ML20033E049

Forwards Status Summary of Implementation of USIs at Plant Consisting of Info Provided by Util in Response to Generic Ltr 89-21,status Summary for Each USI Applicable to Facility & NRC Data Base Printout for Plant

ML20033E049
Person / Time
Site:	Dresden, Byron
Issue date:	02/20/1990
From:	Olshan L Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17202J231	List: ML17202J179 ML20033E049 ML20033E061 ML20033E232 ML20033E239 ML20033F189
References
REF-GTECI-A-09, REF-GTECI-A-44, REF-GTECI-A-47, REF-GTECI-A-49, REF-GTECI-EL, REF-GTECI-RV, REF-GTECI-SY, TASK-A-09, TASK-A-44, TASK-A-47, TASK-A-49, TASK-OR GL-89-21, NUDOCS 9002260061
Download: ML20033E049 (2)

Text

,

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UNITED STATES DD I

NUCLEAR REGULATORY COMMISSION WASMiseGTON, D. C. 30665 February 20, 1990

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Docket Nos. 50-454

)

and 50-455 MEMORANDUM FOR:

File FROM:

Leonard N. 01shan, Project Manager

. )

Project Directorate 111-2 Division of Reactor Projects - III, i

IV, V and Special Projects

SUBJECT:

STATUS OF IMPLEMENTATION OF UNRESOLVED SAFETY ISSUES AT BYRON STATION, UNITS 1 AND 2 l

t Thecurrentimplementationstatusofunresolvedsafetyissues(USIs)at the Byron Station, Units 1 and 2, is set forth in the enclosures to this memorandum.

j Enclosure I contains of a copy of the information provided by the licensee in

.its response to Generic Letter 89-21.

In addition, Enclosure 2 contains a status summary for each USI applicable to this facility. This status sumary is based upon the licensee's response to the Generic Letter, discussions with the licensee, and my review of available l

NRC records and information. Appropriate NRR technical branches have also rr. viewed the USI status summary and this memo.

For those items that are incomplete, my assessment of safety significance is 4

as follows:

USI NUMBER TITLE STATUS A-9 ATWS per 10 CFR 50.62 To be implemented during next refueling outages A-44 Station Blackout Licensee's 4/17/89 response is under staff review A-47 Safety Implication of Generic Letter 89-19 was recently sent I

Control Systems in LWR to licensee; response due 3/90 Nuclear Power Plants A-49 Pressurized Thermal Licensee's 1/17/86 response is under Shock staff review for Unit 2 1

)

quaut-4 (sry)

a j

File 2

February 20, 1990 L

' is a copy of the staff's data base printout for the Byron facility.

It reflects the staff's assessment of USI ir.plementation for all 27 USIs.

It is based on review of the licensee's response to Generic Letter i

89-21, and evaluation by project managers, the USI team, and NRR technical

'l l

staff.

It is my conclusion that.there is no urgent safety concerns for these USIs while awaiting completion, j

i b.:

Leonard N. 01shan, Project Manager Project Directorate III-2 Division of Reactor Projects - Ill, j.-

IV, Y and Special Projects

Enclosure:

1

- As stated

-l cc w/ enclosures:

R. Wessman K. Eccleston R. Herman R. Scholl W. Paulson P. Shemanski 1

1 5

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E. tJ c t. 0 5 0 g E 1.

@ Commonwealth Edison 72 West Adams Stteet. Chicago, Ilknois_

l A00'ebs Fieply to Post O'hCe Box 767 Chicago Ittnois 60c90 0767 November 28, 1989 1

Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation i

U.S. Nuclear Regulatory Commission l

Hashington, DC 20555 i

Subject:

Byron Station Units 1 and 2 l

Braidwood Station Units I and 2 i

Response to Generic letter 89-21 l

NRC Docket Nos. 50-454/0 5 and 50-456/457 j

Reference:

Generic Letter 89-21 Request for Information Concerning Status of Implementation of Unresolved Safety Issue (USI) Requirements i

J

Dear Dr. Hurley:

1 Generic letter 89-21 was issued as a part of the NRC's continuing effort to validate staff understanding regar. ding implementation of unresolved 1

safety issues (USIs). An important aspect of this effort is to ensure that the licensee and the NRC agree on the status of USI resolution implementation at each facility.

This letter presents Commonwealth Edison (Edison's) response to Generic Letter 89-21, for the Byron and Braldwood Stations.

Byron and Braidwood stations have reviewed the guidance'provided in the Generic Letter.

The attached of USIs tables (Enclosure i from the generic letter) hrve been updated to reflect the current status of USI implementation.

The generic letter's guide for updating USI status was utilized for this process.

Accordingly, Attachments "A" and "B" of this letter provide the status of the efforts at Byron and Braidwood Stations, respectively.

Pleae direct any questions regarding this response to this office.

Very truly yours, (Y

k S.C. Hunsader Nuclear Licensing Administrator

-09tth 00e4 891228 PDR ADOCK 05000454 P

PNU

/sc1:0351T:28 Attachments: A-Byron Station B-Braidwood Station y

g WW cc:

A.B. Davis-RIII Resident inspectors-Bf/BH s

L. Olshan-NRR C C a n,l e 4D0

,1 ATTACIWWWT "A" Lg LYnON STATION EPQ.M ' ]

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WIIItESRVED SAFETY ISSUES FOR ist8'It A FTP.At TEC8mlCAL RESOLUTitin IInt BEEN AC3ffEV IISI/MPA' NWHER TITLE REF. fl0CIMENT APPLICAOILITY STAitf5/94TE* -

_REsguisl5 A-1 Water Manner SECY 84-119 All c 10/81 a

i MMEG-09?7. Rev.1 filREC-0993. Ilev. I IIUREG-0737 Item l

I.A.7.3 SRP revfsfens A-2/

AsymmetrfC' R1owdown IIUREG-0609 PMt c el 3/87 synew sen MPA D-10 Loads on Reactor Primary GL 44-04. 60C-4 s2 11/86 sucT1on 3.9.2.4 Coolant Systems A-3 Westinghouse Steen RUREG-0444 W-PWit c s1 6/85 anC ACCErTANCE LTR.

Generator Tube Integrfty SECY 86-97 U2 II/M J M s/22/s5 I

l SECY 88-772 GL 85-02

,3 (no regetrements)

A-4 CE Steam Generator Tube IIUREG-0044. SECY 96-97 CE-PtfR Integrity SECY 88-272 i

Gl. 85-02 (IIe requframents)

A-5 88W Steese Generator IIUREr.-0844. SECY 96-97 481f-PWP

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Tube Integrity SECY 48-272 GL 85-02 (lle %frements) -

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I X-6 Mart I Contafanent IstREG-04n8 Part I-glm

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Short-Term Program i

5 i

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• C - COMPLETE j

NC - NO CHAIIGF5 NECE55AltY NA - IIOT APPLICA8tE j

I - INCOMPLETE E - FVALUATIIIS ACTIGNS RFolfTRFR i

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USI/MPA

_MIMRTR TITLE REF. HOEtPEltT APPLICASitITT STATU5/ SATE

• REMARES A-7/

Mark I long-Tern MREG-0661 Mark I-9NR NA i

l D-01 Program NUREG-0661 Suppl. 1 l

R 79-57

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A-8 Mark II Contefament NUREG-0908 Merk II-9WR NA Pool Dynamic _ toads NUREG-0487. Suppl. 1/2 i

NUREG-0802 l

SPP 6.2.1.1C GDC 16 l

A-9 AntIcfpated Trans1ests NUREG-0460. Vol. 4 AlI I e1-3/9e scsutset.s Nfthout Scram 10 CFR 50.62 s2-10/90 scantTTre 2-15-89

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A-10/

BNR Feedwater Mozzle NUREG-0619 DNR NA MPA B-75 Cracking Letter free DG Efsenhut i

dated 11/13/80 l

GL al-11 i

b A-II Reactor Vessel Material NUREG-0744. Rev. I All c st-to/s4 staca ssaa #4 Tooghness in CFR 50.60/

v2-1I/s6 Pc. 5-3 n2-26 A-l?

Fracture Teoghness of NUREG-0577. Rev. I PNP c 2/s2 BYaos sta l

Steam Generator and SRP Nevision c-13 Reactor Coolant Pump 5.3.4 Supports f

A-17 Systems Interactfees Ltr: DeYoung to All E 10/92 ct,as-20 Ifcensees - 9/72 Irz arsrons i

NUREC-Il74. NUREG-sonntTime 10/27/s9 i

1229. NURES/CR-3922 NURES/CR-4761. NUREG/

CR-4470. GL 89 !

(No regsfrements)

A-74/

Opaliffcation'of Cirts MNIEG-0500. Rev. I All c et to/s4 synon sea sectice-I l

MPA B-60 IE Safety-Nelated StP 3.11 c2 11/s6 3.11 Equipment 10 CFR 50.49 svacassen #5

-GL 82-09. GL 94-24 ssaa #6 i

ssaa #7 I

l SL 85-15 l

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TITLE REF. Il0CIMENT APPtIC40ltiTY STATW5/94TE*-

RF951RR5 A-26/

Reactor Vessel P essure DN! Letters to PWR c s1-10/84 avnen sta j.

MPA B-04 Transient Protection tfcensees 9/76 sz-11/s6 SECTIOu 5.2.2.2 NUREG-0224 mfRrG.0371 SRP 5.2 i

GL 88-11 A-31 Residual Heat Removal MMEG-0606 All Ots After c 2/s2 avnen sua secTIon Shutdown Requirements RG 1.1!3..

01/79.

