responds to generic ltr 89-21 re status of implementation of usi … · 2018. 2. 7. · usi/mpa...

ACCELERATED D UTION DEMONSTRATION SYSTEM

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8912050090 DOC.DATE: 89/ll/28 NOTARIZED: NOFACIL:50-220 Nine Mile Point Nuclear Station, Unit 1, Niagara Powe

AUTH.NAME AUTHOR AFFXLIATXONTERRY,C.D. Niagara Mohawk Power Corp.

RECIP.NAME RECIPXENT AFFILIATIONDocument Control Branch (Document Control Desk)

DOCKET05000220

R

SUBJECT: Responds to Generic Ltr 89-21 for request for info re statusof implementation of USI requirements.

DXSTRIBUTION CODE: A012D COPXES RECEIVED:LTR ENCL SXZE:TITLE: Generic Ltr 89-21 Response,Xmplementation of Unresolved SafetyNOTES

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INTERNAL: ACRSNUDOCS-ABSTRACTWESSMAN,R PD13

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TOTAL NUMBER OF COPIES REQUIRED: LTTR 10 ENCL 10

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al 7 NIAGARAH O MOHAWKNIAGARAMOHAWKPOWER CORPORATION/301 PI&INFIELDROAD, SYRACUSE, N.Y. 13212/TELEPHONE (315) 474-1511

November 28, 1989NHPlL 0459

U.S. Nuclear Regulatory CommissionAttn: Document Control DeskWashington, D.C. 20555

Re: Nine Mile Point Unit 1

Docket No. 50-220DPR-63

Gentlemen:

On October 19, 1989, the Nuclear Regulatory Commission issued a request forinformation concerning status of implementation of Unresolved Safety Issue(USI) requirements (Generic Letter 89-21).

Enclosure 1 to this letter tabulates the requested status of implementation ofUSIs for which a final technical resolution has been achieved and which areapplicable to Nine Mile Point Unit 1. The status provided is based on NiagaraMohawk's judgement of what actions were necessary to resolve the technicalissue's for Unit 1 and is not meant to imply compliance with all referencedguidance documents.

JMT/mjd7998G

Very truly yours,

rNIAGARA HOHANK PO R CORPORATION

C. D. TerrVice President

Nuclear Engineering and Licensing

xc: Regional Administrator, Region IMr. R. AD Capra, DirectorMr. R. E. Hartin, Project ManagerHr. N. A. Cook, Resident InspectorRecords Management

S912050090 89112SPDR ADOCK,05000220P PDC

A

Il II

ENCLOSURE

1'SI/MPA

NUMBER

A-1

TITLE

Water Hammer

-REF. DOCUMENT

SECY 84-119NUREG-0927, Rev. 1

NUREG-0993, Rev. 1

NUREG-0737 ItemI.A.2.3SRP revisions

APPL ICABILITY

All

STATUS/DATE* REMARKS

'MI training Item I.A.2.3incorporated (TMI ItemStatus Letter 4/8/89).'ore Spray System WaterHammer Issue still open.Niagara Mohawk submittedan evaluation on July 6,1989 (NMP1L 0418).Awaiting NRC SER.

'A-2/ Asymmetric BlowdownMPA D-10 Loads on Reactor Primary

Coolant Systems

NUREG-0609GL 84-04, GDC-4

V

PWR NA NMPl is a BWR

A-3 Westinghouse SteamGenerator Tube Integrity

NUREG-0844SECY 86-97SECY 88-272GL 85-02(No requirements)

W-PWR NMPl is a BWR.

A-4 CE Steam Generator TubeIntegrity

NUREG-0844, SECY 86-97 CE-PWRSECY 88-272GL 85-02(No requirements)

NA NMP1 i s a BWR.

*C — COMPLETE

NC — NO CHANGES NECESSARYNA — NOT APPLICABLEI — INCOMPLETEE — EVALUATING ACTIONS REQUIRED

7998G

USI/MPANUMBER

A-5

TITLE

BKH Steam GeneratorTube Integrity

REF. DOCUMENT APPLICABILITY

NUREG-0844, SECY 86-97 BE(H-PHRSECY 88-272GL 85-02(No requirements)

STATUS/DATE* REMARKS

NMPl is a BWR.

E

A-6

A-7/D-01

Mark I ContainmentShort-Term Program

Mark I Long-TermProgram

NUREG-0408

NUREG-0661NUREG-0661 Suppl. 1

GL 79-57

Mark I--BHR

Mark I-BHR

C/1-22-85

C/1-22-85

See NRC's SER dated1/22/85.

The Commission's letterdated 1/22/85 documented~their review and accepta~of the NMP1 Mark Icontainment program.

A-8 Mark II ContainmentPool Dynamic Loads

NUREG-0808 Mark II-BHRNUREG-0487, Suppl. 1/2NUREG-0802SRP 6.2.1.1CGDC 16

NA NMP1 is a Mark I BHR.

A-9 Anticipated TransientsWithout Scram

NUREG-0460, Vol. 410 CFR 50.62

Al 1 Completed during 1988/89outage.

*C — COMPLETE


7998G

USI/MPANUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS

A-10/ BHR Feedwater NozzleMPA B-25 Cracking

NUREG-0619 BHR

Letter from DG Eisenhutdated 1 1/13/80GL 81-11

C/6/84 'ee Niagara Mohawk'sresponse to NUREG-0619commitments datedDecember 29, 1980.

'iagara Mohawk mademodifications to feed-water,low flow controlsystem to reduce/eliminate thermal cyclingon feedwater nozzles in1984.

'iagara Mohawk is makioperational changes tofurther reduce/eliminatethermal cycling on feed-water nozzles beginningwith the startup fromthe current outage.

A-11

A-12

Reactor Vessel MaterialToughness

Fracture Toughness ofSteam Generator andReactor Coolant PumpSupports

NUREG-0744, Rev. 1

10 CFR 50.60/82-26

NUREG-0577, Rev. 1

SRP Revision5.3.4

Al 1

PHR NA

Vessel material propertiesinspected in accordancewith Tech Spec 4.2.2.b.

NMPl is a BHR.

* C — COMPLETENC — NO CHANGES NECESSARYNA — NOT APPLICABLE

I — INCOMPLETEE — EVALUATING ACTIONS REQUIRED

7998G

USI/MPANUMBER

A-17

TITLE

Systems Interactions

REF. DOCUMENT APPLICABILITY

AllLtr: DeYoung tolicensees-9/72NUREG-1174, NUREG-1229, NUREG/CR-3922,NUREG/CR-4261, NUREG/CR-4470, GL 89-18(No requirements)

STATUS/DATE*

NC

REMARKSC

~ ~

Generic Letter 89-18 wasissued for information only.No specific action orwritten response wasrequired. No action hasbeen taken.

A-24/ Qualification of ClassMPA B-60 lE Safety-Related

Equipment

A-26/ Reactor Vessel PressureMPA B-04 Transient Protection

A-31 Residual Heat RemovalShutdown Requirements

A-36/ Control of Heavy LoadsC-10, Near Spent FuelC-15

*C — COMPLETE


7998G

NUREG-0588, Rev. 1

SRP 3.1110 CFR 50.49GL 82-09, GL 84-24GL 85-15

DOP Letters toLicensees 8/76NUREG-0224NUREG-0371SRP 5.2GL 88-11

NUREG-0606RG 1.113,RG 1.139SRP 5.4.7

NUREG-0612SRP 9.1.5GL 81-07, GL 83-42,GL 85-11Letter from DG

Eisenhut dated12/22/80

Al 1

PWR

All OLs After01/79

All

NA

NA

Safety Evaluation docu-menting NMP1 complies with10CFR50.49 issued on1/10/85.

NMPl is a BWR.

NMP1's Operating Licensewas issued prior to 1/79.

Safety Evaluationdocumenting NMPl compliewith NUREG-0612, Section5.1.1 and 5.3 (Control ofHeavy Loads-Phase I) issuedon 3/5/85.


A-39

A-40

Determination of SRVPool Dynamic Loadsand Pressure Transients

Seismic DesignCriteria

NUREG-0802NUREGs-0763,0783,0802NUREG-0661SRP 6.2.1.1.C

SRP Revisions, NUREG/ AllCR-4776, NUREG/CR-0054,NUREG/CR-3480, NUREG/CR-1582; NUREG/CR-1161,NUREG-1233, NUREG/4776NUREG/CR-3805NUREG/CR-5347NUREG/CR-3509

C/1-22-85 For Hark" I Containments;SRV acceptance criteria ispresented in NUREG-0661and dealt with as part ofUSI A-7.

