ML20033E239
| ML20033E239 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Zion  |
| Issue date: | 02/22/1990 |
| From: | Chandu Patel Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17202J231 | List: |
| References | |
| REF-GTECI-A-09, REF-GTECI-A-44, REF-GTECI-A-46, REF-GTECI-A-47, REF-GTECI-A-49, REF-GTECI-EL, REF-GTECI-RV, REF-GTECI-SC, REF-GTECI-SY, TASK-A-09, TASK-A-44, TASK-A-46, TASK-A-47, TASK-A-49, TASK-OR GL-89-12, NUDOCS 9003090467 | |
| Download: ML20033E239 (2) | |
Text
g r
U.
so urg{og.
- q UNITED STATES 8,
i NUCLEAR REGULATORY COMMISSION'
.J f
. %i
- I^
[.;
WASHINGTON, D. C. 20555
.[..
February 22, 1990-p
~
Docket Nos. 50-295 and 50-304 MEMORANDUM FOR:
File FROM:
Chandu P. Patel, Project Manager Project Directorate III-2 m*
Division of Reactor Projects - III, IV, V and Special Projects
SUBJECT:
STATUS OF IMPLEMENTATION OF UNRESOLVED SAFETY ISSUES AT 1
ZION STATION UNITS 1 AND 2 The current implementation status of unresolved safety issues (US!s) at the Zion-Station is set forth-in the enclosures to this memorandum. contains a copy of the information provided by the licensee in its response to Generic letter 89-12. contains a status sumary 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 NRC records and information. Appropriate NRR technical branches have also reviewed the. USI--status sumary and this memo.
1 In addition. enclosure 3 is a copy of the staff's data base printout for Zion facility.
It reflects the staff's assessment of USI implementation for.all 27
-USIs.
It is based on review of the licensee's response to Generic Letter 89-21, and evaluation by project managers, the USI team, and NRR technical staff.
For those items that are incomplete my assessment of schedule is as follows:
A-9 ATWS Rule 1
i The implementation date for the system modifications required by the ATWS i
Rule is scheduled for spring 1991 refueling outage for Unit I and for spring 1990 refueling outage for Unit 2.
The staff approved the i
L plant-specific design in May 1989. However, in September 1989, the licensee changed its design to make Zion design consistent with Byron and Braidwood (previously approved by staff) which slipped the schedule for Unit 1.
Uni. ' is still scheduled to be in compliance by Spring 1990.
A-44 Station Blackout The modifications needed to satisfy this rule have been appropriately comitted to by CECO and will be installed in 1990. The-necessary procedures will be revised one year after a safety evaluation is issued by NRR. The Safety Evaluation is scheduled for December 1990.
Based on this, the implementation should be completed by December 31, 1992, 9Alb9 0 W 7-d Qr o
1 File
. p.
February 22, 1990 l-Aa46 Verification of Seismic Adequacy of Mechanical and Electrical Equipment 1 CECO has committed to participate in the SQUG program and to conduct the.
required plant walkdown in 1990 through 1991 for Unit 2, and in 1991-through 1992 for Unit 1 and common equipment, contingent upon the staff's approval of the Generic Implementation Procedure (GIP) submitted by SQUG'
'in the. middle'of 1989.
A-47 Safety Implications Control Systems CECO is expected to respond by March 1990 on this subject.
1:
1 A-49 Pressurized Thermal Shocks
- j Zion Unit 2 meets the. screening criteria for 10 CFR 50.61.
Zion Unit 1 l-does not meet the requirement of 10 CFR 50.61 as;it is in present rule.-
L.
However, on December 28, 1989, the staff has proposed (54 FR 52946) to change PTS Rule.
Zion Unit I will be in compliance with the revised? rule when it becomes effective.
~
j
/Sl Chandu P. Patel, Projec't Manager.
Project Directorate 111-2 Division of Reactor Projects.- III, IV, V-and Special Projects l
Enclosures:
As stated cc w/ enclosures:
I K. Eccleston l-1 l-l DISTRIBUTION l
Docket File PDIII-2 r/f l..
LLuther-
.CPatel N
l OFFICIAL RECORD COPY Document Name: ZION US!