11/ss 5.4.3 RG I. I ',9

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SRP 5.4.7 i

A-36/

Control of Heavy Loads RUREG-0612 All c-PnASE I C-10 Near Spent Feel SRP 9.1.5 UI-Io/s4 avnen seu C-15 R 81-07. GL 83-42 82-11/s6 sECTION 9.1.5 l

GL 85-11 sYace ssra #5 c-m u II svuon s ua #6 t

ted

"* '-I 12/22/90 A-39 Determination of SRV MmEG-0802 SMR mA

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Pool Dynamic Leeds NUREGs-0763.0783.0002

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and Pressure Transtents NUREG-0661 i

SPP 6.2.I.I.C A-40 Sefsmic Design SRP.Revisfoes. NUREC/

All c 2/s2 sYnon sua Criterfa CR-4775. NUREG/CR-0054 c-14 & 15 NURES/CR-34EKf. NUREG/

CR-1582. NIIRES/CR-Il61, MmFG-1233. HUREG-477o RUNEG/CR-3005 MmEG/CR-5347 NUREG/CR-3509 A-47/

Pfpe Cracks la telling MREG-0313. Rev. I SIR mA MPA F-% Water Reactors NUREG-0313. Rev. 2

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UNRESOLVED SAFETV 155UES FOR WNfCN A FINAL TECHNICAL RESOLWTiftil fut$ OEEN AOffEVE8 t!$I/NPA i

meRIER TITLE:

REF. NGCUENT APPLICABILITY STATUS /ORTE*

REfulRR5 A-1 Water Hanner SECY 84-119" All c 11/s3~

syron.sta pg.

NUREG-09?7 Iter..I 13-24 eraid e

NUREC-0993. Rev. I NUREG-0737. Item sua pg. c-1 l

7.A.7.3.

SRP revfsfees-l t

A-2/

Asynenetric Riewdsun M8 REG-0609 PNR c 11/s3 I

MPA D-10 Loads on Reector Prfeuery GL 44-04. GOC-4 syro, saa pg.

i Coolant Systems 3-34 state ood SER pg. C-1 I

A-3 WestIaghouse Steam NUREG-0944 4-PWR c 11/89 co==tt= ente in Generator Tube Integrfty SECf 86-97 respon e to ct. 85-e2. +

SECY 88-772 completed.

.GL 85-02 (No regefrements) 4 A-4 CE Steam Generator Tube IIUREG-0844. SECY 85-97 CE-PNR mA

[

Integrity SECY 88-272 GL 85-02 (No requirements) i t

A-5

1. 4W Steam Generator NURER-0844. SECY EIS-97 98ti-PNP na l

Tube Integrity SECY 88-272 GL 85-02 (No llegeframents)

{

f E

A-6 Mart I Cont =1anent IIUREG-0408 Part I-8WR nA i

Short-Term Program I

i 1

• C - COMPLETE NC - No CNANGF5 NECESSARY NA - IIOT APPLICA9tE I - INCOMPLETE E - FVAt.U4 TING ACTIONS.NEf)WTilED

= -.

.. ~

4.-

USI/MPA NtJMRFP TITLE' REF. IIOCtPENT ApFLICASILITY STATW5/9 RTE

• REDEIARS A-7/

Mark I long-Tem fiUREG-0661 b rk t-9WR NA D-01 Progran NUREG-0661 Suppl. I GL 79-57 l

A-8 Mark II Contefament MUREG-OROR Merk II-9WR NA Pool Dynamic toads NUREG-0487. Suppl. 1/2 FUREG-0802 SPP 6.2.1.lC E 16 A-9 Anticipated Transfents NUREG-0460. Vol. 4 All I st-3/91 schedule submitted Nfthout Scram 10 CFR 50.62 52-3/9e 2-15-89 A-10/

BWR Feedwater Nozzio NUREG-0619 NWR mA arsidwood sEn pg.

MPA B-?S Cracking Letter free DG Eisenhut c-1 dated 11/13/80 GL 81-11 A-Il Reactor Vessel Materfal NUREG-0744, Rev. I All c 9/s6 mraidweed saa ~

Toughness 10 CFR 50.60/

s pp. I pg. 5-4 i

82-26 A-12 Fracture Toughness of NUREG-0577. Rev. 1 PNP mA Steam Generator and SRP #evs ion s

Reactor Coolant Ptay 5.3.4-i Supports i

A-17 Systems Interactfons Ltr: GeVoung '.o All E 10/93 ct,as-20

[

lfcensees - 9/72 Irs me.po e NUREC-II74, NUREG-s.6=ttted 10/27/s9 1229. NUREG/CR-3922 NURES/CR-4761. NUPEG/

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

A-24/

Que11fication of Class NHREG-0508 Rev. 1 All c lo/s6 st idwood sua MPA B-60 IE Safety-Related SRp 3.11 sepp. 2 pg. 3-16,17 EquIpuent le CFR 58.49 GL 82-09. El M.

GL 85-15 i

t

M 3

USI/MPA m#mER TITLE

_REF. DOCIMENT APPLIC488LITY STATUS / BATE

• RF9RlRr5 A-26/.

Reactor. Vessel Pressure DN! Letters to PWR C I1/83

'syree sEm pg. 5-4 MPA B-04 Transfent Protectlen Lfcensees A/76 staidwood saa NUREG-0224 pg. C-1 NUREC-4371 SRP 5.2 GL 88-11 A-31 Residual Neat Removal,

M MEG-0606 All Ols After c 11/83 nyron sta pg. 5-2 Shutdown Negsfrements RC 1.113 01/79.

11/ss (natural 23,24 mreidwood RG I.139 cire=tation) sea pg. c-1 SRP 5.4.7 A-36/

Control of Neevy Loads NUREG-0612 Aly c 9/s6 staidwood sua C-10, Near Spent Fuel SRP 9.1.5 5=PP 1 PR 9-1 C-15 GL 81-07. GL 83-42 GL 85-11 Letter,from 06 Eisenhet dated 12/22/80 A-39 Determination of SRV M MEG-0807 OWR NA Pool Dynamic loads NUREGs-0763.0783,0002 and Pressure Transients NUREG-0661.

SPP 6.2.1.1.C A-40 Sefssde Des"gn SRP Reefstems. NUREG/

Aly c 11/s3 syron saa pg.

Criteria CR-4776. NUREG/CR-0954, c-14,15 NUREG/CR-3480. NURES/

semid - d sta CR-1582. NURES/CR-II6l.

PE C I NURFG-1233. HUREG-4776-MMEG/CR-3805 NUREG/CR-5347 NUREG/CR-3509 A-47/

Pipe Cracks in Soflin9 MmEG-0313. Rev. I BNR NA MPA B-05 Water Reactors:

NUREG-0313. Rev. 2 GL 81-03. GL A8-01

~

4-ItsI/ PPA -

_ NUMBER.

TITtf REF. MCtpENT APPL ICAsif.ITY STATUS /ORTE*

RE M A-43 Conta f suneet E-. gi.,.y NUPEG-0510 All c 11/s3 syr= m c-15 Senp Performance NUREG-0869. Rev. 1 NUREG-0897. R.G.I.87 C-16 syrea m (Rev. 9). SRP 6.2.2

1. 5 ps. 6-4, seat h ;

E 85-22 sua c-1 I

No Neguirements A-44 Statten 81ackout' RG 1.155 All r O=plement 1 year mespense==beltted NUREG-1932 after unc review) 4/17/89 l

NUREG-1109 l

10 CFR 50.63 l

A-45

$8stdown Decay Neat SECT 81R-260 All E 10/93 Analysis reestas Removal Requirements NUREG-1289 to be completed.

IIUREG/CR-5230 cL ss-20/Ips neepense>

SECY 88-260 eebettted 10/27/s9 (No require w s) r A-46 Seismic Ovaltffcatten NUREG-1c30 All I

snec proeided se=eeic; of Equipment in NUREG-1211/

3 8 "P "'* * ** 8 '" " ' !

Operating Plants E 87-82. R 87-03

'" 6/3/s8: m received 7/29/88. ses; A-47 Safety Impitcatten NUREG-1217. NgpEG-All a 3/w see 1 e beitted 12/es !

of Centrol Systems 1218 der nec m iew E 89-19

[

l i

A-48

,J._;

Centrol IS CFP 50.44 All, except

~-

Measures and Effects.

SECY 89-172 PNRs witai of ",li;,

Ourns large dry on Safety Egsfpment contafsuments I

l A-49 Presserfred fleermal WGs 1.154. 1.99 Pim I.

hitiaz Nec mtw

%eck-SECT 82-465

• I f

SECT R3-288 s

SECT 81-687 19 CFR 50.61/

E 88-11

EwcLotulLE 1

i F

PLANT Byron 1/2 DOCKET N0(S).

50-454/455

.FROJECT MANAGER L. 01shan TECHNICAL CONTACT A. Serkiz USI NO. A-1 TITLE Water Hamer MPA NO. N/A TAC NOS, None ISSUES

SUMMARY

l ThisUnresolvedSafetyIssue(USI)wasresolvedinMarch1984 with the publication of NUREG-0927, " Evaluation of Weter Hamer in Nuclear Power Plants

- Technical Findings Relevant to Unresolved Safety Issue A-1."

Also on March

.15, 1984, the EDO sent the Comissioners SECY 84-119 titled, " Resolution of Unresolved Safety Issue A-1, Water Hamer."

In SECY 84-119, the staff concluded that the frequency and severity of water hamer occurrences had been significantly reduced through.(a) incorporation of design features such as keep-full systems, vacuum breakers, J-tubes, void detection systems, and improved venting procedures; (b) proper design of feed-water valves and control systems; and (c) increased operator awareness and training. Therefore, the resolution of USI A-1 did not involve any. hardware or design changes on existing plants.

ItdidinvolveStandardReviewPlan(SRP) changes (forward fits) and a comprehensive set of guidelines and criteria to i

evaluate and upgrade util.ity training programs (per TMI Task Action Plan Item I.A.2.3).

In addition, the assumption was made that for BWRs with isolation condensers (ICs) a reactor-vessel high water-level feedwater pump trip was in place or being installed. This was necessary because calculated values had postulated an IC failure by water hamer that opened a direct pathway to the environment.

