ML19332E797

Responds to Generic Ltr 89-21 Re Implementation Status of USIs

ML19332E797
Person / Time
Site:	Perry
Issue date:	12/01/1989
From:	Kaplan A CLEVELAND ELECTRIC ILLUMINATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-89-21, PY-CEI-NRR-1088, NUDOCS 8912120160
Download: ML19332E797 (7)

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THE, CLEVELAND ELECTRIC ILLUMINATING COMPANY i

• P.o. 90X 91 3 PERRY, oHlo 44001 B TELEPHONE (216) 2643737 5 ADDRESB 10 CENTER ROAD FROM CLEVELAND: 4741260 3 TELEX: 241699 )

ANSWERBACK: CEIPRYo j Al Kaplan Serving The Best location in the Nation I

PERRY NUCLEAR POWER PLANT vet entum ,

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December 1, 1989 I PY-CEI/NRR-1088 L I

1 U.S. Nuclear Regulatory Commission Document Control Desk Vashington, D. C. 20555 l

Perry Nuclear Power Plant l Docket No. 50-440 1 Implementation of Unresolved Safety Issues (Generic Letter 89-21)

Gentlemen:

( The subject Generic Letter requested completion of the attached Table 1 to

determine Unresolved Safety Issue (USI) implementation status. PNPP status on I

each USI is shown in accordance with your instructions.

If you have any questions, please feel free to call.

Very truly yours G

Al Kaplan Vice President Nuclear Group AKinjc cc: T. Colburn P. Hiland i Region III hg

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UNRESOLVED SAFETh ISSES FOR latfCN A FINAL TECINilCAL RESOLUTION HAS SEEN ACHIEVED i .

(ISI/MPA #EsIAAKS NUM8ER TITLE REF. 90C N NT APPLICABILITY STATUS /DATE* .

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- l A-1 Water Hammer SECY 84-tI9 AlI g- m .

l MIEEG-0927.'Rev. 1 l IRIREG-0993. Rev. I i

NilREG-0737 Item I.A.P.3 i l

SAP revisiens

! A-2/ Asymmetric Slowdown MMEG-0609 PWR wa ya MPA D-10 Loads on Reactor Primary Gt. 94-04. GDC-4 Coolant Systems l A-3 Westinghouse Steam NISEE-0044 W-PWR WA WA Generator Tube Integrfty SECY 86-97

! SECY.88-77; I GL 35-02

! (We requirements)

A-4 CE Steam Generator Tube MftEG-0044. SECY 86-97 CE-PWR ,

WA WA Integrity SECT 88-272 l St. 85-02 (We regelrements) l A-5 B&W Steam Generator MS-0044. SECY 86-97 44W-PWP

1. A **

Tube Integrity SECY 88-272 l GB. 85-02 (IIe acquirements)

E Mark I-8 Wit WA WA i A-6 Mark I Corstafament IlllREG-0408 ,

j Short-Tens Program 1

• C - COMPLETE

kC - NO CHAfIGES NECESSARY NA - NOT APPLICA8tE

. I - INCOMPLETE E - EVALUATING ACTIOlis REQUIRED 1

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PT-G IMEEl-10Mt.

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1%e 2 of 6 s USI/MPA -

NUMBER TITLE REF. DoctrENT APPLICA81LITY STATUS /DATE* RENURS A-7/ Mark I long-Term NUREG-0661 Mark I-BMt wA WA D-01 Program NUREG-0661 Suppl. I R 79-57

• A-8 Mark II Containment NUREG-0808 Merk II-BWR WA Pool Dynamic Loads WA NUREG-0487, Suppl. 1/2 NUREG-0802 '

SPP 6.2.1.1C GOC 16 A-9 Anticipated Transients NUREG-0460. Vol. 4 All c;5/11/89 RxMmrE l Wfthout Scram 10 CFR 50.62 l

A-10/

SWR Feedwater Mozzle NUREG-0619 BWR MPA B-25 Cracking Letter from DG Eisenhut RXMetE 2 dated 11/13/80 R 81-11 1