( }f PDil{l-2:PMfT. @k PDII LA PDill-2:PD LLup /
CPatel JCraig Y
%/2?/90 1 A%/90 L
m" 7
y Comm:n;;ealth EdWon-kWL LubWKE 2!
Q_".f Aodro68 Reply to. Post ORce B;x 767 Li
(
.1 Om First Nabonal Ptara. Cheago. Ilknois i
j Chcagoclihnois 60690 0767 November 30. l'989 L'
L Director of-. Nuclear Reactor Regulation U.S.-Nuclear Regulatory Commisston Mail Station PI-137 Hashington, DC 20555
Subject:
. Zion Nuclear Power Station, Units 1 and 2
' License Nos. DPR-39 and DPR-48 s
Responses to Generic Letter 89-21 NRC Docket Nos. 50-295 and 50-304
Reference:
October 19, 1989 Letter from J.G. Pantlow i
-to All Holders of Operating Licenses
Dear Sir:
Generic Letter 89-21 was issued as part of the NRC's continuing effort to. validate staff. understanding regarding the implementation of i
responses-to Unreviewed Safety Issues (USI). One aspect of this effort is to l
ensure that the~ licensee and NRC personnel agree on the status of USI i
resolution Implementation at each facility.
This letter provides Commonwealth j
Edison's response, in part, to the USI's.
However, some of these issues have l
required extensive research and the responses for those are not available at this time..Specifically, the responses to USI's A-1,
-3, -12 and -43 are still under review and will be provided later.
In a telephone conversation on November 29, 1989 between,P. Shemanski (NRR) and G. Trzyna (CECO), it was agreed that the submittal of the additional information for these open USI's could be deferred until December 13, 1989.
The Attachment to this letter provides the status of the items in a j
table that completes Enclosure 1 of the Generic letter.
1 q
Please direct any further questions that you may have regarding this issue to this office.
Very truly yours, l
} [. Y- -=-- $
p G.E. Trzyna i
Nuclear Licensing Administrator l
cc:
Senior Resident Inspector - Zion I
Chandu Patel - NRR 0422T-h{Ub500~
L J
ZION CENECATING STATION ATTACHMENT USI/MPA NUM8FR TITLE REF. DOCtFFNT APPLICAditITY STATUS /DATE' RDenRetS A-7/
Mark I long-Term MUREG-0661 Nrk I-8WR gg D-G1 Program NUREG-0661 Suppl. I GL 79-57 A-8 Mark II Containment NOREG-0908 Marf li-8WR wa Pool Dynamic Loads NOREG-0487, Suppl. 1/2 NUREG-0802 SPP 6.2.1.1C GDC 16 September 8, 1989 letter i A-9 Anticipated Transient.s NUREG-0460, Vol. 4 Ali I
Ceco tg NRR transmitted Implementation dates:
Spring 1990 - Unit 2 A-10/
BWR Feedwater Isozzle NUREG-0619 MWR f"$
NA Fall, 1990 - Unit I MPA P-25 Cracking Letter from DG Eisenhut dated 11/13/80 GL 81-11
- December 20, 1985 letter tre-A-Il Reac'or Vessel Paterial NOREG-0744 Rev. 1 All E
Ceco to NRC advised that. 50 Touchness 10 CFR 50.60/
Ib date will'be 1994.
Tht.
87-26 date could change as new specimens are withdrawn.
A-12 fracture Toughness of MUREG-0577. Rev. I PWP
_ nove.ber 13, 1989 letter fr-Steam Generator and SRP Revision Reactor Coolant Pump 5.3.4 CECO to NRR transmitted recs Unit 2 specimen data.
Supper's Response will be provided tr A-17 Systems Interactiens Ltr: DeYoung 'o A13 12/13/89.
y Ifcensees - 9/77 C-il74, MWEG-Station response wus trase!.
1229, MMEG/CR-3927, in IFE response to cL88-20 on October 27, 1989.
NUREG/CR 8761. M9EG/
CR-4470 GL 89-18 (No requirements)
C November 2I, 1984 letter fr.