IMPLEMENTATION AND STATUS

SUMMARY

(PLANT SPECIFIC):

In the Byron SER (NUREG-0876, February 1982), the staff concluded that the t

licensee satisfied the requirements of TMI Item I.A.2.3 (page 13-24 of the SER). Licensee comitted to I.A.2.3 in letter dated October 5,-1981.

Section 10.4.7 of the Byron SER (NUREG-0876, February 1982) concludes that the safety-related portion of the condensate and feedwater system meets _

NUREG/CR-1606 with respect to its design and testing for prevention of damaging water hammer. Additional discussion on A-1 is provided on page C-7 of NUREG-0876. Thus, A-1 was resolved prior to licensing.

L l

l l

J

A

REFERENCES:

Bryon 1/2 A-1 1.

RE0VIREMENT DOCUMENTS:

TITLE NUDOCS NO.

DATE Letter froin Denton to Utilities, 8403150310 03/05/84

" Notice of Issuance and Availability NUREG-0927 Rev. 1, Safety Issue A-1" 2,.

IMPLEMENTATION DOCUMENTS:

TITLE NUDOCS NO DATE NUREG-0927

• Evaluation of Water 8306060413 05/31/83

-llammer in Nuclear Power Plants-i Technical Findings Relevant to Unresolved Safety Issue A-1" l

~HUREG-0993 Rev. 1 8306060418 March 1984

" Regulatory Analysis for.

for USI A-1, Water Hamer" SRP Sections:

3.9.3, 3.9.4, 5.4.6, 5.4.7, 6.3, 9.2.1, 9.2.2, l..

10.3, and 10.4.7 i

SECY-84-119 " Resolution 03/15/84 l'

of Unresolved Safety A-1, Water Hammer" Letter from Licensee 8110130351 October 5, 1981 Byron SER February 1982 l.

3.

VERITICATION DOCUMENTS:

TITLE NUDOCS NO.

DATE 1

i l-l 1

7 1

l

i PLANT' Byron 1/2 DOCKET N0(S).

50-454/455 1

PROJECT MANAGER L. 01shan TECHNICAL CONTACT Jai Rajan US1 NO. A-2 TITLE Asymetric Blowdown Loads in RCS MPA NO. D-10 TAC NOS.

56375 (Byron 1)

. ISSUES

SUMMARY

This USI was resolved in January 1981 with the publication of NUREG-0609,

" Asymmetric Blowdown Loads on PWR Primary Systems."

In October 1975, the NRC notified each operating PWR licensee of a potential I

safety. problem concerning the fact that asymmetric LOCA loads had not been considered in the design of any PWR piping system.

In June 1976 the NRC informed each PWR licensee that it was required to reassess the reactor vessel support design of its facility. The staff expanded the scope of the problem in i

l January 1978 with a request for additional information to all PWR licensees.

NUREG-0609 provided guidance for these analyses.

For operating PWRs, Nulti-Plant Action (MPA) Item D-10 was established by NRC's Division of Licensing for implementation purposes.

l During the course of the work on USI A-2, it was demonstrated that there were only a very limited number of break locations which could give rise to signifi-cant' loads. Subsequently, after substantial new technical work, it was demon-strated that pipes would leak before break and that new fracture mechanics j

techniques for the analyzing of piping failures assured adequate protection against failures in primary system piping in PWRs (Generic Letter 84-04). This was' reflected in a revision of General Design Criteria (GDC)-4 (Appendix A to j

10 CFR Part 50) published in the Federal Reg'pW TTshed in the Federal Register ister in final form on April 11, 1

1986, and in a subsequent revision to GDC 3 j

on July.23, 1986.

In addition, it has also been satisfactorily demonstrated in l

the course of the A-2 effort that there is a very low likelihood of simultaneous i

pipe loading with both LOCA and safety shutdown earthquake (SSE) loads.

I Therefore, the last revision of GDC-4 represented the final technical action of NRC regarding the issue of asymmetric blowdown loads issue in PWRs primary

. coolant main loop piping.

]

IMPLEMENTATION AND STATUS

SUMMARY

-(PLANT SPECIFIC):

Section 3.9.2.4 of the Byron SER (NUREG-0876, February 1982) provides a i

discussion of A-2.

An exemption from GDC-4 regarding the need to analyze large primary loop pipe ruptures as structural design basis was issued for Byron 2 on October 28, 1985.

Since Dyron I was operating at that time it did not receive the exemption.

However, the restraints and deflectors were removed on Byron 1 in 3/87, after the rule was-changed. The restraints and deflectors for Unit 2 were removed prior to licensing, 6

l*

i

REFERENCES:

Byron 1/2 A-2 1.

REQUIREMENT DOCUMENTS:

TITLE NUDOCS NO.

DATE l

Generic Letter " Evaluation of Primary Systems for Asymmetric LOCA Loads" 01/20/78 Task Action Plan A-2, " Asymmetric Blowdown Loads on Reactor Primary Coolant System," NUREG-0371 Task Action Plans for Generic Activities 11/78 "Asynnetric Blowdown Loads on PWR Primary Systems," NUREG-0609 US NRC NRR 01/81 i

GDC-4, " Environmental and Dynamic Effects Design Basis" GL 84-04, " Safety Evaluation o' Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops."

2.

' IMPLEMENTATION DOCUMENTS:

TITLE NUDOCS NO.

DATE s

GDC-4 Exemption, Byron 2 8511040476 October 28, 1985-l 3.

VERIFICATION DOCUMENTS:

TITLE NUDOCS NO.

DATE l

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i PLANT Cyron 1/2 DOCKETN0(S).

50-454/H5 PROJECT MA!!AGER L. 01shan TECHNICAL CONTACT E. Murphy

]

USI 140. A 3 A-4, and A-5 TITLE Stean Generator Tube Integrity MPA NO.

TAC NOS.

ISSUES SUW.ARY:

USIs A-3, 4, and 5, were resolved in September 1988 with the publication of

)

NUREG-0844 "NRC Integrated Program for the Resolution of Unresolved Safety

(

!ssues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." USIs A-3, i

A-4, and A-5 did not result in new generic requirements for industry in view of 1

the small potential for reducing risk.

1 Steam generator tube integrity was designated an unresolved safety issue in 1978 after it became apparent that steam generator tubes were subject to widespread degradation, frequent leaks, and occasir,nal ruptures (i.e., gross failures). USI Task Action Plans A-3, A-4, and A-5 were established to evaluate the safety significance of these problems for Westinghouse, Combustion Engineering, and Babcock & Wilcox steam generators, respectively. These j

studies were later combined into a single effort because PWR vendors were all experiencing many of the same problems.

1 NUREG-0844 provides a generic risk assessment that indicates that risk from I

steam generator tube rupture (SGTR) events is not a significant contributor to the total risk at a given site, nor to the total risk to which the general public is routinely exposed. This finding is considered indicative of the effectiveness of licensee programs and regulatory requirements for ensuring steam generator tube integrity in accordance with 10 CFR Part 50, Appendices A and B.

NUREG-0844 also identifies a number of staff-reccmmended actions that can further imprnve the effectiveness of licensee programs in ensuring the integrity of steam gererator tubes and in mitigating the consequences of a SGTR. As'part of the integrated program, the staff issued Generic Letter 85-02 encouraging licensees of PWRs to upgrade their programs, as necessary, to meet the intent of the staff-recommended actions; however, such recommended actions do not t

constitute NRC requirements. The staff's assessment of licensee responses to Generic Letter 85-02 was provided to the Commission in SECY 86-97.

IMPLEMENTAT10t: AND STATUS SUtTARY (PLAtlT SPECIFIC):

The Byron SER (t;URE0-0876, February,1982) on pages 5-21 and C-9, and supplement No. S to the Byron SER, issued October 1984, on page 5-2 discusses the integrity of the steam generator tubes.

Licensee responded to GL 85-02 in letters dated June 17, 1985 and August 22, 1985. Responses were found to be acceptable, but no acceptance (SER) was sent to licensee in writing.

a

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Byron 1/2

REFERENCES:

A-3, 4, 5 1.

_ REQUIREMENT DOCUMENTS:

TITLE NUDOCS NO.

DATE NUREG-0844, "NRC Integrated September 1988 o

Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity" Generic Letter 85-02 04/17/85 SECY-86-97, Steam Generator-USI Program - Utility Responses to Staff Recommendations in Generic Letter 85-02 03/04/86 2.

R EMENTATION DOCUMENTS:

TITLE NUDOCS NO.

DATE Licensee Letter 8506240527 6/17/85 Licensee Letter 8508260099 8/22/85 Byron SER February 1982 Byron SSER 6 October 1984 3.

VERIFICATION DOCUMENTS:

TITLE NUDOCS NO.

DATE i

4 i

.)

I

[

PLANT Byron 1/2 DOCKET N0(S). 50-454/455 PROJECT MANAGER L. 01shan TECHNICAL CONTACT J. Mauck USI NO. A-9

' TITLE ATWS per 10 CFR 50.62 MPA NO.

TAC NOS. 59077 and 63258 ISSUES SUtHARY:

t This USI was resolved in June 1984 with the publication of a final rule (10 CFR 50.62) to require improvements in pli.nts to reduce the likelihood of failure of L

the reactor protection. system (RPS) to shut down the reactor following L

' anticipated transients and to mitigate the consequences of an anticipated

-transientwithoutscram(ATWS) event, j

The rule includes the'following design-related requirements:

50.62(C)(1),

o diverse and independent auxiliary feedwater initiation and turbine trip (for all PWRs; 50.62(C)(2), diverse scram systems for CE and B&W reactors; 50.62 C)(3) alternate rod injection (ARI) for BWRs; 50.62(C)(4); standby liquid control system (SLCS) for BWRs; and 50.62(C)(5). a tomatic trip of recirculation pumps under conditions indicative of.an ATWS :or 9Rs.

lnformation requirements and an implementation schedule are also specified.