A-Il Reac'or Vessel Material NUREG-0744. Rev. I All w^

i Toughness RXMWfE 3

• 10 CFR 50.60/ '

82-26 4

A-12 Fracture Toughness of *NUREG-0577, Pev. 1 PNP wA WA

) Steam Generator and SRP Nevision Reactor Coolant Pump 5.3.4 -  :

Supports i A-17 Systeus Interacffons Ltr: DeYoung to All IC IRC resolvei this ismae 11censees - 9/72 with no specific action NUREC-1174. NL9EG- tequired, however, in- l 1229 NUREG/CR-3922, sights from the A-17

• NURES/CR-4761, NL9EG/ program are being an-  ;

CR-4470, R 89-18 siderei as part of -

(No regufrements) the IMPP IFE proge m . I

, A-24/ Ovalf fIcation of Class - MREG-0588, Rev.1 All - c;9/85 RnMwig 4

! MPA B-60 IE Safety-Related SRP 3.11  :

Equipment 10 CFR 50.49  :

R 82-09. R 84-24 .

l E 85-15 i a

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Attadment Page 3 of 6 ..

USI/MPA STATUS /DATE* REMARKS' TITLE REF. DOCtMENT . APPLICABILITY.

NUMRER DDR Letters'to PWR. N/A WA A-26/ Reactor Vessel Pressure MPA B-04 Transient Protection Licensees 8/76 NUREG-0224-NUREG-4371 SRP 5.2 GL 88-11 J

Residual Heat Removal NUREG-0606 All Ots After uc  %

A-31 01/79.

Shutdown Requirements RG 1.113 RG 1.139 SRP 5.4.7 Control of Heavy Loads NUREG-0612 All C;12/85 Myses <mpleted per-  :

A-36/ SRP 9.1.5 1etter dated 11/8/82, C-10, Near Spent Fuel GL 81-07, GL 83-42, Procedures impleamted .

C-15 12/85.

GL 85-11 Letter from DG Eisenhut dated e 12/22/80 C;11/85 A-39 Determination of SRV NUREG-0802 BWR m w ieted per j Pool Dynamic Loads and Pressure Transients NUREGs-0763,0783,0802 NUREG-0661 ft 1 [

SRP 6.2.1.1.C Seismic Design SRP Revisions MUREG/ All N/A Althot@ A iO is not app--

A-40 CR-4776, NUREG/CR-0054, licable, design of above Criteria NUREG/CR-3480, NUREG/ gtu mi vertical tanks CP-1582,NUREG/CR-1161, vill be A w as part ~..

NUREG-1233, NUREG-4776 of the IMP IPE for ex- '

P4 REG /CR-3805 temal evets (ImE).

RU9FG/CR-5347 MREG/CR-3509 .

Bh? I;6/92 m

• D ' *I8' Pipe Cracks in Boiling NUREG-0313. Rev. 1 - t A-42/ NUREG-0313. Rev. 2

• U ****'8 h MPA B-05 Water Reactors ment m RV nozzles per GL 81-03* GL 88-01 FT-GUNR-104L dated 7/31/89.

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USI/ PPA TITLE APPLICA8ILITY NUMBER REF. DOCLMENT STATUS /DATE* REMARKS A-43 Containment Emergency NUREG-0510, All PC None Sump Perfonnance NUREG-0869, Rev. 1 NUREG-0897, R.G.I.82 (Rev. 0), SRP 6.2.2 GL 85-22 No Requirements

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A-44 Station Blackout RG 1.155 All I;SER + 1 yr IWDUIE 5 NUREG-1032 NUREG-1109 10 CFR 50.63 A-45 Shutdown Decay Heat SECY 88-260 All iC I w DUIE 6 Removal Requirements NUREG-1289 NUREG/CR-5230

SECY 88-260 -

2 (No. requirements)