A-74f Cualification of Class Pf0 REG-0588, Rev. 1 ATI PPA B.60 lE Safety-Related SRP 3.11 NRC to CECO transmitted SER Equi ment 10 CFR 'C.49 First Regional inspection of l
program in January 1985.
c ZION GENERATING STATION
' ATTACHMENT USI/MPA NilMBER TITLE REF. DOCIMENT APPLIC48itITY 5YATUS/DATE*
RFPgnams A-26/
Reactor Vessel Pressure DDR Letters to PIR C
Tech Specs were implementee MPA B-04 Transient Protection Licensees 8/76 on April 28, 1980. They we<i NUREG-0224 recently revised by Amendmet NUREG-0371 No. I10/99.
SRP 5.2 GL 88-11 A-31 Pesidual Heat Removal NUEEG-0606 All Ots After NA Shutc*own Requirements RG 1.113, 01/79.
Control of Heavy Loads NUREG-0612 All C-10, Near Spent fuel SRP 9.1.5 E
r o
1-C-15 GL 81-07, GL 83-42' surficient protection free dropped loads.
Eisenhut dated 12/22/80 A-39 Deterspination of SpV I!! MEG-OP02 BIR NA Pool Dynamic loads Nf*FGs-0763,0783,0sn2 and Pressure Transients NUREG-0661 SPP 6.2.1.1.C A-40 Seismic Design SRP Revisions. NUREG/
All NA Addressed in accordance wi Criteria CP-4776. NUREG/CR-0054, response to USI-A-46.
WOREG/CR-34En, M MEG /
)
CR-1582. NUREG/CR-il61, NtWEG-1233, NUREG-4776 l
NImEG/CR-3805 NtmEG/CR-5347 NOREG/CR-3509 A-42/
Pipe Cracks in Boiling NIREG-0313. Rev. 1 BWR NA MPA 8-05 Water Reactors NUREG-0313. Rev. 7 GL 81-03, GL 88-01
ZION GENERAT*% STATION 4
ATTACHMENT USI/MPA NUMBER TITLE REF. DOCimENT APPLICASILITY STATUS /DATE*
REpmRR$
A Con'ainment Emergency MtmEG-0510, Aly Response will be provided Sump Perforinance NUREG-0869. Rev. I by 12/43/89.
i NUREG-0897 R.G.I.87 h
(Rev. 0), SRP 6.2.2 GL 85-22 No Neguirements 4
A-44 Station Blackout RG 1.155 All E
Review of proposed modif1(
M REG-1932 tion is scheduled to be 88UREG-1109 Co*Pleted by June, 1990.
10 CFR 50.63 A-45 Shutdown Decay Heat SECY R8-260 All E
Letter submitted to NRC on Removal Requirements NUREG-1289 october 27, 1989 in respon IIUREG/CR-5230 to IPEGL88-20.
SCCY 88-260 (No requirements)
A-46 Setsmic Gualification NUREG-1(W All E
Ceneric Sa(ety Evaluation of Equipment in NimEG-1211/
was issued on July 29, 19r i
Operafing Plants GL 87-02, GL 87-03 Several plant walkdown, are pending through 1992.
A-47 Safety implicatinn NUREG-1217, NOREG-AIT E
Response due to MRC by of Control Systems 1218 March, 1990.
~
GL 89-19 A-48 Ilydrogen Control 10 CF# 50.44 All, except NA Measures amt Effects SECY 89-177 PWRs with of Hydrogen Burns large dry on Safety Estelpment containments i
i A-49 Pressurized 1hermal PGs 1.154, 1.91 PWR E
SegQegsgeg-Fbock SECY 82-465 b tween CECO belongs to 4.
SECY R3-288 s&w - owner's croup which
' SECY 81-687 has been estabitehed to 10 CFR 50.61/
address the issue schedeb 4
GL f18-11 to update NRR in Febreaty.
4 1990.
+v
.v.-~
,_e;,
a--.,
y
.r....
.-,,y
_.m.__-_,
..w-.w
.,---,.-s_
__m.__mm__,,m.___,_.m,m_
__,,,___,_,._,__,m_
______..,_.,._____._,.___..,__m._m,
~
.o.,
..m.
.. m. 3. a. w.
--7 ATTACHMENT UNRESOLVED SAFETY ISSIES FOR leffC84 A FIP.Al TECIWilCAL RESOLUTIO tlSI/MPA L
. NUMBER T!TLE REF. IIOCUMENT APPL ICA81LITY STATUS /DATE*
REM 4Rets A-1 Water Hassner SECY 84-119 All MREG-09?7. Rev. I Response will be provid.
mfREG-0993 -Rev. 1 by 12/13/89.