IMPLEMENTATION AND_ STATUS

SUMMARY

(PLANT SPECIFIC):

t Staff SER issued June 13, 1989.

By letter dated August 16, 1989 licensee proposed four changes to ATWS design based on its human factors engineering i

review.

Staff approved these changes in September 12, 1989 letter.

1mplementation schedule for next refueling outages: 3/90 for Unit 1, 10/90 for Unit 2.

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REFERENCES:

Byron 1/2 A-9 1.

REQUIREMENT DOCUMENTS:

3';f NUDOCS NO.

DATE NUREG-0460, and Supplements.

03/80

" Anticipated Transients Without i

Scram for Light Water Reactors" Federal Register Notice i

49FR26045(10-CFR50.62) 06/26/84 i

J 2.

IltPLEMENTATION DOCUMERTS:

TITLE NUDOCS NO.

DATE SER 8906190272 June 13, 1989 Letter on Changes 8909200051 September 12, 1989 i

?

i 3.

ERIFICATIONDOCUMENTS:

TITLE NUDOCS NO.

DATE 1

7 e

f a

PLANT Byron 1/2 DOCKET N0(S). 50-454/455 L

PROJECT MANAGER L. 01shan TECHNICAL CONTACT B. Elliott USI NO. A-11 TITLE Reactor Vessel Materials Toughness I

MPA NO.

TAC N05.

ISSUES

SUMMARY

This USI was resolved in October 1982 with the publication of NUREG-0744,

' Pressure Vessel Material Fracture Toughness.". NUREG-0744 was issued by Generic 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 Letter 82-26 was required.

Because of the remote possibility that nuclear reactor pressure vessels designed to the ASME Boiler and Pressure Vessel Code would fail, the design of nuclear facilities does not provide protection against reactor vessel failure.

Prevention of reactor vessel failure depends primarily on maintaining the reactor vessel material fracture toughness at levels that will resist brittle fracture during plant operation. At service times and operating conditions typical of_ current operating plants, reactor vessel fracture toughness 1

l properties provide adequate margins of safety against vessel failure; however, as plants accumulate more and more service time, neutron irradiation reduces I

the material fracture toughness and initial safety margins, l-A spendix G to 10 CFR Part 50 requires that the Charpy upper shelf energy '

tiroughout the life of the vessel be no less than 50 ft-lb unless it is demonstrated that lower values will provide margins of safety against failure equivalent to those provided by Appendix G of the ASME code. USI A-11 was initiated to address the staff's concern that some vessels were projected to l

have beltline materials with Charpy upper shelf energy less than 50 ft-lb.

NUREG-0744 provides a method for evaluating reactor vessel materials when their Charpy upper shelf energy is predicted to fall below 50 ft-lb. Plants will use-the prescribed method when analysis of irradiation damage predicts that the charpy upper shelf energy is below 50 ft-lb.

IMPLEMENTATION AND STATUS

SUMMARY

(PLANT SPECIFIC):

On page 5-3 of Byron SSER-4 (NUREG-0876), May 1984, the staff approved the l-revised heatup and cooldown curves for Byron 1.

These revised curves were issued with the license on 10/31/84 as part of the tech. specs. Byron 2 curves i

l were issued with Byron 2 license on 11/6/86.

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REFERENCES:

Byron 1/2 A-11 1.

REQUIREMENT DOCUMENTS:

TITLE-NUDOCS NO.

DATE NUREG-0744, Revision 1

" Pressure 10/82 Vessel Material Fracture Toughness

• Generic Letter 82-26. " Pressure Yessel Material Fracture Toughness" 11/12/82 2.

IMPLEMENTATION DOCUMENTS:

TITLE NUDOCS NO.

DATE Byron SSER 4 May 1984 3.

VERIFICATION 00 cut 4ENTS:

TITLE NUDOCS NO.

DATE 4

e

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PLANT Byron 1/2 DOCKETN0(S). 50-454/455 PROJECT MANAGER L. 01shan TECHNICAL CONTACT R. Johnson (RES) i USI NO. A-12 TITLE Potential of Low Fracture Toughness and Lamellar i

Tearing in PWR SG and RCP Supports MPA NO.

TAC N05. None ISSUES SUMARY:

This USI was resolved in October 1983 with the publication of NUREG-0577,

" Potential of Low Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports." The resolution contained no backfit requirements; it only applied to plants with a new construction permit issued after October 1983.

Standard Review Plan Section 5.3.4 was issued at the same time this USI was resolved.

i The concern in this USI, as the title indicates, was the potential of low fracture toughness of some materials selected for fabrication of steam generator (SG) and reactor coolant pump (RCP) supports in operating PWRs.

Lamellar tearing was also of concern.

Fracture toughness is a measure of a material's resistance to fracture in the presence of a previously existing crack. Generally, a material is considered to have adequate fracture toughness if it can withstand loading to its design limit in the presence of detectable flaws under stated conditions of stress and temperature.

The modifications to address this USI could involve maintaining minimum temperature around the supports above its fracture transition temperature, or l

total replacement of existing SG and RCP supports with supports fabricated of material grade which has a higher Charpy upper shelf energy and a lower transition temperature. Analysis cerformed for the resolution of this USI determined that, even with the failure of the SG and RCP supports, the amount l-of incremental release of radioactivity would not be sufficiently high enough l

to justify any modification in terms of increasing the toughness of these supports. This conclusion is based on a value-impact analysis documented in Appendix C of NUREG-0577.

IMPLEMEllTATION,AND STATUS SUtHARY (PLANT SPECIFIC):

D On page C-13 of Byron SER (NUREG-0876, February 1982), staff concluded that

[

plant could be operated safely before the ultimate resolution of USI A-12.

. Ultimate resolution incorporated fix on plants requesting CP after October 1983; thus, no fix required for Byron.

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REFERENCES:

Byron 1/2 A-12 l

i 1.

REQUIREMENT DOCUMENTS:

i TITLE -

NUDOCS NO.

DATE NUREG-0577, Rev. 1, " Potential 10/83 of Low Fracture Toughness and Lane 11ar Tearing in PWR Steam Generator and Reactor Coolant l

Pump Supports" i

2.

IMPLEMENTATION DOCUMENTS:

TITLE NUDOCS NO.

DATE Byron SER February 1982 i

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I 3.

VERIFICATION D0,CUMENTS:

TITLE NUDOCS NO.

DATE I

N/A r

F 4

e

?

-u PLANT Byron 1/2,

.DOCKETN0(S). 50-454/455 FROJECT MANAGER L. 01shan TECHNICAL CONTACT D. Thatcher USI NO. A-17 TITLE Systems interactions in Nuclear Power Plants __

MPA NO.

TAC N05. None

-ISSUES

SUMMARY

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

GL89-18hadtwoenclosureswhich(a)outlinedthebasesfortheresolutionof USI A-17, and (b) provided five general lessons learned from the review of the overall systems interaction issue. The staff anticipated that licensees would review this information in other pcograms, such as tie Individual Plant Examination (IPE)forSevereAccidentVulnerabilities. Specifically, the staff expected that insights concerning water intrusion and flooding from internal sources, as described in the appendix to NUREG-1174, would be considered in the IPE program.

Also considered in the resolution of this USI was the expectation that licensees would continuesto review information on events at operating nuclear. power plants in accordance with the requirements of TM1 Task Action Planitem1.C5(NUREG-0737).

IMPLEMENTATION AND STATUS

SUMMARY

(PLANT SPECIFIC):

Byron applied for CP in September 1973; thus, it did not receive the September 1972 letter regarding flooding.

IPE Response submitted 10/27/89. Byron IPE due 9/92.

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REFERENCES:

Byren -1/2 ~

A-17 m

- - 1 '.

REOUIREMENT DOCUMENTS:.

'i i

-TIT! E '-

NUDOCS NO,_

DATE Generic Letter.89-18 09/06/89 i

l NUREG-1174 " Evaluation of May 1989

~ Systems Interactions in Nuclear Power Plants" L-NUREG-1229 " Regulatory Analysis August 1989 l -_

for Resolution of USI A-17" 19 N

NUREG/CR-3922 " Survey and January 1985 Evaluation of System Interaction Events and Sources"

.NUREG/CR-4261 " Assessment-of June 1986 i

~ System Interaction Experience in.

l Nuclear Power Plants" l

L NUREG/CR-4470 " Survey and-August 1986 F:

Evaluation of Vital 1

l Instrumentation'and Control J

L

. Power Supply Events" 1

NRC Letters to Licensees-9/72 Informing Licensees of Staff Concerns Regarding Potential-Failure of Non-Category I

' Equipment l

2.

If1PLEMENTATION DOCUMENTS:

J TITLE NUDOCS NO.

DATE g

IPE Response 8911080207 October ?7, 1989 3.:

VERIFICATION DOCUMENTS:

w

' TITLE NUDOC H0.

DATE c

-h y

o-T n

1. N,,

!h PLANT Byron 1/2 DOCKET N0(S).

50-454/455 PROJECT MANAGER L. Oisban TECHNICAL CONTACT P. Shemanski US1 NO. A-24 TITLE Qualification of_ Class 1E Equipment MPA NO.

TAC NOS. 57088(Byron 1)

ISSUES

SUMMARY

This USI was resolved in July 1981 with the publication of NUREG-0588, Pevision 1, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment." Part I of the report is the original NUREG-0588 that was issued for comment; that report, in conjunction with the Division of Operating Reactor (00R) Guidelines, was endorsed by a Commission Memorandum and Order as the interim position on this subject until " final" pcsitions were established in rule making. On January 21, 1983 the Commission amended 10 CFP 50.a9 (the rule), effective February 22, 1983, to codify existing qualification methods in national standards, regulatory guides, and certain NPC publications, including NUREG-0588.

