFPT 2A M/A station to match RFP 2A AND the RFP 2B flow match.

• Matches the input AND output of Pump A M/A Station by performing the following: • Depresses the PF key AND read the controller output (PF lamp lit) • Depresses the PF key so the input to the controller is displayed (PF

lamp is off) • Adjusts the manual output lever until the input AND output are

matched on P603 panel. • Monitors RWL, RFPT 2A discharge pressure and RFPT 2A & 2B speed.

• Places RFP A M/A station in AUTO • As required, adjusts RFP A Speed Control Bias Setting to maintain RFPT

2A and 2B speed within 100 RPM

• Informs SRO that the 2A RFPT is in service in Automatic control.

SRO • Informs ATC to increase power with Recirc to 75% power.

Simulator operator proceeds to the next event at the Chief Examiner’s direction



Event Description: SSAC 2C Trips - 2B fails unloaded, 2A must be manually started.


7 Min

Simulator Operator, at direction of the Chief Examiner, ACTIVATE: (RB-2) malfunction mfP51_222C. ENSURE Event Trigger EGP51-3 ACTIVATES the following when the "B" air compressor “red” light is on: loP51-C001BG1 to OFF, loP51-C001BR2 to ON, mfP51_222B, EGP51-4. ONCE "2B"SSAC is placed in PTL, ENSURE Event Trigger EGP51-4 DELETES: loP51-COO1BG1, loP51-C001BR2, and mfP51_222B.

All The following annunciators are received:

• PANEL 2H11-P700 SYSTEM TROUBLE, (650-225) • AIR CMPSR 2C TRIPPED/SHUTDOWN, (700-233)

BOP

• Acknowledges 2H11-P700 panel alarm on 2H11-P650 • Acknowledges AIR CMPSR 2C TRIPPED/SHUTDOWN, (700-233), on

2H11-P700 and notifies the SRO of the alarm. • Confirms/starts “2B” SSAC on P650 panel. • Determines air pressure is continuing to decrease • Starts “2A” SSAC on P650 panel. • Dispatches SO/Maintenance locally to investigate “2C” SSAC tripping and

“2B” SSAC not maintaining air pressure. • As time allows, contacts Health Physics after starting Standby SSACs. • May receive CONTROL BLG SERVICE AIR PRESS LOW, (700-222).



Event Description: SSAC 2C Trips - 2B fails unloaded, 2A must be manually started.


Simulator Operator: 2 minutes after being dispatched to investigate SSAC 2C tripping and SSAC 2B not maintaining air pressure (unless SSAC 2A was started), INFORM the BOP; SSAC 2C has a low oil pressure shutdown indication AND IF SSAC 2B IS STILL RUNNING that SSAC 2B is running unloaded and you cannot locally get it to “LOAD”. OR IF SSAC 2B IS NOT RUNNING that SSAC 2B is off and NOT in automatic mode locally (auto light not lit). Simulator Operator: If requested to crosstie air, REPORT Unit 1 crosstie is not available. IF dispatched locally as Maintenance, if SSAC 2B IS RUNNING, as Maintenance, request SSAC 2B Shutdown for troubleshooting.

SRO

• Notifies SO/Maintenance (if BOP has not) to investigate SSAC 2C tripping and SSAC 2B not maintaining air pressure and initiates a condition report.

• Directs the BOP to start the 2A SSAC IAW AIR CMPSR 2C TRIPPED/SHUTDOWN, (700-233).

• Directs the BOP to start the SSAC 2B. • May request to crosstie air with Unit 1.

BOP • Places SSAC 2B switch to PTL on P650 panel. • As time allows, dispatches SO to perform step 7.5.1.6 of 34SO-P51-001-2.

Simulator operator proceeds to the next event at the Chief Examiner’s direction.



Event Description: RCIC steam line break with failure to auto close, manual works.


15 Min.

Simulator Operator: ENTER RB-3 to starts the following RCIC steam leak:

• mfE51_250 (RCIC Leak) 100/3

• svoE51075 (F008 stuck open)

• svoD11174 D11-K609A Rx Bldg Pot Contam Area Vnt 10 ramp 6

• svoD11175 D11-K609B Rx Bldg Pot Contam Area Vnt 10 ramp 6

• svoD11176 D11-K609C Rx Bldg Pot Contam Area Vnt 10 ramp 6

• svoD11177 D11-K609D Rx Bldg Pot Contam Area Vnt 10 ramp 6 NOTE: It takes approximately 3.0 minutes for the 601-321, “Leak Det Diff

Temp High” to alarm. When, LEAK DET DIF TEMP HIGH, (601-321) alarms ENSURE Event Trigger (EGE51-2) ACTIVATES.

• mf60141026 Rx Bldg Pot Contam Area Radiation High (Annunciator On) approximately 60 seconds later.

BOP

• Determines 2T41-B009, Secondary Stm. Chase Cooler is NOT running • Monitors Steam Tunnel temps

• Enters 34AB-T22-001-2, ‘Primary Coolant System Pipe Break – Rx Building.

• Dispatches operator to ATTS for Steam Chase Temperatures





BOP Acknowledges LEAK DET DIF TEMP HIGH, (601-321) and reports to SRO. Enters 601-321 and starts investigating.

• Confirms which area is producing alarm using Temperature Recorder on 2H11-P614: • R604, Pt. 115 is reading approximately 32°F delta-T • R608, Pt. 117 is reading approximately 32°F delta-T

• Enters 34AB-T22-003-2, ‘Secondary Containment Control’

SIMULATOR OPERATOR: ENSURE Event Trigger (EGE51-2) is ACTIVATED when alarm, LEAK DET DIFF TEMP HIGH, (601-321) illuminates. NOTE: It will take about 30 minutes for a Group 1 isolation to occur. At

direction of the Chief Examiner, if P603-2 “Group 1 System A Trip” alarms, IMMEDIATELY DELETE mfE51_250.

SRO

• Enters the 31EO-EOP-014-2, ‘Secondary Containment’ EOP Flowchart on Secondary Containment High Differential Temperature.

• Has an operator monitor Sec Cont. Temps. • Has operators monitor systems for source of the leak.

BOP

Receives RX BLDG POT CONTAM AREA RADIATION HIGH, (601-426) and reports to SRO, then enters 601-426 and starts investigating.

• On panel 2H11-P600, confirms radiation levels on 2D11-K609A - D on 2D11-R605 recorder

• On panel 2H11-P606, monitors radiation levels on 2D11-K609 A - D, R/B Pot Contamination Vent Exh Rad Mon A through D.

• Enters 34AB-T22-003-2, Secondary Containment Control.





SIMULATOR OPERATOR: If 2E51-F007 is CLOSED DO NOT ENTER (RB-10): SIMULATOR OPERATOR: When EOP SC flowchart is ENTERED and Chief Examiner concurs ENTER (RB-10) mfE51_113 (RCIC Auto Isolation E51-F008)

• ENSURE Event Trigger (EGE51-1) is ACTIVATED when F007 is manually closed:

• svoD11174

• svoD11175

• svoD11176

• svoD11177

• mf601411026

ATC

As time allows the crew will enter the following ARPs: • RCIC STEAM LINE DIFF PRESS HIGH, (602-302) • RCIC TURBINE TRIP, (602-301) • RCIC ISOLATIOM SIGNAL LOGIC A, (602-307) • RCIC ISOLATIOM SIGNAL LOGIC B, (602-313)

Time:_____________





ATC

Responds to failure of RCIC Isolation: • Places 2E51- F008 control switch to CLOSE. • Places 2E51- F007 control switch to CLOSE

(Critical Task is to position 2E51-F007 control switch to close within 5 minutes of receiving 602-302, 602-307 and 602-313 listed above OR may close before alarms are received).

Time:_____________

• Informs SRO of failure of 2E51-F008 to close • Dispatches operator/Maintenance to investigate the leak.

SRO

Reviews Tech Specs section 3.6.1.3, Primary Containment Isolation Valves’ Condition A.1 and A.2 and determines:

• 2E51-F008 Inop • 2E51-F007 must be closed and deactivated within 4 hours AND • 2E51-F007/F008 penetration must be verified ISOLATED every 31

days. • As time allows, contacts the Shift Support Supervisor to draft a Danger

Tagout for 2E51-F007. Enters TS RAS for RCIC 3.5.3 Condition A, which requires verifying HPCI is operable within 1 hour and restoring RCIC in 14 days.




Event Description: 2B Loop of Core Spray experiences high discharge pressure (valve leakage). When ARP is performed, system pressure is restored.


10 Min

At the Chief Examiner’s direction, Simulator Operator enters (RB-4) malfunction mf60111061 Window 25 Core Spray B Disch Pipe Press High (ON) and aoE21-R600B to 465 psig.

BOP

• Enters ARP 34AR-601-125-2 • Confirm validity of alarm using Disch Press indicator, 2E21-R600B (~465

psig) • Confirm the following valves are CLOSED

• 2E21-F037B, Testable Check Bypass Vlv • 2E21-F005B, Inbd Discharge Vlv

• Closes 2E21-F004B, Outbd Discharge Vlv • Opens 2E21-F005B • Closes 2E21-F005B, to reseat • Opens 2E21-F004B to open

SRO • Directs operator to enter 601-113 ARP • Reviews TS 3.5.1 • Reviews TS 3.6.1.3

SRO

Enters TS 3.5.1, ECCS Operating, and determines: TS 3.5.1.A requires the ECCS pump to restored to operable status in 7 days

NOTE: If addressed, 2E21-F004B is NOT a PCIV and TS 3.6.1 for PCIVs.



Event Description: 2B Loop of Core Spray experiences high discharge pressure (valve leakage). When ARP is performed, system pressure is restored.


Simulator Operator, ENSURE Event Trigger EGE21-6 activates when 2E21-F015B is placed to open position. This will clear CS B Disch Press High alarm & return 2E21-R600B to normal.

BOP

• IAW ARP 601-125, may perform the following: • Slightly opens 2E21-F015B to lower CS Discharge pressure • Confirms 601-125 clears • When pressure is approximately 100 psig, closes 2E21-F015B. • If 601-114, Core Spray B Jockey Pump Sys Water Level Low, is

received directs an SO to vent the “B” Loop of Core Spray IAW 34SV-SUV-017-2.

• Monitors Core Spray B Loop pressure for subsequent increases.

NOTE: If pressure is NOT relieved, then a follow-up question on PCIV

leakage TS 3.6.1.3.D may be appropriate.

NOTE: If pressure IS relieved on Core Spray Loop B, and pressure is NOT

monitored, then a follow-up question on a high pressure condition of the inner system piping may be appropriate.

Simulator Operator, at the Chief Examiner’s request, proceeds to the next event.



Event Description: One (1) Reactor Pressure ATTS trip unit causes a half scram and Control Rod 22-27 scrams in due to a blown fuse. The control rod fuse is repaired (Time Compress) and the rod is withdrawn.


20 min At the Chief Examiner’s direction, Simulator operator, ENTER (RB-5) malfunctions mf60211154, mfC12_26_22-27 and svoB21036 (final value of 1200 with ramp of 1000).

ALL

The following annunciators are received: REACTOR VESSEL HIGH PRESSURE TRIP, 603-105 REACTOR AUTO SCRAM SYSTEM A TRIP, 603-117 CRD ACCUMULATOR PRESS LOW OR LEVEL HIGH, 603-148 RMCS / RWM ROD BLOCK OR SYSTEM TROUBLE, 603-239 (when control rod 22-27 is selected) ROD DRIFT, 603-247 ECCS/RPS DIVISION I TROUBLE, 602-110

ATC

• Determines that reactor pressure has not changed. • Announces to SRO that a half-scram in the “A” channel has occurred due

to an invalid high reactor pressure signal. • Informs the SRO that control rod 22-27 has scrammed in.

Simulator Operator: When dispatched to check the ATTS panel reactor pressure instruments, report ONE MINUTE LATER that 2B21-N678A has a red trip light and gross failure light illuminated.

SRO

• Dispatches personal to the ATTS panels to determine which reactor pressure instrument has tripped.

• Contacts maintenance to repair ATTS card 2B21-N678A (2B21-N078A is transmitter).

• Contacts maintenance to check and replace the fuse in the RPS “B” channel for control rod 22-27.

Simulator Operator: Five minutes after being dispatched to repair the “B” RPS channel fuse for control rod 22-27, DELETES mfC12_26_22-27; this will cause the scram light for the rod to extinguish. Using time compression, as maintenance report that control rod 22-27 had a fuse blown, which has been replaced.





SRO

Refers to the following Tech Specs: LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation, and determines that 2B21-N678A requires entry into RAS 3.3.1.1.A to place the channel in trip or the “A” trip system in trip in 12 hours LCO TS 3.2, Power Distribution Limits, has STA confirm compliance Power Distribution Limits. IF thermal limits are not acceptable, the STA will consult with Reactor Engineering on further actions.

Simulator Operator: After being dispatched to repair ATTS card 2B21-N678A AND AFTER the SRO has determined the Tech Spec RAS, DELETE svoB21036 and mf60211154, THEN REPORT to the SRO that time compression has been used and that 2B21-N678A has been repaired and returned to service.

NOTE: The following annunciators and actions may not be taken in the same

sequence as listed below.

ATC

• Addresses annunciator, Reactor Auto Scram System A Trip, 603-117. • Confirm scram group A 1 2 3 4 lights for Trip System A on panel

2H11-P603 are extinguished. • Determine the cause of the trip. • Attempt to correct or bypass the cause of the trip. • Using the Process Computer obtains an OD-7 and determines that

control rod movement has occurred. (May request STA to perform OD-7) (May reset half scram before performing OD-7).





ATC

• Resets RPS Channel A using 2C71-S5, Reactor Scram Reset switch, on panel 2H11-P603, per step 5.2.3 of 34AR-603-117-2.

• Determines that section 4.7 of 34AB-C11-004-2 is required to be used to recover control rod 22-27.

• Notifies Rx Engineering or STA to check thermal limits if the SRO has not already contacted them.

ATC

• Addresses annunciator “Rod Drift,” 34AR-603-247-2 • At panel 2H11-P603, confirms that one or more Rod Drift lights are

illuminated on the full core display. • Selects the drifting rod and confirms that RPIS indicates the rod is not

at an even reed switch position. • Notifies the Shift Supervisor and the STA • Refers to 34AB-C11-004-2, “Mis-positioned Control Rods,” for

recovery of drifting OR mis-positioned control rod. • When directed by the Shift Supervisor, resets the rod drift using the

Rod Drift Alarm reset switch on Panel 2H11-P603.

ATC

• Refers to Attachment 1 of 34AB-C11-004-2 for the proper actions to take. • Are >4 rods mispositioned? NO • Is the reactor sub-critical? NO • Is reactor power < LPSP (21%)? NO • Are Thermal Limits acceptable? YES • Performs Action 1 • Refer to Attachment 2 for restoration steps.





Simulator Operator: As the team gets to the point of needing Attachment 2, provide the team with the marked up copy of Attachment 2 of 34AB-C11-004-2 When contacted as Reactor Engineering with the following question from 34AB-C11-004-2, “Contact Reactor Engineering to determine what thermal limits were exceeded during the event AND what recovery actions are necessary.” Answer NO thermal limits have been exceeded and the recovery method will be notch withdrawal of the rod to position 24 and then continuous withdrawal to position 48, is allowed.

The Simulator Operator will provide a marked up copy of attachment 2 (2 pages) of 34AB-C11-004-2.

ATC

• Refers to Attachment 2 to recover the control rod. • Withdraws the control rod to position 48 using the Rod Movement

switch and Rod Out Notch Override switch (RONOR). • Performs coupling check on control rod

Simulator operator proceeds to the next event at the Chief Examiner’s direction.



Event Description: Earthquake with a loss of Unit 1 SAT 1D.


10 Min

Simulator Operator, at Chief Examiners direction, ACTIVATE (RB-6) (Earthquake – malfunctions to (ON): mf65702209 Window 30 SEISMIC PEAK SHOCK RECORDER HIGH G LEVEL & mf65702227 Window 48 SEISMIC INSTRUMENTATION TRIGGERED & mf65321987 Window 18 230 KV BREAKER TRIP (ANNUNCIATOR ON)

ALL

The following annunciators are received: • 2H11-P657 SYSTEM TROUBLE, (650-224) • SEISMIC PEAK SHOCK RECORDER HIGH G LEVEL, (657-030) • SEISMIC INSTRUMENTATION TRIGGERED, (657-048) • 230 KV BREAKER TRIP, (653-218)

BOP

• 2H11-P657 SYSTEM TROUBLE, (650-224), alarm on 2H11-P650 panel

• 230 KV BREAKER TRIP, (653-218) on 2H11-P653 panel • Communicates both alarms to the SRO

SRO Dispatches the BOP to Panel 2H11-P653 and 2H11-P657

BOP

• Enters 230 KV BREAKER TRIP, 653-218, on 2H11-P653 panel and informs the SRO that PCBs 179440 & 179450 supplying power to SAT 1D have tripped open.

• As time permits, may notify GCC of tripped PCBs





BOP

Informs the SRO of the Seismic alarms and enters ARPs: 34AR-657-030-2 and 34AR-657-048-2 to perform the following actions: NOTE: Actions for both ARPs are the same, except for checking the power

supply.

• Dispatches Unit 1 RO to panel 1H11-P701 to check for further indication of a seismic event by monitoring Peak Shock Annunciator, 1L51-R620, for 12.7 Hz amber lights (> 0.08g, OBE) and 12.7 Hz red lights (> 0.15g, DBE)

Simulator Operator: After one minute, Notifies Unit 2 Control Room that you were in the Reactor Building and felt the floor vibrating.

BOP

• May have the Unit 1 RO check the following: • Peak Shock Annunciator, 1L51-VDC-R620, plugged in on panel 1H11-

P701 • BRKR 3 on 120/208V Essential AC Cab., 1R25-S065

• May have I & C refer to Seismic Instrumentation Earthquake Response Manual, SX-18271, for guidance in analyzing seismic data.

• Enters 34AB-Y22-002-0, Naturally occurring Phenomenon • May inform the Shift Manager to evaluate an Emergency Classification

SRO Directs the BOP to enter 34AB-Y22-002-0, Naturally occurring Phenomenon, if not already entered.





Simulator Operator: Immediately after being dispatched to check the Shock recorder on 1H11-P701 panel, inform the team that the following lights are illuminated:

• 12.7 Hz Amber lights (> 0.08g, OBE) AND

• 12.7 Hz Red lights (> 0.15g, DBE) AND

• Unit 1 has scrammed AND

• SAT 1D de-energized, 1B EDG failed to start, U1 will attempt to start the EDG locally.

(Immediately is appropriate since this indicator is on Unit 1, but right next to the Unit 2 SRO desk. It is NOT simulated in the Simulator since it is a Unit 1 only instrument.) NOTE: With SAT 1D de-energized, 34GO-OPS-013-2, Normal Plant

Shutdown, should NOT be entered.

SRO

• Determines that all electrical power is NOT available • Directs the crew to reduce reactor power to between 40% and 50% RTP

per 34GO-OPS-005-2, Power Changes (GO TO EVENT 7 FOR REACTOR POWER REDUCTION)

• Contacts switchyard maintenance to assist in switchyard damage assessment

• Contacts Maintenance to inspect Independent Spent Fuel Storage Installation (ISFSI) for damage

• Within one hour , dispatches personnel to locally close or confirmed closed the following valves (if not performed by the BOP): • 1P11-F167, CST Sump to Radwaste Drain • 1P11-F3002, Condensate Transfer Pumps and Sample Sink Drain Line

to Yard • 2P11-F051, Retaining Wall Drain • 2P11-F100, Transfer Pump Wall Drain

• Dispatches personnel to inspect the plant for damage





NOTE: These actions are redundant to the SROs and either can perform.

BOP

• Enters 34AB-Y22-002-0 and performs the following actions: • Determines that all electrical power is available • Informs SRO of the requirement to enter 34GO-OPS-005-2, Power

Changes, and reduce reactor power to between 40% and 50% RTP per 34GO-OPS-005-2, Power Changes (GO TO EVENT 7 FOR REACTOR POWER REDUCTION)

• Contacts switchyard maintenance to assist in switchyard damage assessment

• Contacts Maintenance to inspect Independent Spent Fuel Storage Installation (ISFSI) for damage

• Within one hour , dispatches personnel to locally close or confirmed closed the following valves (if not performed by the SRO): • 1P11-F167, CST Sump to Radwaste Drain • 1P11-F3002, Condensate Transfer Pumps and Sample Sink Drain Line

to Yard • 2P11-F051, Retaining Wall Drain • 2P11-F100, Transfer Pump Wall Drain

• Dispatches personnel to inspect the plant for damage



Event Description: Reduce reactor power to between 40% & 50% due to earthquake.


8 Min. SRO • Directs ATC to decrease reactor power to 40% to 50% by decreasing

Recirc flow. Power decreases should be made as recommended by the STA/Reactor Engineering at a rate exceeding 10 MWe/min.

NOTE: May get the RBM UPSCALE, (603-202) and ROD OUT BLOCK, (603-238) alarm, if a peripheral control rod is not selected. This is expected and the operator may select a peripheral rod at this time. May also get Alarm HEATER TROUBLE, 650-135 alarm. This is expected at this power level.

ATC

• IAW 34SO-B31-001-2 (step 7.1.5) & 34GO-OPS-005-2, the ATC decreases Recirc pump speed, not to exceed 10 MWE per minute by depressing the LOWER SLOW or LOWER MEDIUM pushbuttons on the Master (P603 panel) or Individual controls (P602 panel) until reactor power is ~70%.

• If using Individual Controls, pump speed increases will alternate between the “A” & “B” Recirc pumps to prevent excessive flow mismatches.

• Monitors power decrease by observing APRM and generator output

indications. • Notifies SRO of attaining 40% to 50% reactor power.

Simulator Operator enters the next event after power has been decreased between 40% to 50% or at the Chief Examiner’s request.



Event Description: Emergency Depress prior to Torus level reaches 98”.


Simulator Operator, after reactor power is reduced between 40-50% RTP and at Chief Examiners direction, ACTIVATE (RB-1) Torus leak at (3/4”) 0.75”/min) svoT48140 (70/.75), svoT48142 (50/10), svoT48143 (50/10), svoT48147 (50/10), svoT48148 (50/100) LOOK ahead, This leak will be modified at 142 inches in the Torus (EGT48-1) and ALL ADS SRVs fail closed. Simulator Operator: After one minute, Notifies Unit 2 Control Room that you were in the Reactor Building and felt the floor vibrating.

ALL

The following annunciators are received: • 650-224, PANEL 2H11-P657 SYSTEM TROUBLE • 657-086, TORUS S-W AREA INSTR SUMP LVL HIGH • 657-087, TORUS N-W AREA INSTR SUMP LVL HIGH • 657-088, TORUS N-E AREA INSTR SUMP LVL HIGH • 657-089, TORUS S-E AREA INSTR SUMP LVL HIGH • 657-104, TORUS S-W AREA INSTR SUMP LVL HIGH-HIGH • 657-105, TORUS N-W AREA INSTR SUMP LVL HIGH-HIGH • 657-106, TORUS N-E AREA INSTR SUMP LVL HIGH-HIGH • 657-107, TORUS S-E AREA INSTR SUMP LVL HIGH-HIGH • 657-013, TORUS N-E AREA INSTR SUMP LVL HIGH-HIGH-HIGH • 657-031, TORUS S-E AREA INSTR SUMP LVL HIGH-HIGH-HIGH • 657-049, TORUS N-W AREA INSTR SUMP LVL HIGH-HIGH-HIGH • 657-067, TORUS S-W AREA INSTR SUMP LVL HIGH-HIGH-HIGH

BOP • Reports multiple alarms to SRO indicating a break in the Reactor Building. • Directs SO/Maintenance to investigate the leak.





SRO • Directs BOP to 2H11-P657 panel. • When above alarms are reported, directs operator to monitor Torus water

level and then if lowering, enter 34AB-T23-004-2, Torus Water Level.

Simulator Operator: Four minutes after being dispatched to check for leaks in the Torus section of the Reactor Building, report to the crew: A leak has been identified on the “2A” Core Spray line between the Torus and the first Core Spray isolation valve.

ALL • 602-235, TORUS WATER LEVEL HIGH/LOW, annunciates • Recognizes that Torus level is decreasing.

SRO

• Dispatches personnel to determine the location of the Torus leak. • If not already directed, directs NPO to enter 34AB-T23-004-2, Torus

Water Level, and to monitor Torus water level. • Enters the PC EOP Flowchart when Torus level decreases to

146 inches. • May determine that water will not be added to the Torus until the cause of

the low Torus level is identified and controlled. • Enter SC EOP flowchart for SC area water levels being high.

BOP • If NOT already performed, dispatches personnel to the Torus area AND

the Reactor Building diagonals to determine the source of the water loss (if the leak location has not already been reported).





SRO

• IAW the PC flowchart, prior to water level reaching 98 inches, determines that the reactor is required to be shutdown and enters the RC flowchart at point A.

• Assigns the ATC to perform RC-1. • Assigns the BOP operator to perform RC-2 and RC-3. • Enters 31EO-EOP-010-2, RC EOP flow chart if RWL decreases below 3

inches. • Directs RWL Band of 3 to 50 inches.

ATC


RWM display. • Notifies SRO of rod position check. • Places SDV isolation valve switch to "isolate" & confirms closed. • If not tripped, places the Recirc pumps at minimum speed. • Inserts SRMs and IRMs. • Shifts recorders to read IRMS, when required. • Ranges IRMS to bring reading on scale. • Notifies the SRO when the above actions are complete.





BOP


• Checks ECCS Injection Systems • Ensures FW Master Controller setpoint reduces to 9 inches and

output reduces to 25% of previous value (will not due to low power)

• Set down does not auto function (low power), manually reduces FW Master Controller setpoint to approximately 9 inches.

• When feed flow is less than the capacity of the S/U level control valve (≈ 1.5 mlbm/hr), then: • Opens 2N21-F125. • Places 2C32-R619, FW S/U level control valve controller, in Auto,

set at approximately 9 inches. • Closes 2N21-F110. • Will control RWL and with SRO permission will raise RWL to

32 to 42 inches.

BOP

• Performs RC-3 consisting of: • Monitor RPV pressure. • Confirm proper operation of pressure control system (TBV, LLS, etc.). • If necessary, allow RPV pressure to exceed 1074 psig then cycle any

SRV to initiate LLS. • Maintain RPV pressure between 1074 and 800 psig. • Notify SRO of pressure control system operation.

SRO

• If the need to Emergency Depressurize is recognized in time, then Anticipates Emergency Depressurization. • Assign an operator to fully open all Main Turbine Bypass Valves.

• Directs ATC/BOP to place HPCI in Pull To Lock prior to 110 inches

Torus level





ATC/BOP

• On the DEHC panel • Select the Control > Bypass Valve screen. • Insert a ramp rate of 100, then press OK. • Insert a bypass valve position of 100, then press OK. • Checks that the Bypass Valve Jack status is active. • Recognizes that 3 Bypass Valves open. • Reports to the SRO that 3 Bypass Valves are open. • Places HPCI Aux Oil pump in the Pull To Lock position.

ATC • Provides periodic updates on Torus level to the SRO.

SRO • Transitions to CP-1 and orders 7 ADS valves open for Emergency Depress. (GO TO EVENT 9)



Event Description: ALL ADS valves fail to open when Emergency Depress (ED) is required, EOP-108 used to ED.


Simulator Operator, failure of ALL ADS SRVs to open, have been active since the start of scenario.

ATC/BOP

• Places the switches for 7 ADS valves in the open position. Determines that NONE of the ADS valves open.

• Either informs SRO or continues opening other SRVs. Only 4 SRVs (LLS) will open

• Notifies SRO that 4 SRVs open and failure of ALL ADS SRVs fail to open. (Critical Task – Critical task is met when 4 SRVs (LLS SRVs) have been opened and 31EO-EOP-108 has been entered). • The amber lights for the SRVs will NOT illuminate if pressure has been


NOTE: If RPV pressure is below approximately 300psig, the SRV amber lights will not illuminate for SRV position confirmation. The operator can verify that the SRVs have opened by observing SRV tailpipe temperature increase.

SRO

• If the Operator notifies the SRO that 7 SRVs failed to open and that only 4 LLS valves are open, the SRO directs the operator to enter 31EO-EOP-108.

• The amber lights for the SRVs will not illuminate if pressure has been


ATC/BOP

• Depending on Reactor Water Level prior to opening ADS valves, RWL may swell to above 60 inches, requiring the operator to enter 34AB-C32-001-2, Reactor Water Level Above 60 inches. Operator secures all injection other than CRD.

With Chief Examiners Permission the Scenario will be terminated when Reactor pressure is within 50 psig of Torus pressure or as directed by the Chief Examiner.



Initiating Conditions: Unit 2 is operating at 60% RTP. 2B21-F013D, SRV 2D, is inoperable for LLS, RAS written.

Turnover Continue startup and place RFPT 2A in service, starting at step 7.1.11.2.10 of 34SO-N21-007-2. Once RFPT 2A is in service, notify Maintenance to perform walk down of RFPT 2A piping. Once Maintenance walk down is complete, increase Reactor power to 75% RTP.

Summary: • Event 1: Normal; The “2A” RFPT will be placed into service IAW 34SO-N21-007-2.

• Event 2: Component; SSAC 2C Trips - 2B fails unloaded, 2A manually started. The "2C"

Station Service Air Compressor (SSAC) trips due to an equipment malfunction. "2B" SSAC auto starts but does not supply any air to the system because its loader system is malfunctioning. The crew will manually start "2A" SSAC, which does return the air system to normal pressures.

• Event 3: Component/TS; A RCIC steam line will break outside of Primary Containment. The outboard isolation valve is failed open and cannot be closed. Both isolation valves fail to automatically close on an automatic isolation signal and must be manually closed. (Critical Task) The SRO addresses Tech Specs for inoperable Primary Containment Isolation Valve.

• Event 4: Component/TS; 2B Loop of Core Spray experiences high discharge pressure (valve leakage). When ARP is performed, system pressure is restored.

• Event 5: Component; One (1) Reactor Pressure ATTS trip unit causes a half scram and a control rod to scram in due to a blown fuse. The control rod fuse is repaired (Time Compress) and the rod is withdrawn.

• Event 6: Major; Earthquake requiring scram prior to 98” in Torus with a loss of Unit 1 SAT 1D.

• Event 7: Reactivity; Reduce reactor power to between 40% & 50% due to earthquake.

• Event 8: Major; Emergency Depress prior to 98” in Torus. (Critical Task)

• Event 9: Component; ALL ADS valves fail to open when Emergency Depress is required, EOP-108. (Critical Task)


NRC DRAFT Facility: E. I Hatch Scenario No.: 10-3 Op-Test No.: 2016-301 Critical Tasks • Manually isolate the RCIC isolation valves within 5 minutes of receiving 602-302 and 602-313.