USI A-40 being resolvedas -part of A-46.

A-42/ Pipe Cracks in BoilingMPA B-05 Hater Reactors

* C — COMPLETENC — NO CHANGES NECESSARYNA — NOT APPLICABLE

I — INCOMPLETEE — EVALUATING ACTIONS REQUIRED

7998G:

NUREG-0313, Rev. 1

NUREG-0313, Rev. 2

GL 81-03, GL 88-01

BHR I/Next RefuelOutage

NMPl's response to.GenericLetter 88-01 is found inletters dated 7/28/88,8/25/89 and 9/6/89.Requirements of GL 88-01are scheduled to beincorporated into the ISI

-program prior to the nextrefuel outage per the7/28/89 letter. The NMPlTechnical Specificationswere revised to conform tothe positions delineatedin GL 88-01 by AmendmenNo. 107. The SafetyEvaluation ofImplementation ofNUREG-0313, Rev. 1 wasissued on 6/6/84.


A-43

A-44

A-45

A-46

Containment EmergencySump Performance

Station Blackout

Shutdown Decay HeatRemoval Requirements

Seismic Qualificationof Equipment inOperating Plants

NUREG-0510,NUREG-0869, Rev. 1

NUREG-0897, R.G. 1.82(Rev. 0), SRP 6.2.2GL 85-22(No requirements)

RG 1.155NUREG-1032NUREG-110910 CFR 50.63

SECY 88-260NUREG-1289NUREG/CR-5230SECY 88-260(No requirements)

NUREG-1030NUREG-1211/GL 87-02, GL 87-03

Al 1

Al 1

Al 1

Al 1

NC 'o requirements.

'rocedural Chan es-within 12 months ofnotification in accordancewith 10CFR50.63(c)(3).~'atter Modifications ~(As required) within 24months. of noti fi cati onin accordance with10CFR50.63(c)(3).

'osition documented inletter dated 4/13/89.

Development of an NMP1 IPEis scheduled for July 30,1993.

NMPl is a member of theSeismic QualificationUtility Group (SQUG)-US~A-46 is being resolved asgroup solution with NMPlas the pilot plant of theprocess.

*C — COMPLETE


7998G


C

4

A-47

A-48

A-49

Saf ety Imp 1 i cati onof Control Systems

Hydrogen ControlMeasures and Effectsof Hydrogen Burnson Safety Equi pment

Pressurized ThermalShock

NUREG-1217, NUREG-1218GL 89-19

10 CFR 50.44SECY 89-122

RGs- 1.154, 1.99SECY 82-465SECY 83-288SECY 81-68710 CFR 50.61/GL 88-11

All

All, exceptPHRs withlarge drycontainments

PWR NA

NMP1 is currentlyevaluating the requirementsof GL 89-19 and is expectedto provide its response by3/19/90.

'ontainment inerted.'ontainment Atmosphere

Dilution System (CAD) addedper modification Nl-72-0 .

'afety Evaluationdocumenting NMP1 compliewith 10CFR50.44(c)(3)(ii)issued on 4/29/85.

NMPl is a BWR.

* C — COMPLETE

NC — NO CHANGES NECESSARY

NA — NOT APPLICABLEI — INCOMPLETEE — EVALUATING ACTIONS REQUIRED

7998G

'PLANT NMP-I

PROJECT MANAGER Robert E. Martin

USI NO. A-I TITLE Hater Hewer

DOCKET NO(S). 50-220

TECHNICAL CONTACT A. Serkiz

MPA NO. ~NA TAC NOS.

ISSUES SUMMARY:

This Unresolved Safety Issue (USI) was resolved in March 1984, with thepublication of NUREG-0927, "Evaluation of Water Hammer in Nuclear Power Plants- Technical Findings Relevant to Unresolved Safety Issue A-l." Also on March15, 1984, the EDO sent the Commissioners SECY 84-119 titled, "Resolution ofUnresolved Safety Issue A-l, Water Hammer."

In SECY 84-119, the staff concluded that the frequency and severity of waterhammer occurrences had been significantly reduced through (a) incorporation ofdesign features such as keep-full systems, vacuum breakers, J-tubes, voiddetection systems, and improved venting procedures; (b) proper design of feed-water valves and control systems; and (c) increased operator awareness andtraining. Therefore, the resolution of USI A-I did not involve any hardware ordesign changes on existing plants. It did involve Standard Review Plan (SRP)changes (forward fits) and a comprehensive set of guidelines and criteria toevaluate and upgrade utility training programs (per TMI Task Action Plan ItemI.A.2.3). In addition, the assumption was made that for BWRs with isolationcondensers ( ICs) a reactor-vessel high water-level feedwater pump trip was inplace or being installed. This was necessary because calculated values hadpostulated an IC failure by water hammer that opened a direct pathway to theenvironment.

IMPLEMENTATION AND STATUS SUMMARY PLANT SPECIfIC):

By letter dated April 13, 1978, the licensee responded to the Comission'sletter dated February I, 1978 regarding the need for a feedwater pump trip onreactor high water level. By letter dated December 15, 1978, the staffinformed the licensee that after review of their April 13, 1978 response andsubsequent verbal communication with members of the licensee's organization,the staff had determined that a trip of the motor driven feedwater pumps onhigh reactor water level was not necessary to assure safe operation of NineMile Point Unit 1.

However, several years later, in response to the issue of the hi-level tripas a TMI Action Item the licensee committed by letter dated April 1, 1982, toinstall, prior to star tup from the next refueling outage, a reactor vesselhigh level trip of the motor driven feedwater pumps. For the purpose ofUSI A-1, the issue was completed prior to the July 24, 1984 startup by theinstallation of a reactor vessel high level trip of the moter driven feedwaterpumps.

In addition to the scope of the issue dealt with by USI A-l, a concern wasraised during Safety System Funtional Inspection 88-201 regarding thepotential for water hammer during star tup of the core spray system during aLOCA.

-2-

By letter dated March 28, 1989, the licensee forwarded to the NRC theiranalysis for the potential for water hammer during startup of the core spraysystem during a LOCA.

By letter dated April 18, 1989, the licensee forwarded to the NRC the statusof TMI action plan items. TMI issue number I.A.2.3 (Administration ofTraining Program) is shown as being completed.

By letter dated July 6, 1989 Niagara Mohawk submitted a response to a staffrequest for additional information for Safety System Functional InspectionUnresolved Item 88-201-2C regarding the potential for Waterhammer duringstartup of the core spray system during a LOCA. The staff has reviewed thelicensee's response and found it acceptable.

I REFERENCES:

1. RE UIREMENT DOCUMENTS:

TITLE

NMP-1A-1

NUDOCS NO. DATE

Letter from Denton to Utilities,"Notice of Issuance andAvailability NUREG-0927 Rev. 1,Safety Issue A-1"

NUREG-0927 "Evaluation of WaterHammer in Nuclear Power Plants-Technical Findings Relevant toUnresolved Safety Issue A-1"

NUREG-0993 Rev. 1"Regulatory Analysis forfor USI A-l, Water Hammer"

SRP Sections: 3.9.3, 3.9.4,5.4.6, 5.4.7, 6.3, 9.2.1, 9.2.2,10.3, and 10.4.7

SECY-84-119, "Resol utionof Unresolved Safety A-l,Water Hammer"

2. IMPLEMENTATION DOCUMENTS:

TITLE

Letter from T. Ippolito (NRC)to D. P. Disc (NMPC)

Letter from D. P. Disc (NMPC)to D. Eisenhut (NRC)

Letter from C. D. Terry (NMPC)to NRC

Letter from L. Burkhardt (NMPC)to NRC


8403150310

8306060413

8306060418

NUDOCS NO.

7812290049

8204060097

8904050067

8904260240

8907110345

03/05/84

05/31/83

March 1984

03/15/84

DATE

12/15/78

04/01/82

03/28/89

04/18/89

07/06/89

3. VERIFICATION DOCUMENTS:

TITLE NUDOCS NO. DATE

PLANT NMP-1

PROJECT MANAGER Robert E.Martin


TECHNICAL CONTACT J. Kudrick.

USI NO. A-6 TITLE Mark I Containment Short Tenn Pro ram

MPA NO. TAC NOS.

ISSUES SUMMARY:

This USI was resolved in December 1977 with the publication of NUREG-0408,"Mark I Containment Short-Term Program Safety Evaluation Report."