The rule is besed on the 00R Guidelines and NUREG-0588. These provide guidance on (a) how to establish environmental service conditions, (b) how to select methods which are considered appropriate for qualifying the equipment in different areas of the plant, and (c) such other areas as margin, aging, and documentation. NUREG-0588 does not address all areas of qualification; it does supplement, in selected areas, the provisions of the 1971 and 1974 versions of IEEE Standafd 323.

The rule recognizes previous qualification efforts completed as a result of Commission Memorandum and Order CLI-80-21 and also reflects different veesions :EEE 323, dependent on the date of the construction permit Safety Evaluation Report (SER). Therefore, plant-specific requirements may vary in accordance with the rule.

In summary, the rosolution of A-24 is embodied in 10 CFR 50.49.

A measure of whether each licensee has implemented the resolution of A-24 may therefore be found in the determination cf compliance with 10 CFR 50.49. This was addressed by 72 SERs for operating plants issued shortly after publication of the rule and subsequently in operating license reviews pursuant to Standard Review Plan Section 3.11.

This was further addressed by the first-round environmental qualification inspections conducted by the NRC.

1HPLEPEPTAT10NANDSTATUSSUPMARY(PLANTSPECIFIC):

0 F.0 SER published in Byron SSER 5, issued October 1984. SSER's 6, February 1985 and 7. November 1986, provided follow-up discussion.

10 CFR 50.49 required that all equipmert be qualified by March 31, 1985 (extensions could be granted to Novembre 30,1985). Byron 1 met the rule and had all equipment qualified by March 31,_1985.

No letter was sent by the licensee confirming this. By letter dated 10/15/06, licensee stated that the EQ issue was satisfied for both units; Unit 2 was in compliance with 10 CFh 50.42 prior to receiving its OL.

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REFERENCES:

Byron 1/2 L

A-24 1 ~.

REQUIREMENT DOCLHENTS:

TITLE NUDOCS NO.

DATE 00R " Guidelines for. Evaluating o

Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety. Related E_lectrical Equipment" 12/79 l

. Commission Memorandum and-Ordar, L.

- CLI-80-21, on D0R Guidelines and NUREG-0588 05/23/80

+

L NUREG-0588,! Revision l' 07/81 10.CFR 50.49 (48 FR 2730-2733) 01/21/83 l.

Standard and Review Plan'3.11, Environmental Qualification of Mechanical and Electricel Equipment 07/81

' 2.-

It!PLEMENTATION DOCUMENTS:

p TITLE NUDOCS NO.

DATE L.

L Byron ' SSER ' 5.

October -1984 Byron SSER 6

- February' 1985 -

Byron SSER 7

. November 1986 Letter from licensee 8610220203 10/15/86 i

n

-3.

VERIFICATION DOCUMENTS:

TITLE:

NUDOCS N0.

DATE Inspection-Report 8803070340 02/29/88

' 50-454/88-03 &

4 50-455/88-03

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I1 PLANT Byron 1/2 -

DOCKETN0(S).

50-454/455-PROJECT MANAGER L. 01shan TECHNICAL CONTACT. Chu Liana-USI:NO. A-26 TITLE Reactor Vessel Pr_e,ssure Transient Protection MPA NO..

-TAC NOS. -None

ISSUES

SUMMARY

This USI was resolved in September 1978 with the publication of NUREG-0224,

~

" Reactor-Vessel. Pressure Transient Protection for PWRs," and Standard Review

. Plan Section 5.2.

The licensees of all operating PWRs.were requested to

. provide an overpressure prevention-system that could be used whenever.the.

plants were in startup orl shutdown conditions. The issue affected all operating and future plants, and the staff established MPA B-04 for~ implementing the solution at operating PWRs.

Since 1972, there have been numerous reported incidents of pressure transients in PWRs where-technical specification pressure _and temperature limits ~have been

-exceeded. The majority of these events occurred while the reactors were in a solid-water condition during startup or shutdown and at relatively low reactor

-vessel temperatures. :Since=the reactor vessels have less toughness at lower 1 emperatures, they are more-susceptible to brittle fracture under these condi-t tions than at norma 1' operating-temperatures.

In-light of the' frequency of'the reported transients and the associated potential for vessel damage, the NRC staff concluded that measures should be 'sken to minimize the number of future-transients and reduce their severity.

a Generic. Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel-Materials and-its Impact on Plant Operations," was published July 12, 1988. This ger.eric letter provides guidance regarding review of pressure-temperature limits and-indicates that licensees may.have to revise low-temperature-overpressure protection setpoints.

IMPLEMENTATION AND STATUS

SUMMARY

(PLANT SPECIFIC):

On pages 5-2 through 5-5 -on Byron SER (NUREG-0816, February 1982), the staff approved Byron overpressure protection.

Implementation prior to licensing.

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REFERENCES:

Byron 1/2 A-26'

.1.

REQUIREMENT DOCUMENTS:

TITLE NUDOCS NO.

DATE NUREG-0224

" Reactor Yessel-

. Pressure Transient Protection j

for PWRs."'

9/78 NRC Letters to Licensees Informing Licensees of Staff Concerns'Regarding Overpressure Low-Temperature Conditions in PWRs-August 1976:

l Generic Letter 88-11. "NRC 7/12/88 y

PositiononSediationEmbrittlement of Reactor: Vessel Materials and

'Its 7mpac.t on Plant Operations" Standard Review Plan y

Section 5'2 L

2..

IllPLENENTATION DOCUMENTS-TITLE NUDOCS NO.

DATE Byron SER February.1982 l.-

L 3.

VERIFICATION-DOCUMENTS:

TITLE

'NUDOCS NO.

DATE L

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m PLANT. Byron' 1/2 DOCKETN0(S).

50-454/455 PROJECT MANAGER -L. 01shan-TECHNICAL CONTACT R. Jones

USI NO.3 A-31 TITLE RHR Shutdown Requirements MPA NO.

TAC NOS.

56199 and 63239 (Natural Cire

> ISSUES

SUMMARY

This USI'was resolved in May 1978 with the publication of Standard Review' Plan (SRP)- Section 5.4.7.

Only those plants expected to receive an operating

.-license after January 1, 1979 were affected by this resolution. The USI involved establishment of criteria for the' design and operation of systems i

,necessary:to; take a power reactor from normal operating conditions to cold shutdown.

. SRP Section 5.4.7 stated that, for purposes of implementation, plants would be

. divided into three classes:

Class I would require full compliance for o

. Construction Permit (CP) or Preliminary Design Approval (PDA) applications which were docketed on or after January 1,1978. Class 2 required a partial L

implementation _for all plants for which CP or PDA applications were docketed I:

- before January 1, 1978, and for which an Operating License (OL) issuance was.

[1

- expected on or after January 1,-1979. Class 3 affected all operating reactors and all other plants for which issuance of the 0L was expected before January l

. 1, 1979. The extent to which Class 3 plants would require implementation was l.

' based'on the combined staff review of related plant _ features.'

In general, the p

outcome.of these evaluations were that only plants receiving an OL af ter January -

L 1,1979 were affected by this USI resolution, and there were no backfits to L1 operating plants that had' received-an operating license before January 1,1979.

j

. IMPLEMENTATION AND STATUS

SUMMARY

(PLANT SPECIFIC):

p' Pages 5-22 through 5-26 of Byron SER (NUREG-0876,. February 1982)-discuss Byron's conformance to BTP RSB'5-1. The SER stated that Byron met.BTP RSB 5-1 with the exception of conducting a Natural Circulation test or demonstrating

~ that the results of the-Diablo Canyon tests apply to Byron. Closecut of Natural Circulation issue was done by-letter dated November 4, 1986 which li stated that the results of the Diablo. Canyon tests were applicable to Byron; thus, no testing on Byron was required.

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REFERENCES:

1 Byron 1/2 A-31 1.

LREQUIREMENT DOCUMENTS:

a TITLE

- NUDOCS NO.

DATE J

NUREG-0800 " Standard Review Plan,"

5/78-i SRP-Section 5.4.7-

.NUREG-0606 " Unresolved Safety N

-Issues Summary"'

Regulatory' Guide 1.139, " Guidance

'for Residual Heat. Removal"

~

Regulatory Guide 1.113' L,c 2.;

IMPL,EMENTATION DOCUMENTS:

g

.,T,ITLE NUDOCS NO.

DATE Byron SER February 1982

. Letter on Natural 8811090159 November 4, 1988

~ Circulation Testing s

3; VERIFICATION DOCUMENTS:

TITLE NUDOCS NO.

DATE

.s

}

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PLANT 1 Byron 1/2

' DOCKET N0(S). 50-454/455 y

- PROJECT MANAGER L. 01shan' TECHNICAL CONTACT J. Wermiel USI N0.1 A-36 TITLE Control of Heavy Loads, Phases I & II' MPA-NO. ' C-10. C-15 TAC N05. None ISSUES

SUMMARY

-l 1This USI was~ resolved in July-1980 with the publication of NUREG-0612.)" Control of Heavy Loads at Nuclear Power Plants," and Standard Review Plan (SRP Section 9.1.5.

The staff established MPAs C-10 and C-15 for the implementation of

_ Phases.-I and II,_ respectively, of the resolution of this issue at operating plants.

j In nuclear power plano, heavy loads may be handled in several plant areas.- If j

these. loads were to drop in certain locations in the plant, they may impact.

s>ent fuel, fuel in the core, or equipment that may be required to achieve safe slutdown and continue decay heat removal.

USI A-36 was established to systematically examine staff licensing criteria and the adequacy of measures in-effect at operating plants, and to recommend necessary changes to ensure the R

'l safe handling of heavy' loads. The guidelines proposed in NUREG-0612 include definition of safe load paths, use of load handling procedures, training of

_l Jerane operators, guidelines on slings and special lifting devices, periodic inspection and maintenance for the crane, as well as various alternatives.

By Generic Letters dated December 22, 1980, and February 3,1981 (Generic Letter 81-07), all utilities were requested to evaluate their plants against the guidance of NUREG-0612 and to provide their submittals in two parts: Phase I (six mcnth response) and Phase II (nine month response).

