ML19332C921

Forwards Response to Generic Ltr 89-21 Re Implementation Status of USI Requirements

ML19332C921
Person / Time
Site:	FitzPatrick
Issue date:	11/17/1989
From:	Brons J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-MI, REF-GTECI-SC, TASK-***, TASK-OR GL-89-21, JPN-89-075, JPN-89-75, NUDOCS 8911290175
Download: ML19332C921 (9)

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123 Main Street -  ;

F Wtute Plains, New York 10601 914 081.O 40 Q .7 .

t M h f' Johti C. Brons I f 4ff Authority. RfM22 P"' l a

November 17 ,1989 JPN-89-075 ,

U.S. Nuclear Regulatory Commission .l ATTN: Document Control Desk Mall Station P1-137 Washington, D.C. 20555

SUBJECT:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Response to Generic Letter 89-21 Implementation Status of i Unresolved Safety issue Requirements Referenco: NRC Generic Letter 89-21, " Request for Information Concerning Status of implementation of Unresolved Safety issue (USI)

Requirements," dated October 19,1989.

Dear Sir:

The referenced Generic Letter requested that the Authority provide information .

concerning FitzPatrick specific implementation of NRC unresolved safety issues (USI). Attachment I to this letter provides the closure status of each USl Identified in Generic Letter 89-21.

Should you or your staff have any questions regarding this matter, please contact Ms. S. M. Toth of my staff.

Very truly yours, <

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John C. Brons Executive Vice President Nuclear Generation jj cc: See next page.

l (Il I'tl l 8911290175 891117 i PDR ADOCK 05000333 i p- PDC

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p r. 1 cc: U.S. Nuclear Regulatory Commission -

475 Allendale Road -

King of Prussia, PA 19406 ,

Office of the Resident inspector U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 ,

Mr. David E. LaBarge

- Project Directorate 1-1 Division of Reactor Projects 1/11 U.S. Nuclear Regulatory Commission -

Mail Stop 14 B2 Washington, D.C. 20555 e

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New York Power Authority James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 DPR-59 i

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UNRESOLVED SAFETY ISSUES FOR WHICH A FINAL TECHNICAL RESOLUTION HAS BEEN ACHIEVED

- USI/MPA NUMBER TITLE REF.-DOCUMENT . APPLICABILITY l STATUS /DATE* ' REMARKS

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A-1 Water Hanner 'SECY 84-119 .

All C- Note 1.-

NUREG-09?7, Rev. 1 11/4/86 NUREG-0993. Rev. 1 NUREG-0737 Item I.A.P.3 SRP revisions 4

A-2/ Asyneetric R10wdown NUREG-0609 PWR MPA D-10 Loads on Reactor Primary GL 84-04. GDC-4 NA Cociant Systems ,

A-3 Westinghouse Steam NUREG-0844 W-PWR

Generator Tube Integrity SECY 86-97 NA i

SECY.88-?72 GL 85-02 (No requirements)

A-4 CE Steam Generator Tube NUREG-0844, SECY 86-97 CE-PWR gg Integrity SECY 88-272-GL 85-02 (No requirements).

A-5 R8W Steam Generator NUREG-0844. SECY 86-97 R&W-PHP NA Tube Integrity SECY 88-272 GL 85-02 (No Requirements) 4

E i  %-6 Mark I Containment NUREG-0408 Park I-BWR c- Note 2*

Short-Term Program 09/28/78

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• C.- COMPLETE NC - NO CHANGFS NECESSARY NA - NOT ArPLICABLE Is- INCOMPLETE E - FVALUATING ACTIONS RET)UIRED

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.USI/MPA .

NUMBER TITLE. REF.. DOCUMENT APPLICABILITY- . STATUS /DATE* REMARKS" A-7/ Mark I long-Term NUREG-0661 Mark I-BWR C-NUREG-0661 Suppl._.1 Note 3..

D-01 Program GL-79-57 -12/12/84 A-8 Mark II Containment NUREG-0808 .

Mark II-BWR.

NUREG-0487,. Suppl. 1/2 NA

' Pool Dynamic loads NUREG-0802 SPP 6.2.1.1C GDC 16

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A-9 Anticipated Transfents NUREG-0460, Vol.'4 'All I Note 4.

Without Scram 10 CFR 50.62 A-10/ BWR Feedwater Nozzle NUREG-0619 BWR C Note 5.