(Event 3)

• Emergency Depress prior to Torus level reaches 98”. (Event 8)

• The ATC will open an additional (4) valves and enter 31EO-EOP-108 to complete the ED. (Event 9)

ES 301-4 Attributes Required Actual Items 1. Total Malfunctions 5-8 6 1. SSAC 2C Trips - 2B fails unloaded, 2A manually

started (Event 2) 2. RCIC steam line will break (Event 3) 3. 2B Loop of CS high pressure (Event 4) 4. ATTS Scram with CR blown fuse (Event 5) 5. Earthquake with SAT 1D loss (Event 6) 6. All ADS valves fail (Event 9)


1-2 1 1. All ADS valves fail (Event 9)

3. Abnormal Events 2-4 4 1. SSAC 2C Trips - 2B fails unloaded, 2A manually started (Event 2)

2. RCIC steam line will break (Event 3) 3. ATTS Scram with CR blown fuse (Event 5) 4. Earthquake with SAT 1D loss (Event 6)

4. Major Transients 1-2 2 1. Earthquake requiring scram prior to 98” in Torus (Event 6)

2. Emergency Depress prior to 98” in Torus (Event 8)

5. EOPs entered, requiring substantive actions

1-2 2 1. RC (Event 6) 2. PC (Event 7)


0-2 1 1. Emergency Depress prior to 98” in Torus (Event 7)


2-3 3 1. RCIC steam line will break (Event 3) 2. Emergency Depress prior to 98” in Torus

(Event 8) 3. ALL ADS valves fail to open when Emergency

Depress is required, EOP-108. (Event 9)


SHIFT TURNOVER

Safety Focus

UNIT 1 STATUS


UNIT 2 STATUS



Division I

Division II

Green Yellow

Orange

Red

Scheduled evolutions : � Place RFPT 2A in service, starting at step 7.1.11.2.08 of 34SO-N21-007-2

� Once RFPT 2A is in service, Maintenance will perform walk down of RFPT 2A piping

� Once Maintenance walk down is complete, increase Reactor power to 75% RTP


shift :

� None


Active tagouts : � 2B21-F013D, SRV 2D, Caution Tag


of 28


NRC DRAFT Facility: E. I Hatch Scenario No.: 10-04 Op-Test No.: 2016-301 Examiners: Operators: SRO RO BOP Initial Conditions. Unit 2 is operating at 90% RTP. 2D11-K615B, Offgas Post Treatment Radiation Monitor, failed Downscale. RAS written. Turnover: Hydrogen Seal Oil System is currently aligned with the Emergency Seal Oil pump in service. After turnover restore to Normal Hydrogen Seal Oil System Lineup starting at step 7.3.3.2.5 of 34SO-N42-001-2. Once Hydrogen Seal Oil is returned to Normal lineup, increase reactor power to 100% RTP using the Recirculation System.

Event No.


Event Description

1 N/A N (BOP) Restoration To Normal Hydrogen Seal Oil System Lineup starting at step 7.3.3.2.5 of 34SO-N42-001-2

2 mfB31_135B R (ATC) TS (SRO)

Recirc ‘2B’ ASD Cell Bypass; “2B” Recirc speed decreases resulting in a flow mismatch

3 mf65031541 mfN61_73

C (ATC) RFPT loop seal failure requiring manual opening of bypass valve to prevent Main Turbine trip on low vacuum

4 mf70022407 C (BOP) ‘2C’ SSAC high temp condition ‘2A & 2B’ SSACs are manually started by BOP

5

mfB31_39A mfB31_45A rfB31_29 mf60213160

C (ATC) TS (SRO)

Recirc Pump “A” seal leakage causes DW leakage. Pump is secured and isolated

6 mf65702209 mf65702227

C (BOP) TS (SRO)

A leak develops from the HPCI Torus suction line downstream of 2E41-F051 and can be isolated by closing 2E41-F051 (Critical Task)

7 mfS11_161 M (ALL) Loss of Offsite Power

8 mfR43_239C mfR43_49B mfR43_62A

C (ATC) 2C EDG fails to auto tie, 1B EDG will NOT operate & 2A EDG will tie after resetting Shutdown (Critical Task to energize at least one emergency bus)

9 rfE51_234 C (BOP) HPCI 2E41-F001 stuck closed, RCIC 2E51-F013 Fails to Auto Open requiring manual operation of 2E51-F013 to inject prior to RWL reaching -180” (Critical Task)




Event Description: Restoration To Normal Hydrogen Seal Oil System Lineup starting at step 7.3.3.2.5 of 34SO-N42-001-2.


10 Mins

Simulator Operator, IF contacted as SO, report the following: • Seal Oil Vacuum Pump Separator Tank oil level is between the High

AND Low Oil Level on 2N42-LG-R306, Separator Tank Sight Glass • 2N42-F042, Vacuum Tank Inlet Valve, open • Oil level in the Vacuum Tank is above the bottom of the lower

observation window • Seal Oil Vacuum Pump shaft has been manually rotated.

BOP

• Enters Restoration To Normal Hydrogen Seal Oil System Lineup at step 7.3.3.2.5 • Starts 2N42-C001, H2 Seal Oil Vacuum Pump, (2H11-P651).

Simulator Operator, WHEN contacted as SO, 2N42-R302, Vacuum Tank Pressure, is 26.5 in. Hg and steady.

• Confirms 2N42-R302, Vacuum Tank Pressure, stabilizes above 26 in. Hg. by contacting SO locally.

• Starts 2N42-C003, Recirc H2 Seal Oil Pump. • Starts 2N42-C004, Main H2 Seal Oil Pump. • Secure 2N42-C002, Emergency Seal Oil Pump, as follows:

• Holds control switch for ESOP in OFF Pull To Lock position until pump stops as indicated by its extinguished red AND green lights

• Alarm EMERG SEAL OIL PUMP RUNNING, (651-216) will clear after the SO resets the local panel.

• AFTER pump stops, releases control switch. • Confirm its green light is illuminated.

Simulator Operator, when contacted as SO, report the following: • 2N42-R300, Seal Oil Pump Disch Press, is 105 psig • 2N42-R301, Seal Oil / Machine Gas dP, pressure, is 8 psig above

generator gas pressure. • If asked, inform BOP the Emergency Seal Oil Pump shaft has stopped

rotating.



Event Description: Restoration To Normal Hydrogen Seal Oil System Lineup starting at step 7.3.3.2.5 of 34SO-N42-001-2.


Simulator Operator, when contacted as SO, report the following: • DELETE mf65121574, EMERGENCY SEAL OIL PUMP RUNNING

(ANNUNCIATOR ON) • REPORT Local alarm, EMERG SEAL OIL PUMP RUNNING, (N43-

105) has been reset

BOP

• Confirms the following by contacting SO locally: • 2N42-R300, Seal Oil Pump Disch Press, is 100 - 110 psig • 2N42-R301, Seal Oil / Machine Gas dP, pressure, is being maintained

7 - 9 psig above generator gas pressure.

Simulator Operator, at the Chief Examiner’s request, PROCEEDS to the next event.



Event Description: “2B” Recirculation Pump ASD Cell will auto bypass.


10 min Simulator Operator, At Chief Examiner’s direction, press (RB-2) to activate:

• mfB31_135B, Recirc ASD B Cell Bypass

SRO

• Direct maintenance be contacted to determine cause of Recirc speed decrease

• Evaluates TS LCO/RAS Recirculation Loops Operating and enters TS 3.4.1 Condition A.1 to restore within limits within 24 hours.

NOTE: With the ASD 2B Cell Bypassed, the operator may enter 34AB-B31-

001-2, Reactor Recirculation Pump(s) Trip, Recirc Loops Mismatch, OR ASD Cell Bypass”, first due to Recirc flow mismatch.

ATC

• Respond to ASD B TROUBLE, (602-208), Annunciator • Diagnose Recirc pump speed mismatch • Confirm that an ASD cell has failed and is bypassed using SPDS or by

having it verified locally at the ASD cabinet. • Enter 34AB-B31-001-2, Reactor Recirculation Pump(s) Trip, Recirc

Loops Mismatch, or ASD Cell Bypass

• Confirm the “SPD HLD RESET” pushbutton is illuminated

• Enter section III of 34AB-B31-001-2

• If Recirc speed mismatch >35%, inform SRO of 1 hour limit to balance Recirc Pump flows (will determine < 35% mismatch)

If the SRO decides to increase Recirc flow to match flow, then the following

(increase) actions will be taken, otherwise skip this part and go to the (decrease) actions.

SRO If asked, Reactor Engineering guidance for increasing power is:

• “Limit the rate of power change to 10 MWe/minute”



Event Description: “2B” Recirculation Pump ASD Cell will auto bypass.


ATC • If directed by SRO to increase: • Depress the respective “B” SPD HLD RESET pushbutton indicating

lamp • Increase speed of the Recirc Pump 2B per 34SO-B31-001-2, Reactor

Recirc System, by depressing SLOW Raise or MEDIUM Raise pushbutton on ASD 2B.

• After 34AB-B31-001-2 has been addressed entered, the operator will depress the Fault Reset pushbutton to reset the ASD B TROUBLE, (602-108), alarm

If the SRO decides to decrease Recirc flow to match flow, then the following

actions will be taken, otherwise skip.

ATC

• If directed by SRO to decrease: • Depress the respective “B” SPD HLD RESET pushbutton indicating

lamp • Decrease speed of the Recirc Pump 2A per 34SO-B31-001-2, Reactor

Recirc System, by depressing SLOW Lower or MEDIUM Lower pushbutton on ASD 2A.

• After 34AB-B31-001-2 has been addressed entered, the operator will depress the Fault Reset pushbutton to reset the ASD B TROUBLE, (602-208), alarm

NOTE to Examiners: TS SR 3.4.1.1 requires Recirc flow mismatch be less than 5% if operating at greater than or equal to 70% rated core flow, and mismatch be less than 10% if operating at less than 70% rated core flow.

Simulator Operator, When the crew has diagnosed that possibly, an ASD Cell is bypassed, then with Chief Examiner’s concurrence: Call the Control Room as I&C with the following message: The Recirculation Pump ASD 2B has automatically bypassed a cell, conditions at the ASD are normal for this condition.




Event Description: RFPT loop seal failure.


10 Min

SIMULATOR OPERATOR: At the direction of the Chief Examiner, ACTIVATE (RB-3)

• mf65031541 “RFP Loop Seal Level Low (Annunciator On)”

• mfN61_73 “Main Condenser Air In-leakage”

NOTE: It takes approximately 28 minutes for vacuum to decrease to 25.9”.

ATC • Recognize RFP LOOP SEAL LEVEL LOW, (650-319), annunciator

ATC

• Respond to annunciator RFP LOOP SEAL LEVEL LOW, (650-319)

• Dispatches an SO to 2H21-P216 to confirm 2N22-F398, RFP Bracket Drain Loop Seal Fill Valve is open.

• Monitors vacuum at 2H11-P650, on 2N21-R602. • Dispatches an SO to confirm seal water lineup and pressures IAW

34S0-N21-007-2, Condensate and Feedwater System.

BOP • Monitors Inlet Flow to Stack on 2N62-P600

• Receives INLET FLOW TO STACK HIGH, (600-020) • Monitors Inlet flow to Stack at 2N62-P600, on 2N21-R604

SRO May direct entry into 34AB-N61-002-2, Main Condenser Vacuum Low, abnormal.

BOP If directed, may REDUCE reactor power per 34GO-OPS-005-2, Power Changes, to establish and maintain vacuum greater than 25 in. Hg.





SIMULATOR OPERATOR: Wait 4 minutes from being dispatched, NOTIFY the ATC of ONE of the following messages: MESSAGE 1 IF 2N22-F265 is still OPEN REPORT:

• Having difficulty locating 2N22-F398, RFP Bracket Drain Loop Seal Fill Valve but still attempting to locate.

• Seal water pressures are normal and/but the lineup check is NOT completed.

• NO water is coming out of weep holes.

MESSAGE 2 IF 2N22-F265 has been CLOSED REPORT:

• 2N22-F398, RFP Bracket Drain Loop Seal Fill Valve was closed and now is open

• Seal water pressures are normal and/but the lineup check is NOT completed.

• NO water is coming out of weep holes.

SIMULATOR OPERATOR: ENSURE Event Trigger EGN21-2 deletes mfN61_73 when one of the following occurs:

• Condenser vacuum drops to 25.9”

• 2N21-F265 is closed Then 45 seconds later, DELETES mf65031541.

ATC

• With seal water pressures normal, will be required to close 2N21-F265, RFP Loop Seal Outlet Isol Vlv after 5 minutes of alarm RFP LOOP SEAL LEVEL LOW, (650-319), being received.

• With vacuum degrading may close 2N21-F265 sooner to reverse degrading vacuum condition.





ATC • Opens 2N21-F265 when RFP LOOP SEAL LEVEL LOW, (650-319) clears.

BOP • Alarm INLET FLOW TO HOLDUP LINE HIGH, (600-020), clears when flow returns to normal.

SIMULATOR OPERATOR; At Chief Examiners direction, proceed to the next event.



Event Description: ‘2C’ SSAC high temp condition and secured. ‘2A & 2B’ SSACs are manually started.


5 Min.

SIMULATOR OPERATOR: At the direction of the Chief Examiner ACTIVATE (RB-4)

• mf70022407 Control Bldg Aftclr B001C Disch Temp High (Alarm On)

• aoP52-R600 “Service Air Pressure”

• aoP51-R600 “Control Air Pressure” NOTE: SSAC 2B will fail to auto start on lowering pressure. Event Trigger (EGP51-2) will remove this failure when SSAC 2B is manually started.

SIMULATOR OPERATOR: When the operator dispatches a SO locally , wait 2 minutes, then NOTIFY operator that local temp on 2P51-R312C is 125°F and 2P51-R302C is reading 385°F. If requested by operator, report standby SSAC oil levels are normal and Aftercooler/Intercooler drains have been cycled.

BOP

Responds to annunciator CONTROL BLDG AFTCLR B001C DISCH TEMP HIGH, (700-216):

• Sends SO to locally confirm temperature is > 120 degrees F on 2P51-R312C and 2P51-R302C.

• Starts the 2A and/or 2B Service Air Compressor

• Secures the 2C Service Air Compressor by placing its control switch in Pull-to-Lock when the local report is given or earlier as directed by the SRO.

• Dispatches SO/Maintenance to investigate high temperature alarm.



Event Description: ‘2C’ SSAC high temp condition and secured. ‘2A & 2B’ SSACs are manually started.


SIMULATOR OPERATOR: When the operator secures the 2C Service Air Compressor, ENSURE Event Trigger (EGP51-1) is ACTIVATED: DELETES the following:

• aoP52-R600 ‘Service Air Pressure’

• aoP51-R600 ‘Control Air Pressure’ mf70022407 ‘Control Bldg Aftclr B001C Disch Temp High’ (Annunciator On) 30 seconds later.

BOP Annunciator CONTROL BLDG AFTCLR B001C DISCH TEMP HIGH, (700-216) clears.




Event Description: Recirc Pump “A” seal leakage causes DW leakage. Pump is secured and isolated


12 Min

Simulator Operator, with Chief Examiners Permission:

• Call control room BOP operator and tell them to stay on the phone until directed to hang up.

• ACTIVATE: (RB-5) to start Recirc Pump 2A seal leakage, mfB31_39A Final 100, Ramp of 10

ATC

• Receives Annunciator, PUMP A SEAL STAGING FLOW HIGH/LOW, (602-122) in approximately 3 minutes

• Observes and compares 2B31-R603A, Seal A No. 1 pressure indicator, AND 2B31-R602A, Seal A No. 2 pressure indicator, to determine the Inner (#1) seal has failed (No. 1 pressure normal with No. 2 pressure > normal).

• Notifies SRO that the Inner (#1) Seal on Recirc A has failed • Directs a System Operator to confirm seal water flow to Pump A is

between 1.6 to 2.2 gpm locally (this indication not available in the Control Room).

• Assigns an extra operator perform 34SV-SUV-019-2, Surveillance Checks, to determine the magnitude of leakage.

• May monitor Drywell pressure.

Simulator Operator, After the ATC operator has started executing steps in the “Pump A Seal Staging Flow High/Low” ARP, THEN Instruct the BOP operator to hang up the phone.





Simulator Operator; After 3 minutes, NOTIFY the Chief Examiner AND then press (RB-1) to ACTIVATE: mfB31_45A (10/20), Recirc A Outer Seal Failure.

(**IMPORTANT**) • Ensure the ATC responds to the “Outer Seal A Leak Detection Flow

High,” (602-116) and performs the next set of actions (Trip and isolate the “2A” Recirc Pump). ENSURE the BOP operator is the person assigned to vent the Drywell.

ALL

• Receives Annunciators: • OUTER SEAL A LEAK DETECTION FLOW HIGH, (602-116) • DRYWELL/TORUS RCDR R627 TEMP HIGH, (650-204) • PANEL 2H11-P654 SYSTEM TROUBLE, (650-214) • PANEL 2H11-P657 SYSTEM TROUBLE, (650-224) • DRYWELL FLOOR DRAINS SUMP LEAK HIGH-HIGH, (602-408)

ATC

• Observes and compares 2B31-R603A, Seal A No. 1 pressure indicator AND 2B31-R602A, Seal A No. 2 pressure indicator, to determine the outer seal has now failed (No. 1 and No. 2 seal pressure decreases).

• Notifies SRO that the Outer (#2) Seal on “2A” Recirc has also failed. • Has an extra operator perform 34SV-SUV-019-2, Surveillance Checks, to

determine magnitude of leak. • Monitors 2D11-R630, Fission Products Monitor recorder, to determine IF

primary system coolant is leaking from seals. • Monitors Drywell pressure. • PANEL 2H11-P657 SYSTEM TROUBLE, (650-224) • MULTIPOINT TEMPERATURE RCDR 2T47-R626 TEMPERATURE

HIGH (657-025).





ALL

• Receives the following alarms • PRIMARY CONTAINMENT PRESS HIGH, (603-115) • DRYWELL FLOOR DRAINS SUMP LEAK HIGH, (602-408)

SRO • Directs Operator to Check DW Leakage • Directs the BOP to vent the DW with SBGT, when DW pressure

approaches 0.65 psig.

Simulator Operator; As the operator checking DW leakage, report:

• DW Equipment drain leakage is stable at 1.7 gpm • Floor drain leakage has increased from 0.8 gpm to 12.1 gpm.

SIMULATOR OPERATOR; If the crew inserts a Reactor Scram during Event 5, PROCEED to Major Event 7.





SRO

• Directs operator to shutdown and isolate Recirc pump A • Enters Tech Specs: • 3.4 Reactor Coolant System • 3.4.1 Recirculation Loops Operating

• Condition: 3.4.1.A, Requirements of the LCO not met.

(i.e. single loop operations) Required Action: Satisfy requirements of the LCO. Completion time: 24 hrs.

• Condition: 3.4.4.A, Unidentified leakage not within limit OR total leakage not within limit. Required Action: Reduce leakage to within limits. Completion time: 4 hours.

• Condition 3.4.4.B, Unidentified leakage increase not within limit. Required Action: Reduce leakage increase to within limits. Completion time: 4 hours.

• Notifies STA or Reactor Engineering that new limits apply for APLHGR, MCPR, LHGR, and APRMs Simulated Thermal power - New Setpoints are required within 24 hours.





ATC

• Per (602-116) and 34SO-B31-001-2, Recirc System, Performs one of the following to secure the Reactor Recirc Pump: • Either (actions for Immediate Shutdown)

• Place the ASD “A” control switch in Pull to Lock (P602) • Depress the ASD “A” Shutdown pushbutton (P602)

• OR (Actions for Shutdown in a Timely Manner) • Depress the ASD “A” Shutdown pushbutton (P602) • Confirm ASD A speed ramps to ~370 RPM (2B31-R660) or 22%

(2B31-R661) • Confirm Recirc Pump A goes to 0 RPM (2B31-R660) or 22%

(2B31-R661) and drive coasts to 0 gpm (2B31-R614 or R617) • Confirm ASD A Start pushbutton illuminates • Place ASD “A” control switch 2B31-S002A to Pull-To-Lock

(P602) • Closes 2B31-F031A, Reactor Recirc A Pump Disch Vlv • Closes 2B31-F023A, Reactor Recirc A Pump Suction Vlv

• Dispatches SO to close Seal Injection To Pump A Header Isolation

Valve, 2B31-F008A. • Enters 34AB-B31-001-2 “Reactor Recirculation Pump(s) TRIP, Recirc

Loops Flow mismatch, Or ASD Cell Bypass.” • Notify Plant Management and Load Dispatcher

ATC

• Receives and Acknowledges the following Annunciators: • RECIRC LOOP A OUT OF SERVICE, (602-127) • ASD “A” TRIP WARNING, (602-101) • ASD “A” TROUBLE, (602-108) • ASD “A” FATAL FAULT, (602-102) • RBM DOWNSCALE, (603-211) • ROD OUT BLOCK, (603-238) • HEATER TROUBLE ALARM, (650-135) (may come in and clear)





Simulator Operator; AFTER the suction and discharge valves are closed for the “2A” Recirc pump, an you are requested to close 2B31-F008, THEN:

• MODIFY mfB31_45A (Final: 100 Ramp: 1000),”Recirc A Outer Seal Failure."

AND THEN press (RB-9) to ACTIVATE:

• rfB31_29, Recirc mini purge B31-F016A closure (simulates B31-F008A being closed)

• mf60213160, “Outer Seal A Leak Detection Flow High – Annunciator Off (602-116)

AND THEN REPORT that 2B31-F008A is closed.

Simulator Operator; If another DW leakage check is requested, report:

• Drywell floor drain leakage has decreased to 1.8 gpm.



Op-Test No.: 2016-301 Scenario No.: 10-04 Event No.: 6 Page 17 of 28 Event Description: A leak develops from the HPCI Torus suction line downstream of 2E41-F051

and can be isolated by closing 2E41-F051.


Simulator Operator, at Chief Examiners direction, ACTIVATE (RB-6) Torus leak at (3/4”) 0.75”/min) svoT48140 (70/.75), svoT48142 (50/10), svoT48143 (50/10), svoT48147 (50/10), svoT48148 (50/100)

ALL

The following annunciators are received: • PANEL 2H11-P657 SYSTEM TROUBLE, (650-224) • TORUS S-W AREA INSTR SUMP LVL HIGH, (657-086) • TORUS N-W AREA INSTR SUMP LVL HIGH, (657-087) • TORUS N-E AREA INSTR SUMP LVL HIGH, (657-088) • TORUS S-E AREA INSTR SUMP LVL HIGH, (657-089) • TORUS S-W AREA INSTR SUMP LVL HIGH-HIGH, (657-104) • TORUS N-W AREA INSTR SUMP LVL HIGH-HIGH, (657-105) • TORUS N-E AREA INSTR SUMP LVL HIGH-HIGH, (657-106) • TORUS S-E AREA INSTR SUMP LVL HIGH-HIGH, (657-107) • TORUS N-E AREA INSTR SUMP LVL HIGH-HIGH-HIGH,

(657-013) • TORUS S-E AREA INSTR SUMP LVL HIGH-HIGH-HIGH,

(657-031) • TORUS N-W AREA INSTR SUMP LVL HIGH-HIGH-HIGH,

(657-049) • TORUS S-W AREA INSTR SUMP LVL HIGH-HIGH-HIGH,

(657-067)

BOP • Reports multiple alarms to SRO indicating a break in the Reactor Building. • Directs SO/Maintenance to investigate the leak.





SRO • Directs BOP to 2H11-P657 panel. • When above alarms are reported, directs operator to monitor Torus water

level and then if lowering, enter 34AB-T23-004-2, Torus Water Level.

Simulator Operator: Four minutes after being dispatched to check for leaks in the Torus section of the Reactor Building, report to the crew: A leak has been identified on the HPCI suction line and I can’t tell where the leak is exactly.

ALL • TORUS WATER LEVEL HIGH/LOW, (602-235) annunciates • Recognizes that Torus level is decreasing.

SRO

• Dispatches personnel to determine the location of the Torus leak. • If not already directed, directs NPO to enter 34AB-T23-004-2, Torus

Water Level, and to monitor Torus water level. • Enters the PC EOP Flowchart when Torus level decreases to

146 inches. • May determine that water will NOT be added to the Torus until the cause

of the low Torus level is identified and controlled. • Enter SC EOP flowchart for SC area water levels being high.

BOP If NOT already performed, dispatches personnel to the Torus area AND the Reactor Building diagonals to determine the source of the water loss (if the leak location has NOT already been reported).





ATC

• Enters 34AB-T23-004-2, Torus Water Level, and performs the following: • Dispatches personnel to the Torus AND SE Diagonal to determine

source of leakage • Notifies SRO of ECCS TS requirements when closing valves. • Closes 2E21-F019B, Torus Suction Vlv • Closes 2E11-F065B, Torus Suction Vlv • Closes 2E11-F065D, Torus Suction Vlv

Simulator Operator: ENSURE Event Trigger EGE41-4 MODIFIES the following: svoT48140 (r:0.0), svoT48142 (0/10), svoT48143 (0/10), svoT48147 (0/10), svoT48148 (0/10)

• Checks Torus water level and determines the Torus level is continuing to decrease with the above valves closed

• Closes 2E41-F051, HPCI Torus Suction Vlv • Opens 2E21-F019B, Torus Suction Vlv • Opens 2E11-F065B, Torus Suction Vlv • Opens 2E11-F065D, Torus Suction Vlv

• Checks Torus water level and determines the Torus level has stopped decreasing

• Enters 34AB-T23-001-2, Loss of Primary Containment Integrity, AND 34AB-T22-003-2, Secondary Containment Control.

• Notifies SRO that the Torus leak stopped after 2E41-F051 was closed.





SRO

• Reviews TS 3.6.2.2, Suppression Pool Water Level • IAW TS 3.6.2.2 Condition A

• Must restore within 2 hours • Reviews TS 3.3.5.1, ECCS Instrumentation • IAW TS 3.3.5.1 Condition D,

• Only applicable if HPCI is not aligned to the Suppression Pool, Declares HPCI inop within one hour

• Reviews TS 3.5.1, ECCS/RCIC • IAW TS 3.5.1 Condition C,

• Declares HPCI inoperable, • Must verify within one hour that RCIC is operable by administrative

means • Must restore HPCI to operable status within 14 days




Event Description: Loss of Offsite Power with EDGs failures.


15 Mins

SIMULATOR OPERATOR: At the direction of the Chief Examiner, ACTIVATE (RB-7), mfS11_161, “Loss of Offsite Power”.

ALL Recognize a Loss Of Offsite Power.

NOTE: With a Loss of Offsite Power, the pertinent operator RC-2 actions

are limited to checking ECCS.

SRO • Calls for maintenance support in restoring all emergency 4160 VAC buses. • Enters the RC EOP flowchart and directs a RWL band of +3-+50 inches.

NOTE: The crew may start HPCI before –35” RWL or DW pressures

increases to 1.85psig and therefore not be aware the auto initiation signal has failed.

ATC May enter 34AB-R22-003-2, Station Blackout, until the appropriate EDGs are supplying power to emergency buses.



Event Description: Loss of Offsite Power with EDGs failures.


SRO Once power is restored, directs operator to place Torus Cooling in service.

BOP

(Placard)

• Enters 34SO-E11-010-2, Residual Heat Removal or uses placard on 2H11-P601. • Places RHRSW in service • Prelubes RHRSW B pump • Overrides 2E11-F068B Low Discharge Pressure Interlock • Positions 2E11-F068B to 45% OPEN

- RHR HX B Diff Press Low (601-215) alarms • Starts RHRSW pump B • Places 2E11-F068B Low Discharge Pressure Interlock switch to

normal position. • Positions 2E11-F068B to obtain < 4400 GPM AND < 450 PSIG • RHR HX B Diff Press Low (601-215) alarm clears

BOP

(Placard)

• Place RHR loop B in Suppression Pool Cooling • Does NOT position the 2/3 Core Height Permissive switch.

(RWL will NOT be lowered to below 2/3 core height) • Does NOT place the Containment Spray valve Control switch in the

manual position. (LOCA signal is not present) • Confirm open 2E11-F048B, HX Bypass Vlv. • Close 2E11-F047B, Hx Inlet Vlv. • Confirm open 2E11-F003A, HX Outlet Vlv • Start RHR pump 2B

- SEC System Auto Initiation Signal Present (650-234) alarms - Auto Blow Down CS or RHR Press Permissive (602-312) alarms - RHR Flow Low (601-215) alarms

• Open 2E11-F028B, Torus Spray or Test Vlv. • Throttle open 2E11-F024B, Full Flow Test Line Vlv. and establish

RHR flow of less than or equal to 7700 gpm (R603B) - RHR Flow Low (601-215) alarm clears

• Open 2E11-F047B, Hx Inlet Vlv. • Close 2E11-F048B, Hx Bypass Vlv.

• Reports to SRO that RHR has been placed in Suppression Pool Cooling mode.




Event Description: 2C EDG auto starts but fails to tie/must lower then raise frequency to energize bus (fails to auto tie), 1B EDG will NOT run & 2A EDG will start & tie after manipulating Start switch.


SRO

• Directs ATC to address the plant electrical systems and enter the following: • 34AB-R22-003-2, Station Blackout • 34AB-R43-001-2, Diesel Generator Recovery

The ATC may start first with any of the Diesel Generators.

ATC Reviews EDG 2A annunciators and determines EDG 2A failed to Start.

IAW 34AB-R43-001-2, Diesel Generator Recovery, for EDG 2A : • Determines the EDG is not running • Determines the Auto Start System Operative Light is lit. • Places EDG 2A Start/Stop switch to START • Confirms EDG 2A has started and tied • Reports to SRO Bus 2E is now energized

ATC

Reviews EDG 1B annunciators and determines alarms 652-211, Lube Oil Press Low and 652-229, Emergency Engine Shutdown.

SIMULATOR OPERATOR: When called as the SO to investigate the EDGs, wait two minutes and report:

• EDG 1B has a break on the oil pump discharge line

ATC

IAW 34AB-R43-001-2, Diesel Generator Recovery, for EDG 2C : • Determines the EDG is running • Confirm Normal and alternate supply breakers OPEN • Lowers EDG frequency to 57 Hz, then Raise to 60 Hz (Critical Task) • Confirms EDG 2G output breaker closed • Reports to SRO Bus 2G is now energized



Event Description: 2C EDG auto starts but fails to tie/must lower then raise frequency to energize bus (fails to auto tie), 1B EDG will NOT run & 2A EDG will start & tie after manipulating Start switch.


The following can be performed in any order.

• As time allows, directs the SSS to perform the following: • RPS MG Set 2A - restarted • SSAC 2A local breaker - reclosed • Division I Station Service Battery Chargers

(may have been previously performed) • Vital AC Alternate Power Supply returned to service

• As time allows, directs the SSS to perform the following: • RPS MG Set 2B - restarted • RPS Alternate Supply from 2B Essential Cabinet - restarted • SSAC 2B local breaker - reclosed • Division II Station Service Battery Chargers

(may have been previously performed) • Vital AC Battery Charger returned to service




Event Description: RCIC 2E51-F013 Fails to Auto Open requiring manual operation of 2E51-F013 to inject prior to RWL reaching -180”


The malfunction for this event was in at the beginning of the scenario (rfE51_234 Final Value of BYPASS). The operator may start RCIC OR HPCI first, either is acceptable.

BOP • Starts RCIC for level control by performing the following at 2H11-P602 panel:

• Depresses RCIC Manual Initiation P/B • Confirms/Opens 2E51-F046 • Confirms/Starts Barom Cndsr Vac Pump • Confirms/Opens 2E51-F045 • Confirms/Opens 2E51-F019 • Confirms/Closes 2E51-F019 at flow > 79.3 gpm • Realizes 2E51-F013 did not auto open • Opens 2E51-F013 (Critical Task to maintain RWL >-180”) • Adjusts controller for desired flow and with SRO permission will raise

RWL to 32 to 42 inches

With Chief Examiners Permission, the Scenario will be terminated when Reactor water level is controlled in band and One Emergency bus is energized or as directed by the Chief Examiner.

of 28

NRC DRAFT Scenario Summary

Facility: E. I Hatch Scenario No.: 10-04 Op-Test No.: 2016-301

Initiating Conditions: Unit 2 is operating at 90% RTP. 2D11-K615B, Offgas Post Treatment Radiation Monitor, failed Downscale. RAS written.

Turnover Hydrogen Seal Oil System is currently aligned with the Emergency Seal Oil pump in service. After turnover restore to Normal Hydrogen Seal Oil System Lineup starting at step 7.3.3.2.5 of 34SO-N42-001-2. Once Hydrogen Seal Oil is returned to Normal lineup, increase reactor power to 100% RTP using the Recirculation System.