The objectives of the Mark I short-term program were: (a) to examine thecontainment system of each BWR facility with a Mark I containment design toverify that it would maintain its integrity and functional capability whensubjected to the most probable hydrodynamic loads induced by a postulateddesign-basis LOCA, and (b) to verify that licensed Mark I BWR facilities couldcontinue to operate safely, without undue risk to the public health and safetyuntil such time as a methodical, comprehensive long-term program is conducted.

The NRC staff used a safety factor of at least two to failure for the weakeststructural or mechanical component in the Mark I containment system in judgingthat containment integrity and functions would be assured under most probabledesign-basis LOCA-induced hydrodynamic loads.

As indicated in NUREG-0408, the staff required full implementation of thecalculation of the hydrodyiiamic loads and structural analysis as an interimmeasure until complete implementation of the long-term program had beenachieved. In NUREG-0408 the staff concluded that the objectives of the Short-Term Program had been satisfied, thus documenting the basis for resolving thissafety issue. This issue is considered complete for all affected BWRs.

IMPLEMENTATION AND STATUS SUMMARY PLANT SPECIFIC :

By letter dated September 3, 1976, October I, l976, and October 14, 1976, thelicensee provided information to the NRC regarding the resolution of USI A-6.

On February 28, 1978 an Exemption from GDCSO was granted to all affectedlicensees. These exemptions concerned a minimum margin of safety of two inthe containment design as part of the short term program. This was deemed anadequate basis for continued operation until the completion of the Long TermProgram.

An Order for Modification of License and Grant of Extension of Exemption wasissued on January 13, 1981. This Order and Exemption was extended by afurther Order dated January 19, 1982 to require completion of modifications tomeet the Long Term Program defined in NUREG-0661.

For purposes of documenting a programmatic endpoint of USI A-6 for NMP-1, theShort Term Program, the issuance of the Exemption on February 28, 1978 isutilized.

REFERENCES: NMP-1A-6


TITLE

NUREG-0408, "Mark I ContainmentShort Term Program SafetyEvaluation Report" (See Table I-2for letters to BWR licenseesrequesting action)

NUDOCS NO. DATE

10/77


TITLE

G. Rhode (NMPC) to G. Lear (NRC)

Exemption from GDC 50 oncontainment design minimum marginof safety

TA Ippolito (NRC) to DP Disc (NMPC)issuing Order and Extension ofExemption

NUDOCS NO. DATE

10/14/76

02/28/78

01/13/81

D. Vassallo (NRC) to D.P. Disc (NMPC)issuing Order modifying the 1/13/81Order 01/19/82

3. VERIFICATION DOCUMENTS


'LANT NMP-I DOCKET NO(S). 50-220

PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Kudrick

OSI NO. A-7 TITLE Mark I Loo Term Pro ram

MPA NO. TAO NOS. 07942

ISSUES SUMMARY:

This USI was resolved in August 1982 with the publication of Supplement 1 toNUREG-0661, "Safety Evaluation Report, Hark I Containment Long-Term Program"and Standard Review Plan Section 6.2.1.1.C. For operating BWRs, MPA D-Ol wasestablished for implementation purposes.

The focus of this USI was the suppression pool hydrodynamic loads, associatedwith a postulated LOCA, which had not explicitly been included in the originalMark I containment design. The issue was identified during large-scale testingof a Mark III containment design. The staff addressed this issue in NUREG-0661,published in July 1980, and in Supplement 1 to NUREG-0661, published in August1982.

The objective of the long-term program (LTP) was to establish the design-basisloads that are appropriate for the anticipated life of each Mark I BWR facilityand to restore the originally intended design-safety margins for each Mark Icontainment system. The principal thrust of the LTP was the development ofgeneric methods for defining suppression pool hydrodynamic loadings and theassociated structural assessment techniques for the Mark I configuration. Onthe basis of experimental and analytical programs conducted by the Mark IOwners Group, it was determined that the hydrodynamic load definition pro-cedures, with some modifications defined in NUREG-0661, provided a conservativeestimate of these loading conditions. Thus, the requirements associated withthis USI were concerned with the structural assessment of Hark I containmentsand related structures to the hydrodynamic loads defined by the staff in theLTP.

In January 1981, the staff issued "Orders For Modification of License and Grantof Extension of Exemptions" to each licensee of a Mark I plant. The ordersrequired the licensees to assess the suppression pool hydrodynamic loads inaccordance with General Electric documents and NUREG-0661 on a definedschedule. For some plants, the implementation schedule was extended by asubsequent order.


On January 19, 1982 the NRC issued an Order extending the completion date forthe Mark I Long-Term Program at the Nine Mile Point Unit No. 1 facility toread: "Prior to the start of Cycle 8 at the completion of your Spring 1983refueling outage."

By letter dated January 5, 1983, the licensee committed to complete Mark ILong-Term Program modifications prior to the start of the Cycle 8 in 1984.The staff's letter of 1/07/83 provided the staff's position on interpretationof the required completion date for cycle 7 and the beginning of cycle 8 tomeet the requirement of the Order. The licensee.has recently orally confirmedthat these modifications were completed prior to the start of cycle 8 on06/13/84.

As additional background information concerning the staff'spost-implementation review of the plant unique analysis report, it is notedthat the staff forwarded to the licensee the results of its review made withthe assistance of Srookhaven National Laboratory and Franklin Research Center.The staff concluded that the modifications made by the licensee were inaccordance with the generic acceptance criteria contained in Appendix A ofNUREG-0661. Where deviations from the acceptance criteria specified inNUREG-0661 had been taken, they were found acceptable by the staff.

'REFERENCES:


TITLE NUDOCS NO.

NMP-1A-7

DATE

NUREG-0661, "Safety EvaluationReport, Mark I Containment LongTerm Program"

NUREG-0661, Supplement 1

Orders for Modification to Licensefor Applicable Licensees


TITLE

T. A. Ippolito (NRC) to D. P. Disc(NMPC) issuing Order and Extensionof Exemption

Letter from D. P. Disc (NMPC)to T. A. Ippolito (NRC)

D.B. Vassallo (NRC) to D. P. Disc(NMPC) issuing Order modifying the1/13/81 Order

Letter from C. V. Mangan (NMPC)to D. B. Yassallo (NRC)

Letter from D. B. Vassallo (NRC)to G. K Rhode (NMPC)

Letter from D. B. Vassallo (NRC)to B. G. Mooten (NMPC)

Letter C. D. Terry (NMPC)to NRC

NUDOCS NO;

8112230144

8301070265

8301180521

8502040083

8912050090

07/80

08/82

1981

DATE

1/13/81

12/11/81

1/19/82

01/05/83

01/07/83

01/22/85

11/28/89

3. YERIFICATION DOCUMENTS:


'PLANT NMP-1 OOCKET NO(S). 50-220

PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Mauck

USI NO. A-9 TITLE ATWS er 10 CFR 50.62

MPA NO. A-20 TAC NOS. 67506

ISSUES SUMMARY:

This USI was resolved in June 1984 with the publication of a final rule (10 CFR50.62) to require improvements in plants to reduce the likelihood of failure ofthe reactor protection system (RPS) to shut down the reactor followinganticipated transients and to mitigate the consequences of an anticipatedtransient without scram (ATWS) event.

The rule includes the following design-related requirements: 50.62(C)(1),diverse and independent auxiliary feedwater initiation and turbine trip for allPWRs; 50.62(C)(2), diver se scram systems for CE and B8W reactors; 50 .62(C)(3)alternate rod injection (ARI) for BWRs 50.62(C)(4); standby liquid controlsystem (SLCS) for BWRs; and 50.62(C)(5I, automatic trip of recirculation pumpsunder conditions indicative of an ATWS for BWRs. Information requirements andan implementation schedule are also specified.

IMPLEMENTATION AND STATUS SUMMARY (PLANT SPECIFIC):

By letter dated April 1, 1987, the licensee submitted information to'demonstrate the adequacy of the ARI, SLCS and RPT. Additional information wassubmitted by the licensee regarding the ARI and RPT on July 6, 1988.

By letter dated August 31, 1988, the staff forwarded to the licensee a safetyevaluation supporting the licensee submittals on compliance with the ATWS ruleregarding the ARI and RPT.

On October 31, 1988, Tech Specs Amendment No. 101 was issued (in response tothe licensee's application dated March 7, 1988, as supplemented on Apri 1 13,1988). It revises the Tech Specs for the Liquid Poison System to satisfy therequirements of 10 CFR 50.62.