NUREG-0737 ltem I.A.2.3 SRP revisions l
A-2/
Asynenetric Blowdown NUREG-0609 PWR MPA D-10 Loads on Reac. tor Primary GL 14-04, 600-4 Internal CECO review pre-C Coolant Systems formed in April, 1984 verified that acceptable A-3 Westinghouse Steam NUREG-0644 W-PWR equipment leak detection Generator Tube ' integrity SECY 86-97 was installed & no respen SECY 88-772 was necessary.
GL 85-02 Response will be providet i
(No requirements) by 12/13/89.
A-4 CE Steam Generator Tube NUREG-0844, SECY 86-97 CE-PWR Integrity SECY 88-272 NA r
4 CL 85-02 (No requirements)
A-5 B8W Steam Generator NUREG-0844. SECY 86-97 98W-PWR NA Tube Integrity SFCY 88-272 GL 85-02 (No Requfrements)
E i
X-6 Mark i Containment NUREG-0408 Mark I-8WR NA Short-Term Program
- C - COMPLETE NC - NO CHANGES NECESSARY NA - NOT APPLICABLE I - INCOMPLETE F - FVALUATING ACTIONS REQUIRED e
,e-
.i,-y-w-m e,,
,,w*,"-wo~#
- y
--ques
-*e.essw--
2s,w e
2.-
e-a-u r,ew,.
v
,,,-~,vw
=,-
,-e--weww_-w-==,-
w-=*=m_"-w+=c=_m-mm-s.m--<+-,_mw-_
. - - = =-
[
I
_ ) Commonwealth Edison oni F. s Nat.onsmus CNeage ano,s A33'eis red'y to FisIDIce Boi 767 CN:s;: anots 60690 0767 December 13, 1989 i
r I
Director of Nuclear Reactor Regulation US Nuclear Regulatory Comunission Mail Station F1-137 Washington, DC 20555 i
Subject:
Zion Nuclear Power Station, Units 1 and 2 i
License Nos. DPR-39 and DPR-48 NRC Docket Nos. 50-295 and 50-304 i
Supplemental Response to CL 89-21 Reference -November 30, 1989 letter from G.E. Trayna to NRR i
Dear Sir:
The letter indicated in the reference above provided the status of the majority of the issues that were discussed in Generic Letter 89-21. Also contained in the letter was a request for'an extension of the due date until December 13, 1989 in order to provide the status of USI/MPA Numbers A-1, 3, 12 and 43.
l t
The status of these remaining open items is provided below.
U$1/MPA NLPet R TITLE STATU1/DATE CO M NTS A.)
Water Manner C
All 8 Steam Generators were modified to install "J" tubes on the feed ring to prevent water hammer. The last $G i
was modified in Apr41, 1962 1
A-3 Westinghouse $ team Generator C
June 17, 1985 letter from t
Tube Integrity G.t. Alexander to H.R. Denton provided Zion response A.12 Fracture Toughness of Steam g
Generator and seactor Coolant Pump Supports l
A.43 Containment Emergency Sump NA Performance
.e--09ttria+7& Jg.
y c
.i
MI 1
k, J
Director of Nuclear Reactor Regulation 2-December 13, 1989 l
Please direct any additional questions that you may have regarding
[
this matter to this office.
Very truly yours, Clenn E. Trayna Nuclear Licensing Admittistrator
/Imws0448T l
cc Chandu Patel-NRR Senior Resident Inspector >2 ion t
r i
f i
e i
,e
PLANT ZION.STATICW UNITS.1 AND.2..
DOCKET N0(S). 50-295 4 50 304.
-PROJECT PANAGER C..P. Patel TECPNICAL CONTACT A..Serkir USl NO. A-1 TITLE Water. Hammer..
MPA NO. N/A-TAC NOS.
ISSUES
SUMMARY
i This Unresolved Safety Issue (USI) was resolved in Maren 1984 with the publication of NUREG-0927, " Evaluation of Water Hammer in Nuclear Power Plants
- Technical Findings Relevant to Unresolved Safety issue A-1."