Summary:

• Event 1: Normal; The BOP will return Hydrogen Seal Oil to a Normal lineup IAW the system operating procedure.

• Event 2: Reactivity/Component/TS; A “2B” Recirculation Pump ASD Cell will auto bypass. This will result in a flow and power reduction. The ASD Cell Bypass and Recirc flow mismatch will be addressed by annunciator response procedures and an abnormal procedure. The SRO will address TS for Recirc flow mismatch.

• Event 3: Component; The RFPT loop seal will experience a low level condition causing

condenser vacuum to degrade. The ATC operator will isolate the loop seal drain which stops condenser vacuum decreasing. Once the loop seal level is restored, the operator returns the loop seal system to normal lineup.

• Event 4: Component; The “2C” Station Service Air Compressor (SSAC) will receive a high discharge air temp alarm. The operator will start the 2A & 2B SSAC’s and shutdown 2C SSAC.

• Event 5: Component/TS; The “2A” Reactor Recirculation pump seals will sequentially develop a failure that results in reactor coolant leaking into the Drywell (DW). Initially, only the Number 1 seal fails, followed a few minutes later by a Number 2 Seal failure. The ATC operator will trip and isolate the “2A” Reactor Recirc pump. The SRO will address TS for coolant leakage and single Recirc loop operation.

• Event 6: Component; A leak develops from the HPCI Torus suction line downstream of 2E41-F051 and can be isolated by closing 2E41-F051. Torus water level will continue to decrease until 2E41-F051 is isolated. (Critical Task)

• Event 7: Major; The plant will experience a Loss of Offsite power.

• Event 8: Component; 2C EDG fails to auto tie to the emergency bus and must have its frequency lowered, then raised to force the 2C EDG output breaker to close. 2A EDG will tie after Shutdown is manually reset. (Critical Task to energize at least one emergency bus)

• Event 9: HPCI F001 stuck closed and RCIC 2E51-F013 Fails to Auto Open requiring manual operation of 2E51-F013 to inject prior to RWL reaching -180”. (Critical Task)

of 28

NRC DRAFT Critical Task List


Critical Tasks

• A leak develops from the HPCI Torus suction line downstream of 2E41-F051 and can be isolated by closing 2E41-F051. Torus water level will continue to decrease until 2E41-F051 is isolated.

• Starts/ties either 2A or 2C EDG after a LOSP

• Maintains RWL above TAF or opens 7 ADS valves prior to reaching -180 inches

ES 301-4 Attributes Required Actual Items 1. Total Malfunctions 5-8 8 1. ASD 2B Cell bypass (Event 2)

2. RFPT Loop Seal failure (Event 3) 3. SSAC 2C High temperature (Event 4) 4. Recirc Seal leakage (Event 5) 5. Leak from HPCI Torus suction line (Event 6) 6. LOSP (Event 7) 7. 2 EDG failures (Event 8) 8. HPCI F001 fail shut/ RCIC F013 manual open

(Event 9) 2. Malfunctions After

EOP Entry 1-2 2 1. Loss of EDGs (Event 8)

2. HPCI F001 fail shut/ RCIC F013 manual open (Event 9)

3. Abnormal Events 2-4 4 1. Diesel Generator Recovery (Event 8) 2. Reactor Recirculation Pump(S) Trip, Or Recirc

Loops Flow Mismatch (Event 5) 3. Scram (Event 7) 4. Pipe Break in Sec. Containment(Event 6)

4. Major Transients 1-2 1 1. Station Blackout (Event 7) 5. EOPs entered,


1-2 2 1. RC (Non-ATWS) 2. Primary Containment


0-2 0 1. None


2-3 3 1. Leak from HPCI Torus suction line (Event 6) 2. Ties either 2A or 2C EDG after a LOSP

(Event 8) 3. Maintains RWL above TAF or opens 7 ADS

valves before -180 inches RWL (Event 9)

ILT-10 NRC Operating Exam Scenario 4

SHIFT TURNOVER

Safety Focus

UNIT 1 STATUS

Plant Conditions: Unit 1 is operating at 100% power Activities in progress: Maintaining Rated Thermal Power

UNIT 2 STATUS



Division I

Division II

Green Yellow

Orange Red

Scheduled

evolutions :

� Hydrogen Seal Oil System is currently aligned with the Emergency Seal Oil pump in service. After turnover restore to Normal Hydrogen Seal Oil System Lineup starting at step 7.3.3.2.5 of 34SO-N42-001-2.

� Once Hydrogen Seal Oil is returned to Normal lineup, increase reactor power to 100% RTP using the Recirculation System.

Surveillances due

this shift :

� None

Inop Equipment: � 2D11-K615B, Offgas Post Treatment Radiation monitor has failed

Downscale. IAW TS/TRM, I&C has placed 2D11-K615B in the trip condition while maintaining the function of 602-405, Post Treatment O/G Radiation Hi-Hi-Hi/Inop. A Caution Tag is attached to 2D11-K615B switch.

Active tagouts : � None


of 30

Appendix D Scenario Outline Form ES-D-1 NRC DRAFT

Facility: E. I Hatch Scenario No.: 10-05 Op-Test No.: 2016-301 Examiners: Operators: SRO RO BOP Initial Conditions. Unit 2 is operating at approximately 5% RTP preparing to swap Steam Packing Exhausters IAW step 7.3.1 of 34SO-N33-001-2. 2B21-F013M is inop for its mechanical lift setpoint, RAS written. Turnover: Continue with the Startup IAW 34GO-OPS-001-2 at step 7.4.1.

Event No.


Event Description

1 N/A N (BOP) Swap operating Steam Packing Exhausters IAW step 7.3.1 of 34SO-N33-001-2.

2 N/A R (ATC) Withdraw Control Rods to increase power.

3 mfT41_147 C (BOP)

Reactor Building Exhaust fan failure with the Standby Exhaust fan failing to start. Standby Exhaust fan manually started or SBGT fan started to re-establish Rx. Bldg. dP. (Critical Task)

4 mfC11_299 C (ATC) CRD Flow Controller fails in Auto requiring manual operation to re-establish CRD flow.

5 N/A C (BOP) TS (SRO)

HPCI Room fire

6 mfN21_88B I (ATC) Feedwater pump 2B cooling water controller failure.

7 rfC11_141 TS (SRO)

The Backup SDV valves will close due to a small air leak on 2C11-F040 requiring the SRO declare a TS Required Action Statement.

8 N71_70 M(ALL) Con Bay Flooding

9 loB21-F022BG1 loB21-F022BR2 loB21-F022BMG1 loB21-F022BMR2 loB21-F028BG1 loB21-F028BR2 loB21-F028BMG1 loB21-F028BMR2 loB21-F022B_AC1 loB21-F022B_DC1 loB21-F028B_AC1 loB21-F028B_DC1

C (BOP) ‘B’ MSL fails to isolate automatically on low vacuum, manual isolation required

10 diE51-F045 C (ATC) 2E51-F045 stuck closed, unisolate & inject with HPCI or ED prior to -180 inches. (Critical Task)




Event Description: Swap Steam Packing Exhausters IAW 34SO-N33-001-2, Seal Steam System.


3 Min BOP

• Places 2N33-C001 Stm Pkg Exh Blower 2A Control Switch to the OFF position.

• Opens 2N33-F025B, SPE Blower Disch Vlv, UNTIL Red OPEN indicating light ILLUMINATES.

• Places 2N33-C001 Stm Pkg Exh Blower 2B the Control Switch to the RUN.

• Throttles OPEN 2N33-F025B, SPE Blower Disch. Vlv, UNTIL Steam Packing Exhauster Vacuum indicates between 10-20 inches of water vacuum as read on 2N33-R601B.

• Closes 2N33-F025A, SPE Blower Disch Vlv. • Places the control switches in STOP for 2N33-F025A & B.

Simulator Operator enters the next event after surveillance is complete OR at the Chief Examiner’s request.



Event Description: Withdraw control rods to ~9% power and transfer the Reactor Mode Switch to Run


12 Min

SRO Directs ATC to continue rod withdrawal (46-23) to approximately 9% power

ATC

• Starting at Step 31, withdraws all control rods within a step to the withdraw limit

• Initials for control rod withdrawal • Dates for control rod withdrawal • Notifies SRO that reactor power is approximately 9%

SRO Directs:

• ATC confirm all APRMs indicate between 7% AND 10%. • BOP confirm operable APRM DOWNSCALE trips are clear

ATC Reports all APRMs indicate between 7% AND 10%.

BOP

Confirms operable APRM DOWNSCALE trips are clear by performing the following at the APRM ODAs at 2H11-P608:

• Depress the “ETC” key • UNTIL “TRIP STATUS” option ILLUMINATES. • Depress “TRIP STATUS” key,

THEN confirm “APRM FLUX DOWNSCALE ALARM” is NOT active.

SRO Directs BOP or ATC to confirm all IRMs are NOT Upscale.

BOP/ATC

Confirms no IRMs are UPSCALE by observing: • 2H11-P606 upscale lights not illuminated OR • 2H11-P603 upscale lights not illuminated OR • 2H11-P603 annunciator 603-221, “IRM Upscale” not illuminated



Event Description: Withdraw control rods to ~9% power and transfer the Reactor Mode Switch to Run


SRO

Confirms APRM/OPRM Operability: • at least three APRM channels per RPS Trip System are OPERABLE

AND • at least 2 "APRM TWO-OUT-OF-FOUR-VOTER-CHANNELS" per

RPS Trip System are OPERABLE. AND

• at least 3 OPRM channels per RPS Trip System are OPERABLE Confirms surveillances are current:

• 57SV-C51-001-0, APRM Functional Test • 57SV-C51-005-0, APRM Calibration • 57SV-C51-003-0, APRM Two Out of Four Logic Module FT

NOTE: APRM status can be confirmed by the turnover sheet or by directing STA to confirm operability.

Simulator Operator: If contacted for APRM status, inform the SS all APRMs are operable and all surveillances are current.

SRO Directs the ATC (or observes) the following annunciators are CLEAR:

• 603-232, MAIN STEAM LINE PRESS A LOW • 603-233, MAIN STEAM LINE PRESS B LOW

ATC If directed, reports annunciators 603-232 and 603-233 are clear.




Event Description: Reactor Building Exhaust fan (2T41-C007A) failure, broken coupling.


15 Min

At Lead Evaluators direction the Simulator Operator ENTERS (RB-3) mfT41_147, Rx Bldg Exhaust Fan A fails & standby does not auto start

On the 2H11-P650 panel alarms 650-214, Panel 2H11-P654 Panel System Trouble and 650-224, Panel 2H11-P657 Panel System Trouble are received

BOP • Announces the alarm to the SRO

SRO • The SRO repeats the alarm and dispatches the BOP operator to the 2H11-P654 & P657 panels.

NOTE: This failure will require entry into the SC EOP Flow Chart as well as two Abnormal procedures. The following sequence may be slightly out of order because of this.

If the operator enters 657-081 first, then the following actions will be addressed, otherwise skip to the next alarm.

BOP • Acknowledges 657-081, ‘RB Exhaust Fan 2T41-C007A/B Flow Low’ alarm and enters 657-081.

• Confirms the Standby Rx Bldg Vent Exhaust Fan has not automatically started (IT SHOULD HAVE AUTO STARTED), the operator will place the control switch for 2T41-C007A to OFF AND 2T41-C007B to Run.

• Confirm the following dampers are OPEN: • 2T41-F044A, R/B Inboard Isol Dampers Inacc. Areas Exhaust Fans Disch

on 2H11-P657. • 2T41-F044B, R/B Outboard Isol Dampers Inacc Areas Exhaust Fans Dish

on 2H11-P654. • 2T41-F028, Rx Bldg Vent Filter D005 Inlet Damper on

2H11-P657. • Confirm RB Exhaust flow on 2T41-R618 point 2 is indicating

approximately 6.5 KCFM.

• Once RB Exhaust and Supply flows have stabilized, confirms 654-001, RB Inside To Outside Air Diff Press Low, clears and notifies SS.





If the operator enters 654-001 first, then the following actions will be addressed.

BOP

• Acknowledges 654-001, ‘RB Inside To Outside Air Diff Press Low’ alarm and enters 654-001 and performs the following actions:

• Confirms Reactor Building to outside air differential pressure, is less than -0.06" WC, as indicated on 2T46-R604B, Sec Cnmt Diff Press B, red pen, 2H11-P700.

• Determines that the 2T41-C007A, ‘Rx Bldg Exhaust’ fan red light is illuminated and the 2T41-C001A, ‘Rx Bldg Supply’ fan red light is illuminated.

• Determines RB Exhaust flow on 2T41-R618 point 2 is indicating approximately 0 KCFM.

• Informs the SS that the 2T41-C007A, ‘Rx Bldg Exhaust’ fan is running (Red light illuminated) and the 2T41-C001A, ‘Rx Bldg Supply’ fan is running (Red light illuminated) and that Rx Bldg dp is 0”.

• Dispatches a SO/Maint to confirm 2R25-S065, BRKR 15 is closed and to investigate why the RB Exhaust fan has a low flow condition.

• Enters 34AB-T22-002-2, ‘Loss of Secondary Containment Integrity’ and 34AB-T22-003-2, ‘Secondary Containment Control’, and notifies Management of the plant conditions.

BOP

• Enters 34AB-T22-003-2, ‘Secondary Containment Control’. • Monitors secondary containment parameters. • Notifies the SS to enter 31EO-EOP-014-2, EOP Secondary

Containment flowchart due to low RB dp. • Dispatches a SO/Maint to investigate the low RB dp. • Enters 34AB-T22-002-2, ‘Loss of Secondary Containment Integrity’.





SRO

• Dispatches a SO/Maint to determine why the Rx Bldg Fan experienced low flow and why the standby fan did not start.

• Enters 31EO-EOP-014-2, EOP Secondary Containment flowchart.

Simulator Operator waits 2 minutes and reports to the SRO, as a SO, that Breaker 15 at 2R25-S065 is closed. Also report as Maintenance that the RB Exhaust fan A shaft is broken and the fan motor is making a loud noise.

SRO Directs operator to start SBGT with suction from the Rx Bldg or restart RB Ventilation using the 2T41-C007B, ‘RB Exhaust’ fan.

If the SRO directs SBGT started, the following actions will be taken, otherwise skip this part.

BOP • Starts the 2A or 2B SBGT with a suction from the Rx Bldg by performing

the following:

• At 2H11-P657 for 2A or 2H11-P654 for 2B performs the following action IAW the SBGT Placard or 34SO-T46-001-2, SBGT System.

• Opens 2T46-F001A(B) • Places SBGT A (B) in RUN position. • Confirms 2T46-F002A (B) OPENS. • Confirms SBGT A (B) HTR Red Light ILLUMINATES. • Confirms SBGT Flow increases to 1500 - 4000 SCFM.

SRO • Orders the Rx Bldg. Ventilation system re-started IAW 34SO-T41-005-2.

If the SRO directs RB Ventilation started the following actions will be taken, otherwise skip this part.

BOP • Enters 34SO-T41-005-2 and performs the following actions:

• Momentarily depresses Rx Bldg Supply Fans Reset. • Momentarily depresses Rx Bldg Recirc Fan 2T41-B017 Reset.





Simulator Operator, when requested to perform the following, inform the operator that the Unit 1 pushbuttons have been depressed.

BOP

• Requests U1 to depress 1T41/2T41, Rx Bldg/Rf Flr ISOL DMPR RESET A pushbutton, panel 1H11-P657.

• Requests U1 to depress 1T41/2T41, Rx Bldg/Rf Flr ISOL DMPR Reset B pushbutton, panel 1H11-P654.

• Momentarily depresses 2T41-D005, Reactor Building Filter, Deluge reset. • Open 2T41-F028, Rx Bldg Vent Filter D005 Inlet Damper. • Confirm Open/Open 2T41-F044A, Rx Bldg Inboard Isol Dampers

Inaccessible Areas Exhaust Fans Disch.

BOP

• Confirm Open/Open 2T41-F044B, Rx Bldg Outboard Isol Dampers Inaccessible Areas Exhaust Fans Disch, panel 2H11-P654.

• Place in RUN 2T41-C007B, Rx Bldg Vent Exhaust Fan. • Place in STBY 2T41-C007A, Rx Bldg Vent Exhaust Fan.

(Fan Broke may be left in off) • Confirm Open/Open 2T41-F011A, Rx Bldg Inboard Isol Dampers Supply

Fans Disch.

BOP

• Confirm Open/Open 2T41-F011B, Rx Bldg Outboard Isol Dampers Supply Fans Disch, panel 2H11-P654.

• Place in RUN 2T41-C001A or 2T41-C001B, Rx Bldg Supply Fan. • Place in STBY 2T41-C001B or 2T41-C001A, Rx Bldg Supply Fan.

BOP • Dispatches a SO to perform the following:

• ADJUST 2T41-FC-R027 to maintain Rx Bldg Vent Exhaust Flow approximately 6.5 KCFM as indicated by the green pen on 2T41-R618, flow recorder, located on 2H11-P657.

• ADJUST 2T41-FC-R022 to maintain Rx Bldg Vent Supply Flow

approximately 5.3 KCFM as indicated by the red pen on 2T41-R618, flow recorder, located on 2H11-P657.

• While maintaining approximately 6.5 KCFM for Rx Bldg Vent Exhaust

Flow and • 5.3 KCFM Bldg Vent Flow respectively, • ADJUST Flow Controllers 2T41-FC-R027 and 2T41-FC-R022 to obtain

0.25 inches water pressure on 2T46-DPR-R604A&B.





BOP • Informs the SRO that the Rx. Bldg. dP has returned to normal

SRO • If SBGT was started, directs SBGT returned to Standby and exits the SC EOP Flowchart

NOTE: It is not intended to wait until SBGT is returned to Standby, therefore, at Lead Evaluator’s direction, move on to the next Event



Event Description: CRD Flow Controller fails in Auto requiring manual operation to re-establish CRD flow.


10 Min At the Chief Examiner’s direction, Simulator Operator ENTERS (RB-4), mfC11_299.

ATC

• Receives CRD HYD TEMP HIGH, (603-140) alarm. • Determines that the CRD Flow Control Valve A has closed. • Determines 2C11-R600, CRD Flow Controller, output is at minimum and

has failed downscale. • Notifies SRO that the CRD Flow Controller has failed downscale. • Notifies I & C (if SRO does NOT) to investigate 2C11-R600.

ATC

• Enters: • CRD HYD TEMP HIGH, (603-140) • 34AB-C11-001-2, Loss Of CRD System

• Places 2C11-R600 controller in Manual. • Increases output of controller until CRD flow is approximately 50 gpm.

SRO

• Dispatches I & C to investigate 2C11-R600. • Dispatches a SO to monitor CRD drive temperatures. • Directs operator to perform actions of the ARP and 34AB-C11-001-2.

Simulator Operator, if asked, AFTER 3 minutes from being dispatched, report

to the SRO, as an SO, that 2C11-R600 FCV does NOT have any air leaks and appears to be functioning properly. Simulator Operator, if asked, AFTER 5 minutes from being dispatched, report to the SRO, as I&C, that 2C11-R600 FCV appears to be functioning properly with a zero signal going to the FCV and the FCV is closed. Simulator Operator, if asked to swap FCV, AFTER 5 minutes from being dispatched, report to the SRO, as I&C, that 2C11-R600 FCV is receiving a zero signal from the controller and it appears the flow controller is malfunction. Simulator Operator, after 5 minutes from being dispatched, report to the SRO, as I&C, that 2C11-R600 controller has malfunctioned and will need repair.



Event Description: CRD Flow Controller fails in Auto requiring manual operation to re-establish CRD flow.


When requested and if the CRD Hyd Temp High alarm (603-140) is still illuminated, Simulator Operator, as SO checking CRD Temps, reports that 2 CRD drives are >250°F. 30-11 at 260°F, 22-35 at 265°F and some others are slowly increasing. If CRD Hyd Temp High alarm (603-140) is NOT illuminated, Simulator Operator, as SO checking CRD Temps, reports that 2 CRD drives had been >250°F but are now below 250°F and trending down.

ONCE Event 4 is complete, DELETE mfC11_299 and notify the SRO that 2C11-R600 is now repaired.

SRO • Directs ATC to return 2C11-R600 to Automatic operation.

ATC

• Returns CRD System to normal configuration by performing step 7.1.2.29 of 34SO-C11-005-2 or by performing step 4.9 of 34AB-C11-001-2: • Manually adjusts 2C11-R600 to match setpoint • Places 2C11-R600 to Auto • Confirms system parameters are normal • Confirms 603-140, CRD Hyd Temp High, alarm is clear

• Reports to SRO that CRD flow controller has been restored to normal

configuration. Simulator Operator, at the Chief Examiners direction, PROCEEDS to the next

event.



Event Description: HPCI Room Fire.


11 min

Simulator Operator;

Prior to placing the Mode Switch to the Run position AND with Chief Examiner’s direction, NOTIFY the Shift Supervisor, as the SO on rounds, that heavy smoke is coming from the Unit 2 HPCI room and you cannot see what is actually burning. You are leaving the area.

SRO • Directs BOP to enter 34AB-X43-001-2, Fire Procedure, and make the announcement.

BOP

• Enters 34AB-X43-001-2, Fire Procedure, and performs the following: • Sounds the fire alarm (SIREN) over the plant PA system,

(repeats 2 times – may get U1 operator to handle repeats) • Notifies Rad Waste Control room that Reactor Building floor drain

sump pumps are in Automatic • Notifies Radiation Protection to respond to the fire area

• Transitions to section 8.0 for Fire in ‘Reactor Bldg.’ • Stops the following equipment at Panels 2H11-P654 & P657:

• 2T41-C001A, Rx Bldg Supply Fan. • 2T41-C001B, Rx Bldg Supply Fan. • 2T41-C007A, Rx Bldg Exhaust Fan. • 2T41-C007B, Rx Bldg Exhaust Fan. • 2T41-C002A, Refuel Flr Vent Supply Fan. • 2T41-C002B, Refuel Flr Vent Supply Fan. • 2T41-C005A, Refuel Flr Vent Exh Fan. • 2T41-C005B, Refuel Flr Vent Exh Fan.

• Acknowledges annunciators at Panels 2H11-P654 & P657: • Refueling Floor Outside Air Diff Press Low (657-001) • RB Inside to Outside Air Diff Press Low (654-001)

Simulator Operator, Wait 2 minutes and NOTIFIES the Shift Supervisor, as the Fire Brigade Leader, to place HPCI Aux Oil pump in PTL-Off to prevent HPCI from starting while Fire Brigade is in the HPCI room.

BOP

• Transitions to section 8.5 for Fire in ‘South Half of Reactor Bldg.’ • Transitions to section 8.5.6 for Fire in ‘HPCI Room’

• Places 2T41-B005A HPCI Pump Rm Cooler in Off, P657. • Places 2T41-B005B HPCI Pump Rm Cooler in Off, P654. • Places HPCI Aux Oil pump in PTL Off, P601.

• Dispatches a SO to open Brkr 6 in 2R25-S002





Simulator Operator When directed to open Brkr 6 in 2R25-S002, insert malfunctions mfE41_235A and mfE41_235B (RB-5)

• Closes the following HPCI Valves on P601 • 2E41-F002, HPCI Isolation Valve • 2E41-F003, HPCI Isolation Valve • 2E41-F041, HPCI Torus Suction Valve • 2E41-F042, HPCI Torus Suction Valve

Simulator Operator, Once the above HPCI valves are closed, NOTIFY the Shift Supervisor that the fire in the HPCI room is OUT. A storage locker under the stairwell was on fire. We put out the fire without spraying water on any HPCI components. Ventilation and HPCI can be restored to normal.

SRO

• Announces to the crew the fire is out. • As time allows, directs the operator to leave the HPCI Aux oil pump in

PTL Off. • Refers to TS for HPCI INOP

• 3.5 ECCS and RCIC System • 3.5.1 ECCS Operating Enters a 1 hour RAS per TS 3.5.1.C.1 to verify by admin means RCIC is operable and per TS 3.5.1.C.2 to restore HPCI to operable status within 14 days.

• As time allows, Refer to TRM 3.7.2.D.1, ECCS and RCIC Room Coolers and immediately declares HPCI INOP.

SRO • Directs BOP to start Rx Bldg Ventilation System.

BOP

• Starts Rx Bldg Ventilation IAW 34SO-T41-005-2, Section 7.1 • Momentarily Depress the following pushbuttons:

• 2T41-SR1, Rx Bldg Supply Fans Reset. • 2T41-SR3, Rx Bldg Recirc Fan 2T41-B017 Reset at 2H11-P657. • 2T41-D005, Reactor Building Filter, Deluge reset at 2H11-P657.





Simulator Operator, when requested to depress U1 Rx Bldg. / Rf Flr Isol Dmpr Reset A & B, DEPRESS RB-9 rfT41_299 and then WAITS one minute and INFORMS Unit 2 BOP that U1 Rx Bldg./Rf Flr Isol Dmpr Reset A & B have been reset.

BOP

• Requests Unit 1 operator to depress: • 1T41/2T41, Rx Bldg/Rf Flr Isol Dmpr Reset A pushbutton, at

1H11-P657. • 1T41/2T41, Rx Bldg/Rf Flr Isol Dmpr reset B pushbutton, at

1H11-P654.

BOP

• Confirms OPEN the following dampers: • 2T41-F028, Rx Bldg Vent Filter D005 Inlet Damper at 2H11-P657. • 2T41-F044A, Rx Bldg Inboard Isol Dampers Inaccessible Areas

Exhaust Fans Disch at 2H11-P657. • 2T41-F044B, Rx Bldg Outboard Isol Dampers Inaccessible Areas

Exhaust Fans Disch at 2H11-P654. • Place in RUN 2T41-C007B, Rx Bldg Vent Exhaust Fan at

2H11-P657 (2T41-C007A shaft broke). • Place in STBY 2T41-C007A, Rx Bldg Vent Exhaust Fan at

2H11-P657 (left in off).

BOP

• Confirms: • in RUN 2T41-B017, Fan 1(Fan 2) Rx Bldg Recirc at 2H11-P657. • in STBY 2T41-B017, Fan 2(Fan 1) Rx Bldg Recirc at 2H11-P657. • OPEN/OPEN 2T41-F011A, Rx Bldg Inboard Isol Dampers Supply

Fans Disch at 2H11-P657. • OPEN/OPEN 2T41-F011B, Rx Bldg Outboard Isol Dampers Supply

Fans Disch, at 2H11-P654.

BOP

• Aligns fan in: • RUN 2T41-C001A OR 2T41-C001B, Rx Bldg Supply Fan. • STBY 2T41-C001B OR 2T41-C001A, Rx Bldg Supply Fan.

• Verifies Rx Bldg Vent Exhaust Flow approximately 6.5 KCFM as indicated by the green pen on 2T41-R618 located on 2H11-P657

• Rx Bldg Vent Supply Flow approximately 5.3 KCFM as indicated by the red pen on 2T41-R618, flow recorder, located on 2H11-P657.

BOP • Verifies flow is maintaining 0.25 inches water pressure on

2T46-DPR-R604A&B. • Sends SO to verify setpoints of local temperature controllers.





• Verifies in RUN fan: • 2T41-B007, Rx Bldg Cooler Working Floor EL 130 North • 2T41-B008, Rx Bldg Cooler Working Floor EL 130 South • 2T41-B026, Rx Bldg Primary Cooler Mn Steam Pipe Chase • 2T41-B010, Rx Bldg Cooler Equip Room EL 164 • 2T41-B011, Rx Bldg Cooler Operating Floor EL 185 • 2T41-B014, Rx Bldg Cooler RWCU Pmp/Hx EL 158 • 2T41-B015, Rx Bldg Cooler Operating Floor EL 185 • 2T41-B016, Rx Bldg Cooler Working Floor EL 203

• Verifies in AUTO 2T41-B009, Rx Bldg Secondary Cooler Mn. Steam Pipe Chase.



Op-Test No.: 2016-301 Scenario No.: 10-05 Event No.: 6 Page 16 of 30 Event Description: 2B RFPT Cooling water controller failure. TC repair & return to Automatic.


8 Min At the Chief Examiner’s direction, Simulator Operator, INSTRUCT the BOP operator by phone to stay on the line until told to hang up, THEN ENTER: (RB-6) mfN21_88B, Feedwater Pump Lube Oil Cooling System Failure.

ALL

The following alarms will annunciate: • RFPT 2B BRG OIL TEMP HIGH, (650-315) • RFPT 2B BRG TEMP HIGH, (650-333) • RFP/COND BRG METAL TEMP HIGH, (650-112), (approximately 1

minute later if 2B RFPT PSW TCV is NOT opened in a timely manner)

NOTE: The ATC may immediately place the controller in manual IAW 31GO-OPS-021-0, Manipulation and Control of Equipment OR NMP-OS-007-001, Conduct of Operations Standards and Expectations, responding to a failed controller.

ATC

• Addresses the high temp annunciator, pulling the ARP and confirms temperatures:

• Dispatches BOP to panel 2H11-P655, checks all temperature indicators on

2N32-R616 to determine actual oil temperatures. • Confirms that RFPT 2B Oil Temp controller, 2P41-R606, on panel 2H11-

P650 is adjusted for 110 to 130°F.

ATC

• Recognizes the automatic function of the controller has failed, closing the cooling water valve.

• Places the controller in manual, depresses the open/increase pushbutton, opening the valve. Oil temperatures begin decreasing and the alarm extinguishes.

SRO Notifies maintenance of the 2P41-R606, RFPT temperature controller, problem.



Op-Test No.: 2016-301 Scenario No.: 10-05 Event No.: 7 Page 17 of 30 Event Description: The Backup SDV valves will close due to a small air leak on 2C11-F040

requiring the SRO declare a TS Required Action Statement.


10 Min

Simulator Operator, at the Chief Examiner’s direction, ENTERS (RB9) rfC11_141, SDV Outboard Valves close AND 3 minutes later ENSURES Event Trigger EGC11-4 ACTIVATES mf60311307, SDV Not Drained, alarm.

ALL • When the SDV NOT DRAINED, (603-119), alarm is received, recognizes

that the SDV Outboard Valves have closed. (May recognize prior to alarm by scanning the control boards)

ATC

• Enters 603-119 and performs the following: • Determines that 2C11-F035A, 2C11-F035B and 2C11-F037 have

closed. • Determines status of all Scram Valves (blue lights are not lit) on P603

display. • Determines status of SCRAM VLV PILOT AIR HDR PRESS

HIGH/LOW, (603-131), (NOT LIT) • Determines if any Rod Drift lights on P603 (None). • Confirms Scram Disch Vol Isol Test Switch in Normal. • Dispatches SO to the CRD drives to check for leaking Scram Outlet

Valves. • Dispatches SO/Maintenance to determine if an air leak exists on the

SDV valve piping.


Op-Test No.: 2016-301 Scenario No.: 10-05 Event No.: 7 Page 18 of 30 Event Description: The Backup SDV valves will close due to a small air leak on 2C11-F040

requiring the SRO declare a TS Required Action Statement.


Simulator Operator, if a SO was sent to the solenoid power switches, 2 minutes later, REPORTS that the Power supply switches in the RPS M/G Set room for 2C11-F040 solenoids are turned to “ON”. Simulator Operator, wait 3 minutes, then REPORT as a SO that’s been dispatched to check for air leaks, there is an air leak on the piping from 2C11-F040 on the 130’ Rx. Bldg. It appears that the copper piping was bumped by someone or something, causing a crimp and small leak in the piping.

ATC • Notifies the SRO that there is an air leak and crimped piping at 2C11-F040.

SRO

• Enters Tech spec 3.1.8 Condition A which requires the SDV line to be isolated within 7 days.