The licensee states orally that it considers the implementation of 10 CFR50.62/USI A-9 to be complete by February 28, 1990. This is the date bywhich the licensee has verified the mixing concentration for the Liquid PoisonSystem in order to implement the requirements of License Amendment no. 101.

'EFERENCES:


TITLE

NMP-1A-9

NUDOCS NO. DATE

NUREG-0460, and Supplements, ,

"Anticipated Transients WithoutScram for Light Water Reactors"

Federal Register Notice49 FR 26045 (10 CFR 50. 62)

03/80

06/26/84


TITLE

Letter from T. E. Lempges (NMPC)to NRC


Letter from M. C. Haughey (NRC)to C. V. Mangan (NMPC)

Letter from M. F. Haughey (NRC)to C. V. Mangan (NMPC)



TITLE

NUDOCS NO.

8704070373

8807120406

8809060097

8811090428

8912050090

NUDOCS NO.

DATE

04/01/87

07/06/88

08/31/88

10/31/88

11/28/89

DATE

PLANT NMP-1 DOCKET NO(S). 50-220

PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT K. Wichman

US I NO. A-10 TITLE BWR Feedwater Nozzle Crackin

MPA NO. 8-25 TAC NOS; 08499 and 72944

ISSUES SUMMARY:

This issue was resolved in November 1980 with the publication of NUREG-0619,"BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking." MPAB-25 was established by NRC's Division of Licensing for implementationpurposes.

Inspections of operating BWRs conducted up to April 1978 revealed cracks in thefeedwater nozzles of 20 reactor vessels. It was determined that cracking wasdue to high-cycle fatigue caused by fluctuations in water temperature withinthe vessel in the nozzle region.

By letter dated November 13, 1980, Darrell G. Eisenhut provided licensees witha copy of NUREG-0619. The 1'etter stated that NUREG-0619 provided the resolu-tion of the staff's generic technical activity USI A-10, which resulted fromthe inservice discovery of cracking in feedwater nozzles and control rod drivereturn line nozzles. NUREG-0619 describes the technical issues, GeneralElectric and staff studies and analyses, and the staff's positions and require-ments. Licensees were required to respond, pursuant to 10 CFR 50.54(f), thatthey would meet implementation dates indicated in NUREG-0619.

Generic Letter 81-11 was subsequently issued to provide technical clarificationto the November 13, 1980 letter, to clarify that it had been sent to PWR

licensees for information only, and that no response was required from PWR

licensees.


By letter dated December 29, 1980 the licensee responded to NUREG-0619 toconfirm that they would meet the implementation dates indicated in NUREG-0619.

By letter dated July 10, 1981, the NRC accepted the December 29, 1980 proposedactions regarding the implementation of A-10 per NUREG-0619 and requestedadditional information. By letter dated October 29, 1981, the Commissionnotified the licensee that no further correspondence was necessary on thesubject of A-10 implementation and the Commission was satisfied with thelicensee's Commitment to meet the intent of NUREG-0619. The licenseeindicates orally that the initial implementation of its responses toNUREG-0619 was completed on June 4, 1983 when the feedwater low flow controlmodification no. Nl-8269 was completed. This modification was directed atreducing the thermal cycling on the feedwater nozzles.

( ~J

NHP-1A-10

In addition to the scope of the issue dealt with by USI A-10 the followingactions were taken.

(a) By letter dated December 23, 1986, the licensee forwarded a report of theinservice inspection of the feedwater nozzles.

(b) By letter dated triarch 21, 1989, as supplemented tray 5, 1989, and byinformation presented in an April 18, 1989 meeting the licensee requested adeferment to the NUREG-0619 commitment to remove a feedwater nozzle sparger.

(c) By letter dated September 26, 1989, the staff forwarded to the licensee asafety evaluation in which it was concluded that there is reasonableassurance that the facility can be safely operated during the next two cycleswith the feedwater nozzle "A" in its current condition.

(d) In addition, the licensee states that operational changes are being made tofurther reduce/eliminate thermal cycling on feedwater nozzles beginning with thestartup from the current outage.

'EFERENCES'.

RE UIREMENT DOCUMENTS:

TITLE

Letter from D. Eisenhuttransmitting NUREG-0619,"BWR Feedwater Nozzle andControl Rod Drive ReturnLine Nozzle Cracking,"resolution of A-10 tolicensees

Generic Letter 81-11, "BWR

Feedwater Nozzle and ControlRod Drive Return Line NozzleCracking (NUREG-0619)"

NUDOCS NO.

NMP-1A-10

DATE

11/13/80

02/20/81


TITLE

Letter from N. P. Disc (NMPC)to D. G. Eisenhut (NRC)

Letter from T. A. Ippolito (NRC)to D. P. Disc (NMPC)

Letter from T. A. Ippolito (NRC)to D. P. Disc (NMPC)

Letter from C.Y. Mangan (NMPC)to T. E. Murley (NRC)

Letter from NMPC to NRC


NUDOCS NO.

8101050087

8107220366

8111200008

8701020108

8903280112

8905190123


8912050090

Letter from M. L. Slosson (NRC) 8910030451

DATE

12/29/80

07/10/81

10/29/81

12/23/86

03/21/89

05/05/89

09/26/89

11/28/89

3. YER IF I CATION DOCUMENTS:


'LANT NMP-I DOCKET NO(S). 50-220

PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT B. Elliott-USI NO. A-11 TITLE Reactor Vesse1 Materials Toe hoess

MPA NO. A007 TAC NOS. 07445

ISSUES SUMMARY:

This. USI was resolved in October 1982 with the publication of NUREG-0744,"Pressure Vessel Material, Fracture Toughness.". NUREG-0744 was issued byGeneric Letter 82-26 and provided only a methodology to satisfy the require-ments of 10 CFR Part 50, Appendix G. No licensee response to Generic Letter82-26 was required.

Because of the remote possibility that nuclear reactor pressure vesselsdesigned to the ASME Boiler and Pressure Vesse'l Code would fail, the design ofnuclear facilities does not provide protection against reactor vessel failure.Prevention of reactor vessel failure depends primarily on maintaining thereactor vessel material fracture toughness at levels that will resist brittlefracture during plant operation. At service times and operating conditionstypical of current operating plants, reactor vessel fracture toughnessproperties provide adequate margins of safety against vessel failure; however,as plants accumulate more and more ser vice time, neutron irradiation reducesthe material fracture toughness and initial safety margins.

Appendix G to 10 CFR Part 50 requires that the Charpy upper shelf energythroughout the life of the vessel be no less than 50 ft-lb unless it isdemonstrated that lower values will provide margins of safety against failureequivalent to those provided by Appendix G of the ASME code. USI A-ll wasinitiated to address the staff's concern that some vessels were projected tohave beltline materials with Charpy upper shelf energy less than 50 ft-lb.NUREG-0744 provides a method for evaluating reactor vessel materials when theirCharpy upper shelf energy is predicted to fall below 50 ft-lb. Plants will usethe prescribed method when analysis of irradiation damage predicts that thecharpy upper shelf energy is below 50 ft-lb.IMPLEMENTATION AND STATUS SUMMARY PLANT SPECIFIC:

Licensee states that vessel material properties are inspected in accordancewith TS 4.2.2.b. Licensee has not identified a condition approaching the50 ft-lb level and therefore has no need to utilize NUREG-0744 methods.

"

REFERENCES:


TITLE

NUREG-0744, Revision 1, "PressureVessel Material Fracture Toughness"

Generic Letter 82-26, "PressureVessel Material Fracture Toughness"

NUDOCS NO.

NMP-1A-11

DATE

10/82

11/12/82


TITLE


NUDOCS NO.

8912050090

DATE

11/28/89


TITLE NUDOCS NO. DAVE

r

I ~


PROJECT MANAGER Roberr E. Martin TECHNICAL CONTACT D. Thatcher

OSI NO. A-17 TITLE S stems Interactions in Nuclear Power Plants

NPA NO. TAC NOS.

ISSUES SUMMARY:

Generic Letter (GL) 89-18, dated September 6, 1989, was sent to all powerreactor licensees and constitutes the resolution of USI A-17. The genericletter did not require any licensee actions.

GL 89-18 had two enclosures which (a) outlined the bases for the resolution ofUSI A-17, and (b) provided five general lessons learned from the review of theoverall systems interaction issue. The staff anticipated that licensees wouldreview this information in other programs, such as the Individual PlantExamination ( IPE) for Severe Accident Yulnerabilities. Specifically, the staffexpected that insights concerning water intrusion and flooding from internalsources, as described in the appendix to NUREG-1174, would be considered in theIPE program. Also considered in the resolution of this USI was the expectationthat licensees would continue to review information on events at operatingnuclear power plants in accordance with the requirements of TMI Task ActionPlan Item I.C.5 (NUREG-0737).

