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ES-401 FORM ES-401-1

BWR SRO EXAMINATION OUTLINE

Facility: GRAND GULF NUCLEAR STATION Date of Exam: 6 FEBRUARY 2004 K/A CATEGORY

POINTS

TIER GROUP K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G *

POINT TOTAL

1. 1 6 3 2 8 3 4 26 Emergency &

Abnormal 2 0 2 3 4 5 3 17 Plant

Evolutions TIER TOTAL

6 5 5 12 8 7 43

1 1 1 3 3 0 3 3 1 1 3 4 23 2.

Plant 2 1 1 1 0 3 2 0 2 2 0 1 13 Systems

3 0 0 0 0 0 1 1 1 0 1 0 4 TIER

TOTAL 2 2 4 3 3 6 4 4 3 4 5 40

CAT 1 CAT 2 CAT 3 CAT 4 3. Generic Knowledge & Abilities 5 4 2 6 17

Note: 1. Ensure that at least two topics from every K/A category are sampled within each tier

(i.e., the Tier Totals in each K/A category shall not be less than two) 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 exam must total 100 points.

3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant specific priorities.

4. Systems / evolutions within each group are identified on the associated outline. 5. The shaded areas are not applicable to the category tier. 6.* The generic 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. 7. On the following pages, enter the K/A numbers, a brief description of each topic, the

topics importance ratings for the SRO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above.

REVISION 0 11/5/2003 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION FEBRUARY 2004

BWR SRO EXAMINATION OUTLINE EMERGENCY & ABNORMAL PLANT

EVOLUTIONS - TIER 1 GROUP 1

ES-401-1

E/APE #/NAME/SAFETY FUNCTION K1

K2

K3

A1

A2

G TOPIC(S) IMP SRO/RO/BOTH

REC#

RELATED K/A

ORIGIN NOTES:

295003 Partial or Complete Loss of AC Power/ 6 CFR41.7

02 Given plant conditions describe the difference of how loads on BOP and ESF busses are removed and subsequently restored during undervoltage conditions.

3.1 801q001

BOTH AK1.02: 3.4AK1.03: 3.2 AK2.03: 3.9 AK3.01: 3.5 AK3.03: 3.6 AA1.01: 3.8

NEW

295006 SCRAM / 1 CFR41.5

03 Given conditions of a reactor scram, describe the response of the Turbine Pressure Control System.

3.7 802q002

BOTH AK2.07: 4.1AA2.04: 4.1

NEW

295007 High Reactor Pressure / 3 CFR41.6

01 Given Reactor pressure, determine systems available to inject into the RPV for level control.

3.2 803q003

BOTH AK2.03: 3.2AK2.04: 3.3

NEW

295009 Low Reactor Water Level / 2 CFR41.4/41.5/41.7/43.5

02 Given a steam flow / feed flow mismatch and plant conditions, determine the reactor water level response and response of Reactor Water Level control.

3.7 804q004

BOTH MODNRC 8/2002

295010 High Drywell Pressure / 5 CFR41.4/41.5

05 Given plant parameters, determine the affects on Drywell Pressure. (Loss of cooling to the Drywell Chilled Water System with the plant at power.)

3.8 805q005

BOTH 223001K6.01: 3.8 A4.12: 3.6

MOD NRC 3/1998

295013 High Suppression Pool Water Temp. / 5 CFR41.10/43.2/43.5

01 During a surveillance operating RCIC, determine how often Suppression Pool Temperature is required to be monitored and the threshold for alternate actions.

4.0 876q076

SRO AA1.02: 3.9 2.1.33: 4.0 2.4.4: 4.3

MOD NRC 8/2002

295014 Inadvertent Reactivity Addition / 1 CFR41.1/41.2/41.6/43.6

2. 1.30

With the reactor in startup conditions such that the reactor is close to criticality, what are the operator actions if a high worth control rod is withdrawn.

3.4 806q006

BOTH AA1.04: 3.3AA2.02: 3.9 AA2.03: 4.3 2.1.2: 4.0

NEW Pilgrim event2/2003

295015 Incomplete SCRAM / 1 CFR41.6/43.5

04 Given control panel indications, determine the cause preventing full insertion of control rods under scram conditions.

3.7 807q007

BOTH AA1.01: 3.9AA1.02: 4.2 2.1.31: 3.9

NEW

295016 Control Room Abandonment / 7 CFR41.7

05 Given a loss of DC electrical power, describe the status of operation of the Safety Relief Valves operated from the Remote Shutdown Panels.

2.9 808q008

BOTH NEW

295017 High Offsite Release Rate / 9 CFR41.11/41.13/43.4

01 With a release of radioactive material in progress, determine the response of systems to protect the safety of control room personnel and maintain habitability.