• May inform Maintenance to correct the associated air leak.




Event Description: Circ Water Leak Causing Con Bay Flooding/Loss of Vacuum


15 min Simulator Operator: At Lead Evaluator’s direction, press (RB-7) to activate: mfN71_70, Circulation Water System Leak (0 – 100,000 gpm), Final 100, Ramp 1000.

All • Recognizes that Condenser Water Box A R602A dP is approaching 0.

SRO • As time allows, dispatches a SO to locally investigate the reduced Circ Water dP and for a possible leak.

SRO

• When 650-164, Condenser Room Flooding, alarm is received, directs the ATC enter 34AB-C71-001-2, Scram Procedure, and scram the reactor.

• Assigns the ATC to perform RC-1. • Assigns the BOP operator to perform RC-2 and RC-3. • Enters 31EO-EOP-010-2, RC EOP flow chart • Assigns a RWL band between 3” and 50”

Simulator Operator: Ensure EGC71-1 INSERTS mfN61-73 when mode switch is taken to Shutdown, (Final of 100, Ramp 1000) AND EGC71-14 INSERTS overrides both CRD pumps (diC11B-S3A & S3B to STOP).

ATC


RWM display. • Notifies SRO of rod position check. • Places SDV isolation valve switch to "isolate" & confirms closed. • If not tripped, places the Recirc pumps at minimum speed. • Inserts SRMs and IRMs. • Ranges IRMS to bring reading on scale. • Notifies the SRO when the above actions are complete.





BOP


• Checks ECCS Injection Systems and secure as necessary. • Ensures FW Master Controller setpoint reduces to 9 inches and output

reduces to 25% of previous value. • IF set down does not auto function, then manually reduces FW Master

Controller setpoint to approximately 9 inches.

NOTE: When BOP attempts to start RCIC GO TO EVENT 9 for actions.

BOP

• When feed flow is less than the capacity of the S/U level control valve (≈ 1.5 mlbm/hr), then: • Opens 2N21-F125. • Places 2C32-R619, FW S/U level control valve controller, in Auto, set

at approximately 9 inches. • Confirms closed 2N21-F110.

BOP

At 2H11-P603 maximizes CRD flow by: • May Reset the LOCA pushbuttons for the pump being started. • Places 2C11-R600, CRD Flow Control, in manual at zero output • Starts CRD pump 2C11-C001A(B) • Realizes the selected CRD pump will not start • As time allows, notifies SRO that Both CRD pumps will not start • As time allows, dispatches SO/Maintenance to investigate Reset the

LOCA pushbutton for the Standby CRD Pump





EXAMINER NOTE: SRVs actuate in LLS at 1120 psig and then control pressure between 850 - 990 psig.

BOP

• Performs RC-3 consisting of: • Monitor RPV pressure. • If necessary, allow RPV pressure to exceed 1074 psig then cycles any

SRV to initiate LLS. • If necessary, verifies LLS actuates at 1120 psig • Maintain RPV pressure between 1074 and 800 psig. • Notify SRO of pressure control system operation.

SRO • Enters 31EO-EOP-010-2, RC EOP flow chart • If time permits, directs the ATC to reset the scram.

ATC • If directed, resets the scram IAW 34SO-C71-001, “Scram Procedure”.




Event Description: ‘B’ MSL fails to isolate automatically on low vacuum, manual isolation required


Simulator operator ensures EGB21-1 & EGB21-2 for deleting overrides for MSIV lights is ACTIVATED when the 2B21-F022B AND F028B is positioned to close.

ALL • Recognizes the B MSL failed to isolate on low condenser vacuum and places both control switches for 2B21-F022B and F028B to close.

SRO • Directs operator to place B MSL control switches, 2B21-F022B and

F028B, to close and confirm valves indicate close.

BOP • Places B MSL control switches, 2B21-F022B and F028B, to close and

confirms valves indicate closed.

SRO As time allows, directs entry into 34AB-N71-001-2, Circulating Water System Failure

BOP • Confirms Closed MSIVs • Confirms Closed MSL Drains • Confirms close 2N11-F004A and 2N11-F004B, RSSV's




Event Description: 2E51-F045 stuck closed, unisolate & inject with HPCI or ED prior to -180 inches.


BOP

• Starts RCIC for level control by performing the following: • Depresses RCIC Manual Initiation P/B • Confirms Open 2E51-F046 • Confirms Start of Barom Cndsr Vac Pmp • Confirms / Open 2E51-F045 (realizes 2E51-F045 did NOT open and

manually attempts to open valve) • NOTIFIES SRO 2E51-F045 will NOT open • Dispatches SO/Maintenance to investigate 2E51-F045

SRO Directs BOP to unisolate HPCI and place in-service IAW 34SO-E41-001-2, HPCI System.





NOTE: The BOP will perform the following at 2H11-P601 panel.

BOP

IAW 34SO-E41-001-2 Section 7.4.10, Unisolating HPCI With Initiation Signal Present; • Places 2E41-C002-3, HPCI Aux Oil Pump, in PULL-TO-LOCK,

panel 2H11-P601 • Places the Auto Isolation Signal A keylock switch to the RESET position

AND returns to NORMAL. (Verifies Auto Isolation Signal white light is EXTINGUISHED)

• Places the Auto Isolation Signal B keylock switch to the RESET position AND returns to NORMAL. (Verifies Auto Isolation Signal white light is EXTINGUISHED)

• Confirms the following alarms are cleared: • HPCI ISOLATION TRIP LOGIC A INITIATED (601-115) • HPCI ISOLATION TRIP LOGIC B INITIATED (601-121)

• Determines which HPCI valve to use:

• IF necessary, warm AND pressurize the HPCI Steam Line by performing either: step 7.4.10.5.1 (Pressurizing HPCI steam supply with 2E41-F002) OR step 7.4.10.5.2 (Pressurizing HPCI steam supply with 2E41-F003)

(Step 7.4.10.5.1) • Confirms closed 2E41-F002, Inbd Steam Isol Valve • Opens 2E41-F003, Outbd Steam Isol Valve • Opens 2E41-F054, Drain Pot Trap Byp Valve • Slowly throttles open 2E41-F002 • Fully Opens 2E41-F002 when turbine steam inlet pressure

(2E41-R602) is within 50 psig of reactor pressure on 2B21-R623A or B (P601 panel), then places control switch to stop position.

• 601-217, HPCI Isolation Vlv F002/F003 Not Fully Open will clear • Closes 2E41-F054, Steam Line Drain Valve when annunciator

601-110, HPCI Turbine Inlet Drain Pot Level High is clear NOTE: 601-110 may not alarm.





(Step 7.4.10.5.2) • Confirms closed 2E41-F003, Outbd Steam Isol Valve • Opens 2E41-F002, Inbd Steam Isol Valve • Opens 2E41-F054, Drain Pot Trap Byp Valve • Slowly throttles open 2E41-F003 • Fully Opens 2E41-F003 when turbine steam inlet pressure

(2E41-R602) is within 50 psig of reactor pressure on 2B21-R623A or B (P601 panel), then places control switch to stop position.

• 601-217, HPCI Isolation Vlv F002/F003 Not Fully Open will clear • Closes 2E41-F054, Steam Line Drain Valve when annunciator

601-110, HPCI Turbine Inlet Drain Pot Level High is clear NOTE: 601-110 may not alarm.

SRO

• Per CP-1 • Order BOP to Inhibit ADS (if not previously done) • Orders BOP to start ALL RHR & Core Spray pumps. • Orders emergency depressurization once RWL decreases

below –155” and prior to –180”. • Orders all available Table L-1 systems injecting until RWL raises above –

155”. • As time permits, directs Torus Cooling to be placed in service.

BOP • Starts ALL RHR & Core Spray pumps (P601) by placing switches to start IAW placard.





ATC

• Opens 7 ADS valves prior to RWL reaching -180” by: (Critical Task) • Placing switches for the ADS valves to OPEN. • Confirms ALL ADS valve red lights illuminate. • Confirms ALL ADS valve yellow lights illuminate. • Confirms Reactor pressure is decreasing. • Notifies the SRO that ALL ADS valves are open.

BOP

• Opens RHR and Core Spray injection valves open once the Reactor Pressure Low 500 psig alarm illuminates.

• Verifies injection from Core Spray and RHR pumps begins as soon as reactor pressure decreases below the shut off head of the pumps.

• When RWL is restored above Top OF Active Fuel throttles flow for C/S and RHR per the SRO directions.

ATC

NOTE: The operator may place torus cooling in service by using the Placard that's available or using the appropriate section of the procedure. These steps assume the Placard is used. The A and/or B loop of RHR may be used depending on Torus temperature. The following steps are written assuming "B" loop and "B" pump is used. If/When "A" loop is used, substitute "A" for "B" for valves and if "B" pump is not used substitute "A", "C", or "D" for "B" pump.

ATC

• Enters 34SO-E11-010-2, Residual Heat Removal • Places RHRSW in service

• Prelube RHRSW pump • Overrides 2E11-F068B (A) Low Discharge Pressure Interlock • Positions 2E11-F068B (A) to 45% OPEN • Starts RHRSW pump B (A) • Places 2E11-F068B (A) Low Discharge Pressure Interlock switch to

normal position. • Positions 2E11-F068B (A) to obtain < 4400 gpm AND

< 450 psig





ATC

• IF desired to start a SECOND RHRSW pump, • Throttles 2E11-F068B (A) to achieve max flow rate

(not to exceed 4400 GPM). • Opens 2E11-F068B (A) an additional 5%. • Starts second RHRSW Pump. • Positions 2E11-F068B (A) to obtain < 8800 gpm AND < 450 psig

ATC

• Places RHR B (A) Loop in Torus cooling per the placard by performing the following steps: • Opens 2E11-F048B (A) • Closes 2E11-F047B (A). • Opens 2E11-F003B (A). • Starts RHR Loop B (A) pump • Opens 2E11-F028B (A) • Receives annunciator Auto Blowdown CS OR RHR Press • Receives annunciator "SEC System Auto Initiation Signal Present." • Throttles OPEN 2E11-F024B (A) • Opens 2E11-F047B (A) • Ensures RHR flow is < 11,500 GPM, THEN Closes 2E11-F048B • Notifies the SRO that RHR "B" (A) loop is in service • May place the second pump in service.

With Chief Examiners Permission, the Scenario will be terminated when reactor water is being controlled using HPCI or RHR/CS following the Emergency Depressurization, with reactor pressure below 50 psig or as directed by the Chief Examiner.

of 30

NRC DRAFT Scenario Summary


Initiating Conditions: Unit 2 is operating at approximately 5% preparing to swap Steam Packing Exhausters per step 7.3.1 of 34SO-N33-001-2. 2B21-F013M is inop for its mechanical lift setpoint, RAS written.

Turnover Continue with the Startup IAW 34GO-OPS-001-2 at step 7.4.1. Summary:

• Event 1: Normal; Swap Steam Packing Exhausters IAW 34SO-N33-001-2, Seal Steam System.

• Event 2: Pull control rods to continue Startup to ~ 9% power, raise pressure set to 920.

• Event 3: Component; Reactor Building Exhaust fan failure with the Standby Exhaust fan failing to start. The operator takes manual control to place the Standby Exhaust fan or SBGT in service to re-establish the required Reactor Building dP. (Critical Task)

• Event 4: Component; CRD Flow Controller fails closed causing a loss of normal CRD flow. The ATC will place the controller in manual and restore CRD flow.

• Event 5: Component; HPCI Room Fire.

• Event 6: Instrument; Feedwater pump 2B cooling water controller will fail requiring the ATC to maintain cooling water manually.

• Event 7: The Backup SDV valves will close due to a small air leak on 2C11-F040 requiring the SRO declare a TS Required Action Statement.

• Event 8: Major; Condenser Bay Flooding is received which de-energizes 4160V 2A & 2B, requiring the operator to insert a manual scram. ‘B’ MSL fails to automatically close; manual works.

• Event 9: ‘B’ MSL fails to isolate automatically on low vacuum, manual isolation required.

• Event 10: Component; RCIC F045 valve stuck closed, must un-isolate HPCI for injection or ED and inject with Low Pressure Systems prior to -180 inches. (Critical Task)

of 30

NRC DRAFT Critical Task List


Critical Tasks

• Standby Exhaust fan manually started or SBGT fan started to re-establish Rx Bldg dp. (Event 3)

• RCIC F045 valve stuck closed, must un-isolate HPCI for injection or ED and inject with Low Pressure Systems prior to -180 inches. (Event 9)

ES 301-4 Attributes Required Actual Items 1. Total Malfunctions 5-8 7 1. Reactor Building Exhaust fan failure (Event 3)

2. CRD Flow Controller fails closed (Event 4) 3. HPCI room fire (Event 5) 4. RFP 2B cooling water failure (Event 6) 5. Condenser Bay flooding (Event 8) 6. MSIV line fails to isolate (Event 9) 7. 2E51-F045 failure (Event 10)


1-2 2 8. MSIV line fails to isolate (Event 8) 9. 2E51-F045 failure (Event 9)

3. Abnormal Events 2-4 3 1. 34AB-T22-002-2, Loss of Sec Cont Integrity (Event 3)

2. 34AB-X43-001-2, Fire Procedure, (Event 5) 3. 34AB-C71-001-2, Scram Procedure (Event 8)

4. Major Transients 1-2 1 1. Condenser Bay Flooding-Scram (Event 8) 5. EOPs entered,


1-2 1 1. RC (Non-ATWS) (Event 8)


0-2 1 1. CP-1, (Event 10)


2-3 2 1. Stby Exh fan manually started or SBGT fan started to re-establish Rx Bldg dp. (Event 3)

2. 2E51-F045 failure/un-isolate HPCI or the Reactor Emergency Depressurized prior to RWL reaching -180 inches. (Event 10)

ILT-10 NRC Operating Exam Scenario 5

SHIFT TURNOVER

Safety Focus

UNIT 1 STATUS

Plant Conditions: � Unit 1 is operating at 100% power � Activities in progress: Maintaining Rated Thermal Power

UNIT 2 STATUS

Plant Conditions: � Unit 2 is operating at approximately 5% preparing to swap Steam Packing Exhausters per step 7.3.1 of 34SO-N33-001-2.

� FWLC System is operating in dP Mode.


Division I

Division II

Green Yellow

Orange Red

Scheduled evolutions : � Continue with the Startup IAW 34GO-OPS-001-2 at step 7.4.1. � Continue with control rod withdrawal at Step 31 46-23.


shift :

� None

Inop Equipment: � None

Active tagouts : � None


Southern Nuclear Company

Operations Training Job Performance Measure (JPM)

DRAFT CR-SIM 1 (RO & SRO-I)

Title Withdraw CRs: Drifting Rod (Alt Path)

Author: Anthony Ball

Media Number: CR-SIM 1 2016-301

Time: 15.0 Minutes

Reviewed By: Date:

Reviewed by Instructional Technologist or designee Date

Approved By

Date

Course Number

Program Name

OPERATIONS TRAINING

Media Number

CR-SIM 1 2016-301

Rev. No. Date Reason for Revisions Author’s Initials

Sup’s Initials

00 09/16/98 Initial development SCB DHG

01 03/05/99 Revised to correct inaccurate procedure number. SCB DHG

02 10/20/99 Upgrade format, procedure revision RAB RSG

03 11/06/00 Include objective number RAB DHG

04 03/19/02 Include initial operator statement RAB RAB

05 03/17/05 Deleted “S” from procedure numbers, changed Revision and Rev. numbers to “Current Version”. Changed to Nuclear Plant Operator. Updated Simulator IC# and initial conditions.

ARB DHG

06 06/27/05 Revised Initial License statement for successful completion

RAB RAB

07 06/30/06 Remove Response Cues RAB RAB

08 09/30/10 Added HU Pass/Fail criteria. Updated Southern Company logo.

ELJ CME

09 10/28/10 Revised simulator setup and rod group to be withdrawn. Changed the rod that drifts to 30-23

ELJ CME

09.1 10/17/11 Reviewed JPM against current procedure. Added Fundamental

question to new Attachment 1. Correct rod group and rods to be

moved. Replaced malfunction with the correct one.

MMG ALS

9.2 Reviewed JPM against current procedure. Changed

“Media Number” to CR-SIM 2 2016-301. After NRC

Exam, Media Number will be changed LR-JP-25031

ARB

Line Contributors The following individuals contributed to the development of this lesson plan.

Rev. No. List of Contributors

09.1 MMG

CR-SIM 1 2016-301 Page 1 of 6

UNIT 1 ( ) UNIT 2 (X)

TASK TITLE: Withdraw CRs: Drifting Rod (Alt Path)

JPM NUMBER: CR-SIM 1 2016-301

TASK STANDARD: The task shall be completed when the Reactor has been manually scrammed per 34AB-C11-004-2.

TASK NUMBER: 001.010

OBJECTIVE NUMBER: 001.010.A

PLANT HATCH JTA IMPORTANCE RATING: RO 3.57 SRO 3.52

K/A CATALOG NUMBER: 201003A203 K/A CATALOG JTA IMPORTANCE RATING :

RO 3.40 SRO 3.70

OPERATOR APPLICABILITY: Nuclear Plant Operator (NPO)

GENERAL REFERENCES: Unit 2

34GO-OPS-065-0 34AB-C11-004-2 (current versions)

REQUIRED MATERIALS: Unit 2

34GO-OPS-065-0 (current version) 34AB-C11-004-2 (current version) Control Rod Movement Sequence Sheet

APPROXIMATE COMPLETION TIME: 15.0 Minutes SIMULATOR SETUP: REFER TO SIMULATOR SETUP SHEET ON THE

FOLLOWING PAGE

CR-SIM 1 2016-301 Page 2 of 6

SIMULATOR SETUP

Simulator Initial Conditions: 1. RESET the Simulator to 7% RTP, Xfer to RUN and leave in FREEZE. 2. INSERT the following MALFUNCTIONS:

MALF # TITLE FINAL VALUE

RAMP RATE

ACT. TIME

mfC12_24 30-15 Control Rod Failure (Drift Out) 30.23 N/A 00000

3. Take the Simulator OUT OF FREEZE and PERFORM the following MANIPULATIONS:

A. Sign off all rods in step 25 up to 38-31 (student will move 38-31 and 30-15).

B. Ensure that drive water dP is 260 psid and stable.

4. PLACE the Simulator in FREEZE until the INITIATING CUE is given. 5. ESTIMATED Simulator SETUP TIME : 15 Minutes NOTE: The simulator operator will act as second verifier for rod movement and read the

pre-job brief to the operator.

UNIT 2

READ TO THE OPERATOR INITIAL CONDITIONS:

1. A normal plant startup is in progress per 34GO-OPS-001-2, “Plant Startup,” and is currently at Step 7.4.3.

2. Rod withdrawal to achieve 6-7% on the APRMs is in progress.

3. Rods in Step 25 of the Pull Sequence are being withdrawn.

4. Rod Worth Minimizer is operable and has been loaded with the correct movement sequence, which has been approved by the Reactor Engineering Supervisor.

5. The pre-job brief has been completed.

INITIATING CUES:

Withdraw Controls Rods in Step 25 to their withdraw limit.

CR-SIM 1 2016-301 Page 4 of 6

STEP #

PERFORMANCE STEP STANDARD SAT/UNSAT (COMMENTS)

(** Indicates critical step)

For INITIAL Operator Programs: For OJT/OJE ; ALL PROCEDURE STEPS must be completed for Satisfactory Performance. For License Examinations; ALL CRITICAL STEPS must be completed for Satisfactory Performance.

(1) The standard for human performance tools, safety, PPE, and other pertinent expectations is considered met provided any deviations are minor and have little or no actual or potential consequence. Errors may be self-corrected provided the action would not have resulted in significant actual or potential consequences. Reference: NMP-TR-111, “On-The-Job Training and Task Performance Evaluation”

START TIME:__________

1. Operator identifies the procedure needed to perform the task.

Operator has IDENTIFIED the correct procedure as 34GO-OPS-065-0.

2. Operator reviews the procedure’s precautions and limitations.

Operator has REVIEWED the precautions and limitations.

PROMPT: WHEN the operator addresses an approved copy of the Control Rod

Movement Sequence Sheet, GIVE the operator the Control Rod Movement Sequence Sheet.

3. Operator identifies the materials that are required.

Operator has IDENTIFIED and OBTAINED Control Rod Movement Sequence Sheet.

NOTE: The operator may select any control rod in Rod Step 25, although the

operator should proceed in consecutive order.

IF THEN

PASS

� Human performance tools, safety, PPE met (1), AND � For initial trg all steps completed correctly OR � For continuing trg, critical steps (if used) completed

correctly

� Mark the JPM as a PASS

FAIL � Above standards not met � Mark the JPM as a FAIL

CR-SIM 1 2016-301 Page 5 of 6

STEP #



4. Select a control rod in Rod Step 25. (Rod 38-31)

At panel 2H11-P603, the push-button is DEPRESSED on CONTROL ROD SELECT Matrix for selected control rod in Rod Step 25.

5. Withdraw the control rod to Position 08.

At panel 2H11-P603, ROD MOVEMENT CONTROL switch is used to withdraw the rod.

6. Confirm the proper control rod movement.

At panel 2H11-P603, the operator VERIFIES that rod position indicator indicates “08” for rod moved in previous step on Four-Rod Display and/or RWM.

7. Complete the line, for the selected rod, on the Control Rod Movement Sequence sheet.

On the Control Rod Movement Sequence sheet, on the line for the selected rod (Withdrawn side of sheet), the operator has:

Filled in INIT block.

Filled in DATE block.

** 8. Select the next control rod in Rod Step 25. (Rod 30-15)

At panel 2H11-P603, the push-button is DEPRESSED on CONTROL ROD SELECT Matrix for selected control rod in Rod Step 25.

** 9. Withdraw the control rod to Position 08.

At panel 2H11-P603, ROD MOVEMENT CONTROL switch is used to withdraw the rod.

NOTE: 30-15 is the control rod that will drift.

10. Confirm the proper control rod movement.

At panel 2H11-P603, the operator VERIFIES that rod position indicator indicates drifts past “08” and the ROD DRIFT alarm illuminates.

NOTE: The operator may drive the control rod in using the EMERGENCY IN

switch. To allow the operator to address 34AB-C11-004-2, the simulator operator may hold the rod in with the EMERGENCY IN switch.

CR-SIM 1 2016-301 Page 6 of 6

STEP #



** 13. Manually scram the Reactor. At panel 2H11-P603, REACTOR SCRAM PUSHBUTTONS are depressed or the REACTOR MODE SWITCH in placed in shutdown.

NOTE: The task is to scram the Reactor. If the operator scrams the Reactor and

continues with scram actions, the evaluator may stop the JPM by stating that another operator will take care of scram actions.

PROMPT: IF the operator asks whether or not to scram the Reactor, DIRECT the

operator to respond as the procedure directs.

END TIME:__________

NOTE: The terminating cue shall be given to the operator when:

− With no reasonable progress, the operator exceeds double the allotted time.

− Operator states the task is complete.

PROMPT: AFTER the task is completed, ASK the fundamental question in

Attachment 1 (Can be asked any time prior to completing all JPMs).

TERMINATING CUE: We will stop here.

CR-SIM 1 2016-301 Page 7 of 6

STEP #



Attachment 1 (Fundamental Question) Question: During normal operations (steady state with no evolutions in progress), what would tell the OATC that a control rod was drifting out? Answer: The first indication would be a “Rod Drift” alarm. Then the rod that is drifting would have a red light on the Full Core Display. The rod could then be selected on the Four Rod Display. May also pick up upscale alarms.



DRAFT CR-SIM 2 (ALL)

Title:

Perform RC-2, HPCI Injection (Alternate Path)



Time: 15 Minutes

Line Review By (N/A for minor revisions)

N/A

Date:

Reviewed by Instructional Technologist or designee (N/A for minor revisions)

N/A

Date:

Approved By (Training Program Manager or Lead Instructor)

Date:

This page intentionally left blank

of 14

Course Number N/A

Program Name OPERATIONS TRAINING

Media Number CR-SIM 2 2016-301


Sup’s Initials

3.0

Revised LR-JP-00505 to be an Alternate Path JPM

for use on 2016-301 NRC Exam. Reviewed JPM

against current procedure. Changed “Media

Number” to CR-SIM 2 2016-301. After NRC

Exam, Media Number will be changed to new LR-

JP-00505D (Alt Path).

ARB

of 14



CR-SIM 2 2016-301 Page 5 of 14

UNIT 1 () UNIT 2 (X)

TASK TITLE: Perform RC-2, HPCI Injection (Alternate Path)


TASK STANDARD: The task shall be completed when the operator has verified RFPTs

are controlling RWL, then recognizing RFPTs have tripped and manually starts HPCI with a flow controller failure to restore RWL to between +3 and +111 inches.


OBJECTIVE NUMBER: H-OP005.005

PLANT HATCH JTA IMPORTANCE RATING: RO X.XX SRO X.XX

K/A CATALOG NUMBER: 206000A3.05 K/A CATALOG JTA IMPORTANCE RATING :

RO 4.3 SRO 4.3


GENERAL REFERENCES: Unit 1 Unit 2

NA 34AB-C71-001-2, Ver 12.5

34SO-E41-001-2, Ver 30

REQUIRED MATERIALS: Unit 1 Unit 2

NA 34AB-C71-001-2, Ver 12.5

34SO-E41-001-2, Ver 30

APPROXIMATE COMPLETION TIME: 15 Minutes SIMULATOR SETUP: Refer to simulator setup sheet on the following page.

CR-SIM 2 2016-301 Page 6 of 14

SIMULATOR SETUP

Simulator Initial Conditions:

1. RESET the Simulator to 90% power or SNAP 612 and leave in FREEZE. 2. ACTIVATE THE FOLLOWING EVENT TRIGGERS:

Trigger # DESCRIPTION CONDITIONS

EGE41-1 Inserts E41_106 - HPCI flow controller output fails low aoE41-R610.aivToPanel >.65

EGC11-6 Inserts mfC11-299 – CRD Flow Control Fails Low, trips

both RFPTS after 30 seconds diC71A-S1.aivToPanel!=3

3. INSERT the following MALFUNCTIONS:

Activator MALF # TITLE FINAL VALUE

RAMP RATE

EGE41-1 mfE41_106 HPCI flow controller output fails low

EGC11-6 mfC11-299 CRD Flow Control Fails Low

EGC11-6 mfN21_87A A Feedwater Pump trip TRIP

EGC11-6 mfN21_87B B Feedwater Pump trip TRIP

ST 0 mfE51_110 RCIC Turbine Trip TRIP

ST 0 loE51_F045G1 Steam to RCIC turbine OFF

ST 0 loE51_F045R2 Steam to RCIC turbine OFF

ST 0 mfB21_132 ADS Fails To Initiate


A. Isolate RCIC (2E51-F007 and 2E51-F008 switches to close) and acknowledge alarms. B. OPEN 2E51-F054 to bleed pressure from RCIC. C. When RCIC inlet pressure =0 psig, CLOSE 2E51-F054.

4. PLACE the Simulator in FREEZE until the INITIATING CUE is given.

CR-SIM 2 2016-301 Page 7 of 14

5. PLACE DANGER TAGS on the following equipment:

MPL # COMPONENT TAGGED POSITION

2E51-F007 RCIC Steam Isolation Valve CLOSED

2E51-F008 RCIC Steam Isolation Valve CLOSED

6. ESTIMATED Simulator SETUP TIME : 7 Minutes

UNIT 2


1. Unit 2 is currently at approximately 90%.

2. Due to a problem with the Main Turbine, the shift crew is preparing to manually scram the reactor.

3. RCIC is danger tagged out.

4. Another operator has been assigned to perform RC-1.

5. Pre-Job Brief is NOT required.

INITIATING CUES:

After the unit has been scrammed, PERFORM RC-2 and RC-3.

CR-SIM 2 2016-301 Page 9 of 14

STEP #




IF THEN

PASS


correctly



(1) The standard for human performance tools, safety, PPE, and other pertinent expectations is considered met provided any deviations are minor and have little or no actual or potential consequence. Errors may be self-corrected provided the action would not have resulted in significant actual or potential consequences. (AG-TRN-01-0685 Section 6.5.3 provides examples)

Start Time:_________

NOTE: IF HPCI is started prior to completion of all RC-2 actions, the operator may not have time to complete the remaining RC-2 actions.

NOTE: The step sequence of this JPM may not match the RC-2 actions performed

by the operator. This is acceptable as long as the RC-2 steps are performed correctly.

AT THIS TIME , the SIMULATOR OPERATOR PERFORMS Steps 1 through 5 of RC-1, and then

REPORTS to the SS that ALL Control Rods have fully inserted.

NOTE: After the Control Rod status has been reported to the SS, no further RC-1 actions are required to be performed.

The EVALUATOR will act as the Shift Supervisor and acknowledge reports as necessary.

1. Perform scram choreography for RC-2.

The Operator reports to the SS current RWL AND that RFPTs will be used to control RWL.

2. Perform scram choreography for RC-3.

The Operator reports to the SS current reactor pressure AND that EHC is being used to control Reactor Pressure.

CR-SIM 2 2016-301 Page 10 of 14

STEP #



3. CHECK ECCS Injection Systems and secure as necessary. (RC-2)

The Operator has DETERMINED that HPCI is NOT running and is NOT needed for injection.

4. ENSURE FW Master Controller setpoint reduces to 9 inches and output reduces to 25% of previous value. (RC-2)

On 2H11-P603, the Operator EVALUATES the Master Feedwater Level Controller and RECOGNIZES the controller has set-down to 9 inches.

5. IF set down does not auto function, THEN MANUALLY REDUCE FW Master Controller setpoint to approximately 9 inches. (RC-2)

On 2H11-P603, the Operator EVALUATES the Master Controller Setpoint has automatically reduced to 9”.

6. OPEN 2N21-F125. (RC-2)

On 2H11-P650, the Operator CONFIRMS 2N21-F125 is OPEN.

7. PLACE 2C32-R619, FW S/U LVL CONTROL VLV Controller, in AUTO, set at approximately 9 inches. (RC-2)

On 2H11-P603, the Operator VERIFIES 2C32-R619 is set at 9 inches and in AUTO with the AUTO button illuminated.

8. CLOSE 2N21-F110. (RC-2)

On 2H11-P650, the Operator TAKES the switch for 2N21-F110 to close, green light illuminates.

PROMPT: As Shift Supervisor, GIVE the operator a RWL band of +3 and +50 inches.

NOTE: SIMULATOR INSTRUCTOR: EGC11-6 will trip both RFPTs 30 seconds after the Rx. Mode switch is placed to Shutdown.

PROMPT: Following the RFPTs tripping, if necessary, As Shift Supervisor, GIVE

the operator a RWL band of +3 and +50 inches.

9. Determine that RFPT have tripped and RWL is decreasing.

The Operator has diagnosed that both RFPs are tripped.

CR-SIM 2 2016-301 Page 11 of 14

STEP #



ALTERNATE PATH STARTS HERE (Step 10 which is following the loss of both RFPTs).

NOTE: The following JPM steps are written assuming the Operator starts HPCI

using the HPCI Placard.

10. IF required, depress the High Water Level Reset Pushbutton (Placard Step 1)

Operator DETERMINES that a High RWL condition has not occurred (no white light). It is permitted to depress the pushbutton but is not required.

11. OPEN 2E41-F059 (Placard Step 2)

At 2H11-P601, the Operator OPENS LUBE OIL CLG WTR VLV, 2E41-F059, red light illuminated.

12. START Barom Cndsr Vac Pump (Placard Step 3)

At 2H11-P601, the Operator STARTS HPCI BAROM CNDSR VACUUM PUMP, 2E41-C002-2, red light illuminated.

PROMPT: IF the operator addresses posting High Radiation Areas, as the Shift

Supervisor, INFORM the operator that Health Physics is posting the areas.