Also on March 15, 1984, the EDO sent the Commissioners SECY 84-119 titled, " Resolution of Unresolved Safety issue A-1, Water Hammer."
In SECY B4-119, the staff concluded that the frequency and severity of water hammer 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 procedurest (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, it did involve Standard Review Plan (SRP) changes (forward fits) and a comprehensive set of guidelines and criteria to evaluate and upgrade utility training programs (per TMI Task Action Plan item I.A.2.3).
In addition, the assumption was made that for RWRs 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 hammer that opened a direct pathway to the environment.
IMPLEMENTATION.AND. STATUS.SUMWARY.(PLANT. SPECIFIC):
All steam generators were modified to install "J" tubes on the feed ring to prevent water hammer. The last steam generator was modified in April 1982.
I i
l
L REFEREWCES:
' Zion Station Units 1 & 2 A-1 1..
REQUIREMENT DOCUNENTS:
TITLE NUDOCS NO.
DATE Letter from 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 Hammer in Nuclear Power Plants-Technical Findings Relevant to Unresolved Safety, Issue A-1" NUREG-0993 Rev. 1 830E060418 March 1984 "Reoulatory Analysis for.
for USl A-1, 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, " Resolution 03/15/84 of Unresolved Safety A-1, Water Hammer" LetterfromG.Trzyma(CECO)toNRC 12/13/89 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS td).
DATE l
.5
IL i
1 PLANT ZICW. STATION UNITS.1 AND 2 DOCKET N05, 50-295 8.50 304 PROJECT MANAGER C..P..Patel USI NO. A,-2 TITLE: Asymmetric Blowdown load.on Reactor Primary Coolant. System i
MPA'NO. D-10 TAC NOS. 08872/08873 I
ISSUES
SUMMARY
]
This USI was resolved in January 1981 with the publication of NUREG-0609, i
" Asymmetric Blowdown loads on PWR Primary Systems."
l 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 reouired to reassess the reactor vessel i
support design of its facility. The staff expanded the scope of the problem in January 1978 with a request for additional information to all PWR licensees.
HUREG-0609 provided guidance for these analyses.
For operating PWRs, Multi-Plent Action (MPA) Item 0-10 was established by NRC's Division of Licensing for implementation purposes.
During the course of the work on 051 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, af ter substantial nw technical work, it was demon-strated that pipes would leak before break and that new fracture mechanics techniques for the analyzing of piping failures assured adequate protection j
sgainst 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 r
10 CFR Part 50) published in the Federal R.egister in final form on April 11, 1986. and in a subsequent revision to GDC-4 published in the Federal Register on iuly 23, 1986, in addition, it has also been satisfactorily demonstrated in l
the course of the A-? effort that there is a very low likelihood of simultaneous pipe loading with both LOCA and safety shutdown earthouake (SSE) loads.
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 conlant rain loop piping.
IMPLEMENTATION.AND STATUS
SUMMARY
.(PLANT. SPECIFIC):
The staff issued a Safety Evaluation of Westinghouse Topical Reports dealing i
with elimination of postulated pipe breaks in PWR primary loops (Generic
. Letter 84-04)
The staff concluded that an acceptable technical basis was providad so that the l
asymmetric blowdown loads resulting from double ended pipe breaks in main coolant loop piping need not be considered as a design basis for certain plants l
in Westinghouse Owner's Group, provided the leakage detection system at the facility was sufficient to provide adequate margin to detect the leakage from the postulated circumferential throughwall flaw utilizing the guidance of l
Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systers," with the exception that the airborne particulate radiation monitor did not have to be seismically qualified.
At least one leakage detection system l
-with a sensitivity capable of detectino 1 gpm in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> had to be operable.
L The licensee performed an evaluation of leak detection system and concluded in L
April 1984, that it met NRC's criteria.
However, the licensee did not send any l
-letter to the staff until November 30, 1989 response to GL 89-21 on USI. The issuance of revised GDC-4 in April 1986, covered this issue generically.
l
l_
REFERENCES:
Zion Station Units 1 & 2 A-2 1.
REQUIREMENT DOCUMENTS TITLE NUDOCS NO.