3.9 809q009

BOTH AK3.05: 3.6 NEW

PAGE 1 TOTAL TIER 1 GROUP 1 1 1 2 3 2 1 PAGE TOTAL # QUESTIONS 10

REVISION 1 11/14/2003 PAGE 1 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION

FEBRUARY 2004 BWR SRO EXAMINATION OUTLINE

EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1

CONT. ES-401-1

E/APE #/NAME/SAFETY FUNCTION K1

K2

K3

A1

A2

G TOPIC(S) IMP SRO/RO/BOTH

REC#

RELATED K/A

ORIGIN NOTES:

295023 Refueling Accidents / 8 CFR41.4/41.5/41.10/43.5/43.7

02 With a Refueling outage in progress, determine the effects of a loss of Fuel Pool Cooling and Cleanup on the Fuel Storage pools.

3.1 877q077

SRO NEW

295024 High Drywell Pressure / 5 CFR41.7/41.10/43.5

06 Given a high drywell pressure condition, determine the operation of the Divisional Diesel Generators.

4.0 811q011

BOTH EA1.06: 3.7 NEW

295025 High Reactor Pressure / 3 CFR41.5/41.6/41.7

04 With a rising reactor pressure, determine the response of the RPS and ATWS ARI/RPT.

4.1 812q012

BOTH EK2.01: 4.1 NEW

295026 Suppression Pool High Water Temp. / 5 CFR41.10/41.12/43.4/43.5

2. 3.2

With RHR operating in Suppression Pool Cooling in response to a high Suppression Pool Temperature, describe the basis for contacting Radiation Protection personnel.

2.9 810q010

BOTH 2.1.32: 3.8 NEW

295027 High Containment Temperature / 5 CFR41.9/41.10/43.2/43.5

03 Determine the Containment Temperature at which Emergency Depressurization is required.

3.8 813q013

BOTH EK3.01: 3.8 MODNRC 12/2000

295030 Low Suppression Pool Water Level / 5 CFR41.8/41.10/43.5

02 Determine the Suppression Pool Water level at which ECCS pump NPSH is questionable.

3.8 814q014

BOTH NEW

295031 Reactor Low Water Level / 2 CFR41.2/41.3/41.10/41.14/43.5

01 Given plant conditions and a low reactor water level, determine core cooling mechanism and adequacy.

4.7 815q015

BOTH 2.1.1: 3.82.4.6: 4.0 2.4.18: 3.6

MOD NRC 8/2002

295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 CFR41.10/43.5

2. 4.40

Given ATWS conditions, determine the Emergency Plan Emergency Action Level.

4.0 878q078

SRO 2.4.41: 4.1 NEW Alert vs. Site Area Emergency

295038 High Offsite Release Rate / 9 CFR41.10/41.12/43.4/43.5

02 Given meteorological data, maps and a radioactive release, determine protective action recommendations to be issued.

3.8 879q079

SRO 2.4.44: 4.0 NEW

500000 High Containment Hydrogen Conc. / 5 CFR41.9

01 Determine the bases for the Hydrogen Control leg of EP-3.

3.9 817q017

BOTH MODNRC 8/2002

295031 Reactor Low Water Level / 2 CFR41.7/41.10/43.5

2. 1.31

Determine actual reactor water level when operating from the Remote Shutdown Panels using the associated graphs and given indications.

3.9 816q016

BOTH EK2.01: 4.4EA2.01: 4.6

MOD NRC 8/20021

PAGE 2 TOTAL TIER 1 GROUP 1 3 2 0 3 0 3 PAGE TOTAL # QUESTIONS 11

REVISION 1 11/14/2003 PAGE 2 OF 13 NUREG 1021, REVISION 8 SUPPLEMENT 1

GRAND GULF NUCLEAR STATION

FEBRUARY 2004 BWR SRO EXAMINATION OUTLINE

EMERGENCY & ABNORMAL PLANT EVOLUTIONS - TIER 1 GROUP 1

CONT. ES-401-1

E/APE #/NAME/SAFETY FUNCTION K1

K2

K3

A1

A2

G TOPIC(S) IMP SRO/RO/BOTH

REC#

RELATED K/A

ORIGIN NOTES:

295025 High Reactor Pressure / 3 CFR41.4/41.5/41.14

05 Given plant conditions, describe the response of RCIC to a rising reactor pressure.

3.7 818q018

BOTH 217000A1.04: 3.6

NEW

295017 High Off-Site Release Rate / 9 CFR41.10/41.13/43.2/43.4/43.5

01 With a liquid radwaste discharge required and a discharge permit, determine whether a release is allowed.

3.1 880q080

SRO 2.3.3: 2.9 2.3.6: 3.1

NEW

295015 Incomplete SCRAM / 1 CFR41.1/41.2/41.5

04 Describe the r

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