** 13. Perform the following in rapid succession:

Place 2E41-F001 CS in OPEN,

AND

……………..

AND

……………. (Placard Step 4)

At 2H11-P601, the Operator PLACES CS for the TURB STEAM SUPPLY VLV, 2E41-F001 in the OPEN position.

CR-SIM 2 2016-301 Page 12 of 14

STEP #



14. Perform the following in rapid succession:

…………….,

AND

VERIFY red light illuminates,

AND

……………. (Placard Step 4)

At 2H11-P601, the Operator VERIFIES 2E41-F001 red light is illuminated.

** 15. Perform the following in rapid succession:

…………….,

AND

……………..

AND

START Aux Oil Pump (Placard Step 4)

At 2H11-P601, the Operator STARTS the AUX OIL PUMP, 2E41-C002-3, red light illuminated.

** 16. OPEN 2E41-F006 (Placard Step 5)

At 2H11-P601, the Operator OPENS PUMP DISCHARGE VLV, 2E41-F006, red light illuminated.

17. CONFIRM TCV and TSV OPEN (Placard Step 6)

At 2H11-P601, the Operator CONFIRMS that the:

TURBINE CONTROL VLV

and

TURBINE STOP VLV

are OPEN, red light illuminated.

18. Confirm the HPCI Turbine comes up to speed as directed by the Turbine Flow Controller, 2E41-R612.

At panel 2H11-P601, the operator RECOGNIZES HPCI turbine has NOT increased to rated speed and that the HPCI FLOW CONTROL 2E41-R612 has failed LOW.

NOTE: The following step is NOT sequence sensitive and may be performed at any point during the performance of this JPM.

CR-SIM 2 2016-301 Page 13 of 14

STEP #



** 19. Operator takes manual control of the HPCI Flow Controller.

At panel 2H11-P601, HPCI FLOW CONTROL 2E41-R612 Manual (M) pushbutton has been DEPRESSED, Manual (M) amber light illuminated.

** 20. Operator increases HPCI Turbine speed.

At panel 2H11-P601, HPCI FLOW CONTROL 2E41-R612 level is TAKEN to the Open (O) direction until: PUMP DISCH PRESS 2E41-R601 is greater than Reactor pressure and the desired flow is obtained.

21. CONFIRMS 2E41-F012 CLOSES at flow > 790 gpm (Placard Step 7)

At 2H11-P601, the Operator CONFIRMS MIN FLOW VLV, 2E41-F012 is CLOSED when HPCI pump flow is > 790 gpm, green light illuminated.

22. ADJUST controller for desired flow (Placard Step 8)

At 2H11-P601, the Operator ADJUSTS HPCI FLOW CONTROL, 2E41-R612 to control rate of RWL level increase to maintain >+3 and <+111 inches.

22. Confirm 2E41-F001 is full OPEN (Placard Step 9)

At 2H11-P601, the Operator CONFIRMS 2E41-F001 is fully OPEN.

End Time: _________


− Operator completes step 22 of this JPM.

− With NO reasonable progress, the operator exceeds double the allotted time.

− Operator states the task is complete. TERMINATING CUE: We will stop here.

CR-SIM 2 2016-301 Page 14 of 14

STEP #



Summary of JPM Attributes

JPM CR-SIM 2 2016-301: SUMMARY OF JPM QUANTITATIVE ATTRIBUTES CATEGORY Minimum NRC

Attributes JPM CONTENT

Total Critical Steps

At least 2

5

Step 13 Open 2E41-F001 Opening the steam supply valve is required to provide motive force.

Step 15 Start Aux Oil Pump Required to provide oil pressure to open the TCV and TSV.

Step 16 Open 2E41-F006 Opening the discharge valve is required to allow for injection.

Step 19 Controller in Manual Allows manual control of output

Step 20 Increase output Required to obtain injection

Number of JPM Steps

<30

22

Time to Perform JPM

<45 min

15 min

Normal / Faulted / Alternate Path

Alternate Path RWL is being controlled with RFPTs when a both RFPTs trip, requiring the student to manually start HPCI and inject to the RPV. HPCI will NOT auto start on low RWL or high DW pressure.

Setting (administered)

Simulator

Is LOD “1” or “5” NO NO




Title:

OVERRIDE THE MSIVS IN AN EMERGENCY



Time: 18.0 Minutes

Line Technical Review By (N/A for minor revisions)

N/A

Date:

N/A

Reviewed by Instructional Technologist or designee:

N/A

Date:

N/A

Approved By (Training Program Supervisor, Lead Instructor or Line Supervisor)

Date:

of 14


of 14

3

Course Number N/A




Sup’s Initials

01 08/24/92 General revision and format change WMM SCB 02 08/16/93 General revision, word processor change RAB RSG 03 08/19/94 Modify simulator setup, adjust format RAB SMC 04 03/28/96 Format change RAB DHG 05 04/10/97 Revised based on feedback from the 1996 Requal

annual exam. SCB RSG

06 02/07/00 Format modification, change time allowance based on running average, upgrade to the new simulator operating system

RAB DHG

07 11/02/00 Include objective number RAB DHG 08 03/11/02 Include initial operator statement RAB RAB 09 03/17/05 Deleted “S” from procedure numbers, changed

Revision and Rev. numbers to “Current Version”. Changed to Nuclear Plant Operator. Updated Simulator IC# and MSIV position.

ARB DHG


RAB RAB

11 04/18/06 Remove Response Cues RAB RAB 12 11/06/06 Correct MPLs and Switch labels on Unit 1 BKW RAB 13 06/07/12 Revision for use on 2012-301 CR-SIM 3 and updated

procedure changes. After NRC Exam, Media Number will be changed to LR-JP-14.14.

ARB CME

14

Reviewed JPM against current procedure for use on 2016-301 NRC Exam. Changed “Media Number” to CR-SIM 3 2016-301. After NRC Exam, Media Number will be changed to the next revision of LR-JP-14.14 and U1 material will be added back.

ARB

CR-SIM 3 2016-301 Page 4 of 14

UNIT 1 (X) UNIT 2 (X)

TASK TITLE: OVERRIDE THE MSIVS IN AN EMERGENCY


TASK STANDARD: The task shall be completed when the MSIVs have been opened per 31EO-EOP-111-1/2.





RO 4.20 SRO 4.00



31EO-EOP-111-1 31EO-EOP-011-1 (current versions)

31EO-EOP-111-2 31EO-EOP-011-2 (current versions)


31EO-EOP-111-1 (current version)

31EO-EOP-111-2 (current version)

APPROXIMATE COMPLETION TIME: 18.0 Minutes SIMULATOR SETUP: REFER TO SIMULATOR SETUP SHEET ON THE FOLLOWING

PAGE

CR-SIM 3 2016-301 Page 5 of 14

SIMULATOR SETUP

Simulator Initial Conditions: 1. RESET the Simulator to IC #113 or Snap 613 and leave in FREEZE.

2. ACTIVATE THE FOLLOWING EVENT TRIGGERS:


EGB21-17 Delete mf60321350 Group I System A Trip diA71B-S32.aivToPanel<1

EGB21-18 Delete mf60321351 Group I System B Trip diA71B-S33.aivToPanel<1 3. INSERT the following MALFUNCTIONS:


RAMP RATE

ACT. TIME

mfC11_211 Scram Discharge Volume ATWS (Var) 20 100 00000

mfE41_235A HPCI Fails to Auto Start on Low Level 00000

mfE41_235B HPCI Fails to Auto Start on Hi Drywell Press 00000

mf60321383 Spur Ann -- LOW CONDENSER VACUUM A BYPASS

00000

mf60321384 Spur Ann -- LOW CONDENSER VACUUM B BYPASS

00000

mf60321350 Group I System A Trip 00000

mf60321351 Group I System B Trip 00000

4. INSERT the following SIMULATOR VALUE OVERRIDES (SVO):

SVO # DESCRIPTION FINAL VALUE

RAMP RATE

ACT. TIME

svoB21005 LT-N081A Group I/II Isolation -110 100 00000

svoB21006 LT-N081B Group I/II Isolation -110 100 00000

5. INSERT the following REMOTE FUNCTIONS:

REM # DESCRIPTION STATUS

rfN11045 SJAE A Steam CLOSE

rfB21148 Grp 1 Low Rx Water Level Bypass ORIDE

rfC71279 Group 1 Isolation Oride Jumpers ORIDE

CR-SIM 3 2016-301 Page 6 of !Syntax Error, !

6


A. Take simulator out of FREEZE and perform RC-1, RC-2 and TC-1. B. Place 2N11-F004A & F004B control switches in CLOSED & ENSURE they are

CLOSED. C. Place 2N33-F003 control switch in CLOSED & ENSURE it is CLOSED D. Place the MSIV control switches in CLOSED. E. Enter Remote Functions WHEN Torus temperature reaches 109°F or greater. F. Inhibit ADS and inject SBLC. G. Acknowledge annunciators.

7. PLACE the Simulator in FREEZE until the INITIATING CUE is given. 8. ESTIMATED Simulator SETUP TIME : 15 Minutes

EVALUATOR COPY

UNIT 2


1. The Condenser Low Vacuum and Low RWL isolations have been bypassed.

2. Main Condenser is in operation with Circ Water and Condensate Systems in operation.

3. An ATWS condition exists and 31EO-EOP-011-2 (RCA) is in progress.

4. EHC is in operation. INITIATING CUES:

Open the MSIVs to re-establish Main Condenser as the heat sink using 31EO-EOP-111-2.

CR-SIM 3 2016-301 Page 8 of 14

STEP #




IF THEN

PASS


correctly




START TIME:_________

1. Place MSIV Control Switches to CLOSED.

At panel 2H11-P601, confirms the following OUTBOARD MSIV switches are in CLOSED:

MSIV 2B21-F028A MSIV 2B21-F028B MSIV 2B21-F028C MSIV 2B21-F028D

2. Place MSIV Control Switches to CLOSED.

At panel 2H11-P602, confirms the following INBOARD MSIV switches are in CLOSED:

MSIV 2B21-F022A MSIV 2B21-F022B MSIV 2B21-F022C MSIV 2B21-F022D

3. Confirm A & B RFPTs are tripped. At panel 2H11-P650, Operator has determined BOTH RFPTs are TRIPPED by:

Checking annunciators(650-325 & 326) are ILLUMINATED

CR-SIM 3 2016-301 Page 9 of 14

STEP #



PROMPT: IF the operator addresses isolation bypasses, as the Shift Supervisor, INFORM the operator that the Condenser Low Vacuum and Low Water Level isolations are bypassed.

** 4. Reset Group I Isolation. The GR ISOL RESET switch has been taken to GR I RESET position at panels 2H11-P601 and 2H11-P602.

5. Bypass the High Flow isolation on 2P70-F004 and F005, if necessary.

Operator has determined it is NOT necessary to bypass the High Flow Isolation by:

Checking annunciators

OR

Checking INBD INLET ISOL 2P70-F004 and F005 red lights illuminated on panel 2H11-P700.

NOTE: It is acceptable for operator to have the High Flow Isolation bypassed. IF operator decides to bypass the High Flow Isolation, as the Shift Support Supervisor, INFORM the operator that it has been bypassed.

6. Confirm closed Inboard MSIVs. At panel 2H11-P602, the following INBOARD MSIVs are CLOSED, green light illuminated:

MSIV 2B21-F022A MSIV 2B21-F022B MSIV 2B21-F022C MSIV 2B21-F022D.

** 7. Open Outboard MSIVs. At panel 2H11-P601, the following switches are in OPEN SLO TEST, red light illuminated:


CR-SIM 3 2016-301 Page 10 of 14

STEP #



8. Check differential pressure across Inboard MSIVs.

Operator has determined differential pressure is greater than 200 psid using:

REACTOR PRESSURE indicator A(B,C) 2C32-R605A(B,C) on panel 2H11-P603

AND

MAIN STEAM PRESS "A"("B") 2N32-R654A(B) indicator on Turbine EHC Panel or recorder PRESS TO STOP VLVS 1 & 4 2N11-R601

OR

*Monitor* → *steam pressure* 2N32-K4001A OR 2N32-K4001B.

NOTE: Any valid indication of Reactor Pressure and Steam Line Pressure can be used for Step 7.

9. Confirm/Close valves 2N11-F001A & B.

The operator has ADDRESSED contacting a SO to verify/close SJAE 2A(2B) MAIN STEAM ISOL VALVE at panel 2H21-P216.

PROMPT: WHEN the operator addresses valves 2N11-F001A and B, as the SO, INFORM the operator that the valves 2N11-F001A & B are closed.

10. Confirm the following valves are closed:

2N11-F004A

2N11-F004B

2N33-F003

At panel 2H11-P650, the following valves are CLOSED, green light illuminated:

2ND STG A & B MSR RHTR STM SPLY VLV 2N11-F004A

2ND STG C & D MSR RHTR STM SPLY VLV 2N11-F004B

MAIN STM FEED VLV 2N33-F003.

CR-SIM 3 2016-301 Page 11 of 14

STEP #



** 11. Open the following valves:

2B21-F016

2B21-F019

The following valves are OPEN, red lights illuminated:

MSL DRAIN VLV, 2B21-F016, at panel 2H11-P602

MSL DRAIN VLV, 2B21-F019, at panel 2H11-P601

12. Open the following valve:

2B21-F020

The following valve is OPEN, red lights illuminated:

DRAIN VLV, 2B21-F020, at panel 2H11-P602

13. Verify differential pressure across MSIVs is less than 200 psid.

Operator has determined differential pressure is less than 200 psid using:

REACTOR PRESSURE indicator A(B,C) 2C32-R605A(B,C) on panel 2H11-P603

AND

MAIN STEAM PRESS "A"("B") 2N32-R654A(B) indicator on Turbine EHC Panel or recorder PRESS TO STOP VLVS 1 & 4 2N11-R601

OR

*Monitor* → *steam pressure* 2N32-K4001A OR 2N32-K4001B.

NOTE: Any valid indication of Reactor Pressure and Steam Line Pressure can be used for Step 11.

** 14. Open Inboard MSIVs. At panel 2H11-P602, the following switches are in OPEN SLO TEST, red light illuminated:


CR-SIM 3 2016-301 Page 12 of 14

STEP #



NOTE: It is acceptable if the operator uses the Open Fast Test switch position.

PROMPT: WHEN the operator addresses the Main Condenser Vacuum System, as the Shift Supervisor, INFORM the operator that another operator will place it in service and monitor vacuum.

PROMPT: IF the operator addresses system restoration, as the Shift Supervisor, INFORM the operator that restoration is not required at this time.

END TIME: __________

NOTE: The terminating cue shall be given to the operator when any one of the following is met:

− After JPM step #13 is complete.


− Operator states the task is complete. TERMINATING CUE: We will stop here. EVALUATOR – PICK UP the Initiating Cue sheet.

of 14





At least 2

4

Step 4 Reset Group I Isolation The Group I Isolation is required to be reset to re-open the MSIVs

Step 7 Open the Outboard MSIVs Opening the Outboard MSIVs is required to re-establish the Main Condenser as the heat sink, and to reduce differential pressure across the Inboard MSIVs less than 200 psid.

Step 11 Open 2B21-F016 & F019 The MSL Drain valves are opened to reduce differential pressure across the Inboard MSIVs less than 200 psid.

Step 14 Open the Inboard MSIVs Opening the Inboard MSIVs is required to re-establish the Main Condenser as the heat sink.

Number of JPM Steps

<30

14

Time to Perform JPM

<45 min

18.0 min


Normal Open the MSIVs using 31EO-EOP-11-2.


Simulator


of 14

UNIT 2


1. The Condenser Low Vacuum and Low RWL isolations have been bypassed.

2. Main Condenser is in operation with Circ Water and Condensate Systems in operation.

3. An ATWS condition exists and 31EO-EOP-011-2 (RCA) is in progress.

4. EHC is in operation. INITIATING CUES:

Open the MSIVs to re-establish Main Condenser as the heat sink using 31EO-EOP-111-2.




Title:

REACTOR PRESSURE CONTROL WITHIN BAND WITH BPVs & SR Vs

Author: Error! Reference source not found.


Time: 20.0 Minutes


N/A

Date:

N/A


N/A

Date:

N/A


Date:

of 13

3

Course Number N/A




Sup’s Initials

00 5/26/11 Initial development for use on 2011-301 NRC Exam.

ARB CME

0.1

Reviewed JPM against current procedure for use on 2016-301 NRC Exam. Changed “Media Number” to CR-SIM 4 2016-301. After NRC Exam, Media Number will be changed to a new LR-JPM.

ARB

CR-SIM 4 2016-301 Page 4 of 13


TASK TITLE: REACTOR PRESSURE CONTROL WITHIN BAND

WITH BPVs & SRVs


TASK STANDARD: The task will be completed when the operator is controlling Reactor pressure in a band using SRVs per 34SO-B21-001-2.


OBJECTIVE NUMBER: 200.034.C

PLANT HATCH JTA IMPORTANCE RATING: RO SRO


RO 3.8 SRO 3.9



34SO-N30-001-2 (current version) 34SO-B21-001-2 (current version)


34SO-N30-001-2 (current version) 34SO-B21-001-2 (current version)


PAGE

CR-SIM 4 2016-301 Page 5 of 13

5

SIMULATOR SETUP Simulator Initial Conditions: 1. RESET the Simulator to IC #113 or Snap 614 and leave in FREEZE. 2. INSERT the following MALFUNCTIONS and EVENT TRIGGERS:

KEY MALF # DESCRIPTION FINAL RAMP DELAY

ST-0 mfB21_129A SRV A Fails Stuck 0000

ST-0 mfB21_129B SRV B Fails Stuck 0000

ST-0 mfB21_129C SRV C Fails Stuck 0000

ST-0 mfB21_129D SRV D Fails Stuck 0000

ST-0 mfB21_129E SRV E Fails Stuck 0000

ST-0 mfB21_129F SRV F Fails Stuck 0000

ST-0 mfB21_129G SRV G Fails Stuck 0000

ST-0 mfB21_129H SRV H Fails Stuck 0000

ST-0 mfB21_129K SRV K Fails Stuck 0000

ST-0 mfB21_129L SRV L Fails Stuck 0000

ST-0 mfB21_129M SRV M Fails Stuck 0000

ST-0 mfE41_104 HPCI Turbine Trip 0000

ST-0 mfE51_110 RCIC Turbine Trip 0000

ST-0 mfN21_87A Feedwater Pump A Trip 0000

ST-0 svoB21053 PT-N127A SRV Electrical open 1000 1000 0000

ST-0 svoB21054 PT-N127B SRV Electrical open 1000 1000 0000

ST-0 svoB21055 PT-N127C SRV Electrical open 1000 1000 0000

ST-0 svoB21056 PT-N127D SRV Electrical open 1000 1000 0000

ST-0 mfB21_226A Low-Low Set A Fails Inop 0000

ST-0 mfB21_226B Low-Low Set B Fails Inop 0000

ST-0 mf60131129 Leak Det Diff Temp Hi 0000

ST-0 mf60131135 Leak Ambient Temp Hi 0000

CR-SIM 4 2016-301 Page 6 of 13

6

RB-1 mfN37_134 ALL Bypass Valves Fail CLOSED 9999

EGN37-1 diN32-C001A EHC Pump 2A Switch Trip 9999

EGN37-1 diN32-C001B EHC Pump 2B Switch Trip 9999

ACTIVATE THE FOLLOWING EVENT TRIGGERS:


EGB21-21 SRV B Stuck, opens when control switch is in OPEN diB21-F013B.iivPanel=1

EGB21-22 SRV G Stuck, opens when control switch is in OPEN diB21-F013G.iivPanel=1

EGB21-23 SRV F Stuck, opens when control switch is in OPEN diB21-F013F.iivPanel=1

EGB21-24 SRV D Stuck, opens when control switch is in OPEN diB21-F013D.iivPanel=1

EGN37-1 Trips Both EHC pumps after 3 minutes an_p650t(24) == 1

3. Take the Simulator OUT OF FREEZE and PERFORM the following MANIPULATIONS OR

RUN SCENARIO FILE and EVENT TRIGGER (current rev) CR-SIM 4 2016-301:

A. Perform RC-1 and RC-2 actions such that the 2N21-F110 is closed and the SULCV is in auto and set at +20”.

B. Perform TC-1; ensuring 2N11-F004A and B are closed. C. Lower Pressure Set to 825 psig. D. Ensure Reactor pressure stabilizes at ~825 psig. E. Ensure BYPASS VALVE screen and SPEED screen is called up. F. Place RPV Pressure (B025) in the P603 plasma display.

4. PLACE the Simulator in FREEZE until the INITIATING CUE is given. 5. ESTIMATED Simulator SETUP TIME : 15 minutes

UNIT 2


1. Unit 2 Reactor has been scrammed.

2. Other operators are performing scram actions.

3. A Reactor pressure reduction is required due to an unisolable steam leak in the Reactor Building.

4. You have the sole responsibility for Reactor Pressure Control. INITIATING CUES:

LOWER AND THEN MAINTAIN Reactor pressure between 700 and 800 psig using the Bypass Valves.

CR-SIM 4 2016-301 Page 8 of 13


IF THEN

PASS


correctly



(1) The standard for human performance tools, safety, PPE, and other pertinent expectations is considered met provided any deviations are minor and have little or no actual or potential consequence. Errors may be self-corrected provided the action would not have resulted in significant actual or potential consequences.



Operator has obtained procedure 34GO-OPS-013-2, step 7.5.5 or uses Placard at Turbine Panel.

NOTE: For Steps 2 through 8 all indication will be on panel 2H11-P650 using HMI Screens 2N32-K4001A or B.

2. Selects *Control* / *psi-load* screen

The operator has SELECTED the * Control* screen and also has SELECTED the *psi-load* button on 2N32-K4001A or B.

3. Selects *Ramp Rate* button The operator has SELECTED the *Ramp Rate* button on 2N32-K4001A or B.

CR-SIM 4 2016-301 Page 9 of 13

NOTE: An acceptable Ramp Rate results in a cooldown rate NOT exceeding 100ºF/hr OR RWL exceeding +100 inches.

** 4. Enters a ramp rate The operator has ENTERED a ramp rate >0 and <100 displayed on 2N32-K4001A or B.

** 5. Selects *Pressure* button The operator has SELECTED the *Pressure* button on 2N32-K4001A or B.

** 6. Enters desired target pressure The operator has ENTERED a target pressure between 700 and 800 psig on 2N32-K4001A or B.

** 7. Adjusts *Ramp Rate* AND/OR the *Pressure* setpoints to maintain pressure/cooldown rate.

The operator has VARIED the

ramp rate/pressure to maintain a cooldown rate ≤100° F/Hr OR RWL exceeding 100 inches change on 2N32-K4001A or B.

ALTERNATE PATH STARTS HERE

NOTE: SIMULATOR OPERATOR , WHEN the operator has successfully opened Bypass Valves (BPVs) and reactor pressure is being reduced AND with the Chief Examiners approval, ENTER RB-1 malfunction which FULLY CLOSES ALL Bypass Valves.

8. The failure of ALL BPVs is recognized.

Operator identifies all BPVs closed on 2N32-K4001A or B.

PROMPT: IF the operator reports the failure of the BPVs and EHC pumps, INFORM the operator that you will get Maintenance to investigate, and to control Reactor pressure between 700 – 800 psig.

NOTE: SIMULATOR OPERATOR, ENSURE after 3 minutes, EVENT TRIGGER EGN37-1 TRIPS BOTH EHC Pumps.

CR-SIM 4 2016-301 Page 10 of 13

PROMPT: IF alarm 650-124, Max Combined Flow Limit Limiting, is received, INFORM the operator that another operator will address the ARP.

PROMPT: IF the operator addresses using other systems to control Reactor pressure, INFORM the operator to control Reactor pressure between 700 – 800 psig.

NOTE: The operator can use any of the following procedures to control reactor pressure: • Placard on 2H11-P602 panel • 34SO-B21-001-2, Automatic Depressurization (ADS) And Low-Low

Set (LLS) Systems

9. Operator identifies applicable procedures to perform the task.

Operator has IDENTIFIED the correct procedure as: Placard on 2H11-P602 panel 34SO-B21-001-2, step 7.4.1.

NOTE: ALL SRVs, EXCEPT LLS SRVs, fail to open with the switch.

10. Attempts to OPEN SRV 2B21-F013M.

At Panel 2H11-P602, the operator has PLACED the control switch for 2B21-F013M into the OPEN position, red light ILLUMINATES.

11. The failure of the 2B21-F013M to open is recognized.

At 2H11-P602, the operator has RECOGNIZED the failure of the 2B31-F013M to OPEN, amber light remains EXTINGUISHED.

PROMPT: IF the operator reports the failure of the 2B21-F013M to open, INFORM the operator that you will get Maintenance to investigate, and to control Reactor pressure between 700 – 800 psig.

NOTE: Due to failures, LLS will NOT automatically ARM with Reactor pressure >1074 psig.

CR-SIM 4 2016-301 Page 11 of 13

NOTE: EVENT TRIGGER EGB21-21 will allow 2B21-F013B, LLS valve to be used to control RPV pressure.

NOTE: EVENT TRIGGER EGB21-22 will allow 2B21-F013G, LLS valve to be used to control RPV pressure.

NOTE: EVENT TRIGGER EGB21-23 will allow 2B21-F013F, LLS valve to be used to control RPV pressure.

NOTE: EVENT TRIGGER EGB21-24 will allow 2B21-F013D, LLS valve to be used to control RPV pressure.

NOTE: Any of the LLS valves (2B21-F013B, D, F, or G) will be used to control RPV pressure any will successfully met the intent of Critical Step ** 12. 34SO-B21-001-2 allows opening of any of the LLS valves to control RPV pressure.

** 12. OPEN SRV 2B21-F013B, D, F, or G. At Panel 2H11-P602, the operator has PLACED the control switch for 2B21-F013B, D, F, or G into the OPEN position prior to 1250 psig, red light ILLUMINATED.

PROMPT: IF the operator addresses checking SRV tailpipe temperature on recorder 2B21-R614, at Panel 2H11-P614, INFORM the operator that another operator will check the recorder for proper SRV operation.

** 13. Reactor pressure in band: 700 to 800 psig

The operator CONTROLS pressure in band: 700 to 800 psig (± 50 psig)

CR-SIM 4 2016-301 Page 12 of 13

** 14. Before Reactor decreases below 700 psig, CLOSE SRV 2B21-F013B, D, F, or G.

At panel 2H11-P602 the operator has PLACED the control switch for 2B21-F013B, D, F, or G into the CLOSE position PRIOR to Reactor pressure lowering below its band (-50 psig), green light ILLUMINATED.

NOTE: AFTER the operator has demonstrated proper control of Reactor pressure, INFORM the operator that another operator will continue maintaining Reactor pressure in band.

END TIME: __________

NOTE: The terminating cue shall be given to the operator when any one of the following is met:



− Operator states the task is complete. TERMINATING CUE: Another operator will continue from here. EVALUATOR – PICK UP the Initiating Cue sheet.

CR-SIM 4 2016-301 Page 13 of 13

STEP #







At least 2

7

Step 4 Enters ramp rate Entering a ramp rate is required to lower the pressure reference signal.

Step 5 Selects Pressure Selecting the Pressure button allows the operator to reduce the desired pressure value.

Step 6 Enters target Pressure Entering the target pressure allows the operator to start the pressure reduction to this desired pressure value.

Step 7 Adjusts ramp/rate Adjusts as needed to maintain a cooldown rate ≤100° F/Hr OR RWL exceeding 100 inches.

Step 12 Opens SRV 2B Terminates the pressure increase.

Step 13 Reactor pressure in band Controls pressure in band

Step 14 Closes SRV 2B Prior to exceeding pressure band on lower band

Time to Perform JPM

<45 min

20.0 min


Alternate Path When the Bypass valves fail, the Operator is required to control pressure via an alternate method and use SRVs.


Simulator





Title:

Perform a Manual Initiation of Drywell Sprays (Alte rnate Path)



Time: 10.0 Minutes


N/A

Date:

N/A


N/A

Date:

N/A


Date:

of 13

Course Number N/A



Ver. No. Date Reason for Revisions Author’s Initials

Sup’s Initials

00.0 30Nov15 Initial Development for Simulator only. MCK ALS

0.1

Reviewed JPM against current procedure for use

on 2016-301 NRC Exam. Changed “Media

Number” to CR-SIM 5 2016-301. After NRC

Exam, Media Number will be changed to LR-JP-

00701A (Alt Path).

ARB

.

of 13


Ver. No. List of Contributors

CR-SIM 5 2016-301 Page 5 of 13


TASK TITLE: Perform a Manual Initiation of Drywell Sprays (Alte rnate

Path)


TASK STANDARD: The task shall be completed when the RHR System has been initiated in the Drywell spray mode, per 34SO-E11-010-2 with total RHR flow > 5000gpm but <17,000gpm. Operator is to take action based on Overload protection alarm for the 2A RHR pump by securing the pump.


OBJECTIVE NUMBER: 007.001.O

PLANT HATCH JTA IMPORTANCE RATING:

RO 4.71 SRO 4.05


RO 3.50 SRO 3.50



N/A 34SO-E11-010-2 Ver. 42.0

31EO-EOP-012-2 Ver. 6


N/A 34SO-E11-010-2 Ver. 42.0

APPROXIMATE COMPLETION TIME: 10.0 Minutes SIMULATOR SETUP: Refer to simulator setup sheet on the following page.

CR-SIM 5 2016-301 Page 6 of 13

SIMULATOR SETUP Simulator Initial Conditions: 1. RESET the Simulator to IC # 113 (100%) or SNAP 615 and leave in FREEZE. 2. ACTIVATE THE FOLLOWING EVENT TRIGGERS:


EGE11-31 Inserts 2A RHR Pump Overload Alarm loE11-F021AR2.algToPanel = 1

3. INSERT the following MALFUNCTIONS :

Activator MALF # TITLE FINAL VALUE

RAMP RATE

ST 0 mfE41_107 HPCI Failure to Start (F001 Stuck)

ST 0 mfE51_110 RCIC Turbine Trip

ST 0 mfB21_229B FW Line B Break Inside Containment (Var) 10 100

ST 0 mfR22_183 4KV Buss 2F Fault 1

ST 0 mfR43_167B Diesel Gen Failure to Start 1B 1

4. INSERT the following REMOTE FUNCTIONS:

Activator REM # Description Status

ST 0 rfE11_167 2E11-F017A & B Override 5 Min Timer ORIDE

5. INSERT the following OVERRIDES:

Activator TAG # S/M/L DESCRIPTION Final Value

Ramp Rate

Delay

ST-0 diE11-F028B S Torus Spray OR Test Vlv CLOSE 0

ST-0 loE11-F028BG1 L Torus Spray OR Test Vlv green

light OFF 0

ST-0 loE11-F028BR2 L Torus Spray OR Test Vlv red light OFF 0

ST-0 diE11-F027A S Torus Spray Vlv CLOSE 0

ST-0 loE11-F027AG1 L Torus Spray Vlv green light OFF 0

ST-0 loE11-F027AR2 L Torus Spray Vlv red light OFF 0

CR-SIM 5 2016-301 Page 7 of 13

6. Take the Simulator OUT OF FREEZE and PERFORM the following MANIPULATIONS :

A. Perform first five (5) Steps of RC-1. B. Close 2N21-F006B. Allow the valve to go FULL CLOSE. C. Perform actions of RC-2 to stabilize RWL around 9 inches. D. Place RHR pump 2B in OFF. E. Trip Recirculation Pumps 2A and 2B place in PTL. F. Secure all Drywell cooling fans. G. Allow Simulator to run approximately 10 min to stabilize conditions. H. Acknowledge all Annunciators.