MPA B-25 Cracking Letter from DG Eisenhut 04/27/84 dated 11/13/80 GL 81-11 A-11 Reactor Vessel Material NUREG-0744, Rev. 1 All c Note 6.

Toughness 10 CFR 50.60/ 04/30/86 82-26 A-12 Fracture Toughness of NUREG-0577,.Rev. 1 PWP NA Steam Generator and SRP Revision Reactor Coolant Pump 5.3.4 Supports A-17 Systems Interactions Ltr: DeYoung to All E Note 7.

Ifcensees - 9/72 12/30/90 -

NUREC-1174. NUREG-1229, NUREG/CR-3922, NUREG/CR-4761. NUPEG/

CR-4470, GL 89-18 (No requirements)

A-24/ Qualification of. Class NUREG-0588, Rev. 1 All C Note-8~.

MPA B-50 1E Safety-Related SRP 3.11 03/12/85 Equipment 10 CFR 50.49 GL 82-09. GL 84-24 GL 85-15

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USI/MPA .

NUM8ER TITLE -REFf DOCUMENT ~ .' APPLICABILITY -STATUS /DATE* REMARKS

-A-26/ Reactor Vessel Pressure- por. Letters to PWR NA MPA B-04 Transient Protection Licensees 8/76 NUREG-0224 NUREG-0371 SRP 5.2

-GL 88-11 A-31 Residual Heat Removal NUREG-0606 A11 Ots After NA Shutdown Requirements RG 1.113, 01/79.

RG 1.139'

_SRP 5.4.7 l .A-36/ Control of Heavy Loads NUREG-0612 All C 160te 9

• C-10, Near Spent Fuel SRP 9.1.5- Phase I 02/11/85 C-15 GL 81-07, GL 83-42 Phase II 06/28/85 GL 85-11 Letter from DG Eisenhut dated 12/22/80 A-39 Determination of SRV NUREG-0802 BWR c Note 10.

Pool Dynamic Loads NUREGs-0763,0783,0802 and Pressure Transients NUREG-0661 ~

SPP 6.2.1.1.C A-40 Seismic Design SRP Revisions, NUREG/ All I

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• Criteria CR-4776 NUREG/CR-0054, 09/89 NUREG/CR-3480, NUREG/

CR-1582, NUREG/CR-1161, NUREG-1233, NUREG-4776 NUREG/CR-3805 NUREG/CR-5347 NUREG/CR-3509 A-42/ Pipe Cracks in Boiling NUREG-0313. Rev. 1 BWR c Note 12.

MPA B-05 Water Reactors NUREG-0313, Rev. 2 06/20/84 GL 81-03, GL 88-01 1

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-NUMBER TITLE REF. DOCUMENT . APPLICABILITY' STATUS /DATE* REMARKS A-43 Con +ainment Emergency NilREG-0510 All 'c Sump Performance NIJREG-0869,.Rev. 1 12/3/89 Note 13..

NilREG-0897, R.G.1.87

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(Rev. 0),'SRP'6.2.2 GL 85-22 No Requirements A-44 Station Blackout RG 1.155 All I -- Note 14.-

NilREG-1032 NilREG-1109 s 10 CFR 50.63 A-45 Shutdown Decay Heat- SECY 88-260 All I- 'N o t e ' 1 5'.

Removal Requirements NUREG-1289- 12/30/90 NUREG/CR-5230 SECY 88-260 (No requirements)-

A-46 Seismic Qualification NUREG-1030 All I Note 16.

of Equipment in NUREG-1211/

Operating Plants GL 87-02, GL B7-03 A-47 Safety Implication NUREG-1217, NOREG- All E . Note 17.

of Control Systems 1218'. 3/19/90 GL 89-19 A-48 Hydrogen Control 10 CF# 50.44 All, except NC N te 18.

Measures and Effects SECY 89-17? PWRs with of Hydrogen Burns large dry on Safetv Equipment containments A-49 Pressurized Thermal PGs 1.154, 1.99 PWR' NA Shock SECY 82-465

. SECY P.3-288

.. SECY 81-687 10 CFR 50.61/

GL 88-11 -

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s NOTES <

(Sheet 1 of 2) [

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1. The Authority. received notification by letter, dated November (

4,--1986, 'from the National Academy for Nuclear Training for  ;

,, the accreditation of training programs at ;its nuclear power plants,. James-A. FitzPatrick and Indian Point 3.