Phase I responses

-were to address Section 5.1.1 of NUREG-0612 which covered the following areas:

1.

Definition of safe load paths 2.

. Development of load handling procedures 3.

Periodic ~ inspection and testing of cranes 4.

Qualifications, training and specified conduct of operators 5.

Special lifting devices should satisfy the guidelines of ANSI N14.6.6.

6Property "ANSI code" (as page type) with input value "ANSI N14.6.6.</br></br>6" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..

Lifting devices that are not specially designed should be installed and used in_accordance with the guidelines of ANSI B10.9 7.

Design of cranes to ANSI B30.2 or CMAA-70

]

Phase'17 responses were to address Sections 5.1.2 thru 5.1.6 of NUREG-0612 which m Ted the need for electrical interlocks / mechanical stops, or alterna i aly, single-failure-proof cranes or load drop analyses in the spent fuel pool area (PWR), containment building (PWR), reactor building (BWR), other areas and the specific guidelines for single-failure-proof handling systems.

As stated in Generic Letter 85-11

" Completion of Phase II of ' Control of Heavy Loads-at Nuclear Power Plants' - NUREG-0612," all licensees have completed the requirement to perform a review and submit a Phase I and a Phase II report.

1 Based on the improvements in heavy loads handling obtained from implementation

'of.NUREG-0612 (Phase I), further action was not required to reduce the risks associated with;the handling of heavy loads. Therefore, a detailed Phase 11 review of heavy loads was not necessary and Phase II was considered completed.

~

4 e

u:

, =

REFERENCES:

Byron 1/2 A-36 Wh'ile not a requirement,-NRC encouraged the implementation of any act

< identified in Phase II-regarding the handling of heavy loads that we>-

considered appropriate.

~ IMPLEMENTATION-AND' STATUS sum %RY (PLANT SPECIFIC):'

Phase I was'foundacceptableinByronSSER(NUREG-0876), dated 0ctober1984, on page 9-1. -In Byron SSER 6 (February 1985) on page 9-1, the staff stated -

that based on.its review of' Phase 11 no further action by the licensee was

.necessary.

Initial response in 12/22/80 Generic Letter was submitted 4/7/82.

Resolved prior to-licensing.

1.

REQUIREMENT DOCUMENTS:

TITLE-NUDOCSl(E DATE Letter,-Darrell G. Eisenhut, NRC,

- to all licensees, applicants for OLs and holders of cps transmitting NUREG-0612 and.' staff. positions 12/22/80 Generic Letter 85-11 Hugh L.

-Thompson, NRC, to all licensees for Operating Reactors, " Completion-

- of Phase 11 of ' Control of Heavy Loads at Nuclear Power Plants'

' NUREG-061?"

06/28/85 2.

IMPL&1ENTATION D0CUMENTS:

. TITLE.

NUDOCS NO.

DATE' s

SSER 5 October 1984-

- SSER 6 February 1985 l

ti.

r x

3.-

VERIFICATION DOCUMENTS:

s TITLE NUDOCS NO.

DATE 5

1

1 r

PLANT _ Byron'1/2 DOCKETN0(S).

50-454/455 PROJECT MANAGER.-'L. 01shan

-TECHNICAL CONTACT H. Ashar

USI NO. A-40 TITLE Seismic Design Criteria

+

l-MPA NO.

TAC NOS.

J,SSUES

SUMMARY

l The staff has resolved USI A-40 as documented in NUREG/CR-5347, "Recommenda-b

.tions for Resolution of Public Coments on USI A-40," issued in June 1989, and NUREG-1233,." Regulatory Analysis for USI A-40," issued in September 1989.

For plants not covered under the scope of USI A-46, " Seismic Qualification of Equipment in Operating Plants," the staff concluded that tanks in plants that L

were-subject to licensing review by the staff after 1984 had been reviewed to J

current requirements and found acceptable.

For tanks in plants reviewed during 1980-1984,.the staff identified four plant sites (six units) that were not explicitly reviewed to current requirements. The four plants (Callaway_1/2, Wolf Creek, Shearon Harris 1, and Watts Bar 1/2) are being handled on a plant-specific basis.

L L

USI A-40_ originated in 1977. The basic objectives were (a) to study the seismic

-design criteria,.(h) to quantify the conservatism associated with the criteria, and (c) to recomend modifications to the Standard Review Plan (SRP) if changes l-are justified. Lawrence Livermore National Laboratory (LLNL) completed the 1

study and published its findings in NUREG/CR-1161, " Recommended Revisions to l1 USNRC_- Seismic Design Criteria," dated May 1980. The report recomended speci-

?

fic changes to the Standard Review Plan (SRP). NRC staff reviewed the report and developed some other changes that would reflect the present state of seismic design practices. The resulting SRP changes were issued for public coment in June:1988, and the final _SRP changes are to be published-in October 1989.

es consist of (a) clarification of_ development of site

'The-major _SRP chanfb) justification for use of single synthetic time-history by specific spectra, power spectral density function, (c) location and reductions of input ground '

motion for soil structure interaction, and (d) design of above-ground vertical tanks. Except for item (d), these items do not constitute any additional l

-requirements for current licenses and applications, and thus, no backfitting is

'being required for these items.

However, the revised provisions could be used for margin studies and' reevaluations or individual plant examination for

externalevents(IPEEE).

The participant utilities in the Seismic Qualification Utility Group (SQUG) agreed to implement the changed criteria for flexible vertical tanks for their e

plants. For-the four plants where this issue has to be resolved on an indi-vidual basis a 10 CFR 50.54(f) request-for-information letter has been sent to the affected utilities.

If the information received indicates that large above-ground vertical tanks do not meet the new criteria, plant-specific

-backfits will be considered.

IMPLEMENTATION AND STATUS

SUMMARY

(PLANT SPECIFIC):

On pages c-14 and C-15 of Byron SER (llVREG-0876, February 1982), the staff

cor.cluded that the seismic design basis and seismic design of Byron are iacceptable, and that the resolution of USI A-40 should not effect this

conclusion because the techniques under consideration are essentially these utilized in the Byron review.

!?

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REFERENCES:

Byron 1/2 A p LI.

REQUIREMENT' DOCUMENTS:

g E

L TITLE NUDOCS NO.-

-DATE Regulatory Analysis for NUREG-1233 Sept. 1989 USI-A-40 Recommendations for: Resolution NUREG/CR-5347 June'1989 of Public Cohonents on USI A-40 Standard Review Plan NUREG-0800 To be issued i

Sections 2.5.2, 3.7.1, o.

L 3.7.2, 3.7.3-(Revision 2)

Response of Seismic' NUREG/CR-4776 Feb. 1987

- Category'I Tanks: to Earthquake Excitation Engineering Characteri-NUREG/CR-3805 Feb.-Aug. 1986 zation of Ground Motion, Vols. 3,4,5 I

Proceedings of.the NUREG/CR-0054 June 1986 j'

Workshop on Soil-Structure Interaction-Value Imgact Assessment NUREG/CR-3480 Aug. 1984 for Seismic Design Criteria Seismic Hazard Analysis NUREG/CR-1582 Oct.1981 l

u Application of Methodology, i

l-Results and Sensitivity Studies, Vol. 4 l

Recommended Revision to NUREG/CR-1161 May 1980 Nuclear Regulatory Commission i

Seismic Design Criteria L

Power Spectral Density Functions NUREG/CR-3509 June 1988 ln Compatible with NRC R.G. 1.60

[

Response Spectra l0 2.

APLEMENTAT'ONDOCUMENTS:

.Ti(LE.

NUDOCS NO.

DATE I

l-Request' for Infomation Letters Docket Nos.

May 1989 j

L to Owner's-of Callaway 1&2, Wolf 483, 486, 482, Creek 1, Shearon Harris 1, Watts. 400, 390, 391 Bar.182-i3.

VERIFICATION DOCUMENTS:

h

. TITLE.-

NUDOCS NO.

DATE 1

4

.-?\\>

^

y>,

~

1

' t' ;

R(. ~

VM

_. _ PLANT Byron 1/2 DOCKETN0(S),'50-454/455-t PROJECT MANAGER L.01shan TECHNICAL CONTACT' A. Serkiz I

s V

USI NO.- A-43 TITLE Containment Emergency Sump Performance MPA NO.'

TAC NOS.

ISSUES

SUMMARY

19. USl NO. A-43 TITLE: Containment Emergency Sump Performance The resolution-of this USI was presented to the Commission in October 1985-in-SECY-85-349.

NUREG-0897, Revision 1, " Containment Emergency Sump Performance,"

presentsLthe results of de staff's technical findings. These findings estab-lished a need to revise current licensing guidance on these matters. RG 1.82-b Revision 0 and Standard Review Plan Section 6.2.2, " Containment Heat Removal Systems" were revised to reflect.this new guidance.

No licensee actions were L:

- required.

J Initially, an issue existed concerning the availability of adequate recircula-tioncooling-waterfollowingaloss-of-coolantaccident(LOCA)wnenlong-term-l recirculation)of cooling water from the PWR containmsnt sump, or the BWR residual heat removal system (RHR) suction intake, must be initiated and i

maintained to prevent core melt.

The technical concerns evaluated under USI A-43 were:

(a) post-LOCAadverse

- conditiens resulting from potential vortex formation and air ingestion and-

- subsequentpumpfailure,(b)blockageofsumpscreenswithLOCA-genecated t

insulation debris causing inadequate net oositive suction head (NPSH) on pumps, and-(c) RHR and containment spray pumps inoperability due to possible air, debris, or particulate ingestion on pump seal and bearing systems.