DATE Generic Letter " Evaluation of Primary Systems for Asymmetric LOCA Loads" 01/20/78 Task Action Plan A-2, "Asymetric Blowdown Loads on Reactor Primary Coolant System," NUREG-0371 Task Action Plans for Generic Activities 11/78 "Asymetric Blowdown loads on PWR Primary Systems," NUREG-0609 US NRC NRR 01/81 GDC-4, " Environmental and Dynamic Effects Design Basis" GL 84-04, " Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main loops."
2.
ItiPLEMENTATION DOCUMENTS:
(1) WCAP 9558 Rev. 2 " Mechanistic Fracture 5/81
. Evaluation of Reactor Coolant Pipe containing a Postulated Circumferential Throughwall Crack" (2) WCAP 9787 " Tensile and Toughness Properties 5/81 of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation" (3) Letter from E.P. Rahe (K) to D.G. Eisehut (NRC) 11/10/81
" Westinghouse Reponse to Questions and Coments raised by members of ACRS subcomittee on Metal Components during the Westinghouse Pre-sentation on September 25, 1981."
(5) Letter from G. Trzyna (CECO) to NRC 11/30/S9 l
l i
PLANT ZION STATION UNITS 1 AND 2 DOCKET HOS, 50-295 &.50-304 r
PROJECT MANAGER ( P. Patel USI NO. A-9 TITLE: ATWSRule(10CFR.50.62[
MPA NO. A-20 TAC N05, 59160/59161,
~1SSUES SUMNARY:
This USI was resolved in June 1984 with the publication of a final rule (10 CFR 50.62) to recuire improvements in plants to reduce the likelihood of failure of i
the reactor protection system (RPS) to shut down the reactor following anticipated transients and to mitigate the consecuences of an anticipated transient 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 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 liouid control system (SLCS) for EWRs; and 50.62(C)(5), automatic trip of recirculation pumps under conditions indicative of an ATWS for BWRs.
Information requirements and an implementation schedule are also specified.
I IMPLEMENTATIONAND. STATUS.
SUMMARY
.(PLANTSPECIF,10),:
-The Westinghouse Owners Group developed a set of conceptual ATWS mitigating system actuation circuitry (AMSAC) designs The staff reviewed the design (WCAP-10858) generic to Westinghouse plants.
and issued a safety evaluation on July 6, 1986.
By letter dated June 5,1986, Commonwealth Edison Company (CECO) submitted a detailed design description of the AMSAC proposed for Zion Unit I and P.
The staff 1ssued a safety evaluation on May 22,'1989. The licensee was expected to complete installation by Fall 1989 for Unit 1 and by Spring 1990 for Unit 2.
However, by letter dated September 8, 1989, the staff was informed that Ceco will not be able to implement the change by Fall 1989 for Unit 1.
The proposed implementation schedule for Unit 1 is the Spring 1991 refueling outage. The licensee will still meet the Spring 1990 date for Unit 2.
i 1
l p
l-1
REFERENCES:
Zion Station Units 1 & 2 A-9 1.
REQUIREMENT. DOCUMENTS
_ TITLE NUDOCS.u0.
DATE NUREG-0460, and Supplerents, 03/80
" Anticipated Transients Without Scram for Light Water Reactors" Federal Register Notice 49 FR 26045 (10 CFR 50.6?)
06/26/84 2.
IMPLEMEWTATION DOCUMENTS:
TITLE NUDOCS NO.
DE Letter from P. LeBlond (CECO) to H. Denton 6/5/86 Letter from C. Rossi (NRC) to L. D. Butterfield 7/6/86 (WOG)
Letter from P. LeBlond (Ceco) to H. Denton 1/9/87 Letter from P. LeBlond (CECO) to H. Denton
?/6/87 Letter from R..A. Newton (WOG) to J. Lyon 2/26/87
-Letter from P. LeBlond (CECO) to NRC 7/21/87 Letter from R. A. Newton (WOG) to J. Lyon 8/3/87.
Letter from G. Trzyna (CECO) to NRC
?./26/88 Letter from G. Trzyna (Ceco) to NRC 5/25/88 Letter from G..Trzyna (CECO) to NRC 8909130351 9/8/89 9
I 1
FLANT Zion. Station. Units 1 6.2 DOCKET N0(S).