7. PLACE DANGER TAGS on the following equipment:

MPL # COMPONENT TAGGED POSITION

2E11-F028B Torus Spray OR Test Vlv CLOSED

8. PLACE Protected Equipment tags on the following: NONE


UNIT 2


1. Torus pressure is greater than 11 psig.

31EO-EOP-012-2, PC Primary Containment Control is in progress.

2. 4160 VAC bus 2F is de-energized and cannot be restored.

3. 2E11-F028B, Torus Spray OR Test Vlv, is Danger Tagged out of service.

4. 2E11-F027A, Torus Spray Valve, is bound closed and cannot be opened.

5. Both Reactor Recirc pumps are tripped.

6. All Drywell Cooling fans are secured.

7. Drywell pressure and temperature are within the SAFE region of the Drywell Spray Initiation Limit Curve. (GRAPH 8).

8. Pre-Job Brief is NOT required. INITIATING CUES:

Initiate Drywell sprays using the RHR Loop 2A per 34SO-E11-010-2.

CR-SIM 5 2016-301 Page 9 of 13

STEP #




IF THEN

PASS


correctly





NOTE: This JPM is set up for the Operator to use the CONTAINMENT SPRAY INITIATION PLACARD.

1. IF RWL <2/3 core height, (-193 inches), PLACE the Cnmt Spray Vlv Cntl 2/3 Core Ht Permis keylock in MANUAL OVERRD.

Step 1

Operator VERIFIES that RWL is >-193” and this step is not applicable.

** 2. IF required by EOPs AND LOCA signal present,

PLACE Cnmt Spray Vlv Cntl switch in the MANUAL position.

Step 2

At panel 2H11-P601, the Operator PLACES the CONTAINMENT SPRAY VALVE CONTROL switch to MANUAL, white light illuminated for 2A Loop RHR.

3. Confirm CLOSED 2E11-F015A and/or 2E11-F017A, unless required for core cooling.

Step 3

At panel 2H11-P601, CONFIRM Closed 2E11-F015A and/or 2E11-F017A.

CR-SIM 5 2016-301 Page 10 of 13

STEP #



4. IF power is being provided by EDG, CHECK EDG loading prior to start of RHR pump(s).

Step 4

Operator VERIFIES that EDGs are not providing power to emergency buses.

5. START RHR pump(s) in loop A.

Step 5

At panel 2H11-P601, Operator VERIFIES the 2A RHR pump is RUNNING, red light illuminated.

6. IF TORUS Spray is desired, PERFORM the following:

Step 6

Operator DETERMINES that Torus Sprays are in service on 2B Loop of RHR remainder of step 6 is NOT required.

** 7. IF DRYWELL Spray is desired, PERFORM the following:

OPEN Containment Spray Valve 2E11-F021A.

Step 7.1

At panel 2H11-P601, the Operator OPENS 2E11-F021A CNMT SPRAY INBD VLV.

ALTERNATE PATH STARTS HERE

(When 2E11-F021A goes intermediate (>50% Open), 2A RHR pump indicates an overload condition but the pump fails to trip).

8. Operator identifies annunciator RHR Pump A OVLD/LOCKOUT Relay Trip is alarming.

Operator IDENTIFIES annunciator, RHR Pump A OVLD/LOCKOUT Relay Trip (601-212), is alarming.

NOTE: The Operator may refer to the Annunciator Response Procedure (ARP)

34AR-601-212-2 “RHR PUMP A OVLD/LOCKOUT RELAY TRIP” only step 5.1 is required action.

** 9. IF 2E11-C002A, RHR Pump A has NOT tripped, TRIP 2E11-C002A, panel 2H11-P601,

ARP – Step 5.1

Operator TRIPS 2A RHR pump (2E11-C002A) at Panel 2H11-P601.

PROMPT: IF Operator asks, INFORM the Operator that Maintenance will investigate the problem with 2A RHR Pump.

PROMPT: IF Operator asks, INFORM the Operator to Spray the Drywell.

CR-SIM 5 2016-301 Page 11 of 13

STEP #



PROMPT: IF Operator asks, INFORM the Operator conditions are in the Safe Region of the Drywell Spray Initiation Limit curve.

PROMPT: IF Operator addresses the CAUTION for Drywell Spray and Torus Sprays

being in separate Loops, INFORM the Operator to Spray the Drywell.

10. IF RWL <2/3 core height, (-193 inches), PLACE the Cnmt Spray Vlv Cntl 2/3 Core Ht Permis keylock in MANUAL OVERRD.

Step 1

Operator VERIFIES that RWL is >-193” and this step is not applicable.

** 11. IF required by EOPs AND LOCA signal present,

PLACE Cnmt Spray Vlv Cntl switch in the MANUAL position.

Step 2

At panel 2H11-P601, the Operator PLACES the CONTAINMENT SPRAY VALVE CONTROL switch to MANUAL, white light illuminated for 2B Loop RHR.

12. Confirm CLOSED 2E11-F015B and/or 2E11-F017B, unless required for core cooling.

Step 3

At panel 2H11-P601, the Operator CONFIRMS Closed 2E11-F015B and/or 2E11-F017B.

13. IF power is being provided by EDG,

CHECK EDG loading prior to start of RHR pump(s).

Step 4

Operator VERIFIES that EDGs are not providing power to emergency buses.

** 14. START RHR pump(s) in Loop B.

Step 5

At panel 2H11-P601, the Operator STARTS RHR pump 2B, red light illuminated.

15. IF TORUS Spray is desired, PERFORM the following:

Step 6

Operator DETERMINES that Torus Sprays cannot be placed in service on 2B Loop of RHR so remainder of Step 6 is not required.

** 16. IF DRYWELL Spray is desired, PERFORM the following:

OPEN 2E11-F021B

Step 7.1

At panel 2H11-P601, the Operator OPENS 2E11-F021B CNMT SPRAY INBD VLV.

NOTE: To achieve the proper spray pattern from the nozzles, a Minimum Drywell

Spray flow of at least 5,000 gpm is required.

CR-SIM 5 2016-301 Page 12 of 13

STEP #



** 17. THROTTLE OPEN 2E11-F016B

Step 7.2

The Operator THROTTLES OPEN 2E11-F016B, CNMT SPRAY OUTBD VLV.

** 18. Operator establishes RHR Total flow to provide adequate drywell spray flow.

The Operator ESTABLISHES Total RHR Flow of >5000 gpm.

19. Operator verifies proper response. The Operator OBSERVES drop in Drywell pressure.

20. Refer to 34SO-E11-010-2, AND place RHR Hx into service.

Step 8

Operator REFERS to 34SO-E11-010-2

PROMPT: IF Operator proceeds to place RHR HX into service, INFORM the

Operator another Operator will perform those actions.

END TIME: __________

NOTE: The terminating cue shall be given to the operator when any of the following conditions are met:




of 13





At least 2

6

Step 9 Operator TRIPS 2A RHR pump Action required protecting equipment from potential fire.

Step 11 PLACES switch to MANUAL Action required for opening drywell spray valves

Step 14 STARTS RHR pump 2B Required to initiate drywell sprays.

Step 16 OPENS 2E11-F021B Required to initiate drywell sprays.

Step 17 THROTTLES OPEN 2E11-F016B Required to establish drywell sprays.

Step 18 ESTABLISHES Total RHR Flow Minimum flow is >5000gpm for drywell sprays.

Number of JPM Steps

<30

20

Time to Perform JPM

<45 min

10 min


Alternate While drywell sprays are being aligned in the 2A Loop of RHR the 2A RHR pump receives an overload condition but does not trip. This requires the operator to trip the 2A RHR pump and establish drywell sprays in the 2B Loop of RHR.


Simulator





Title:

Transfer an Emergency 4160 VAC Bus from the Emergency to the Normal Power Suppy



Time: 20 Minutes


N/A

Date:

N/A


N/A

Date:

N/A


Date:

Course Number N/A




Sup’s Initials

01 06/09/89 General revision and format change WGW RSG

02 08/24/89 Revise questions & add LR lesson plans JEM DHG

03 07/01/92 General revision and format change WMM DHG

04 09/23/93 General revision, word processor change RAB RSG

05 07/18/94 Correct for D/G loading, adjust format RAB SMC

06 02/23/95 Correct simulator setup and prompts to include directions for the simulator Operator to place speed drop and voltage regulator to the correct positions, adjust format, include phonetics in the initiating cue

RAB DHG

07 08/17/95 Format change, incorporate student comment about Unit 1 section

RAB SMC

08 07/05/96 Format change RAB SMC

09 10/01/97 Revised based on comments from the 1997 annual exam.

SCB DHG

10 03/05/99 Revised to make steps for direction of synchscope rotation, step 8 (U1) and Step 9 (U2) noncritical. Revised for new simulator malfunction numbers. SCB DHG

11 02/11/00 Format modification, change time allowance based on running average, update to new simulator operating system RAB DHG

12 03/29/00 TLB comment; change step 10 of U2 for voltage requirement of the diesel

RAB SMC

13 11/02/00 Include objective number, TLB comment, correct Step 9 of Unit 1 section to agree with procedure change RAB DHG

14 03/11/02 Incorporate instructor comment on goal for power decrease of Step 12 for Unit 2, include initial Operator statement RAB RAB

15 03/08/05 Documentum revision DNM RAB


RAB RAB


18 09/17/08 Changed note before step 8 from “lightly loaded” to > 500 KW” to

correspond to procedure guidance. JWP RAB

18.1 10/17/11 Reviewed JPM against current procedure. Added Fundamental question to Attachment 1. Added pass / fail criteria. Added 30AC-OPS-003 for resetting LOSP Lockout Relay. Added the following prompt,” WHEN the Operator addresses Manual Start Permissive lights, INFORM the Operator that U-1 light is illuminated and U-2 light is extinguished.”

MMG ALS

18.2 08/24/15 Added “Summary of JPM Attributes” and changed wording to match procedure including procedure step numbers. Deleted Unit 1 section, will perform in Simulator for Unit 2. Deleted fundamental question.

MMG ALS

18.3

Reviewed JPM against current procedure. Changed “Media Number” to CR-SIM 6 2016-301. After NRC Exam, Media Number will be changed back to LR-JP-02711.

ARB

CR-SIM 6 2016-301 Page 5 of 13


TASK TITLE: Transfer an Emergency 4160 VAC Bus from the

Emergency to the Normal Power Suppy


TASK STANDARD: The task shall be completed when the Operator has transferred the “2F” 4160 VAC Emergency Bus power supply from the associated Emergency Diesel to the “D” Startup Transformer, per 34SO-R43-001, with exception of placing the selected Diesel Generator in Standby.




RO 3.07 SRO 3.07


RO 3.60 SRO 3.60



NA 34SO-R43-001-2, Ver 28.2


NA 34SO-R43-001-2, Ver 28.2

APPROXIMATE COMPLETION TIME: 20 Minutes SIMULATOR SETUP: N/A

CR-SIM 6 2016-301 Page 6 of 13

SIMULATOR SETUP

Simulator Initial Conditions: 1. RESET the Simulator to IC # for 100% power or SNAP 616 and leave in FREEZE. 2. INSERT the following REMOTE FUNCTIONS:

Activator REM # Description Status

ST 0 rfR43241 Diesel Gen 1B Engine Control Switch UNIT 2

RB 1 rfR43295 Diesel Gen 1B Engine Remote Speed Droop (0 to 100) 50

RB 2 rfR43188 Diesel Gen 1B Unit 1 Voltage Reg Transfer MAN


A. Open the Normal and Alternate supply breakers to Bus “2F.”

B. Verify the “1B” D/G output breaker closes and match flags on the breaker switch.

C. Place Diesel Generator “B” Volt Select switch in OFF position.

D. Ensure Diesel Generator load is greater than 500 kW. If necessary, start another PSW pump on that bus.


UNIT 2


1. 4160 VAC Emergency Bus “2F” is being supplied from Emergency Diesel Generator “1B.”

2. The condition that caused Bus “2F” to de-energize has been corrected.

3. The normal off-site power supply is available.

4. 34AR-652-202-2, LOSS OF OFF-SITE POWER, is in progress.

5. Diesel Generator “1B” control has been transferred to Unit 2.


Transfer the 4160 VAC Bus “2F” to the normal power supply.

CR-SIM 6 2016-301 Page 8 of 13

STEP #




IF THEN

PASS


correctly





1. Operator determines the correct procedure.

The Operator DETERMINES that procedure 34SO-R43-001-2, section 7.2.2.3 should be used.

2. Confirm power has been restored to Startup Transformers 2C OR 2D. (potential lights for Startup Aux Xfmr 2C OR 2D ILLUMINATED, panel 2H11-P651) (step 7.2.2.3.1)

At panel 2H11-P651, the Operator has VERIFIED that SUT “2D” potential lights are illuminated.

PROMPT: WHEN the Operator addresses the requirements to reset the LOSP Lockout

Relay, as the Shift Supervisor, INFORM the Operator that all requirements are met per NMP-OS-007-001 and 31GO-OPS-021-0.

CR-SIM 6 2016-301 Page 9 of 13

STEP #



** 3. Reset 4160V Bus 2F LOSP Lockout Relay in accordance with NMP-OS-007-001, Conduct of Operations Standards and Expectations, and 31GO-OPS-021-0, Manipulation of Controls and Equipment, AND ***************************** (step 7.2.2.3.2)

At panel 2H11-P652, the Operator RESETS BUS “2F” LOSP LOCKOUT RELAY.

(Switch TURNED clockwise and HELD until its white light illuminates).

4. ****************************************************************************************, AND,

confirm annunciator 652-202, LOSS OF OFF SITE POWER, is clear. (step 7.2.2.3.2)

At panel 2H11-P652, the Operator CONFIRMS annunciator 652-202, LOSS OF OFFSITE POWER, is clear.

** 5. Place Diesel Gen 1B Mode Select Switch in TEST. (step 7.2.2.3.3)

At panel 2H11-P652, the Operator PLACES Diesel Gen 1B Mode Select Switch in TEST.

PROMPT: IF the Operator addresses Diesel Generator “1B” keylock switch, as the SO,

INFORM the Operator the switch is in the Remote Unit 2 position (step 7.2.2.3.4).

PROMPT: WHEN the Operator addresses Manual Start Permissive lights, INFORM

the Operator that U-1 light is extinguished (step 7.2.2.3.4.2). PROMPT: WHEN the Operator addresses adjusting speed droop, the simulator

Operator should TOGGLE REMOTE FUNCTION rfR43295 , (RB 1) “D/G 1B Engine Remote Speed Droop (0 to 100),” to 50. As the SO, INFORM the Operator that speed droop has been adjusted to 50. (step 7.2.2.3.5).

PROMPT: WHEN the Operator addresses the position of the Unit 1 Voltage Regulator

Transfer switch, the simulator Operator should TOGGLE REMOTE FUNCTION rfR43188, (RB 2) “Diesel Gen 1B Unit 1 Voltage Reg Transfer,” to MAN . As the Unit 1 Control Room Operator, INFORM the Operator that the switch is in manual. (step 7.2.2.3.6).

CR-SIM 6 2016-301 Page 10 of 13

STEP #



** 6. Place OR confirm Diesel Gen 1B Voltage Reg Transfer switch to MANUAL, panel 2H11-P652. (step 7.2.2.3.7)

At panel 2H11-P652, the Operator PLACES VOLTAGE REG TRANSFER switch in MANUAL, green light illuminated.

** 7. Place Synch Switch (SSW) in ON for: ACB 135574, 4160V Bus 2F Normal Supply,

OR

ACB 135564, 4160V Bus 2F Alternate Supply. (step 7.2.2.3.8)

At panel 2H11-P652, the Operator PLACES SSW ACB 135574, in the ON position.

NOTE: Diesel Generator B Volt Select switch must be turned on to check DG

voltage.

8. Adjust the Diesel Gen 1B Voltage Adjust switch UNTIL diesel output voltage is equal to 4160V. (step 7.2.2.3.9)

At panel 2H11-P652, the Operator ADJUSTS the “1B” DIESEL GEN B VOLTAGE ADJUST switch as necessary to indicate approximately 4160 volts on VOLTMETER for 4KV BUS “2F”.

NOTE: For the purpose of this JPM, 4000 to 4300 volts is considered matched. NOTE: For performance of this JPM, the Diesel Generator is Heavly loaded ( > 500 KW).

9. Using Diesel Gen 1 B Speed Adjust switch, adjust diesel speed to attain a slow synchroscope rotation in the desired direction (1 to 3 rpm). (step 7.2.2.3.10)

At panel 2H11-P652, the Operator ADJUSTS the DIESEL GEN B SPEED ADJUST switch as needed until the Synchroscope rotation is in the clockwise direction at 1 to 3 rpm.

CR-SIM 6 2016-301 Page 11 of 13

STEP #



10. Using Diesel Gen 1B Voltage Adjust, adjust Diesel Gen 1B output voltage to match the highest phase of the incoming source (Startup Transformer 2D (normal) OR 2C (alternate)). (step 7.2.2.3.11)

At panel 2H11-P652, the Operator ADJUSTS Diesel output voltage using DIESEL GEN B, VOLTAGE ADJUST to match the highest phase of SUT “2D.”

NOTE: The Operator should not exceed 4400 volts on any phase of the Diesel

Generator (Although Step 10 is not a critical step, if Operator exceeds 4400 volts, then Step 10 becomes a critical step and would constitute Operator failure).

NOTE: For the purpose of this JPM, ±100 volts is considered matched.

** 11. WHEN the sychroscope indicates 2 minutes to 12 AND WHEN the synchroscope lights approach the dimmest point, take: ACB 135574, 4160V 2F Normal Supply,

OR

ACB 135564, Alternate Suppy,

To CLOSE. (step 7.2.2.3.12)

At panel 2H11-P652, the Operator TAKES 4160 “2F,” NORMAL SUPPLY ACB 135574 control switch to CLOSE,

(when synchroscope is 2 minutes to 12:00 position and when the synchroscope lights approach the dimmest point, then released to the midposition, red light illuminated.)

NOTE: If the synchroscope for ACB 135574 was not energized the breaker

remains OPEN, green light on.

12. Using the Diesel Gen 1B Speed Adjust switch, decrease the load on diesel to between 400 AND 500 KW WHILE maintaining diesel reactive load between 400 AND 500 KVAR, using Diesel Gen 1B Voltage Adjust switch. (step 7.2.2.3.13)

At panel 2H11-P652, the Operator ADJUSTS the DIESEL GEN B, VOLT ADJUST and SPEED ADJUST switches as necessary to reduce load on Diesel to 400-500 KW while maintaining 400-500 KVAR.

CR-SIM 6 2016-301 Page 12 of 13

STEP #



** 13. Take ACB 135570, Diesel Gen 1B Emergency Supply, control switch to TRIP. (step 7.2.2.3.14)

At panel 2H11-P652,the Operator TAKES DIESEL GEN B, EMERGENCY SUPPLY ACB 135570 control switch to TRIP and released to midposition, green light illuminated.

14. Confirm ACB 135570, Diesel Gen 1B Emergency Supply, green (OPEN) light is ILLUMINATED. (step 7.2.2.3.15)

At panel 2H11-P652, the Operator IDENTIFIES that the DIESEL GEN B, EMERGENCY SUPPLY ACB 135570 green light is illuminated.

15. Place Synch Switch (SSW) in OFF for:

ACB 135574, 4160V Bus Normal Supply

OR

ACB 135564, Alternate Supply. (step 7.2.2.3.16)

At panel 2H11-P652, the Operator PLACES SSW ACB 135574 in OFF.

PROMPT: WHEN the Operator addresses securing of the Diesel Generator, as the

Shift Supervisor, INFORM the Operator that another Operator will shutdown the Diesel Generator.

End Time: _________





of 13


JPM JP02711-18.2: SUMMARY OF JPM QUANTITATIVE ATTRIBUTES CATEGORY Minimum NRC



At least 2

6

Step 3 Reset Lockout Relay Operator must reset Lockout Relay to be able to close the normal supply breaker.

Step 5 EDG in Test Operator must place Mode Select Switch to TEST to be able to parallel sources.

Step 6 Volt Reg to Manual Voltage Reg is transferred to manual to allow voltage to be adjusted.

Step 7 Synch Switch ON Operator must place synch switch for normal breaker to ON to defeat interlock so breaker can be closed.

Step 11 Close Norm Bkr To supply power from normal supply, the breaker must be closed.

Step 13 Open EDG Bkr To un-parallel electrical sources, must open EDG output breaker.

Number of JPM Steps

<30

15

Time to Perform JPM

<45 min

20 min


Normal Operator is told to transfer power from the 1B EDG to the normal supply.


Simulator


of 10



DRAFT CR-SIM 7 (RO Only)

Title:

Conduct a Rod Worth Minimizer (RWM) Functional Test (Failure)



Time: 10.0 Minutes


N/A

Date:

N/A


N/A

Date:

N/A

Approved By (Training Program Supervisor or Lead Instructor)

Date:

of 10

Course Number

N/A

Program Name

OPERATIONS TRAINING

Media Number

CR-SIM 7 2016-301


Sup’s Initials

00 11/15/99 Initial development RAB DHG



03 03/17/05

Deleted “S” from procedure numbers, changed Revision and Rev. numbers to “Current Version”. Changed to Nuclear Plant Operator. Updated Simulator IC# and added “Prompt” for sequence control mode.

ARB DHG

04 06/27/05 Revised Initial License statement for successful completion RAB RAB


05.1 10/17/11

Reviewed JPM against current procedure. Added pass / fail criteria. Added Fundamental question to new Attachment 1. Corrected JPM to reflect what a Rod Group 3 and 4 rod is. Added prompt to tell student what Step a Rod Group 3 & 4 is.

MMG ALS

05.2 2/18/16

Reviewed JPM against current procedure for use on 2016-301 NRC Exam. Changed the Media Number to CR-SIM 7 2016-301. Deleted Attachment 1 (Fundamental Question).

RAG

of 10 Line Contributors

The following individuals contributed to the development of this lesson plan.


05.1 MMG

CR-SIM 7 2016-301 Page 1 of 10


TASK TITLE: Conduct a Rod Worth Minimizer (RWM) Functional

Test (Failure)


TASK STANDARD: The task shall be completed when the operator has successfully conducted the Rod Worth Minimizer Functional Test per 34GO-OPS-001.





RO 3.40 SRO 3.50



N/A 34GO-OPS-001-2, Att 5, Ver 47.1 34GO-OPS-065-0, Ver 12.6 34AB-C11-004-2, Ver 4.1 Control Rod Movement Sequence


N/A 34GO-OPS-001-2, Att 5, Ver 47.1 Control Rod Movement Sequence

APPROXIMATE COMPLETION TIME: 10.0 Minutes SIMULATOR SETUP: Refer to Simulator Setup sheet on the following page.

CR-SIM 7 2016-301 Page 2 of 10

SIMULATOR SETUP

Simulator Initial Conditions: 1. RESET the Simulator to All Rods IN and leave in FREEZE.

2. ACTIVATE the following Event Trigger:


EGC91-01 Inserts Override C91-J001 f2 loC51-RS-50-31B1.algToPanel=1 3. INSERT the following REMOTE FUNCTIONS:

REM # DESCRIPTION STATUS

rfC51_190 RWM Sequence Control ON


A. RESET all annunciators.

B. SELECT a Control Rod from Step 1 and initialize the RWM by pushing the ETC button on the operator display.

C. VERIFY that all RWM rod blocks are clear D. VERIFY the Rod Sequence Selector Switch is in A12 (B12).

E. VERIFY RWM Sequence Control is in the ON position.

4. PLACE the Simulator in FREEZE until the crew assumes the shift. 5. ESTIMATED Simulator SETUP TIME : 10 Minutes

UNIT 2


1. The Reactor is shutdown; the manual scram has been reset. The Reactor Mode Switch is in START & HOT STBY.

2. All systems required for plant startup are in service. Permission for Reactor startup has been given.

3. 34GO-OPS-001-2, Plant Startup, is in progress. INITIATING CUES:

Perform the Rod Worth Minimizer Functional Test using Attachment 5 of 34GO-OPS-001-2.

CR-SIM 7 2016-301 Page 4 of 10

STEP #




IF THEN

PASS


correctly





1. Operator confirms that power is turned on to the Refueling Bridge.

Operator has VERIFIED power is turned on to the Refueling Bridge.

PROMPT: WHEN the operator addresses confirming power on the Refueling Bridge,

as the SO, INFORM the operator that the Refueling Bridge has power.

2. Confirm RMCS/RWM ROD BLOCK OR SYS TROUBLE annunciator is clear.

At panel 2H11-P603, the operator VERIFIES that the annunciator 603-239, RMCS/RWM ROD BLOCK OR SYS TROUBLE, is extinguished.

PROMPT: WHEN the operator addresses the Select Sequence, INFORM the operator

that the currently selected sequence in the RWM is identical to the Control Rod Movement Sheets and conforms to BPWS.

PROMPT: IF the operator addresses bypassed rods and the Change Log of the Control

Rod Movement Sequence, INFORM the operator there are no bypassed rods listed.

CR-SIM 7 2016-301 Page 5 of 10

STEP #



PROMPT: IF the operator addresses RWM's Full Range Sequence Control mode, INFORM the operator that RWM is in Sequence Control mode.

3. Confirm RWM keylock switches are in OPERATE.

At panel 2H11-P603, the Operator’s Display keylock switch is in OPERATE.

Operator IDENTIFIES that the keylock switch at panel 2H11-P616 must also be in OPERATE.

PROMPT: WHEN the operator addresses the Instrument Console keylock switch on

panel 2H11-P616, INFORM the operator it is in the OPERATE position.

4. Confirm Step 01 is displayed on RWM Operator’s Display.

At panel 2H11-P603, the operator has VERIFIED that 01 is displayed on the RWM Operator’s Display.

5. Confirm the Reactor Mode switch is in START & HOT STBY.

At panel 2H11-P603, the operator places or CONFIRMS the Reactor Mode switch is in START & HOT STBY.

NOTE: This step should only require the operator to confirm the mode switch

position. This is one of the initial conditions given to the operator at the beginning of the task.

6. Select a rod from Step 2 of the Rod Worth Minimizer Sequence and verify system response.

At panel 2H11-P603, using the CONTROL ROD SELECT matrix pushbuttons, a rod is SELECTED from Step 2 of the RWM Sequence

AND

At the RWM operator’s display, the operator has CONFIRMED the following messages are displayed: -“SE” “WB” - BLOCK: WITHDRAW”.

NOTE: If the operator selects any rod NOT in Step 1 of the RWM Sequence, they

will still receive the same indications as a Step 2 rod.

CR-SIM 7 2016-301 Page 6 of 10

STEP #



7. Attempt to withdraw the selected rod. At panel 2H11-P603, the operator VERIFIES no rod motion occurred.

PROMPT: WHEN the operator asks the STA/RE for a Rod Group 3 rod, INDICATE that Rod Group 3 starts at Step 7.

8. Select a rod from Rod Group 3 of the Rod Worth Minimizer Sequence and verify system response.

[In our current pull sequence, Rod Group 3 starts at Step 3, rod 02-23]

At panel 1H11-P603, using the CONTROL ROD SELECT matrix pushbuttons, a rod is SELECTED from Rod Group 3 of the RWM Sequence

AND

At the RWM operator’s display, the operator has CONFIRMED the following messages are displayed: -”SE” “WB” -”BLOCK: WITHDRAW”.

9. Attempt to withdraw the selected rod. At panel 2H11-P603, the operator VERIFIES no rod motion occurred.

NOTE: When the operator selects a Rod Group 4 rod, Event Trigger EGC91-01 will ACTIVATE OVERRIDE C91_J001DI to BYPASS.

PROMPT: WHEN the operator asks the STA/RE for a Rod Group 4 rod, INDICATE that Rod Group 4 starts at Step 15.

** 10. Select a rod from Rod Group 4 of the Rod Worth Minimizer Sequence and verify system response.

[In our current pull sequence, Rod Group 4 starts at Step 11, rod 50-31].

At panel 1H11-P603, using the CONTROL ROD SELECT matrix pushbuttons, a rod is SELECTED from Rod Group 4 of the RWM Sequence

AND

At the RWM operator’s display, the operator has CONFIRMED the following messages are displayed: -”SE” “WB” -”BLOCK: WITHDRAW”.

CR-SIM 7 2016-301 Page 7 of 10

STEP #



** 11. Attempt to withdraw the selected rod. At panel 2H11-P603, the operator RECOGNIZES that the control rod moved.

** 12. Insert the control rod and notify the SS that the RWM failed its operability check.

At panel 2H11-P603, the operator INSERTS the control rod to the full in position; AND,

Notifies the SS that the RWM has failed its operability check.

PROMPT: IF addressed by the operator, INFORM the operator that the STA has

confirmed the RWM Scram Buffers are clear and ready to accept data.

END

TIME: __________




− Operator states the task is complete. TERMINATING CUE: Another operator will continue from here.

of 13




Title:

Perform 34IT-T45-001-2, Reactor Building Instrument Sumps Isolation Valve Exercise

Author: Richard A. Greenhouse


Time: xx.x Minutes


Date:


Date:

Approved By (Training Program Supervisor or Lead Instructor)

Date:

of 13


of 13

Course Number N/A




Sup’s Initials

00 2/3/16 Initial development for use on 2016-301 NRC Exam.

RAG

of 13



CR-SIM 8 2016-301 Page 5 of 13


TASK TITLE: Perform 34IT-T45-001-2, Reactor Building Instrument Sumps Isolation Valve Exercise


TASK STANDARD: This task shall be successfully completed when all of the JPM

Critical Steps corresponding to 34IT-T45-001-2, Sections 7.2, 7.3 and 7.4, have been correctly performed to complete the Instrument Sump Isolation Valve exercise.

TASK NUMBER: xxx.xxx

OBJECTIVE NUMBER: xxx.xxx.x


RO x.xx SRO x.xx

K/A CATALOG NUMBER: 268000K1.04 K/A CATALOG JTA IMPORTANCE RATING :

RO 2.7 SRO 2.9



NA 34IT-T45-001-2, Ver 0.9


NA 34IT-T45-001-2, Ver 0.9

APPROXIMATE COMPLETION TIME: xx.x Minutes SIMULATOR SETUP: Refer to Setup sheet on the following page.

CR-SIM 8 2016-301 Page 6 of 13

SIMULATOR SETUP

Simulator Initial Conditions: 1. RESET the Simulator to IC for 100% RTP and leave in FREEZE. 2. Take the Simulator OUT OF FREEZE and PERFORM the following MANIPULATIONS :

A. None


UNIT 2


1. Unit 2 is operating at 100% RTP

2. 34IT-T45-001-2, Reactor Building Instrument Sumps Isolation Valve Exercise, is in progress. Section 7.1 of 34IT-T45-001-2 is complete.

INITIATING CUES:

Perform 34IT-T45-001-2, Reactor Building Instrument Sumps Isolation Valve Exercise, sections 7.2, 7.3 and 7.4.

CR-SIM 8 2016-301 Page 8 of 13

STEP #



IF THEN

PASS


correctly





NOTE To Evaluator: Perform the turn-over for this JPM prior to entering the simulator when possible (i.e allow the student to review the procedure in the Virtual Simulator or Simulator Assessment Room, etc.) Doing this will minimize the time the simulator is not being actively used.

PROMPT: WHEN the operator addresses obtaining permission from the Shift Supervisor, INFORM the operator that permission has been granted.

1. Confirm the Sump Isol switch for 2T45-F005, 2T45-F003, and 2T45-F001 is in CLOSED, panel 2H11-P654.

(Step 7.2.1)

At panel 2H11-P654, the operator confirms the control switch for 2T45-F005, 2T45-F003 and 2T45-F001 is in the CLOSE position.

CR-SIM 8 2016-301 Page 9 of 13

STEP #


2. Confirm the following valves are CLOSED:

• 2T45-F005, RHR N-E Inbd Sump Isol

• 2T45-F003, Torus N-E & S-E Inbd Sump Isol

• 2T45-F001, HPCI Sump Isol

(Step 7.2.2)

At panel 2H11-P54, the operator confirms the following:

2T45-F005 is CLOSED, green light illuminated.