A-46 Seiseic Qualiffcation NUREG-1030 All wA E00DUIE 7

of Equipoent in - NUREG-1211/

j Operating Plants GL 87-02, GL 87-03 A-47 Safety Implication NUREG-1217, NUREG- All E;3/20/99 G.,89-19 ic pu.= due by of Control Systems 1218 3/20/90 GL 89-19 1

j A-48 Hydrogen Control 10 CFR 50.44 All, except I;SER + 6 nos RUDUIE 8 i Measures and Effects SECY 89-172 PWRs with j of Hydrogen Burns large dry

on Safety Equfpment containments i

l A-49 Pressurized Thermal RGs 1.154, 1.99 PNR wA wA Shock SECY 82-465 SECY 83-288 SECY 81-687 10 CFR 50.61/ .;

GL 88-11

, ,s Attcchment 1 PY-CEI/NRR-1088 L Page 5 of 6 CEI Footnotes to USI Tabulation

1. Letter PY-CEI/NRR-0190L, dated 3/19/85 confirms that ATWS design changes were installed prior to fuel load. Licensing closure documented in NRC letter dated 3/15/89, pending completion of 2 pump test of SLCS. Letter PY-CEI/NRR-0993L dated 5/11/89 confirmed completion of the pump test.

2. HEDE-21821 describes the PNPP Feedvater Nozzle /Sparger design. SER Section 3.9.3.1 found this desip acceptable. A PNPP-specific fracture j mechanics analysis (NEDC-30827) on the feedvater nozzles was completed in December 1984, which verified acceptable results to meet the criteria in  ;

Generic Letter 81-11, 4

3. Letter PY-CEI/NRR-0930L dated 11/17/88 provided verification that, based  ;

on the prediction methods in Reg. Guide 1.99, Rev. 2, the adjusted reference temperature at the 1/4 T position in the vessel vall is less than 200 degrees F at the end of plant life, and the shelf energy for i critical materials in the vessel belt line remains above 50 ft-lbs at end l of plant life. Since A-11 addressed plants with shelf energies predicted i

l to fall 1 Gov 50 ft-lbs, this issue is N/A to PNPP. Should future calculat et.4 predict lover shelf energies (below 50 ft-lbs), 10CFR50.60 and 10CFR50 Appendix G require the necessary evaluations be performed.

4. PY-CEI/NRR-0355L dated 9/23/85 provided verification that the EQ program 1 l, was in place and that all equipment within 50.49 scope was qualified.

5. PY-CEI/NRR-0995L, dated 4/17/89, provided results of PNPP analysis and

. proposed changes (implementation required within 1 year after NRC accepts .

plan).

6. NRC resolved this issue with no specific actions required, however, A-45 l concerns on shutdevn decay heat removal are being considered as part of-the PNPP IPE program.

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7. Although A-46 is not applicable to PNPP, the methodology from this issue vill be used to conduct a seismic margins walkdown. This effort vill (1) resolve an ACRS commitment and (2) help resolve the seismic portion of j IPEEE.

8. PNPP-specific resolution of this issue vill be completed following NRC  ;

l safety evaluations of the Hydrogen Control Ovners Group program and a I; topical report on equipment survivability.

1, NJC/ CODED /2886 6

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IT-GUtem-1088L I Attarfaunt Page 6 of 6 NRC GUIDE FOR UPDATING USI STATUS (1) Enclosure 1 lists all unresolved safety issues (USIs) for ,

which a final technical resolution has been achieved. Please review the entire listing for each licensed reactor unit. Where an item is not -

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applicable for your facility, mark "NA" in the status column.

1 (2) Where an item is applicable to your facility, but no changes were necessary, mark "NC" in the status column.

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-l (3) Where an item is applicable to your facility and changes are complete, mark "C" in the status column and indicate month and year implementation was complete, including reference to any supporting documentation. -

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.(4) Where an item is applicable to your facility and is not fully implemented, provide your pro,iected implementation date (month and year) c and a short note identifying the outstanding item (e.g., hardware, ,

l L .procedures, training, Technical Specifications). Mark "I" for incomplete.

i (5) Where a USI resolution was only recently issued, such as A-40 and A-47, and you are evaluating your response, identify expected response date and indicate "E" in the status column.

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