IMPLEMENTATION AND STATUS SUMMARY PLANT SPECIFIC:

Per guidance of GL 89-21 on the status of USIs, no licensee actions wererequired in response to GL 89-18 and accordingly, the licensee stated that noactions were taken.

As background information on flooding issues beyond the scope of GL 89-18 itis noted that By letter dated August 3, 1972, the licensee was requested toreview their facility following an event at guad Cities Unit 1 where floodingcaused degradation of some of the engineered safety features. By letter datedSeptember 29, 1972, the licensee responded to the Augsut 3, 1972 letter andstated that there was no flooding potential for existing engineered safetysystem.

REFERENCES:


TITLE

Generic Letter 89-18

NUREG-1174 "Evaluation ofSystems Interactions in NuclearPower Plants"

NUREG-1229 "Regulatory Analysisfor Resolution of USI A-17"

NUREG/CR-3922 "Survey andEvaluation of System InteractionEvents and Sources"

NUREG/CR-4261 "Assessment ofSystem Interaction Experience inNuclear Power Plants"

NUREG/CR-4470 "Survey andEvaluation of VitalInstrumentation and ControlPower Supply Events"

NRC Letters to LicenseesInforming Licensees of StaffConcerns Regarding PotentialFailure of Non-Category IEquipment

NUDOCS NO.

NMP-1A-17

DATE

09/06/89

May 1989

August 1989

January 1985

June 1986

August 1986

9/72



Letter from T. J. Brosnan (NMPC)to D. J.Skovholt (NRC)

Letter from C. D. Terry, (NMPC) 8912050090

09/29/72

11/28/89


TITLE NUOOC NO. DATE


PROJECT HANAGER Robert E. Martin TECHNICAL CONTACT P. Shemanski

USI NO. A-24 TITLE gualification of Class 1E E ui ment

HPA NO. TAC NOS. 42476

ISSUES SUMMARY:

This USI was resolved in July 1981 with the publication of NUREG-0588, Revision1, "Interim Staff Position on Environmental gualification of Safety-RelatedElectrical Equipment." Part I of the report is the original NUREG-0588 thatwas issued for comment; that report, in conjunction with the Division ofOperating Reactor (DOR) Guidelines, was endorsed by a Commission Memorandum andOrder as the interim position on this subject until "final" positions wereestablished in rule making. On January 21, 1983 the Commission amended 10 CFR50.49 (the rule), effective February 22, 1983, to codify existing qualificationmethods in national standards, regulatory guides, and certain NRC publications,including NUREG-0588.

The rule is based on the DOR Guidelines and NUREG-0588. These provide guidanceon (a) how to establish environmental service conditions, (b) how to selectmethods which are considered appropriate for qualifying the equipment indifferent areas of the plant, and (c) such other areas as margin, aging, anddocumentation. NUREG-0588 does not address all areas of qualification; it doessupplement, in selected areas, the provisions of the 1971 and 1974 versions ofIEEE Standard 323. The rule recognizes previous qualification effortscompleted as a result of Commission Memorandum and Order CLI-80-21 and alsoreflects different versions IEEE 323 dependent on the date of the constructionpermit Safety Evaluation Report (SER . Therefore, plant-specific requirementsmay vary in accordance with the rule.

In summary, the resolution of A-24 is embodied in 10 CFR 50.49. A measure ofwhether each licensee has implemented the resolution of A-24 may therefore befound in the determination of compliance with 10 CFR 50.49. This was addressedby 72 SERs for operating plants issued shortly after publication of the ruleand subsequently in operating license reviews pursuant to Standard Review PlanSection 3.11. This was further addressed by the first-round environmentalqualification inspections conducted by the NRC.


By letter dated May 31, 1984, the licensee forwarded to the NRC the status ofthe Environmental gualification Program for NMP-1.

On January 10, 1985, the NRC issued a Safety Evaluation stating that "NHPC'sEquipment gualification Program is in compliance with the requirements of10 CFR 50.49, that the proposed resolution for each of the environmentalqualification deficiencies identified for Nine Mile Point, Unit No. 1 isacceptable and that continued operation of Nine Mile Point, Unit No. 1 willnot present undue risk to the public health and safety."

The schedule for implementation of 50.49 requirements for certain items wasextended by several NRC letters including the one of March 15, 1985 whichaddressed the emergency condensen isolation valve actuators. The licenseestates orally that initial implementation of all 50.49 requirements wascomplete on July 9, 1986 when these valves were made operable.

REFERENCES:


TITLE

DOR "Guidelines for EvaluatingEnvironmental qualification ofClass 1E Electrical Equipment inOperating Reactors"

NUDOCS NO.

NMP-1A-24

DATE

NUREG-0588, "Interim Staff Positionon Environmental gualification of .

Safety Related Electrical Equipment"

Commission Memorandum and Order,CLI-80-21, on DOR Guidelines andNUREG-0588

NUREG-0588, Revision 1

10 CFR 50.49 (48 FR 2730-2733)

Standard Review Plan 3.11,Environmental qualification ofMechanical and Electrical Equipment


12/79

05/23/80

07/81

01/21/83

07/81

TITLE

Letter C. V. Mangan (NMPC)to D. B. Yassallo (NRC)

Letter D. B. Vassallo (NRC)to B. G. Hooten (NMPC)

Letter H. R. Denton (NRC)to B. G. Hooten (NMPC)


TITLE

NUDOCS NO.

8502010710

8504030322

NUDOCS NO.

DATE

05/31/84

01/10/85

3/15/85

DATE


PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Wermiel

USI NO. A-36 TITLE Control of Heav Loads Phases I St IIMPA NO. C-10 C-15 TAC NOS. 08063

ISSUES SUMMARY:

This .USI was resolved in July 1980 with the publication of NUREG-0612, "Controlof Heavy Loads at Nuclear Power Plants," and Standard Review Plan (SRP) Section9.1.5. The staff established MPAs C-10 and C-15 for the implementation ofPhases I and II, respectively, of the resolution of this issue at operatingplants.

In nuclear power plants, heavy loads may be handled in several plant areas. Ifthese loads were to drop in certain locations in the plant, they may impactspent fuel, fuel in the core, or equipment that may be required to achieve safeshutdown and continue decay heat removal. USI A-36 was established tosystematically examine staff licensing criteria and the adequacy of measures ineffect at operating plants, and to recommend necessary changes to ensure the-safe handling of heavy loads. The guidelines proposed in NUREG-0612 includedefinition of safe load paths, use of load handling procedures, training ofcrane operators, guidelines on slings and special lifting devices, periodicinspection and maintenance for the crane, as well as various alternatives.

By Generic Letters dated December 22, 1980, and February 3, 1981 (GenericLetter 81-07), all utilities were requested to evaluate their plants againstthe guidance of NUREG-0612 and to provide their submitta'ls in two parts: PhaseI (six month response) and Phase II (nine month response). Phase I responseswer e to address Section 5.1.1 of NUREG-0612 which covered the following areas:

1.2.3.

5.

6.

7.

Definition of safe load pathsDevelopment of load handling proceduresPeriodic inspection and testing of cranesqualifications, training and specified conduct of operatorsSpecial lifting devices should satisfy the guidelines of ANSIN14.6.6.Lifting devices that are not specially designed should be installedand used in accordance with the guidelines of ANSI B30.9Design of cranes to ANSI B30. 2 or CMAA-70

Phase II responses were to address Sections 5.1.2 thru 5.1.6 of NUREG-0612which covered the need for electrical interlocks/mechanical stops, oralternatively, single-failure-proof cranes or load drop analyses in the spentfuel pool area (PWR), containment building (PWR), reactor building (BWR), otherareas and the specific guidelines for single-failure-proof handling systems.

As stated in Generic Letter 85-11, "Completion of Phase II of 'Control of HeavyLoads at Nuclear Power Plants' NUREG-0612," all licensees have completed therequirement to perform a review and submit a Phase I and a Phase II report.Based on the improvements in heavy loads handling obtained from implementationof NUREG-0612 (Phase I), further action was not required to reduce the risksassociated with the handling of heavy loads. Therefore, a detailed Phase IIreview of heavy loads was not necessary and Phase II was considered completed.

While not a requirement, NRC encouraged the implementation of any actionsidentified in Phase II regarding the handling of heavy loads that wereconsidered appropriate.