50-295 and.50-304 1
PROJECT MANAGER C. P..Patel TECHNICAL CONTACT B. Elliott.
USI NO. A-11 Til E Reactor vessel. Materials Touchness.
MPA NO. A-7 TAC NOS. 6689/8182 j
MUES
SUMMARY
This USI was resolved in October 1982 with the publication of NUREG-0744,
" Pressure Vessel Material Fracture Toughness.". NUREG-0744 was issueo 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 f ail, 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 1
typical of current operating plants, reactor vessel fracture toughness properties. provide adequate margins of safety against vessel failure; however, as plants accumulate more and more service time, neutron irradiation reduces the material fracture toughness and initial safety margins.
]
Appendix G to 10 CFR Part 50 requires that the Charpy upper shelf energy throughout the life of the vessel be no less~than 50 ft-lb unless it is i
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 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):
By letter dated December 20, 1985, the Commonwealth Edison Company informed thestaffthataccordingtoCEco'sestimate,anupgr-shelfenergyof50ft-1bs l
will be reached at a fluence of approximately 5x10 neutrons /cm. According r
to CECO's estimate, the earliest date that the upper-shelf energy could fall below 50 ft-lbs would be in 1994 This issue will continue to be evaluated as additional surveillance capsule data becomes available.
If that evaluation confirms the validity of the 1994 estimate, then the required notification per Appendix G will be made in accordance with the three-year lead time.
,m..
REFERENCES:
Zion Station Units 1 & 2 A [
1.
. REQUIREMENT DOCUMENTS i
TITLE NUDOCS NO.
DATE NUREG-0774, Revision 1
" Pressure Vessel Materiel Fracture Toughness" Generic l'etter 82-26, " Pressure i
Vessel Material Fracture loughness" 11/12/82 2,
IMPLEMENTATION DOCUMENTS:
- TITL E NUDOCS.NO.
DATE Letter from P. LeBlond to 8512300153 12/20/85 l
H. Denton t
3.-
VERIFICATION. DOCUMENTS:
TITLE NUDOCS NO.
DATE 1
4 e
o t
- )d ;
I PLANT ZION STAT 10tl UNITS 1 AND 2 C0CKET N0(S).
50-495 & 50-304 PROJECT MANAGER C. P. Patel TECHfilCAL CONTACT D. Thatcher USI fiO. A-17 TITLE Systems Interactions'in Nuclear Power Plants PPA NO.
TAC NOS.
ISSUES SUWARY:
Generic Letter (GL) 89-18, dated September 6,1o89, was sent to all power reactor licensees and constitutes the resolution of USI A-17.
The generic letter did not require any licensee actions.
GL 89-18 had two enclosures which (a) outlined the bases for the resolution of 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 programs, such as the Individual Plant Examination (IPE) for Severe Accident Vulnerabilities.
Specifically, the staff expected that insights concernino 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 USl was the expectation that licensees would continue to review information on events at operating nuclear power plants in accordance with the requirenents of TM1 Task Action Plan Item 1.C.5 (NUREG-0737).
IMPLEMENTATION AND STATUS
SUMMARY
(PLAtlT SPECIFIC):
Ceco submitted a response to Generic Letter 88-20 on Individual Plant Examination (IPE) for severe accident vulnerabilities.
Zion IPE will be completed by November 1991.
By letter dated September 28, 1972, the staff requested CECO to review for potential impact of circulating water system and five protection system f ailure on safety-related system.
By letter dated November 17, 1972 CECO responded that there was no impact from these systems on safety-related systems.
l l
i.
REFERENCES:
Zion Station Units 1 & 2 A-17 1
I>1 1.
REQUIREMENT DOCUMENTS:
.{
TITti NUDOCS NO.
DATE Generic letter 89-18 09/06/89 NUREG-1174 " Evaluation of May 1989 Systems Interactions in Nuclear Power Plants" HUREG-1229 "Re9ulatory Analysis Au9ust 1989 for Resolution of US1 A-17" i
NUREG/CR-3922 " Survey and January 1985 i
Evaluation of System Interaction Events and Sources" l
NUREG/CR-4261 " Assessment of June 1986 System Interaction Experience in Nuclear Power Plants" NUREG/CR-4470 " Survey and August 1986 i
Evaluation of Vital Instrumentation and Control Power Supply Events" i
NRC Letters to Licensees 9/72 Informing Licensees of Staff Concerns Regarding Potential i
Failure of Non-Category 1 Equiprent 2.