** 3. Place the Sump Isol switch for 2T45-F005, 2T45-F003, and 2T45-F001, in AUTO:

• 2T45-F005

• 2T45-F003

• 2T54-F001 (Step 7.2.3)

At panel 2H11-P54, the operator places the control switch for 2T45-F005, 2T45-F003 and 2T45-F001 in the AUTO position.

4. Confirm the following valves OPEN:

• 2T45-F005 • 2T45-F003 • 2T45-F001

(Step 7.2.4)


2T45-F005 is OPEN, red light illuminated.



** 5. Place the Sump Isol switch for 2T45-F005, 2T45-F003, and 2T45-F001 in CLOSE:




(Step 7.2.5)

At panel 2H11-P54, the operator places the control switch for 2T45-F005, 2T45-F003 and 2T45-F001 in the CLOSE position.

CR-SIM 8 2016-301 Page 10 of 13

STEP #


6. Confirm the following valves CLOSE:




(Step 7.2.6)





7. Confirm the Sump Isol switch for 2T45-F007 and 2T45-F006 is in CLOSED, panel 2H11-P654.

(Step 7.3.1)

At panel 2H11-P654, the operator confirms the control switch for 2T45-F007 and 2T45-F006 is in the CLOSE position.


• 2T45-F007, Torus N-W & S-W Inbd Sump Isol

• 2T45-F006, RCIC Sump Isol

(Step 7.3.2)




** 9. Place the Sump Isol switch for 2T45-F007 and 2T45-F006 in AUTO:

• 2T45-F007

• 2T45-F006 (Step 7.3.3)

At panel 2H11-P54, the operator places the control switch for 2T45-F007 and 2T45-F006 in the AUTO position.


• 2T45-F007 • 2T45-F006

(Step 7.3.4)




** 11. Place the Sump Isol switch for 2T45-F007 and 2T45-F006 in CLOSE:



(Step 7.3.5)

At panel 2H11-P54, the operator places the control switch for 2T45-F007 and 2T45-F006 in the CLOSE position.

CR-SIM 8 2016-301 Page 11 of 13

STEP #





(Step 7.3.6)




13. Confirm the Sump Isol switch for 2T45-F004 and 2T45-F002 is in CLOSED, panel 2H11-P654.

(Step 7.4.1)

At panel 2H11-P654, the operator observes the control switch for 2T45-F004 and 2T45-F002 is in the CLOSE position.


• 2T45-F004, RHR N-E Outbd Sump Isol

• 2T45-F002, Torus N-E & S-E Outbd Sump Isol

(Step 7.4.2)




** 15. Place the Sump Isol switch for 2T45-F004 and 2T45-F002 in AUTO:

• 2T45-F004

• 2T45-F002 (Step 7.4.3)

At panel 2H11-P54, the operator places the control switch for 2T45-F004 and 2T45-F002 in the AUTO position.


• 2T45-F004 • 2T45-F002

(Step 7.4.4)




** 17. Place the Sump Isol switch for 2T45-F004 and 2T45-F002 in CLOSE:


• 2T45-F002, Torus N-E & S-E Sump Isol

(Step 7.4.5)

At panel 2H11-P54, the operator places the control switch for 2T45-F004 and 2T45-F002 in the CLOSE position.

CR-SIM 8 2016-301 Page 12 of 13

STEP #




• 2T45-F002, Torus N-E & S-E Sump Isol

(Step 7.4.6)




PROMPT: ONCE the operator has completed Step 7.4.6, INFORM the operator that

another operator will complete the restoration section of the procedure.

END TIME: __________





of 13

UNIT 2


1. Unit 2 is operating at 100% RTP

2. 34IT-T45-001-2, Reactor Building Instrument Sumps Isolation Valve Exercise, is in progress. Section 7.1 of 34IT-T45-001-2 is complete.

INITIATING CUES:

Perform 34IT-T45-001-2, Reactor Building Instrument Sumps Isolation Valve Exercise, sections 7.2, 7.3 and 7.4.



DRAFT PLANT 1 (ALL)

Title:

From Outside the MCR, Insert a Manual Reactor Scram (Alternate Path)


Media Number: PLANT 1 2016-301

Time: 14 Minutes


Date:


Date:


Date:

Course Number

N/A Program Name

OPERATIONS TRAINING Media Number

PLANT 1 2016-301


Sup’s Initials

0.0 08/06/15

Initial development. A manual scram by opening RPS breakers will not work forcing the Operator to use the SDV Level Switches in the Reactor Building. Added “Summary of JPM Attributes” and changed wording to match procedure including procedure step numbers.

MMG ALS

0.1

Reviewed JPM against current procedure for use on 2016-301 NRC Exam. Changed Media number to PLANT 1 2016-301 and will be returned to LR-JP-01018A after 2016-301 NRC Exam.

ARB


TASK TITLE: From Outside the MCR, Insert a Manual Reactor Scram

(Alternate Path)

JPM NUMBER: PLANT 1 2016-301

TASK STANDARD: The task shall be completed when the operator has successfully inserted a reactor scram using the SDV Level Switches per 31RS-OPS-001 by tripping magnetrol switches 1C11-N013A & B OR 1C11-N013C & D.




RO 4.20 SRO 4.67


RO 3.6 SRO 3.9



31RS-OPS-001-1, Ver 5.24 31RS-OPS-001-2, Ver 6.23


31RS-OPS-001-1, Ver 5.24 31RS-OPS-001-2, Ver 6.23


UNIT 1


1. A condition has occurred which required the Control Room to be evacuated.

2. The Reactor is NOT shutdown.

3. 31RS-OPS-001-1, Shutdown From Outside the Control Room, is in progress.


Insert a Manual Scram by de-energizing RPS per 31RS-OPS-001-1.

PLANT 1 2016-301 Page 7 of 20

STEP #




IF THEN

PASS


correctly






The Operator has IDENTIFIED the correct procedure as 31RS-OPS-001-1, step 4.4.

2. De-energize RPS.

OPEN the following breakers at RPS distribution panel 1C71-P001:

• 1C71-CB3A

• **********

(Step 4.4.1.1)

At panel 1C71-P001, the Operator OPENS breaker 1C71-CB3A.

3. De-energize RPS.

OPEN the following breakers at RPS distribution panel 1C71-P001:

• **********

• 1C71-CB3B

(Step 4.4.1.1)

At panel 1C71-P001, the Operator ATTEMPTS to open breaker 1C71-CB3B. The Operator RECOGNIZES that the breaker will not reposition.

PLANT 1 2016-301 Page 8 of 20

STEP #



PROMPT: WHEN the Operator attempts to open breaker 1C71-CB3B, INFORM the Operator that the breaker did NOT reposition and can not be moved.

PROMPT: IF the Operator requests direction from the Shift Supervisor, INFORM the

Operator to Insert a Manual Scram per 31RS-OPS-001-1.

NOTE: The Operator may elect to reclose breaker 1C71-CB3A prior to addressing the SDV Level Switches.

ALTERNATE PATH STARTS HERE (Step 4)

NOTE: The critical steps will be either steps 4, 5, 8, and 9 OR steps 6, 7, 10, and 11. Both

switches on one side of the Reactor Building will insert a Full Reactor Scram. To receive credit, the Operator ONLY has to trip both magnetrol switches on one side of the Reactor Building.

** 4. Scram Discharge Volume Level Switches.

At the Scram Discharge Volume, remove the covers to one of the following sets of level switches:

1C11-N013A and ********* OR **************************

(Step 4.4.2.1)

At location 139RER11, the Operator REMOVES the cover from magnetrol switch 1C11-N013A.



********** and 1C11-N013B, OR **************************

(Step 4.4.2.1)

At location 139RER11, the Operator REMOVES the cover from magnetrol switch 1C11-N013B.

NOTE: The Operator may remove the covers for 1C11-N013A and B, then trip their magnetrol

switches and replace their covers. The Operator may stop there or proceed to performing the same actions for 1C11-N013C and D.

PLANT 1 2016-301 Page 9 of 20

STEP #





*************************** OR 1C11-N013C and ************

(Step 4.4.2.1)

At location 139RER03, the Operator REMOVES the cover from magnetrol switch 1C11-N013C.



*************************** OR ********** and 1C11-N013D

(Step 4.4.2.1)

At location 139RER03, the Operator REMOVES the cover from magnetrol switch 1C11-N013D.

** 8. Trip magnetrol switches for the switches with removed covers:

1C11-N013A and ***********, OR **************************

(Step 4.4.2.2)

At location 139RER11, the Operator TRIPS magnetrol switch 1C11-N013A.


********** and 1C11-N013B, OR **************************

(Step 4.4.2.2)

At location 139RER11, the Operator TRIPS magnetrol switche 1C11-N013B.


*************************, OR 1C11-N013C and **********

(Step 4.4.2.2)

At location 139RER03, the Operator TRIPS magnetrol switche 1C11-N013C.

PLANT 1 2016-301 Page 10 of 20

STEP #




*************************, OR ********** AND 1C11-N013D

(Step 4.4.2.2)

At location 139RER03, the Operator TRIPS magnetrol switche 1C11-N013D.

12. Replace the covers to the magnetrol switches.

(Step 4.4.2.3)

Operator REPLACES all removed magnetrol switch covers.

END TIME: __________





UNIT 2


1. A condition has occurred which required the Control Room to be evacuated.

2. The Reactor is NOT shutdown.

3. 31RS-OPS-001-2, Shutdown From Outside the Control Room, is in progress.


Insert a Manual Scram by de-energizing RPS per 31RS-OPS-001-2.

PLANT 1 2016-301 Page 12 of 20

STEP #




IF THEN

PASS


correctly






Operator has IDENTIFIED the correct procedure as 31RS-OPS-001-2 step 4.4

2. De-energize RPS.

OPEN the following breakers at RPS distribution panel 2C71-P001, at 130TGT12.

• 2C71-CB3A

• **********

(Step 4.4.1.1)

At panel 2C71-P001, the Operator OPENS breaker 2C71-CB3A.

PROMPT: WHEN the Operator attempts to open breaker 2C71-CB3B, INFORM the

Operator that the breaker did NOT reposition and can not be moved.

PLANT 1 2016-301 Page 13 of 20

STEP #



3. De-energize RPS.

OPEN the following breakers at RPS distribution panel 2C71-P001, at 130TGT12.

• **********

• 2C71-CB3B

(Step 4.4.1.1)

At panel 2C71-P001, the Operator ATTEMPTS to open breaker 2C71-CB3B. The Operator RECOGNIZES that the breaker will not reposition.

PROMPT: IF the Operator requests direction from the Shift Supervisor, INFORM the

Operator to Insert a Manual Scram per 31RS-OPS-001-2. NOTE: The Operator may elect to reclose breaker 2C71-CB3A prior to addressing

the SDV Level Switches. ALTERNATE PATH starts here (Step 4).

NOTE: The critical steps will be either steps 4, 5, 8, and 9 OR steps 6, 7, 10, and 11. Both switches on one side of the Reactor Building will insert a Full Reactor Scram. To receive credit, the Operator ONLY has to trip both magnetrol switches on one side of the Reactor Building.


Remove the covers to 2C11-N013A & ***********, Magnetrol Switches, at 130RBR15 ******************* **************************** (Step 4.4.2.1)

At location 130RBR15, the Operator REMOVES the cover from magnetrol switch 2C11-N013A.


Remove the covers to ********* & 2C11-N013B, Magnetrol Switches, at 130RBR15 ******************* **************************** (Step 4.4.2.1)

At location 130RBR15, the Operator REMOVES the cover from magnetrol switch 2C11-N013B.

NOTE: The Operator may remove the covers for 2C11-N013A and B, then trip

their magnetrol switches and replace their covers. The Operator may stop there or proceed to performing the same actions for 2C11-N013C and D.

PLANT 1 2016-301 Page 14 of 20

STEP #




Remove the covers ************* ***************************** ******* OR 2C11-N013C & ***********, Magnetrol Switches, at 130RBR23. (Step 4.4.2.1)

At location 130RBR23, the Operator REMOVES the cover from magnetrol switch 2C11-N013C.


Remove the covers ************* ***************************** ******* OR ************ & 2C11-N013D, Magnetrol Switches, at 130RBR23. (Step 4.4.2.1)

At location 130RBR23, the Operator REMOVES the cover from magnetrol switch 2C11-N013D.

** 8. Trip the magnetrol switches.

(Step 4.4.2.2)

At location 130RBR15, the Operator TRIPS magnetrol switche 2C11-N013A.


(Step 4.4.2.2)

At location 130RBR15, the Operator TRIPS magnetrol switche 2C11-N013B.


(Step 4.4.2.2)

At location 130RBR23, the Operator TRIPS magnetrol switche 2C11-N013C.


(Step 4.4.2.2)

At location 130RBR23, the Operator TRIPS magnetrol switche 2C11-N013D.

12. Replace the covers to the magnetrol switches.

(Step 4.4.2.3)

Operator REPLACES all removed magnetrol switch covers.

END TIME: __________

PLANT 1 2016-301 Page 15 of 20

STEP #







of 20


JPM PLANT 1 2016-301: SUMMARY OF JPM QUANTITATIVE ATTRIBUTES CATEGORY Minimum NRC



At least 2

4

Step 4/6 Remove SDV Level Switch Cover Operator removes 1/2C11-N013A(C) magnetrol switch cover to gain access to switch.

Step 5/7 Remove SDV Level Switch Cover Operator removes 1/2C11-N013B(D) magnetrol switch cover to gain access to switch.

Step 8/10 Trip SDV Level Switch Operator Trips 1/2C11-N013A(C) magnetrol switch to de-energize RPS A.

Step 9/11 Trip SDV Level Switch Operator Trips 1/2C11-N013B(D) magnetrol switch to de-energize RPS B.

Number of JPM Steps

<30

12

Time to Perform JPM

<45 min

14 min


Alternate Path When the Operator attempts to insert aReactor Scram by de-energizing RPS, one of the breakers will not open. The Operator must decide to continue with the procedure and insert a Reactor Scram with the SDV Level Switches in the Reactor Building.


Plant




DRAFT PLANT 2 (RO & SRO-I)

Title:

From the Remote Shutdown Panel, Start RHR and Inject into the Reactor

Author:

Richard A. Greenhouse


Time

13 Minutes

Line Technical Review By (N/A for minor revisions) Date:

Reviewed by Instructional Technologist or designee: Date


Date:

Course Number N/A


Media Number PLANT 2 2016-301


Sup’s Initials

11 02/04/00 Format modification, modify title, upgrade to the new simulator operating system

RAB DHG


13 02/26/02 Include initial operaotr statement RAB RAB

14 03/17/05 Deleted “S” from procedure numbers, changed Revision and Rev. numbers to “Current Version,” changed “Reactor Operator” to “Nuclear Plant Operator,” and changed IC #104 for Simulator Setup.

BEB DHG


RAB RAB


17 09/21/09 Added HU steps. CLN ALD

17.1 8/11/11 Reviewed JPM against current procedure. Added pass / fail criteria. Added to ensure Reactor pressure is < 200 psig and RWL is -120” in the simulator setup. Added “Confirms Reactor pressure is less than 449 (500) psig” prior to opening injection valves.

MMG DNM

17.2 Reviewed JPM against current procedure. Changed media number to PLANT 2 2016-301. Removed U1 material and will be replaced once NRC exam is complete.

RAG

PLANT 2 2016-301 Page 4 of 12


TASK TITLE: From the Remote Shutdown Panel, Start RHR and

Inject into the Reactor


TASK STANDARD: The task shall be completed when the operator has successfully started the “B” loop of RHR in the LPCI Mode and injected into the Reactor from the Remote Shutdown Panel per 31RS-OPS-001.





RO 4.20 SRO 4.30



N/A 31RS-OPS-001-2 (current version)


N/A 31RS-OPS-001-2 (current version)

Key for Remote Shutdown Panel (if performed in plant)


PAGE

PLANT 2 2016-301 Page 5 of 12

SIMULATOR SETUP

Simulator Initial Conditions: 1. RESET the Simulator to IC #104 and leave in FREEZE. 2. INSERT the following MALFUNCTIONS:


RAMP RATE

ACT. TIME

mfB31_210A Recirc Loop A Suction Rupture (Var) 10 100 00000

mfE21_202A Core Spray LOCA Signal Failure 00000

mfE11_202B RHR LOCA Signal Failure 00000


A. Turn off Condensate and Condensate Booster Pumps. B. Place Mode Switch in Shutdown. C. Ensure Reactor Pressure is < 200 psig. D. Ensure RWL is about -120”.

4. PLACE the Simulator in FREEZE until the crew assumes the shift. 5. ESTIMATED Simulator SETUP TIME : 10 Minutes

UNIT 2


1. An event has occurred which resulted in the evacuation of the Control Room.

2. Normal power supplies are available.

3. RHR Loop “B” is in Standby.

4. RWL is -120 inches and RHR has not initiated.

5. Reactor pressure is <200 psig.

6. 31RS-OPS-001-2 is in progress. INITIATING CUES:

From the Remote Shutdown Panel, initiate RHR Loop “B” in the LPCI mode and inject into the Reactor per 31RS-OPS-001-2.

PLANT 2 2016-301 Page 7 of 12

STEP #






1. Operator identifies the materials that are required.

Operator identifies the required materials and where to obtain them.

NOTE: ** Indicates Transfer Switches that MUST be positioned to

satisfactorily complete the critical step.

2. Place the following Transfer Switches in EMERG position:

2C82-S53

Step1.0

At Panel 2C82-P001, the Operator has placed 2C82-S53 Transfer Switch in the EMERG position.


2C82-S52

Step 1.0



2C82-S5

Step 1.0


IF THEN

PASS


correctly



PLANT 2 2016-301 Page 8 of 12

STEP #




2C82-S18

Step 1.0



2C82-S8

Step 1.0


** 7. Place the following Transfer Switch in EMERG position:

2C82-S80

Step 1.0



2C82-S9

Step 1.0



2C82-S13

Step 1.0



2C82-S10

Step 1.0



2C82-S1

Step 1.0



2C82-S12

Step 1.0



2C82-S17

Step 1.0


PLANT 2 2016-301 Page 9 of 12

STEP #




2C82-S14

Step 1.0



2C82-S11

Step 1.0


** 16. Place the following Transfer Switches in EMERG position:

2C82-S16

Step 1.0


PROMPT: IF addressed, INFORM the operator RHRSW Pumps “2B” and “2D” are

tripped. PROMPT: AT this time, INFORM the operator that Reactor Recirculation Pumps

“2A” and “2B” have been tripped locally.

17. Confirm Minimum Flow Bypass Valve, 2E11-F007B is open.

Step 4.0

At panel 2C82-P001 the operator VERIFIES that the MIN FLOW BYP VLV, 2E11-F007B, is OPEN, red light is illuminated.

** 18. Start RHR Pump 2E11-C002B.

Step 5.0

At panel 2C82-P001, RHR PUMP, 2E11-C002B, is RUNNING, red light illuminated.

PROMPT: WHEN the operator addresses starting RHR Pump “2D,” INFORM the

operator a System Operator has started it locally.

19. Confirm or open Heat Exchanger Shell Side Bypass Valve, 2E11-F048B.

Step 7.0

At panel 2C82-P001, BYPASS VLV, 2E11-F048B, is OPEN, red light illuminated.

20. Confirm or close the following valve:

2E11-F028B

Step 8.0

At panel 2C82-P001, the Operator has confirmed 2E11-F028B is CLOSED, green light illuminated.

PLANT 2 2016-301 Page 10 of 12

STEP #




2B31-F023B

Step 8.0



2E11-F011B

Step 8.0



2E11-F016B

Step 8.0



2E11-F006A

Step 8.0

At panel 2C82-P001, the Operator has confirmed 2E11-F006A is CLOSED, green light illuminated.


2E11-F006B

Step 8.0



2E11-F006C

Step 8.0

At panel 2C82-P001, the Operator has confirmed 2E11-F006C is CLOSED, green light illuminated.


2E11-F006D

Step 8.0

At panel 2C82-P001, the Operator has confirmed 2E11-F006D is CLOSED, green light illuminated.


2E11-F073B

Step 8.0


29. Confirms Reactor pressure is less than 500 psig

Step 9.0

Confirms Reactor pressure is less than 500 psig

PROMPT: IF addressed, INFORM the operator Reactor pressure is less than 500 psig.

PLANT 2 2016-301 Page 11 of 12

STEP #



** 30. Open the RHR Inboard Injection Valve, 2E11-F015B.

Step 9.0

At panel 2C82-P001, RHR INBD INJ VLV, 2E11-F015B, OPEN, red light illuminated.

31. Confirm or open the RHR Outboard Injection Valve, 2E11-F017B.

Step 9.0

At panel 2C82-P001 operator VERIFIES RHR OUTBD INJ VLV, 2E11-F017B, is OPEN, red light illuminated.

PROMPT: IF addressed, INFORM the operator RHR Pumps “2A” and “2C” have not

auto started. PROMPT: IF addressed (or after operator states injection has been achieved), as the

Shift Supervisor, INFORM the operator that another operator is responsible for raising and maintaining RWL.

END TIME: __________




− Operator states the task is complete.

TERMINATING CUE: We will stop here.

PLANT 2 2016-301 Page 12 of 12



JPM PLANT 2 2016-301: SUMMARY OF JPM QUANTITATIVE ATTRIBUTES CATEGORY Minimum NRC



At least 2

6

Step 7 PLACES switch 2C82-S80 to EMERG Allows control of 2E11-F007B.

Step 8 PLACES switch 2C82-S9 to EMERG Allows control of 2E11-C002B.

Step 9 PLACES switch 2C82-S13 to EMERG Allows control of 2E11-F017B, F015B & C001B.

Step 16 PLACES switch 2C82-S16to EMERG Allows control of 2B31-F023B.

Step 18 STARTS RHR pump 2B Required to initiate LPCI injection.

Step 30 OPENS 2E11-F015B Required to initiate LPCI injection.

Number of JPM Steps

<30

31

Time to Perform JPM

<45 min

13 min


Normal


Plant or Simulator




DRAFT PLANT 3 (ALL)

Title:

Transfer the Vital AC System from Alternate Power to the Inverter



Time: 15 Minutes


Date:


Date:


Date:

Course Number

N/A Program Name

OPERATIONS TRAINING Media Number

PLANT 3 2016-301


Sup’s Initials

01 07/01/92 General revision and format change WMM RSG

02 08/13/93 General revision, incorporate instructor comments, word processor change

RAB RSG

03 11/15/93 Correct standard in step 10, both units RAB SMC

04 07/05/96 Format change, modify time allowance, made steps 5, 7, and 9 critical

RAB SMC

05 09/14/98 Revised based on annual exam comments. SCB DHG

06 02/21/00 Format modification, change title to better reflect task, update K/As

RAB DHG

07 11/02/00 Include objective number, change operator applicability to System Operator

RAB DHG


09 03/08/05 Documentum Revision DNM RAB

10 06/13/05 Revised Initial License statement for successful completion RAB RAB

11 04/11/06 Removed Response Cues RAB RAB

12 09/25/09 Added HU tools ADY ALD

12.1 10/17/11 Reviewed JPM against current procedure. Added Fundamental question to new Attachment 1.

MMG ALS

12.2 01/22/14 Removed fundamental questions and added JPM attribute page. Added procedure step numbers to JPM staps.

JSC ELJ

13 08/06/15 Updated “Summary of JPM Attributes” and changed wording to match procedure including procedure step numbers.

MMG ALS

13.1

Reviewed JPM against current procedure to be used on ILT-10 NRC Exam. Changed Media Number to Plant 3 2016-301. Removed U1 information and once exam is complete will be added back and update the JPM Database.

ARB

PLANT 3 2016-301 Page 5 of 12


TASK TITLE: Transfer the Vital AC System from Alternate Power to the

Inverter


TASK STANDARD: The task shall be completed when Vital AC Power has been transferred from the alternate source to the inverter per 34SO-R25-002.




K/A CATALOG NUMBER: 2620012130 K/A CATALOG JTA IMPORTANCE RATING :

RO 3.90 SRO 3.40



34SO-R25-002-2, Ver. 5.3


34SO-R25-002-2, Ver. 5.3

APPROXIMATE COMPLETION TIME: 15 Minutes SIMULATOR SETUP: N/A.

UNIT 2


1. Vital AC room cooler is in operation.

2. The Vital AC rectifier and inverter have been started.

3. Pre-Job Brief is NOT required.

INITIATING CUES:

Transfer Unit 2 Vital AC from the Alternate Source to the Vital AC Inverter per 34SO-R25-002-2.

PLANT 3 2016-301 Page 7 of 12

STEP #




IF THEN

PASS


correctly





1. Confirm SYNC-MONITOR light is ILLUMINATED. (Step 7.1.4.1)

At the Vital AC Panel, location 112TET12, the Operator CONFIRMS the SYNC - MONITOR clear light is illuminated.

2. Confirm in OR place RETURN MODE Switch in AUTO position. (Step 7.1.4.2)

At Vital AC Panel, 112TET12, the Operator CONFIRMS the RETURN MODE switch is in AUTO.

** 3. Place MANUAL BYPASS switch in BYP-SYNC position. (Step 7.1.4.3)

At Vital AC Panel, 112TET12, the Operator PLACES the MANUAL BYPASS switch in BYP-SYNC position.

4. Confirm SYNC-MONITOR light is EXTINGUISHED. (Step 7.1.4.4)

At Vital AC Panel, 112TET12, the Operator CONFIRMS the clear SYNC-MONITOR light is extinguished.

PLANT 3 2016-301 Page 8 of 12

STEP #



** 5. Place MANUAL BYPASS switch in BYP-TEST position. (Step 7.1.4.5)

At Vital AC Panel, 112TET12, the Operator PLACES the MANUAL BYPASS switch in BYP-TEST.

PROMPT: IF the operator asked for the status of the red INVERTER static switch position light, INFORM the operator that light is ILLUMINATED (Step 7.1.4.6).

6. Confirm the following:

red INVERTER static switch position light is ILLUMINATED; AND

*************** *************** (Step 7.1.4.7)

At Vital AC Panel, 112TET12, the Operator has IDENTIFIED the following indications:

INVERTER red light illuminated. ***************** *****************


*******************

the white ALT LINE static switch position light is EXTINGUISHED;

******************* (Step 7.1.4.7)


**********************. ALT LINE white light extinguished.

**********************


***********************

***********************

SYNC MONITOR light is EXTINGUISHED. (Step 7.1.4.7)


*********************. *********************. SYNC MONITOR clear light extinguished.

NOTE: Step 7.1.4.8 is not applicable.

PLANT 3 2016-301 Page 9 of 12

STEP #



** 9. Place TEST toggle switch in ALT LINE position. (Step 7.1.4.9)

At Vital AC Panel, 112TET12, the Operator PLACES the TEST toggle switch in ALT LINE (down) position.


white ALT LINE static switch position light is ILLUMINATED; AND

**********************

********************** (Step 7.1.4.10)


ALT LINE white light illuminated.

*****************

*****************


**********************

red INVERTER static switch position light is EXTINGUISHED; AND

*********************** (Step 7.1.4.10)


********************

INVERTER red light extinguished.

*********************


***********************

***********************

SYNC MONITOR light is EXTINGUISHED (Step 7.1.4.10)


*******************

*******************

SYNC MONITOR clear light extinguished.

** 13. Place MANUAL BYPASS switch in NORMAL position. (Step 7.1.4.11)

At Vital AC Panel, 112TET12, the Operator PLACES the MANUAL BYPASS Switch in NORMAL.

** 14. Place TEST toggle switch in CENTER position. (Step 7.1.4.12)

At Vital AC Panel, 112TET12, the Operator PLACES the TEST toggle switch in center position.

PLANT 3 2016-301 Page 10 of 12

STEP #



PROMPT: IF the operator asked for the status of the red INVERTER static switch position light, INFORM the operator that light is ILLUMINATED (Step 7.1.4.13).


red INVERTER static switch position light is ILLUMINATED; AND

*******************

******************* (Step 7.1.4.14)


INVERTER red light illuminated.

****************

****************


*********************

white ALT LINE static switch position light is EXTINGUISHED; AND

********************** (Step 7.1.4.14)

At Vital AC Panel, 112TET12, the Operator has IDENTIFIED the following indications: *********************

ALT LINE White light extinguished.

*********************


**********************

**********************

SYNC MONITOR light is EXTINGUISHED. (Step 7.1.4.14)


*********************

*********************

SYNC MONITOR clear light extinguished.

NOTE: Step 7.1.4.15 is not applicable.

PROMPT: WHEN addressed by the operator, as the Shift Supervisor, INFORM the

operator that it is desired to place the Return Mode Switch to Manual.

18. IF desired, place RETURN MODE SWITCH in MANUAL position. (Step 7.1.4.16)

At Vital AC Panel, 112TET12, the Operator PLACES the RETURN MODE switch in MAN.

End Time: ___________

PLANT 3 2016-301 Page 11 of 12

STEP #







of 12


JPM JP02731-13: SUMMARY OF JPM QUANTITATIVE ATTRIBUTES CATEGORY Minimum NRC



At least 2

5

Step 3 BYP-SYNC position Energizes synch circuit to compare inverter output voltage to bypass voltage (checks for out of phase between sources).

Step 5 BYP-TEST position Aligns Alternate power to Static Switch.

Step 9 Toggle Switch to Alt Line Allows Static Switch to supply Alternate power to bus.

Step 13 BYP-Switch to Normal Opens B1 contact so Alternate Power and Inverter power are controlled by the Static Switch (needed for Static Switch to have automatic control).

Step 14 Toggle Switch to Center Places Static Switch in control with power being supplied by the Inverter as long as sufficient voltage is present.

Number of JPM Steps

<30

18

Time to Perform JPM

<45 min

15 min


Normal Operator is told to transfer power from Alternate to the Inverter.


Plant


DRAFT


Operations Training JPM

ADMIN 1 – (ALL ) Title:

SBLC Operable

Author: ANTHONY BALL

Media Number: 2016-301 ADMIN 1

Time: 15.0 Minutes


Date:


Date:


Date

Course Number N/A


Media Number 2016-301 ADMIN 1


Sup’s Initials

0.0 Initial Development. RLS

0.1

Minor revision to match TS and use on ILT-10 NRC Exam. Reviewed JPM against current TS. Changed “Media Number” to 2016-301 ADMIN 1. Modified sodium pentaborate volume, temperature and concentration to plot in region on both figures 3.1.7-1 and 3.1.7-2. This JPM will become a new LR-JPM after NRC Exam.

ARB/ABG

UNIT 1 (X) UNIT 2 ( )

TASK TITLE: SBLC Operable

JPM NUMBER: 2016-301 ADMIN 1

TASK STANDARD: This task shall be successfully completed when the operator has

successfully completed the critical steps and determined that SBLC is outside of Region “A” but inside Region “B” and entry into TS 3.1.7 is required.




RO 3.9 SRO 4.2

K/A CATALOG NUMBER: G2.1.25 K/A CATALOG JTA IMPORTANCE RATING :

RO 3.9 SRO 4.2



Unit 1 TS (current amendment)


Unit 1 TS (current amendment)

APPROXIMATE COMPLETION TIME: 15.0 Minutes SIMULATOR SETUP: N/A

UNIT 1


1. Chemistry has just added Sodium Pentaborate to the Unit 1 SBLC storage tank.

2. Tank level is currently at 3400 gallons and solution temperature is at 70 degrees F.

3. Chemistry has just reported that tank concentration level is 6.5%.

INITIATING CUES:

Determine the status of SBLC for Unit 1.