IMPLEMENTATION AND STATUS SUMMARY PLANT SPECIFIC):

The licensee responded to the NRC December 22, 1980 letter by several lettersbetween May 22, 1981 and November 25, 1985. For the purpose of USI A-36, theissue was completed by the last submittal made by the licensee on January 18,1985. The staff provided acceptance of the licensee's program by a SafetyEvaluation issued on March 5, 1985.

REFERENCES:


TITLE NUDOCS NO.

NMP-1A-36

DATE

Letter, Darrell G. Eisenhut, NRC,to all licensees, applicants forOLs and holders of CPs transmittingNUREG-0612 and staff positions

Generic Letter 85-11, Hugh L.Thompson, NRC, to all licensees forOperating Reactors, "Completionof Phase II of 'Control of HeavyLoads at Nuclear Power

Plants'UREG-0612"


12/22/80

06/28/85

TITLE

Letter D. P. Disc (NMPC)to D. G. Eisenhut

Letter D. P. Disc (NMPC)to D. G. Eisenhut

Letter T. E. Lemoges (NMPC)to D. Eisenhut

Letter T. E. Lemoges (NMPC)to D. Ei,senhut

Letter T. E. Lemoges (NMPC)to D. Eisenhut

Letter T. E. Lemoges (NMPC)to'. Eisenhut

Letter C. Y. Mangan (NMPC)to D. C. Vassallo (NRC)

Letter T. E. Lempges (NMPC)to Vassallo (NRC)


NUDOCS NO.

8108040118

109290460

8208050470

8205040529

8208050470

8310040597

8407310223

8501230433

8503210434

DATE

07/28/81

9/22/81

08/01/82

04/29/82

08/01/82

09/30/83

07/26/84

01/18/85

03/05/85

3 VERIFICATION DOCUMENTS:


0

C ~

PLANT NMP-1 OOCKET NO(S). 50-220

PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Kudrick

OSI NO. A-39 TITLE Determination of SRV Pool Dynamic Loads andTem erature Limits

NPA NO. TAC NOS.

ISSUES SUMMARY:

This USI was resolved with the publication of Standard Review Plan (SRP)Section 6.2.1.1.C, in October 1982. In addition, NUREGs 0763, 0783 and 0802were issued for Mark I, Mark II, and Mark III containments, respectively.

BWR plants are equipped with safety/relief valves (SRVs) to protect the reactorfrom overpressurization. Plant operational transients, such as turbine trips,will actuate the SRV. Once the SRV opens, the air column within the partiallysubmerged discharge line is compressed by the high-pressure steam released fromthe reactor. The compressed air discharged into the suppression pool produceshigh-pressure bubbles. Oscillatory expansion and contraction of these bubblescreate hydrodynamic loads on the containment structures, piping, and equipmentinside containment.

NUREG-0802 presents the results of the staff's evaluation of SRV loads. Theevaluation, however, is limited to the quencher devices used in Mark II and IIIcontainments. With respect to Mark I containments, the SRV acceptance criteriaare presented in NUREG-0661, "Safety Evaluation Report, Mark I Containment andLong-Term Program," and are dealt with as part of USI A-7.

SRP Section 6.2.1.1.C addresses the applicable review cr iteria, since all MarkII and III containment designs are understood to have completed their operatinglicense (OL) reviews subsequent to resolution of this USI and reflection of theresolution in the SRP.

IMPLEMENTATION AND .STATUS SUMMARY PLANT SPECIFIC):

As stated above, the SRV acceptance criteria are presented in NUREG-0661 forNine Mile Point Unit 1, (Mark I containment) and are dealt with as part ofUSI A-7.

REFERENCES:


TITLE

SRP 6.2.1.1.C, Pressure SuppressionType BWR Containments

NUDOCS NO.

NMP-1A-39

DATE

NUREG-0802, "Safety/Relief Valvequencher Loads: Evaluation for8WR Mark II and III Containments,Generic Technical Activity A-39"

NUREG-0661, "Safety Evaluation Report-Mark I Long Term Program"

1982

7/80


TITLE


NUDOCS NO.

8912050090

DATE

11/28/89



*The applicable SRP revision number would depend on the date of the evaluationfor each specific plant.

I+,~ ~

PLANT NMP-I DOCKET NO(S). 50-220

PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT W. Koo--

USI NO. A-42 TITLE Pi e Cracks in Roi1in Water Reactors

NPA NO. 8-06 TAC NOS. 69147

ISSUES SUMMARY:

This USI was resolved in February 1981 with the publication of NUREG-0313,Revision 1, "Technical Report on Material Selection and Processing Guidelinesfor BWR Coolant Pressure Boundary Piping." That NUREG document was issued toall holders of BWR operating licenses or construction permits and to allapplicants for BWR operating licenses. The staff established MPA B-05 forimplementation of the resolution at operating plants.

Pipes have cracked in the heat-affected zones of welds in prima y system pipingin BWRs since mid-1960. These cracks have occurred mainly in Type 304 stainlesssteel, which is the type used in most .operating BWRs. The major problem isrecognized to be intergranular stress corrosion cracking (IGSCC) of austeniticstainless steel components that have been made susceptible to this failure bybeing "sensitized," either by post-weld heat treatment or by sensitization of anarrow heat affected zone near welds.

"Safe ends" that have been highly sensitized by furnace heat treatment whileattached to vessels during fabrication were found to be susceptible to IGSCC inthe late 1960s. Most of the furnace-sensitized safe ends in older plants havebeen removed or clad with a protective material, and only a few BWRs still havefurnace-sensitized safe ends in use. Most of these, however, are in smallerdiameter lines.

Cracks reported before 1975 occurred primarily in 4-inch-diameter recirculationloop bypass lines and in 10-inch-diameter core spray lines. Cracking is mostoften detected during inservice inspections using ultrasonic test techniques.Some piping cracks have been discovered as a result of primary coolant leaks.

NUREG-0313, Revision 1 provided the NRC staff's revised acceptable methods forreducing the IGSCC susceptibility of BWR code class 1, 2, and 3 pressureboundary piping of sizes identified above and safe ends. In addition, itprovided the requirements for augmented inservice inspection of piping withnonconforming materials.

As a result of further IGSCC degradations in larger piping, the staff providedlicensees with additional requirements in several NRC communications (i.e.,Bulletins 82-03, 83-2, and 84-11). The long-term resolution of IGSCC in BWRpiping (including the scope of A-42) was provided in NUREG-0313, R'evision 2which was transmitted to all holders of BWR operating licenses via GenericLetter 88-01.


NMP-1A-42

By letter dated July 1, 1981, July 8, 1981, August 6, 1982, December 20, 1982,and February 17, 1983, the licensee provided information in response toGeneric Letter 81-04.

The safety evaluation on implementation of NUREG-0313, Rev. 1 was issued by thestaff on 06/06/84. For the purpose of responding to the USI A-42, as definedby the scope of GL 81-04/NUREG 0313, Rev. 1, this safety evaluation isconsidered to reflect the resolution of the USI by the licensee. The SE notesthat a certain NRC contractor's review concluded that not all of theNUREG-0313, Rev 1 guidelines had been found to be met. However that reviewdid not encompass later plant specific actions taken in at least four areas.Acknowledging these actions and the replacement of recirculation piping, thestaff's June 6, 1984 transmittal letter then provided the evaluation ofNUREG-0313 Rev. 1 compliance for information and assistance purposes in thelicensee's preparation of responses to staff initiatives (namely GL 84-11)beyond the scope of USI A-42. On this basis the staff's June 6, 1984 safetyevaluation is considered to reflect the programmatic conclusion of USI A-42issues by the licensee and the staff. This date also corresponds to theresolution of MPA B-05 on 06/84.

Going well beyond the scope of USI A-42, it is noted that the licensee respondedto GL 88-01 by letter dated July 28, 1988 as supplemented by letter datedAugust 25, 1989 and September 6, 1989. Technical Specification AmendmentNo. 107 which incorporates the requirements of GL 88-01 was issued on July 7,1989. The licensee plans to implement the requirements of GL 88-01 prior tothe next refueling outage per the July 28, 1989 letter. The licensee'ssubmittals are under review by the staff.To recap the discussion presented above for the 24 GWRs that were operatingwhen GL 81-04 was issued (February 1981, the implementation date for this USIhas been determined to be the date of the letter transmitting the staff'sevaluation of the licensee's response to GL 81-04. Hence, the date this USIwas implemented for NMP-1 is June 6, 1984.