This revised guidance applies only to future construction permits, preliminary design approvals, final design approvals, standardized designs, and applica-L

- tions'for licenses to manufacture. The staff. performed a' regulatory analysis j

to determine if this new guidance should be applied to operating plants. The L

results of this analysis were reported in NUREG-0869 Revision 1, "USI A-43 L

% N tory Analysis," issued in October 1985. The staff concluded that the l

bory analysis does not support any new generic requirements for present N oseas to perform debris assessments.

i IMfLEMENTATION_AND STATUS S,UMMARY (PLANT SPECIFIC):~

Pages C-15 and C-16 of Byron SER (NUREG-0876, February 1982) addresses USI A-43. ' A commitment was made by the Licensee to perform inplace tests to verify sump recirculation capability. This item was closed on page 6-4

' of Byron SSER 5, October 1984 Resolved prior to licensing.

4 T

j p

1 o

v

. PLANTo Byron 1/2

. DOCKET N0(S). '50-454/455 PROJECT MANAGER L. 01shan TECHNICAL CONTACT P. Gill-

=f

- USI NO. A-44 TITLE S_tation Blackout

' MPA NO.

TAC NOS.

68522 and 68523 ISSUES

SUMMARY

i

..This USI was resolved in June 1988 with the publication of a new rule (10

~ CFR 50.63) and Regulatory Guide 1.155.

Station blackout means'the loss of.offsite ac power to the essential and nonessential electrical buses concurrent with turbine trip and the 1

unavailability of the redundant onsite emergency ac power systems. WASH-1400 shon d that station blackout could be an important risk contributor, and opeNting~ experience has indicated that the reliability-of ac power systems L

might be less than originally anticipated. For these reasons, station blackout was designated as a USI in 1980. A proposed rule was published for coment on

- March 21-1986.

A final rule, 10 CFR 50.63, was published on June 21, 1988 and

- becameeffectiveonJuly 21, 1988. Regulatory Guide 1.155 was' issued at the same time as the rule and references an industry guidance document, 3

NUMARC-8700, in order to comply with the A-44 resolution, licensees will be required to:

L maintain onsita emergency ac power supply reliability above a minimum r

level develop procedures and training for recovery from a station blackout

- determine the duration of a station blackout that the plant should be able l

to withstand use an alternate qualified ac power source,-if available to cope with a station blackout i

evaluate the plant's actual capability to withstand and recover from a station blackou'.

~

backfit hardware modifications if necessary to improve coping ability

- Section 50.63(c)(1) of the rule required each licensee to submit a response including.the results of a coping analysis within 270 days from issuance of an 1

operating license or the effective date of the rule, whichever is later.

IMPLEMENTATION AND STATUS

SUMMARY

(PLANT SPECIFIC):

Licensee response was dated April 17, 1989. Review is scheduled to be completed on March 31, 1990.

Licensee committed to change procedures one year after NRC review is completed; thus, March 31, 1991 k

1 g

A.

2 A -

b; 4

k 4

y

REFERENCES:

Byron 1/2 l

A-43

~

1.

REQUIREMENT DOCUMENTS i

. TITLE' NUDOCS NO..

DATE l

NUREG-0869, Rev. 1, "USI

'10/85 A-43 Regulatory Analysis" l

i NUREG-0897, Rev. 1, " Containment 10/85-

. Emergency Sung Performance" GL 85-22, " Potential for. Loss 12/03/85 i

of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage" l

2.

IMPLEMENTATION DOCUMENTS:

TITLE NUDOCS NO.

DATE-

- Byron SER' February 1982 Byron SSER 5 October 1984

+

L 3.-

VERIFICATI0t! DOCUMEttTS:

TITLE NUDOCS NO.

DATE I.

- Inspection Report 8406260354

. June 12, 1984 No.150-454/84-24 and 50-455/84-17 1

l i

4 6

e I

a N

6 L

u.

Byron 1/2

REFERENCES:

A '

1.

REQUIREMENT DOCUMENTS:

i

-TITLE NUDOCS NO.

.DATE 10 CFR 50.63, " Loss of All-l Alternating Current Power" 06/21/88 i

Regulatory Guide 1.155,

" Station Blackout" 08/88

- 2. -

IMPLEMENTATION DOCUMENTS:

TITLE NUDOCS NO.

DATE Letter from licensee

- 8904240429 April 17, 1989L t

l-3.;

VERIFICATION DOCUMENTS:-

r 2'

NUDOCS NO..

DATE TITLE l

i k

h lk r

1 q

4 L

)

1

'5-s r

N a

.i 1

PLAtlT Byron 1/2 DOCKETN0(S). 50-454/455 __

PROJECT MANAGER _ L.'01shan TECHNICAL CONTACT R. Jones USI NO. A-4 5 TITLE Shutdown Decay Heat Removal Requirements

.MPA NO.

TAC NOS.

~

ISSUES

SUMMARY

.USI A-45 was resolved by SECY 88-260, " Shutdown De my cleat Removal Requirements (USI-A *5)," issued September 13, 1988, without imposing any new licensing.

requirementsotherthantheIndtvidualPlantExamination(IPE),asdescribed, below. At.the same time-the staff issued NUREG-1289, " Regulatory and Backfit.-

Analysis: USI A-45."

Since all of the significant USI A-45 results have been found to be highly plant specific, the Comission decided it was not appropriate to propose a single generic corrective action to be applied uniformly to all plants.

+

The Comission is currently implementing the Severe Accident Policy (50 FR 32138) and will require all plants presently operating or under construction to

=i undergo a systematic examination termed the IPE. The reason for this examina-L tion is to identify any plant-specific vulnerabilities to-severe accidents.

'l i'

The-IPE analysis intends to examine and understand the plant emergency pro-

-l cedurgs, design, operations, maintenance, and surveillance, in order to identify vulnerabilities. - The analysis will examine both the decay heat

)

removal systems and those systems used for other related functions. This 2

l<

includes CE plants without power-operated relief valves.

J NRC has decided to subsume A-45 into the IPE program as the most effective way of achieving resolution'of specific plant concerns associated with A-45 l

l

!!!'PLEMENTATION AND, STATUS

SUMMARY

(PLANT SPECIFIC):

l

-IPE. Response submitted 10/27/89.

IPE due 9/92.

J l.

d 1

l

-l

.l l

l l'~,

l m.

7-J

<s4 L

REFERENCES:

Byron 1/2 A 1.

REQUIREMENT DOCUMENTS t

TITLE-NUDOCS NO.

DATE Federal Rei.ister Notice "10 CFR Part 50, ShL down Decay' Heat Removal Requi.ements" r

- NUREG/CR-5230 " Shutdown Decay Heat April 1989 Removal Analysis:

Plant Case Studies and Special Issues Summary Report" NUREG-1289 " Regulatory and Backfit 11/30/88 Analysis-for the Resolution. of -

US! A 55" SECY-88-260 " Shutdown Decay Heat 09/13/88 Removal Requirements 2.

IMPLEMENTATION DOCUMENTS:

TITLE NUDOCS NO.

DATE 3.-

' VERIFICATION DOCUMENTS:

-TITLE NUDOCS NO.

DATE t

i.t. ^

S

3 s

9 i 6

' PLANT Byron 1/2 DOCKET N0(S) < 50-4541!56 PROJECT MANAGER L. 01shan1 TECHNICAL-CONTACT g. g]

+

USI NO. A-47 TITLE _ Safety implication of Control Symns in LWR Nuclear Power Plants

'MPA NO.-

TAC NOS.

o

.t ISSUES'

SUMMARY

I USI A-47 was resolved September 20, 1989, with the publication of Generic l:

Lotter(GL)88-19.

L The generic letter states:

"The staff has concluded that all PWR plants should provide

automatic steam generator overfill protection, all BWR plants (y

.should provide automatic reactor vessel overfill pentection, and

('

that plant procedures and. technical specifications for all plants should include provisions to verify periodically the operability'of the overfill protection and to assure that r

automatic-overfill protection is available to mitigate main feedwater overfeed events during reactor power operation. Also, the system design and setpoints should be selected with the objective of dnimizing inadvertent trips of the main feedwater system during plant startup, normal operation, and protection i

system t.urveillante. The Technical Specifications-recommenda-tions are consistent with the criteria and the risk considera-tiens of the ComisGan Interim Policy Statement on Technical S)ecifiu. tion Imprevement.

In addition, the staff recommends t1st all BWR recipients reassess and modify, if needed, their

operating precedures and operator training to assure that the

operators can mitigate reactor vessel overfill events-that may occur-via the condensate booster pumps during reduced system pressure operation."

L.Also, page 2 of the generic letter provides for additional actions for CE and

.B&W plants. 'The generic letter provides amplifying guidance for licensees, t

The generic-letter requires that licensees provide NRC with their schedule and commitments within 180 days of.the letter's date. The implementation schedule ifor._ actions on which commitments.are made should be prior to startup after the

-first' refueling outage, but no-later than the second refueling outage, beginning 9 months after receipt of the letter.

IMPLEliENTATI0ffAND STATUS

SUMMARY

(PLANT SPECIFIC):

Response due 3/90-a i

r f

m

REFERENCES:

Byron 1/2 A,,

i

'1.

REQUIREMENT DOCUMENTS

~^#

TITLE-NUDOCS NO.

DATE Generic Letter 89-19 09/20/89

" Request for Action Related to Resolution of USI A-47" NUREG<1217 " Evaluation of Safety June 1989

~1mplications of Control Systems in LWR Nuclear Power Plants" NUREG-1218 " Regulatory Analysis July 1989 for Resolution of USI A-47" 2..

IMPLEMENTATION DOCUMENTSj.

TITLE NUDOCS NO.-

DATE 4

l 1

e b

3.

yERIFICATIONDOCUMENTS:

~ TITLE NUDOCS NO.

DATE I

i f

i e

f w

. PLANT Byron 1/2 DOCKETN0(S).

50 454/455 FROJECT MANAGER L. 01shan-TECHNICAL CONTACT B. Elliott

)

. USI fl0.

A TITLE -Pressurized Thermal Shock-j MPA N0..