2016-301 ADMIN 1 Page 7 of 8

STEP #



IF THEN

PASS


correctly





1. Operator enters Unit 1 Tech Specs section 3.1.7.

Operator enters unit 1 Tech Specs section 3.1.7.

2. Plots location on the Unit 1 SBLC Tech Specs Figure 3.1.7-1.

Using Unit 1 Tech Specs the operator plots location on SBLC Figures 3.1.7-1.

** 3. Determines that Sodium Pentaborate is outside the permissible region A of Figure 3.1.7-1 but inside Region B (original lincensing basis limit) IAW U1 Tech Specs.

Using Unit 1 Tech Specs Figure 3.1.7-1 the operator determines that Sodium Pentaborate is outside the permissible region A but inside Region B.

4. Plots location on the Unit 1 SBLC Tech Specs the Figure 3.1.7-2.

Using Unit 1 Tech Specs the operator plots location on SBLC Figures 3.1.7-2.

2016-301 ADMIN 1 Page 8 of 8

STEP #


** 5. Determines that Sodium Pentaborate is outside the permissible region A of Figure 3.1.7-2 but inside Region B (original lincensing basis limit) IAW U1 Tech Specs.

Using Unit 1 Tech Spec Figure 3.1.7-2 the operator determines that Sodium Pentaborate is outside the permissible region A but inside Region B.

** 6. Determine a LCO entry condition into Unit 1 Tech Specs Section 3.1.7.

Using Unit 1 Tech Specs the operator determines that a LCO exists for Section 3.1.7.

NOTE: The operator may identify applicable Unit 1 Tech Spec action statement,

but this is not critical or required.

END

TIME: __________

NOTE: The terminating cue shall be given to the Operator when:


− With NO reasonable progress, the Operator exceeds double the allotted time.

− Operator states the task is complete. TERMINATING CUE: We will stop here. EVALUATOR – PICK UP the Initiating Cue sheet AND ATTACHMENT 1& 2 .

DRAFT



ADMIN 2 – (RO ONLY ) Title:

USE A SYSTEM LOGIC DIAGRAM

Author: A. B. GENEREUX


Time: 20.0 Minutes


Date:


Date:


Date

Course Number N/A




Sup’s Initials

0.0 Initial Development. ABG


TASK TITLE: USE A SYSTEM LOGIC DIAGRAM


TASK STANDARD: The task shall be complete when the operator has determined the

failure condition of a relay using Plant Hatch logic drawings.

TASK NUMBER:




RO 3.5 SRO 3.9



H-27664, H-27671, H-27667, H-27665, H-51689


H-27664, H-27671, H-27667, H-27665, H-51689

APPROXIMATE COMPLETION TIME: 20.0 Minutes SETUP: This JPM may be performed in the Simulator Library, Control Room, Simulator, or

Document Control.

UNIT 2


1. The Unit 2 is at operating at 100% rated power.

2. HPCI is aligned per 34SV-E41-002-2, HPCI Pump Operability.

3. Relay 2E41-K20 fails in the ENERGIZED condition. INITIATING CUES:

Using applicable plant logic drawings, DETERMINE the HPCI System response to a failure of relay 2E41-K20 in the ENERGIZED condition.

2016-301 ADMIN 2 Page 7 of 8

STEP #



IF THEN

PASS


correctly





Provide the Applicant with drawings H-27664, H-27671, H-27667, H-27665 and H-51689

1. Operator locates relay 2E41-K20. Using the relay tabulation table on H-27664 the operator locates 2E41-K20 on drawing H-27667, coordinate F-10.

** 2. Operator determines that 2E41-F004, HPCI CST Suction valve, DOES NOT receive a close signal.

Using the relay tabulation table on H-27664 the operator locates 2E41-F004 on drawing H-27671, coordinates E-1 and E-2. The operator DETERMINES that 2E41-F004 DOES NOT receive a close signal (remains OPEN).

2016-301 ADMIN 2 Page 8 of 8

STEP #


** 3. Operator determines that 2E41-F011, HPCI Test to CST valve DOES receive a close signal.

Using the relay tabulation table on H-27664 the operator locates 2E41-F011 on drawing H-27671, coordinate J-2, the operator DETERMINES that 2E41-F011 DOES receive a close signal. (travels CLOSE)

** 4. Operator determines that 2E41-F008, HPCI Test to CST valve DOES receive a close signal.

Using the relay tabulation table on H-27664 the operator locates 2E41-F008 on drawing H-51689, coordinate D-6, the operator DETERMINES that 2E41-F008 DOES receive a close signal. (travels CLOSE)

END TIME: __________



− With NO reasonable progress, the Operator exceeds double the allotted time.


EVALUATOR – PICK UP the Initiating Cue sheet AND DRAWINGS H-27664, H-27671, H-27667, H-27665 and H-51689.

DRAFT



ADMIN 3 – SRO ONLY Title:

VERIFY FUEL MOVEMENT SHEET



Time: 30.0 Minutes


Date:


Date:


Date

SOUTHERN NUCLEAR OPERATING COMPANY PLANT E. I. HATCH Page 3 of 9 FORM TITLE: TRAINING MATERIAL REVISION SHEET Program/Course Code: OPERATIONS TRAINING Media Number: 2016-301 ADMIN 3

Rev. No.

Date Reason for Revision Author’s Initials

Supv’s Initials

00 03/17/09 Initial development FNF CME

01 06/07/12 Revision for use on 2012-301 NRC Exam. After NRC Exam, Media

Number will be changed to LR-JP-45.33A. ARB CME

1.1

Minor revision for use on 2016-301 NRC Exam. Reviewed

JPM against current procedure. Changed “Media Number”

to 2016-301 ADMIN 3. After NRC Exam, Media Number

will be changed to LR-JP-45.33A.

ARB

2016-301 ADMIN 3 Page 4 of 9


TASK TITLE: VERIFY FUEL MOVEMENT SHEET


TASK STANDARD: The task will be complete when the operator has identified all in-core placement errors of all components listed on Page 1 of the attached Fuel Movement Sheet.



PLANT HATCH JTA IMPORTANCE RATING: RO Not Available SRO Not Available


RO 2.2 SRO 3.9

OPERATOR APPLICABILITY: Senior Reactor Operator

GENERAL REFERENCES: Refuel Floor

34FH-OPS-001-0 (current version) 42FH-ERP-014-0 (current version)

REQUIRED MATERIALS: Refuel Floor

Fuel Movement Sheets 34FH-OPS-001-0 (current version) 42FH-ERP-014-0 (current version)


EVALUATOR COPY

UNIT 2


1. Unit 2 is in a refueling outage.

2. You are the oncoming Refuel Floor SRO.

3. The off-going Refuel Floor SRO asks you to verify the attached Fuel Movement Sheet.

4. The fuel movement sheet, Core Map and pictures of the core cells are available.

INITIATING CUES:

Verify the in-core placement of all components listed on the attached Fuel Movement Sheet.

2016-301 ADMIN 3 Page 6 of 9

STEP #




IF THEN

PASS


correctly



(1) The standard for human performance tools, safety, PPE, and other pertinent expectations is considered met provided any deviations are minor and have little or no actual or potential consequence. Errors may be self-corrected provided the action would not have resulted in significant actual or potential consequences. Reference: NMP-TR-111, “On-The-Job Training and Task Performance Evaluation”.

START TIME: __________

PROMPT: Hand the operator the fuel movement sheet, Core Map and pictures of the

core cells. PROMPT: IF the operator has problems reading the bundle serial numbers, THEN

provide the serial numbers to the operator. PROMPT: IF the operator asks about Spent Fuel Pool verification, THEN tell the

operator he is only responsible for in-core verifications.

1. Obtains the correct procedures. Obtains and reviews 42FH-ERP-014-0, “Fuel Movement” and 34FH-OPS-001-0, “Fuel Movement Operation.”

2016-301 ADMIN 3 Page 7 of 9

STEP #



2. Verify correct loading of Control Cell 38-37.

Operator determines Control Cell 38-37 bundles are the correct bundles and in the correct orientation:

• JLV675

• JLV682

• JLK804

• JLK817

** 3. Determine INCORRECT loading of Control Cell 14-37.

Operator determines Control Cell 14-37 bundles are the correct bundles but in the INCORRECT orientation:

• JLV678

• JLV670

NOTE: Bundles JLV678 and JLV670 are 180º out.



• JLK805

• JLK812


Operator determines Control Cell 38-17 bundle is the INCORRECT bundle but in the correct orientation:

• JLV701

NOTE: Wrong bundle loaded. The correct bundle is JLV698

2016-301 ADMIN 3 Page 8 of 9

STEP #





• JLV708

• JLK807

• JLK820


Operator determines Control Cell 14-17 Double Blade Guide is in the INCORRECT orientation.

END TIME: __________



− With no reasonable progress, the Operator exceeds double the allotted time.

− Operator states the task is complete. TERMINATING CUE: We will stop here. EVALUATOR – PICK UP the Initiating Cue sheet.

UNIT 2


1. Unit 2 is in a refueling outage.

2. You are the oncoming Refuel Floor SRO.

3. The off-going Refuel Floor SRO asks you to verify the attached Fuel Movement Sheet.

4. The fuel movement sheet, Core Map and pictures of the core cells are available.

INITIATING CUES:

Verify the in-core placement of all components listed on the attached Fuel Movement Sheet.

DRAFT



ADMIN 4 (RO ONLY ) Title:

CONDUCT OF OPERATIONS, 34SV-SUV-019-1 SURVEILLANCE



Time: 15.0 Minutes


Date:


Date:


Date

SOUTHERN NUCLEAR OPERATING COMPANY PLANT E. I. HATCH Page 3 of 12 FORM TITLE: TRAINING MATERIAL REVISION SHEET Program/Course Code: OPERATIONS TRAINING Media Number: 2016-301 ADMIN-4

Rev. No. Date Reason for Revision Author’s Initials

Supv’s Initials

1 5/26/11 Made editorial changes from LR-JP-10022 to use on 2011-301 NRC Exam. Changed K/A to G2.1.7. ELJ CME

1.1

Minor revision to match procedure and use on ILT-10 NRC Exam. Reviewed JPM against current procedure. Changed “Media Number” to 2016-301 ADMIN 4. This JPM will become LR-JP-10022 after NRC Exam.

ARB

2016-301 ADMIN 4 Page 4 of 12

UNIT 1 (X) UNIT 2 ()

TASK TITLE: CONDUCT OF OPERATIONS, 34SV-SUV-019-1

SURVEILLANCE


TASK STANDARD: This task will be satisfactorily met when the student has completeted section 7.5 of 34SV-SUV-019-1, SURVEILLANCE CHECKS, and informed the evaluator that Unit 1 drywell cooling system should be placed in “Additional Cooling Operating Mode.”

TASK NUMBER:

OBJECTIVE NUMBER:

JTA IMPORTANCE RATING: K/A CATALOG NUMBER: G2.1.7

RO 4.40 SRO 4.70

OPERATOR APPLICABILITY: Reactor Operator (RO)


34SV-SUV-019-1


34SV-SUV-019-1. Complete previous shift data in step 7.5.4 (149 and 148) Calculators

APPROXIMATE COMPLETION TIME: 15.0 Minutes SIMULATOR SETUP: NOT applicable

UNIT 1

READ TO THE CANDIDATE INITIAL CONDITIONS:

1. Unit 1 is operating at 100 % power.

2. 1T47-R611 is out of service.


4. The SS has directed this surveillance be completed as a paper version.

5. 1T47-R611 PT 14 and 1T47-R612 PT 10 Previous readings:

• 1T47-R611 PT 14 (1T47-N009) was 149ºF • 1T47-R612 PT 10 (1T47-N003) was 148ºF

6. SPDS is available INITIATING CUES:

Complete section 7.5 of 34SV-SUV-019-1, SURVEILLANCE CHECKS, which evaluates drywell temperatures, AND Inform the evaluator of any actions that need to be taken as a result of the readings or results obtained from this surveillance.

2016-301 ADMIN 4 Page 6 of 12

STEP #


PROMPT: GIVE the operator an entire copy of 34SV-SUV-019-1.


NOTE: When the candidate addresses the need for SPDS readings provide

Attachment 1.

1. Determine method for obtaining temperature readings.

Per NOTE “S” of 34SV-SUV-019-1, the candidate determines temperature readings can be obtained from SPDS.

2. Performs step 7.5.1 of 34SV-SUV-019-1.

From the SPDS screen shot, the candidate list the temperature readings on the surveillance sheet with no errors for;

1T47-N001L 120

N004 109

N008, 114

N001M, 114

N005 114


The candidate evaluates the temperatures from step 7.5.1 and determines the maximum temperatute minus the minimum temperature is less than 40ºF.


The candidate evaluates the readings in step 7.5.1. and concludes the highest is less than 275ºF and the lowest temperature is greater than 50ºF.

2016-301 ADMIN 4 Page 7 of 12

STEP #


NOTE: When addressing the temperature readings from the previous reading reply with:

• 1T47-R611 PT 14 (1T47-N009) was 149ºF • 1T47-R612 PT 10 (1T47-N003) was 148ºF.


From the SPDS screen shot, the candidate list the temperature readings on the surveillance sheet for 1T47-N009 176

and 1T47-N003 175

and list the temperatures from the previous readings provided by the evaluator.

** 6. Performs step 7.5.5 of 34SV-SUV-019-1.

The candidate compares the current temperature readings in step 7.5.4 to those from the previous reading and concludes the temperatures DO differ by more than 10ºF AND that a CR must be written


The candidate confirms the maximum reading in step 7.5.4 is less than 275ºF and the minimum is greater than 50ºF AND the maximum temperature minus the minimum temperature of step 7.5.4 is less than 50ºF.

2016-301 ADMIN 4 Page 8 of 12

STEP #



From the SPDS screen shot, the candidate list the temperature readings on the surveillance sheet with no errors for;

1T47-N001J, 199

N001K, 164

N002, 157

N001A, 182

N001B, 187

N010. 154


The candidate determines the maximum temperature from step 7.5.7 minus the lowest temperature from step 7.5.7 is less than 100ºF.


The candidate confirms the maximum reading in step 7.5.7 is less than 275ºF and the minimum is greater than 50ºF.


Using the formula at the bottom of the surveillance page, the candidate calculates the average drywell temperature to be 136.7ºF. (Accept ±1°F due to rounding errors)

2016-301 ADMIN 4 Page 9 of 12

STEP #


** 12. Addresses any additional actions that are required as a result of the average drywell temperature reading.

The candidate informs the evaluator that since average drywell temperature exceeds 135ºF, per note “L” of the surveillance the shift is to place the Drywell Cooling System in “Additional Cooling Operating mode” per 34SO-T47-001-1.

END TIME:__________



− Operator states the task is complete. TERMINATING CUE: That completes this JPM.

Answer Key

Page 10 of 12

7.5 PANEL - INSTRUMENT / TECH SPEC. NOTE REAC MODE

FREQ T/S OR OPER

LIM N D

7.5.1

1H11-P657: - 1T47-R611, Pt 9 (1T47-N001L)

R,S,C (spec only)

1,2,3 c NA

120

Pt 11 (1T47-N004) 109

Pt 13 (1T47-N008) 114

1H11-P654: - 1T47-R612, Pt 9 (1T47-N001M) 114

Pt 11 (1T47-N005) 114

7.5.2 Confirm max minus min in 7.5.1 < 40°F B 1,2,3 c SAT

7.5.3 Confirm max < 275°F and min > 50°F in 7.5.1 B,C 1,2,3 c NA

SAT

7.5.4

1H11-P657: - 1T47-R611, PT 14 (1T47-N009)

R,S 1,2,3 c NA

Previous shift reading 149

Present reading 176

1H11-P654: - 1T47-R612, PT 10 (1T47-N003) Previous shift reading 148

Present reading 175

7.5.5

IF the previous reading differs from the present reading by greater than 10°F OR IF erratic instrument behavior is observed THEN submit a CR to evaluate the Operability of the instruments in step 7.5.4.

(SR 3.3.3.1.1 for 3.3.3.1-1(10))

N/A 1,2,3 c

YES IF CR submitted. YES

NR IF NOT required

7.5.6 Confirm max < 275°F and min > 50°F AND max minus min < 50°F in 7.5.4

(SR 3.3.3.1.1 for 3.3.3.1-1(10)) B 1,2,3 c NA

SAT

7.5.7

1H11-P657: - 1T47-R611, Pt 7, (1T47-N001J)

R,S 1,2,3 c NA

199

Pt 8, (1T47-N001K) 164

Pt 10, (1T47-N002) 157

1H11-P654: - 1T47-R612, Pt 7, (1T47-N001A) 182

Pt 8, (1T47-N001B) 187

Pt 13, (1T47-N010) 154

7.5.8 Confirm max minus min in 7.5.7 < 100°F

(SR 3.3.3.1.1 for 3.3.3.1.1(10)) B 1,2,3 c

SAT

7.5.9 Confirm max < 275°F and min > 50°F in 7.5.7 B,C 1,2,3 c NA

SAT

7.5.10 Average Drywell Temperature

(SR 3.6.1.5.1) L 1,2,3 c < 150°F 136.7

INITIALS

Calculations verified DATE TIME

DW Temp = (7.5.1 TE's ) (0.63) + (7.5.4 TE’s) (0.22) + (7.5.7 TE's ) (0.15) 5 2 6

1

of 12

UNIT 1

READ TO THE CANDIDATE INITIAL CONDITIONS:

1. Unit 1 is operating at 100 % power.



4. The SS has directed this surveillance be completed as a paper version.

5. 1T47-R611 PT 14 and 1T47-R612 PT 10 Previous readings:

• 1T47-R611 PT 14 (1T47-N009) was 149ºF • 1T47-R612 PT 10 (1T47-N003) was 148ºF

INITIATING CUES:

Complete section 7.5 of 34SV-SUV-019-1, SURVEILLANCE CHECKS, which evaluates drywell temperatures, AND Inform the evaluator of any actions (if any) that need to be taken as a result of the readings or results obtained from this surveillance.

of 12

Attachment 1

DRAFT



ADMIN 5 (ALL) Title:

EVALUATE AN RWP AND SURVEY MAP



Time: 10.0 Minutes


Date:


Date:


Date

SOUTHERN NUCLEAR OPERATING COMPANY PLANT E. I. HATCH Page 3 of 8 FORM TITLE: TRAINING MATERIAL REVISION SHEET Program/Course Code: OPERATIONS TRAINING Media Number: 2016-301 ADMIN 5

Rev. No. Date Reason for Revision Author’s Initials

Supv’s Initials

00 08/16/13 Modified 2009-302 NRC Exam & renumbered for

ILT-8 NRC Exam. After exam will be renumbered

and placed into the LOCT & ILT bank.

ARB CME

0.1 Modified 2013-301 NRC Exam to a different area

with different radiation values & renumbered for ILT-

10 NRC Exam. After exam will be renumbered and

placed into the LOCT & ILT bank as a New JPM.

ARB

2016-301 ADMIN 5 Page 4 of 8

UNIT 1 ( ) UNIT 2 (x)

TASK TITLE: Comply with radiation work permit requirements during normal

or abnormal conditions.


TASK STANDARD: The task shall be completed when the operator has determined: the correct survey map, Maximum stay time before dosimetry alarm occurs, and the actions if an alarm occurs on dose accumulated.

TASK NUMBER: N/A

OBJECTIVE NUMBER: N/A

TYPE N/A

PLANT HATCH JTA IMPORTANCE RATING: RO N/A SRO N/A


RO 3.5 SRO 3.6



RWP 16-0004 for Operations RP surveys N.E. Diagonal (U1 & U2) RP surveys S.E. Diagonal (U1 & U2) 60AC-HPX-004-0, Radiation & Contamination Control 60AC-HPX-002, Personnel Dosimetry 34SO-E11-010-2, Attachment 3


RWP 16-0004 for Operations RP surveys N.E. Diagonal (U1 & U2) RP surveys S.E. Diagonal (U1 & U2) 60AC-HPX-004-0, Radiation & Contamination Control 60AC-HPX-002, Personnel Dosimetry 34SO-E11-010-2, Attachment 3


UNIT 2

READ AND GIVE A COPY TO THE OPERATOR INITIAL CONDITIONS:

1 Unit 2 is at 100% power with NO significant problems.

2. 34SV-E11-001-2, Residual Heat Removal Pump Operability, is to be performed this shift on “B” Loop RHR pumps.

3. The current OPS RWP is 16-0004.

4. The RWP and RP Survey Maps are available. INITIATING CUES:

You are assigned to locally perform the pre-start checks for 34SV-E11-001-2, Residual Heat Removal Pump Operability, “B” loop RHR and are to:

• Determine the correct survey map for “B” loop RHR.

• Calculate the maximum stay time before the DAD alarms on dose

accumulated, assuming you: • remain near the RHR pumps and • are at the highest current General Area Dose Rate

• State your required actions if the DAD alarms on dose

accumulated.

2016-301 ADMIN 5 Page 6 of 8

STEP #


For Initial Operator Programs:

For OJT/OJE ; All procedure steps must be completed for Satisfactory Performance. For License Examinations; ALL CRITICAL STEPS must be completed for Satisfactory Performance.


START TIME: _________

PROMPT: AT this time, give the operator the RWP and RP Survey Maps.

** 1. Determine the appropriate survey map.

Operator determines that survey map 133757; U2 S.E. Diag. 87 is the correct map.

NOTE: If operator selects the incorrect map, Critical Step #2 will be INCORRECTLY calculated.

** 2. Determine the maximum stay time before the DAD alarms on dose accumulated.

Operator determines the DAD will alarm on dose accumulated in 30 minutes.

NOTE: Per RWP, DAD set at 25 mr for rounds. Per Survey Map, the max general area dose rate is 50 mr/hr.

IF THEN

PASS


correctly



2016-301 ADMIN 5 Page 7 of 8

STEP #


** 3. Determine required actions if the DAD alarms on dose accumulated.

Operator determines that immediate exit and RP notification is required.

END TIME:__________

NOTE: The terminating cue shall be given to the Applicant when:


− With NO reasonable progress, the Applicant exceeds double the allotted time.

− Applicant states the task is complete. TERMINATING CUE: That completes this JPM. .

DRAFT



ADMIN 6 SRO ONLY Title:

Emergency Classification - Complete NMP-EP-110 Checklist 1



Time Critical: 15.0 Minutes


Date:


Date:


Date

of 15

Course Number N/A




Sup’s Initials

0.0 04/30/07 Initial development CME RAB

1.0 11/04/10 Added Task and HU Pass/Fail criteria. Updated Southern Company logo. ELJ CME

2.0 01/12/11 Updated references to NMP-EP-110 and NMP-EP-111 MCK DNM

3.0 06/07/12 General revision for use on NRC Exam 2012-301. After NRC Exam, Media Number will be changed to LR-JP-25061-03.

ARB CME

3.1 03/07/16

Minor revision to match procedure and use on ILT-10 NRC Exam. Reviewed JPM against current procedure. Changed “Media Number” to 2016-301 ADMIN 6. This JPM will become LR-JP-25061 after NRC Exam.

ARB

2016-301 ADMIN 6 Page 5 of 15


TASK TITLE: Emergency Classification - Complete NMP-EP-110

Checklist 1


TASK STANDARD: This task shall be successfully completed when the event has been classified and NMP-EP-110 Checklist 1, step 1 through step 6, has been correctly completed.




K/A CATALOG NUMBER: Generic 2.4.41 K/A CATALOG JTA IMPORTANCE RATING :

RO 2.30 SRO 4.1

OPERATOR APPLICABILITY: Senior Reactor Operator

GENERAL REFERENCES: Unit 1 & 2

NMP-EP-110, Ver 8.1 NMP-EP-111, Ver 11.0

REQUIRED MATERIALS: Unit 1 & 2

NMP-EP-110, Ver 8.1 NMP-EP-111, Ver 11.0

APPROXIMATE COMPLETION TIME: 15.0 Minutes SIMULATOR SETUP: NA

UNIT 1 & 2

READ TO THE OPERATOR

INITIAL CONDITIONS:

1. You are the On Shift Shift Manager.

2. Unit 1 and Unit 2 have experienced a Loss of Offsite Power (LOSP).

3. All rods fully inserted on the scram signal.

4. For the last 16 minutes, only EDG 2A has been running and is supplying power to 4160 VAC 2E.

5. EDGs 1B and 2C start attempts have not been successful.

6. All other Unit 2 parameters are normal.

7. The following Unit 1 conditions exist:

• All rods fully inserted on the scram signal

• 4160 VAC 1E, 1F and 1G have been de-energized for 16 min.

• EDG 1A and 1C start attempts have not been successful

• All other Unit 1 parameters are normal INITIATING CUES:

Classify the Event by Completing NMP-EP-110 Checklist 1,

Steps 1 through 5.

Raise your hand when you have completed Checklist 1, Step 5.

This JPM is TIME CRITICAL .

Current time is: ___________________

2016-301 ADMIN 6 Page 7 of 15

STEP #




IF THEN

PASS


correctly




NOTE: The CLASSIFICATION must be made within 15 minutes of the initial prompt and the Student states they understand the initial conditions.

NOTE: The Student is expected to obtain a copy of Checklist 1 if the Initiating

Cue is given in the Simulator or Control Room.

1st START TIME:_________


The operator has OBTAINED Check List 1, which is contained in NMP-EP-110.

2. Determine the appropriate Initiating Condition Matrix for classification of the event based on the current operating mode:

HOT IC/EAL Matrix Eval Chart

COLD IC/EAL Matrix Eval Chart

Both HOT & COLD IC/EAL Matrix

(Checklist 1, Step 1)

On Checklist 1, Step 1,

The operator has selected HOT IC/EAL Matrix Evaluation Chart

2016-301 ADMIN 6 Page 8 of 15

STEP #



3. Evaluate the status of the fission product barrier using Figure 1, Fission Product Barrier Evaluation.

Select the condition of each fission product barrier:

Fuel Cladding Integrity

*******************

*******************


On Checklist 1, Step 2.a,

The operator has selected INTACT for Fuel Cladding Integrity.



*******************

Reactor Coolant System

*******************



The operator has selected INTACT for Reactor Cooling System.



*******************

*******************

Containment Integrity



The operator has selected INTACT for Containment Integrity.

6. Determine the highest applicable fission product barrier Initiating Condition (IC). (Checklist 1, Step 2.b.)

On Checklist 1, Step 2.b,

The operator has selected NONE

2016-301 ADMIN 6 Page 9 of 15

STEP #



7. Evaluate AND determine the highest applicable IC/EAL using the Matrix Evaluation Chart(s) identified in step 1 THEN Go To step 4. (Checklist 1, Step 3)

On Checklist 1, Step 3.

The operator has identified SS1

8. Check the highest emergency classification level identified from either step 2b or 3:

Classification *********** (Checklist 1, Step 4)


The operator has selected

Site-Area as the Classification.

** 9. Check the highest emergency classification level identified from either step 2b or 3:

********** Based on IC # (Checklist 1, Step 4)


The operator has selected SS1 for the Based on IC#.

NOTE: It is expected that the IC# be filled in on Checklist 1 (in the above step).

Credit for this step will be given if the proper IC# is announced during the Crew Update announcing the classification to the crew.

10. Remarks (Identify the specific EAL, as needed). (Checklist 1, Step 4)


The operator has written Loss of both SATs and failure of EDGs to power Emerg Buses and at least 1 Emerg Bus not restored within 15 minutes in the space provided.

NOTE: If follow-up questioning reveals that a classification was declared and

based on another IC #, the classification should be evaluated for validity.

2016-301 ADMIN 6 Page 10 of 15

STEP #



11. Declare the event by approving the Emergency Classification. (Checklist 1, Step 5)


The operator has signed their name as the Emergency Director in the space provided.

12. Fill in the Date in the space provided. (Checklist 1, Step 5)


The operator has entered the current Date in the space provided.

** 13. Fill in the Time in the space provided. (Checklist 1, Step 5)


The operator has entered the current Time in the space provided. Time Critical Stop Time: .

NOTE: For this step to be completed considered SAT, the time entered must be within 15 minutes of the time recorded on the Initial Conditions sheet provided to the operator.

1st

END TIME: __________

UNIT 1 & 2

READ TO THE OPERATOR

INITIAL CONDITIONS:

1. You are the On Shift Shift Manager.

2. You have just completed NMP-EP-110, Checklist 1 Steps 1 through 5.

INITIATING CUES:

Complete NMP-EP-110 Checklist 1 Step 6.

Raise your hand when you have completed Checklist 1, Step 6.

This JPM is TIME CRITICAL .

Current time is: ___________________

2016-301 ADMIN 6 Page 12 of 15

STEP #


** Indicates critical step)

PROMPT: WHEN the operator enquires about meteorological conditions, GIVE the

operator the MIDAS Information Sheet if not given earlier when performing a Group JPM.

14. Obtain Meteorological Data (not required prior to event declaration). (Checklist 1, Step 6)

The operator has obtained Meteorological Data (i.e. MIDAS Information Sheet).

** 15. Record the following:

Wind Direction (from)

************

***********

********* (Checklist 1, Step 6)


The operator has entered 130 in the space provided for Wind Direction (from).

** 16. Record the following:

*******************

Wind Speed

***********

***********



The operator has entered 5 in the space provided for Wind Speed.

17. Record the following:

*****************

**********

Stability Class

**********



The operator has entered D in the space provided for Stability Class.

18. Record the following:

****************

***********

***********

Precipitation



The operator has entered 0 in the space provided for Precipitation.

2016-301 ADMIN 6 Page 13 of 15

STEP #



NOTE: The Operator may perform a “Crew Update” and announce what the

Classification is and the IC# the classification is based on.

PROMPT: If the operator addresses performance of Checklist 1 Step 7 Initiate Checklist 2, Emergency Plan Initiation, INFORM the operator that another operator will Initiate Checklist 2.

2nd

END TIME: __________




2016-301 ADMIN 6 Page 14 of 15


*** INSTRUCTOR ANSWER KEY ***

Checklist 1 - Classification Determination

Initial Actions Completed by 1. Determine the appropriate Initiating Condition Matrix for classification of the event based on

the current operating mode:

� HOT IC/EAL Matrix Evaluation Chart (Go To Step 2) to evaluate the Barriers)

� COLD IC/EAL Matrix Evaluation Chart (Go To Step 3)

� Both HOT & COLD IC/EAL Matrix Evaluation Chart apply (Go To Step 2)

Student


a. Select the condition of each fission product barrier:

LOSS POTENTIAL LOSS INTACT

Fuel Cladding Integrity � � �

Reactor Coolant System � � �

Containment Integrity � � �

Student

b. Determine the highest applicable fission product barrier Initiating Condition (IC): Student

(select one) � FG1 � FS1 � FA1 � FU1 �None

3. Evaluate AND determine the highest applicable IC/EAL using the Matrix Evaluation Chart(s) identified in step 1 THEN Go To step 4.

Hot IC# SS1 Unit 1 and/or Cold IC# __________ Unit___ or � None

Student

4. Check the highest emergency classification level identified from either step 2b or 3:

Classification Based on IC# Classification Based on IC# � General � Alert

� Site-Area SS1 � NOUE

� None N/A

Remarks (Identify the specific EAL, as needed): A loss of both SATs and failure of EDGs to power Emerg Buses and at least 1 Emerg Bus not restored within 15 minutes.

Student

5. Declare the event by approving the Emergency Classification.

Student Date: ***** / ***** / **** Time: ****** Emergency Director

Student

6. Obtain Meteorological Data (not required prior to event declaration):

Wind Direction (from) 130 Wind Speed 5 Stability Class D Precipitation 0 Student

7. Initiate Attachment 2, Checklist 2 - Emergency Plan Initiation. ____________

NOTE

Key Parameters should be allowed to stabilize to accurately represent plant conditions prior to classifying an event

es-401 bwr examination outline form es-401-1es-401 bwr examination outline form es-401-1 facility:...

Documents