'EFERENCES:


TITLE

NUREG-0313, Revision 1, "TechnicalReport on Material Selection andProcessing Guidelines for BWR

Coolant Pressure Boundary Piping,"

Generic Letter 81-04, "Implemen-tation of NUREG-0313, Rev. 1 forSelection and Processing Guidelinesfor BWR Coolant Pressure BoundaryPiping (Generic Task A-42)"


TITLE

Letter D. P. Disc (NMPC) toD. G. Eisenhart (NRC)

Letter T. E. Lempges (NMPC)to D. G. Eisenhart (NRC)

Letter C. V. Mangan (NMPC)to D. B. Vassallo (NRC)

Letter C. V. Mangan (NMPC)to D. 8. Vassallo (NRC)



Letter L. Burkhardt (NMPC)to NRC

Letter L. Burkhardt (NMPC)to NRC


NMP-1A-42

NUDOCS NO.

NUDOCS NO.

8107070307

8208130195

8212270265

8302230216

8406200403

8808040296

8909010320

8909200156

8912050090

DATE

07/80

02/26/81

DATE

07/01/81

08/06/82

12/20/82

02/17/83

06/06/84

07/28/88

08/25/89

09/06/89

11/28/89



I C

4

a

a


PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT P. GillUSI NO. A-44 TITLE Station Blackout

NPA NO. A-022 TAG NOS. 66570

ISSUES «SUMMARY:

This USI was resolved in June 1988 with the publication of a new rule (10CFR 50.63) and Regulatory Guide 1. 155.

Station blackout means the loss of offsite ac power to the essential andnonessential electrical buses concurrent with turbine trip and theunavailability of the redundant onsite emergency ac power systems. HASH-1400showed that station blackout could be an important risk contributor, andoperating experience has indicated that the reliability of ac power systemsmight be less than originally anticipated. For these reasons station blackoutwas designated as a USI in 1980. A proposed rule was published for comment onMarch 21, 1986. A final r,ule, 10 CFR 50.63, was published on June 21, 1988 andbecame effective on July 21, 1988. Regulatory Guide 1.155 was issued at thesame time as the rule and references an industry guidance document,NUMARC-8700. In order to comply with the A-44 resolution, licensees will berequired to:

maintain onsite emergency ac power supply reliability above a minimumlevel

develop procedures and training for recovery from a station blackout

determine the duration of a station blackout that the plant should be ableto withstand

use an alternate qualified ac power source, if available, to cope with astation blackout

evaluate the plant's actual capability to withstand. and recover from astation blackout

backfit hardware modifications if necessary to improve coping abilitySection 50.63(c)(1) of the rule required each licensee to submit a responseincluding the results of a coping analysis within 270 days from issuance of anoperating license or the effective date of the rule, whichever is later.IMPLEMENTATION AND STATUS SUMMARY (PLANT SPECIFIC):

By letter dated April 13, 1989, the licensee submitted a response as requiredin Section 50.63(c)(i) of the rule. In its response the licensee stated thatthe Class IE batteries evaluated in accordance with NUMARC 8700 guidelines weredetermined to be inadequate to meet station blackout loads for four hours.The licensee is performing an analysis to determine if additional loadstripping will allow Nine Mile Point Unit I to meet station blackout loads forfour hours or whether additional capacity will be required. The licensee hasnot yet finalized its plans and therefore, the results and plans for meetingstation blackout loads are not yet available.

I

0

o

'EFERENCES:


TITLE

10 CFR 50.63, "Loss of A11A1ternating Current Power"

Regu1atory Guide 1.155,"Station 81ackout"

NUDOCS NO.

NMP-1A-44

DATE

06/21/88

08/88


TITLE


NUDOCS NO.

8904240053

DATE

04/13/89



Ir I

0

en,

'LANT NMP-1 DOCKET NO(S). 50-220

PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT P. Y. Chen

USI NO. A-46 TITLE Seismic qualification of Equipment in OperatingPlants

NPA NO. 8-106 TAG NOS. 69461

ISSUES SUMMARY:

USI A-46 was resolved with the issuance of GL 87-02 on February 19, 1987, whichendorsed the approach of using the seismic and test experience data proposed bythe Seismic gualification Utility Group (SHRUG) and Electric Power ResearchInstitute (EPRI). This approach was endorsed by the Senior Seismic Review andAdvisory Panel (SSRAP) and approved by the NRC staff.The scope of the review was narrowed to equipment required to bring eachaffected plant to hot shutdown and maintain it there for a minimum of 72 hours.The review includes a walkthrough of each plant which is required to inspect equip-ment. Evaluation of equipment will include: (a) adequacy of equipmentanchorage; (b) functional capability of essential relays; (c) outliers anddeficiencies (i.e., equipment with non-standard configurations); and(d) seismic systems interation.

As an outgrowth of the Systematic Evaluation Program (SEP), the need wasidentified for reassessing design criteria and methods for the seismic quali-fication of mechanical equipment and electrical equipment. Therefore, theseismic qualification of the equipment in operating plants must be reassessedto ensure the ability to bring the plant to a safe shutdown condition whensubject to a seismic event. The objective of this issue was to establish anexplicit set of guidelines that could be used to judge the adequacy of theseismic qualification of mechanical and electrical equipment at operatingplants in lieu of attempting to backfit current design criteria for new plants.

Generic Letter 87-02 with associated guidance, required all affected utilitiesto evaluate the seismic adequacy of their plants. The specific requirementsand approach for implementation are being developed jointly by SgVG and thestaff on a generic basis before individual member utilities proceed withplant-specific implementation.


For NMP-1, the licensee performed the seismic review and trial walkdown ofNMP-1 in early 1988. By letter dated September 23, 1988, the licensee advisedNRC that it will complete the work scope contained in Revision 0 of the GIP bythe end of the next planned refueling outage which is dependent on NMP-1 dateof return to operation (scheduled for early 1990).

The next step in the resolution of this issue is dependent on staff action.

'EFERENCES: NMP-1A-46


TITLE

Generic Letter 87-02, "Verifi-cation of Seismic Adequacy ofMechanical and Electric Equipmentin Operating Reactors"

NUREG-1211, "Regulatory Analysisfor Resolution of Unresolved SafetyIssues A-46..."

NUREG-1030, "Seismic Qualificationof Equipment in Operating Plants,Unresolved Safety Issue A-46"

Letter attached with "GenericSafety Evaluation Report on SQUGGIP, Revision 0," from L. Shao(NRC) to Neil Smith (SQUG)

"Generic Implementation Procedure(GIP) for Seismic Verification ofNuclear Plant Equipment," Revision I

"Generic Implementation Procedure(GIP for Seismic Verification ofNuclear Plant Equipment," Revision 0


TITLE

Letter C. D. Terry (NMPC) to NRC


TITLE

NUDOCS NO.

NUDOCS NO.

2809280263

NUDOCS NO.

DATE

02/19/87

02/87

02/87

07/29/88

12/88

06/88

DATE

09/23/88

DATE

1

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PLANT NMP-I

PROJECT MANAGER Robert E. Martin


TECHNICAL CONTACT J. Mauck

USI NO. A-47 TITLE Safety Imp1ication of Controi Systems in LNRNuclear Power Plants

MPA NO. 6113 TAC NOS. 74966

ISSUES SUMMARY:

USI A-47 was resolved September 20, 1989, with the publication ofGeneric'etter

(GL) 88-19.

The generic letter states:

"The staff has concluded that all PWR plants should provideautomatic steam generator overfill protection, all BWR plantsshould provide automatic reactor vessel overfill protection, andthat plant procedures and technical specifications for allplants should include provisions to verify periodically theoperability of the overfill protection and to assur e thatautomatic overfill protection is available to mitigate mainfeedwater overfeed events during reactor power operation. Also,the system design and setpoints should be selected with theobjective of minimizing inadvertent trips of the main feedwatersystem during plant startup, normal operation, and protectionsystem surveillance. The Technical Specifications recommenda-tions are consistent with the criteria and the risk considera-tions of the Commission Interim Policy Statement on TechnicalSpecification Improvement. In addition, the staff recommendsthat all BWR recipients reassess and modify, if needed, theiroperating procedures and operator training to assure that theoperators can mitigate reactor vessel overfill events that mayoccur via the condensate booster pumps during reduced systempressure operation."

Also, page 2 of the generic letter provides for additional actions for CE andB&W plants. The generic letter provides amplifying guidance for licensees.