TAC NOS. 59942 and 63252-ISSUES

SUMMARY

The fina' rule (10 CFR 50.61) on pressurized thermal shock (PTS) was approved by the Commission in July 1985.

Regulatory Guide 1.154, " Format and Content

of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for PWRs,"

was later published in February 1987. Thus, this issue was resolved and new requirements were established, applicable to PWRs only. The rule required that

~

each operating reactor meet the screening criteria provided in the rule or provide supplemental analysis to demor. strate that PTS is not a. concern for the 1

facility.

NeutronirradiationofreactorpreTsurevesselweldandplatematerials decreases the fracture toughness of the materials. The fracture toughness sensitivity toiradiation-induced change is increased by the presence of certain materials such as copper. Decreased fracture toughness makes it more likely l

that, if a severe overcooling event occurs followed by or concurrent with high vessel pressure, and if a small crack is present on the vessel's inner surface, that crack could grow to a size that might. threaten vessel integrity.

Severe pressurized overcooling events are improbable since they require multiple failures and improper operator performance.

However, certain y

precurscr events have happened that could have potentially threatened vessel L

integrity if additional failures had occurred and/or if'the vessel had been

- more highly irradiated. Therefore, the possibility of vessel failure due to a severe pressurized overcooling event cannot be ruled out.

IMPLEMENTATION AND STATUS

SUMMARY

(PLANT SPECIFIC):

Licensee response to 10CFR 50.61 dated January 17, 1986. Staff SER on Byron 1

(

issued November 26,1986.

Byron 2 still under review, p

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i Ji:

n s

J E

4

REFERENCES:

Byron 1/2 A-49 i

1.

REQUIREME,NT DOCUMENTS:~

TITLE NUDOCS NO.

DATE 10 CFR 50.61, " Fracture Toughness-7/85 Requirements for Protection Against

. Pressurized Thermal Shock Requirements" Reg. Guide 1.154, " Format and Content 1/89 i

of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for PWRs" SECY 82-465, " Pressurized Thermal Shock" 11/23/82 SECY 83-288, " Proposed Pressurized

)

l' Thermal Shock Rule" 07/15/83 L

Regulatory Guide 1.154 -

" Format and Content of Plant-Specific Pressurized Thermal-Shock Safety Analysis Reports for Pressurized Water Reactors" 02/87 r

^

Generic Letter 88-11 "NRC Position

>on Radiation Embrittlement of Reactor l

Vessel Materials and Its Impact on Plant Operations" 7/12/88

-2.

IMPLEMENTATION DOCUMENTS:

TITLE NUDOCS-NO.

DATE l

-Licensee' Letter 8601230286' January 17, 1986 j,

Staff SER on Byron 1 8612080012 November 26,-1986 l

l

-l i

l 3.

VERIFICATION DOCUMENTS:

y lw TITLE NUDOCS NO.

DATE r

u l

7 p

(

~

e V

E N c. Lo tu p. 6 l31 j

g Page'No.

1 i

02/02/90 LISTING OF 1NCOMLETE 05) SATA FOR INPUT FROM PROJECT MANA6ERS i

.a 188UE ISSUE K SCRIPilVE NAME IMPLEMENT IMPLEMENT LICENSEE COMENT STAFF COMENT -

INIMMR.

DATE STATUS 1

88 PLMT MME: BYRON 1 A-01 NATER M MER'

//

NC A-02. ASYMETRIC BLONDOM LDADS ON 03/31/87 C RESTRAINT 4 MFLECT N0DIFIChi10NS MM

. REACTOR PRIMRY C00LMT SYSTEMS

'A-03 NESilN6 HOUSE STEM SENERATOR Tutt //

E IN D ONLY INTESRiff-A CE STEM BENERATOR TUK INTESRITY / /

N/A CE PLMTS ONLY A MN STEAM SEERhiDR TUDE.

//

N/A BM PLMIS ONLY INTEGRITY a

A 06 MARK I SHORT-TERM PR06 RAM

/ /' N/A MK 1 DNR SNLY L

A-07 MARK I LONMERM PR06RM

/ /

N/A MK I HR ONLY' l!

A00 MARK 11 CONTAIMENT P0OL DYNAMIC

/ /

N/A MK 11 DNR DNLY LOADS - LON6-TERM PR06AAM.

l A-09 ATNS 03/31/90 1 A 10 DNR FEEDNATER N0!!LE CRACKIN6

/ /

N/A INR ONLY A-!! ' REACTOR VESSEL MATERIALS

/ /

NC

~:

[L TOU6HNE$$

I A 12 FRACTURE TOU6HNESS OF STEM

'/ /

N/A CP AFTER 83 DNLY L

6ENERATOR MD REACTOR C00LMT l

PUMP SUPPORTS A-17

'SYSTEMSINTERACT10N

/ /

NC NOREQUIREMENTS

' A 24 OUALIFICAi!0N OF CLASS 1E 03/31/85 C SAFETY-RELATED EDUlPMENT p

l.

.A-261 REACTOR YESSEL PRESSURE TRANSIENT / /

NC LTOPS AT DL l

PROTECTION

-.A-31

. RHR SHUTDOM REQUIREENTS

/ /

NC LICENSIN6 SER-

A-36 ' CONTROL OF HEAVY LDADS LEAR SPENT. //

E 6L-85-!!ENBED l

FUEL'

[

.A-39

. DETERMINAi!DN OF SAFETY RELIEF

/ /' N/A.

SNR DNLY VALVE POOL DYNANIC LOADS MD.

f J

TEMPERATURE LIM!is A-40 SE!SMIC MS!6N CRITERIA -

/ /

NC SHORT-TERMPROGRAM y

A 42 PIPE CRACKS IN BolLIN6 NATER

//

N/A DNR Oll.Y REACTORS A-43' 'CONTAl MENT EMER6ENCY SUMP

/ /

NC

.IW O ONLY l "

PERFORMANCE A-44~

STAT 10N ILACK00T 03/31/91 1 SER 3/31/90 A 45.

SHUTDONN DECAY EAT REMOVAL.

/ /

E IPE SUBSUMED BY SEVERE ACC i

l~

REQUIREMENTS A-46 SEISMIC QUALIFICAfl0N OF

' //

N/A OLD PLANTS ONLY EQUIPMENT IN CPERATING PLANTS A 47 SAFETY IMPLICAi!ONS OF CONTROL 03/31/90 E EN REQUIREMENTS L

~SYCTEMS-

A-48 HYDROSEN CONTROL MEASURES AND

/ /

N/A N/A DRY CONTAIN 1

l EFFECTS OF HYDROGEN BURNS ON SAFETYEQUIPMENT A 49 PRESSURl!ED THERMAL SHOCK 01/17/86 C j

u Igx'

-i ;

. Page No.-

2-02/02/90 Lill!N6 0F IE OMPLETE U$l DATA

^

FOR INPUT FROM PROJECT MNA8ERS t

!$$UE ISSUE MSCRIPilVE ME IMPLEENT IMPLEENT LICENSEE COMMENT 51AFF COMENT NUMMR :

MTE STATUS

+

' 38 PLMT ME: BYRON 2:

A-01 NATER HAMER

//

NC -

A ASYMETRIC BLOND 0M LDADS ON

/ /- NC REACTOR PRIMRY C00LMT SYSTEMS

-A-03 NESilN8HOUSESIEARSENER/itsiUK //

E 1W0 OIR,Y INTEORITY A-04 CE STEM SENERATOR TUBE INTESRITY //

N/A CE PL MTS ONLY A-05 B&W STEM SEERATOR TUK

//

N/A NN PLMTS ONLY INTEBRITY A-06 MAK ! SHORT-TERM PROGRAM

//

N/A -

R 1 BM DNLY A-07 MRK I LON6-TERM PROGRAM

//

N/A E I DNR DNLY 7

A-08 MARK !! CONTAINMENT P0OL DYMMIC

//

N/A MK !! BNR DNLY LOADS - LONt, TERM PR06RM A-09 ATNS 10/31/90 1 A 10 BNR FEE NATER N0!!LE CRACKIN6

//

N/A BNR ONLY L A-!! ' REACTOR VESSEL MTERIALS

//

NC TOU6HNESS-

'A FRACTURE T006 MESS pF STEM

//

N/A CP AFTER 83 DNLY SENERATOR AND REACTOR COOLANT PUMP SUPPORTS A-17 SYSTEMSINTERACT10N

//

NC NOREQUIREMENTS

.A 24 OUALIFICATION OF CLASS !E

//

NC I-

SAFETY-RELATED E9UIPMENT

[

A-26; REACTOR YESSEL PRESSURE TRMSIENT / /- NC LTOPS AT DL PROTECT!DN A-31. -RHR SHUTDOM REQUIREMENTS

//

NC LICENSIN6 SER A CONTROL OF HEAVY LOADS NEAR SPENT //

NC 6L-85-!! ENDED FUEL 1 A-39 MTERMINATION OF SAFETY RELIEF

//

N/A BM DNLY-I VALVE POOL DYNAMIC LOADS A8 TEMPERATURELIMITS A 40 SEISMIC DESIGN CRITERIA -

//

E SHORT TERM PR06RM A-42 PAPE CRACKS IN 80! LINS NATER

//

N/A BM DNLY REACTORS.

A-43 CONTAIMENT EMER6ENCY SUMP

//

NC INFO ONLY PERFORMNCE A-44 STATION BLACK 0UT 03/31/91 1 SER3/31/90 A 45 SHUTDOM DECAY HEAT REMOVAL

//

NC IPE SUBSUMED BY SEVERE ACC REQUIREMENTS-L>

A-46 SEISMIC 004LIFICAfl0N OF-

/ /.

N/A OLD PLANTS ONLY EQUIPMENT IN OPERATIN6 PLMJ1 A-47-'

SAFETY IMPLICAi!0NS OF CONTROL 03/31/90 E NEN REQUIREENTS

!4 SYSTEMS A-48 ' HYDR 06EN CONTROL MEASURES AG

//

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