2. NYPA: letter, dated September 28, 1978, to the NRC -

Multiple

, Subsequent Safety Relief Valve Actuations Plant Unique -i Assessment.

3.: NRC letter,' dated December 12, 1984, to NYPA, " Post Imple-mentation Audit Review of Unique Analysis Report for Mark I Containment Long Term Program -

Program Found Acceptable"-

([[::JAF-84-364|JAF-84-364]]). .,

The NRC conducted an inspection (89-01)- in January 1989 to determine if the F2.zPatrick plant had modified the. primary containment . boundary in accordance with USI A-7. The inspector concluded that the work was performed using appropriate methods. .This information can be found in an NRC 4

, letter dated February 15, 1989.

4. The Alternate Rod Injection'(ARI) System was considered to be "

in compliance- with 10 CFR 50.62 by NRC letter dated November-18, 1988 ([[::JAF-88-351|JAF-88-351]]).

The FitzPatrick plant was granted an exemption from 10 CFR

50. 62 (c) (4 ) by NRC letter. dated. November 17, 1988, Amendment

'116 ([[::JAF-88-337|JAF-88-337]]).

. A projected completion date for diversity provisions of the  ;

Recirculating Pump RPT logic Trip (RPT)illdesigns is being reassessed be submitted at this. time. change w during the second quarter of 1990.

5.. lBy NRC letter dated July 21, 1986, the FitzPatrick plant,

p. " conforms to the requirements of'NUREG-0619 and Generic Letter l' 81-11."

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6. By letter JPN-86-022 dated April 30, 1986, NYPA informed the NRC lof the analysis performed according to 10 CFR.50 Appendix

H, performance results to limits given in 10 CFR 50 Appendix G, and that the End-of-Life Charpy upper-shelf energy is no L less than 50 ft-lb. This is monitored according to the  ;

L reactor vessel surveillance program.  ;

7. This USI will be considered as part of the FitzPatrick I m Individual Plant Evaluation which is due to the NRC on December 30, 1990 (JPN-89-069, dated October 27, 1989).

p 8. NYPA was informed by the NRC that Equipment Qualification, L relative to safety related equipment, was satisfactory R ([[::JAF-85-073|JAF-85-073]], dated March 12, 1985, SE enclosed). ,

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t NOTES (Sheet 2 of 2). 1 L 9. .- The NRC' issued a Safety Evaluation ([[::JAF-85-044|JAF-85-044]] on February  ;

11,- 1985 which concluded that'the> guidelines)of NUREG-0612, '

Sections 5.5-1 and 5.3 were satisfactory, and Phase I of this issue for the FitzPatr1ck plant was acceptable.  ;

Phase II of this issue was considered complete by the NRC on

= June.28, 1985 as per Generic Letter 85-11 ([[::JAF-85-196|JAF-85-196]]) . l

10 . Reference USI A-7. I

11. Resolution of this item has been included under the scope- of i USI A-46. -US11 A-46 is being resolved-on a_ generic-basis by I the Seismic Qualification Utility Group (SQUG). i

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12. The NRC issued a Safety Evaluation on June 20, 1984. It ',

closed this' issue in. review of response to Generic Letter 81-03. The SER stated that the determination of compliance with Generic Letter 81-03, Revision -1,- guidelines is not germane due to Generic Letters 84-07 and 84-11'..

13. .This issue-was closed as a result of Generic Letter 85-22 dated December 3, 1985.

14 . _ Implementation schedule is to be provided within 30 days of -

NRC notification according to 10 CFR 50.63 (c) (3) (J PN-8 9 -018 , _

dated April-17, 1989.

15. This USI will be considered as part of the FitzPatrick Individual Plant Evaluation, which is due to the NRC on December 30, 1990 (JPN-89-069, dated October 27, 1989).

16 . - This item is being resolved on a generic basis by the SQUG.

Implementation schedule is based on completion of final SQUG Generic Implementation Procedure- (GIP) and issuance of the Safety Evaluation' Report by the NRC.-

17. The expected response date, according to Generic Letter 89-19, is March 19, 1990.

18. This issue is part of the initial plant design Technical Specifications for the FitzPatrick plant.

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