The generic letter requires that licensees provide NRC with their schedule andcommitments within 180 days of the letter's date. The implementation schedulefor actions on which commitments are made should be prior to startup after thefirst refueling outage, but no later than the second refueling outage,beginning 9 months after receipt of the letter.IMPLEMENTATION AND.STATUS SUMMARY (PLANT SPECIFIC :

NMP-1 is current1y evaluating the requirements of GL 89-19 and is expected toprovide its response by March 19, 1990.

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'EFERENCES:C

P

1. RE UIREMENT DOCUMENTS

TITLE

Generic Letter 89-19"Request for Action Relatedto Resolution of USI A-47"

NUREG-1217 "Evaluation of SafetyImplications of Control Systemsin LWR Nuclear Power Plants"

NUREG-1218 "Regulatory Analysisfor Resolution of USI A-47"


TITLE


NMP-1A-47

NUDOCS NO.

NUDOCS NO.

8912050090

DATE

09/20/89

June 1989

July 1989

DATE

11/28/89



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'LANT NMP-1 DOCKET NO(S). 50-220

PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J . Kudrick

USI NO. A-48 TITLE Hydrogen Control Measures and Effects of HydrogenBurns on Safety E ui ment

MPA NO.

ISSUES SUMMARY:

TAC NOS.

The NRC staff concluded April 19, 1989, that USI A-48 is resolved, as stated inSECY 89-122.

USI A-48 was initiated's a result of the large amount of hydrogen generatedand burned within containment during the Three Mile Island (TMI) accident.This issue covers hydrogen control measures for recoverable degraded coreaccidents for all BWRs and those PWRs with ice condenser containments.Extensive research in this area has led to significant revision of the Com-mission's hydrogen control regulations, given in 10 CFR 50.44, publishedDecember 2, 1981.

10 CFR 50.44 requires inerting of BWR Mark I and Mark II containments as amethod for hydrogen control. The BWR Mark I and Mark II reactor containmentshave operated for a number of years with an inerted atmosphere (by addition ofan inert gas, such as nitrogen) which effectively precludes combustion of anyhydrogen generated. USI A-48 with respect to BWR Mark I and II containments isnot only resolved but understood to be fully implemented in the affectedplants.

The rule for BWRs with Mark III containments and PWRs with ice condensercontainments was published on January 25, 1985. The rule required that theseplants be provided with a means for controlling the quantity of hydrogenproduced, but did not specify the control method. In addition, the task actionplan for USI A-48 provided for plant-specific reviews of lead plants forreactors with Mark III and ice condenser containments. Sequoyah was chosen asthe lead plant for ice condenser containments and Grand Gulf for Mark IIIcontainments. Both of the lead plant licensees chose to install igniter-typesystems which would burn the hydrogen before it reached threatening concentra-tions within the containment. Final design igniter systems have been installednot only in both lead plants, Sequoyah and Grand Gulf, but in all other icecondenser and Mark III plants as well. The staff's safety evaluations of thefinal analyses required to be submitted by these licensees by the rule arescheduled for completion in 1989.

Large dry PWR containments were excluded from USI A-48 because they have a, greater ability to accommodate the large quantities of hydrogen associated with

a recoverable degraded core accident than the smaller Mark I, II, III and icecondenser containments. However, this issue has continued to be consideredand, in 1989, hydrogen control for large dry PWR containments was identified asa high-priority Generic Issue (GI) 121. The resolution of GI 121 is beingactively pursued in close coordination with more recent research findings.

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NMP-IA-48

ISSUES SUMMARY (CONT.):

The NRC staff has concluded that USI-A-48 is resolved as stated in SECY89-122. If interested, the report should be consulted for further detailsregarding the relationship of A-48 to other ongoing hydrogen activities.IMPLEMENTATION AND STATUS SUMMARY (PLANT SPECIFIC):

Containment inerted. Containment Atmosphere Dilution System (CAD) added permodification Nl-72-03. Safety Evaluation documenting NMP-1 complies with10 CFR 50.44(c)(3)(ii) issued on April 29, 1985.

The licensee states orally that the capability to inert NMP-1 has existedsince the beginning of its operating life. Therefore the date of theProvisional Operating License, August 22, 1969, is utilized as the initialimplementation date for inerting of the containment.

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REFERENCES:


TITLE

10 CFR 50.44, Standards forCombustible Gas System inLight-Mater-Cooled PowerReactors

SECY-89-122, Resolution ofUSI A-48, "Hydrogen ControlMeasures and Effects ofHydrogen Burns on SafetyEquipment"


TITLE

Letter C. D. Terry, (NMPC) to NRC

NUDOCS NO.

NUDOCS NO.

DATE

12/81

04/19/89

DATE

11/28/89



Vy ~

o~ W ~

~Pa~~~ No. 1

02/06/90

LISTING OF INCOtlPLETE USI DATA

FOR INPUT FRON PROJECT tlANAGERS

ISSUE

HUNBER

ISSUE DESCRIPTIVE HAttE IttPLENEHT INPLEtlENT LICEHSEE COtNEHT STAFF CONNENT

DATE STATUS

NT MANE: HINE NILE POIHT 1lf PLA

A-01

A-02MATER HAtlNER

ASYNNETRIC BLQMDOMN LOADS OH

REACTOR PRIttARY COOLANT SYSTENS

MESTIHGHOUSE STEAN 6EHERATOR TUBE

INTE6RITY

CE STEAN 6ENERATOR TUBE INTE6RITY

B4M STEAN GENERATOR TUBE

IHTE6RITY

NARK I SHORT-TERN PR06RAN

NARK I LONG-TERN PR06RAN

NARK II CONTAINMENT POOL DYHANIC

LOADS - LONG-TERM PROGRAN

ATMS

BMR FEEDMATER NOZZLE CRACKIH6

REACTOR VESSEL NATERIALS

TOUGHHESS

FRACTURE TOU6HNESS OF STEAtl

GENERATOR AHD REACTOR COOLANT

PUttP SUPPORTS

SYSTENS INTERACTION

OUALIFICATIOH OF CLASS 1E

SAFETY-RELATED EQUIPtlEHT

REACTOR VESSEL PRESSURE TRANSIENT

PROTECTION

RHR SHUTDOMN REQUIREMENTS

CONTROL OF HEAVY LOADS NEAR SPENT

FUEL

DETERNIHATIOH OF SAFETY RELIEF

VALVE POOL DYNANIC LOADS AHD

TEttPERATURE LINITSSEISNIC DESIGN CRITERIA-SHORT-TERN PROGRAN

PIPE CRACKS IH BOILIH6 MATER

REACTORS

COHTAINtiENT ENER6EHCY SUNP

A-03

A-04

A-05

A-06

A-07

A-08

A-09

A-10

A-11

A-12

A-17

A-24

A-26

A-31

A-36

A-39

A-40

A-42

A-43

A-44

A-45

A„"46

PERFORNANCE

STATION BLACKOUT

SHUTDOMN DECAY HEAT REttOVAL"

REQUIRENEHTS

SEISNIC OUALIFICATIOH OF

/ / NC

/ / H/A

/ / N/A

SSFI

H/A

H/A

02/28/78 C

06/13/84 C

/ / H/A

i% ~06/04/83 C

/ / NC

/ / H/A

/ / HC

07/09/86 C

/ / H/A

/ / H/A

01/18/85 C

/ / NC SEE A-07

/ / NC

06/06/84 C

/ / HC

12/31/92 I/ / NC

/ / I

IPE

A-47

A-48

EOUIPNENT IH OPERATIH6 PLANTS

SAFETY INPLICATIOHS OF COHTROL 03/19/90 E

SYSTEMS 8 ~~/ay<HYDROGEN CONTROL NEASURES AHD

PMR OHLY

MESTIN6HOUSE ONLY

CE PLANTS ONLY

BRM PLANTS ONLY

DELTA P CONTROL

NK II BMR ONLY

CP AFTER 83 ONLY

HO REQUIRENENTS

EXTENSION GRANTED

PllR ONLY

HEM PLANTS ONLY.SR'L-85-11

ENDED

SUBSUNNED BY A-46

6L-88-01 H/A

INFO ONLY

SER 12/31/90SUBSUNED BY SEVERE ACC

REO UNDER DEVEL

HEM REOUIRENENTS

INERTED

A-49

EFFECTS OF HYDR06EH BURNS ON

SAFETY EOUIPllENT

PRESSURIIED THERNAL SHOCK / / N/A PMR ONLY

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responds to generic ltr 89-21 re status of implementation of usi … · 2018. 2. 7. · usi/mpa...

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