IMPLEMENTATION.0OCUMENTS:
TITLE NUDOCS.N0; DATE Letter from M. H. Richter (CECO) 8911080207 10/27/89 to NRC (DCS)
?
3.
VERIFICATION. DOCUMENTS:
TITLE NUDOC.NO.
DATE I
+
i t
PLANT ZION STATION UNITS 1 AND 2 DOCKET NOS. 50-295 & 50-304 PROJECT MANAGER C. P. Patel USI NO. A-24 TITLE: OUALIFICATION OF CLASS 1E E001PMENT MPA NO. B-60 TAC NOS. 42472/42473 REQUIREMENTS
SUMMARY
1 The publishing of NUREG-0588, Rev. 1, " Interim Staff Position on Environmental Oualification of Safety-Related Electrical Equipment" in July 1981 " Completed the staff resolution of Generic Technical Activity A-24, Qualification of Class i
1E Safety Related Equipment." Part I of the report is the original "For Comment" NUREG 0588 which in conjunction with the DDR Goidelines was endorsed by a Commission Memorandum and Order as the interim position on this subject until " final" positions were established in rule making. On January 21, 1983 the Commission amended its regulations, in 10 CFR 50.49 (the rule), effective February 22, 1983 to codify existing qualification methods in national standards, regulatory guides, and certain NRC publications including NUREG-0588.
The rule is based on the 00R Guidelines and NUREG-0588. Which " provide guidance on (1) how to establish environemntal service conditions, (2) how to select methods which are considered appropriate for qualifying the equipment in different areas of the plant, and (3) other areas such as margin, aging, and documentation." NUREG-0588 does not address all areas of oualificaitont it does supplement, in selected areas, the provisions of the 1971 end 1974 vresions of IEEE Standard 723. The rule recognizes previous qualification efforts completed as result of Memorandum and Order CLI-80-20 and also reflects different IEEE 323 versions dependant on the date of the construction permit SER. Therefore, plant specific requiements may vary in accordance with the rule.
In summary, the resolution 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 e
found in the determination of compliance with 10 CFR 50.49. This was addressed by 72 SERs for operating plants shortly after publication of the rule and subsequently in operating license reviews pursuant to SRP 3.11.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
By Order dated October 24, 1980, the licensee was ordered to qualify all safety-related electrical equipment in accordance with 00R Guidelines or NUREG-0588 by no later than June 30, 1982.
The licensee submitted its program by letters dated May 21, 1981, August 26, 1981 February 16, 1982, and March 24, 1982. The staff issued a Safety Evaluation on December 14, 1982, and identified several deficiencies. A meeting was held on January 25, 1984 to discuss the-licensee's proposed resolution to each of the deficiencies identi-fied and also 20 discuss a justification for continued operation (JCO) for those equipment which were not qualified.
The licensee documented completion of the qualification program by a letter dated April 10, 1984 Final Safety i
Evaluation was issued a November 21, 1984.
First inspection was conducted in January 1985. Several follow-up inspections have been conducted and some l
violations have been identified. However, no Civil Penalty has been accessed.
I
f
,G -
REFERENCES:
Zion Station Units 1 & 2 A-24 l
1.
REQUIREMENT DOCUMENTS T I TL t NUDOCS.NO.
DATE 00R " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Ecuipment in i
Operating Reactors l
NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety Related Electrical Ecuipment 12/79 i
Commission Memorandum and Order, 80-CL1-21, on D0R Guidelines and NUREG-0588 05/23/80 t
Order for Modifications of Licenses Concerning Environmental
' Qualification of Safety-Related Electrical Equipment i
NUREG-0588, Revision 1 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 Electrical Equipment 07/81 2.
IMPLEMENTATION. DOCUMENTS:
TITLE NUDOS NO.
DATE LetterfromS.-Varga(NRC)to L. De1 George 12/14/82 Letter from F. Lentine (CECO) to 8405020309
[
.H. Denton (NRC) 04/10/84 Letter to D. Farra (CECO) 8412030693 11/21/84
}
from S. Varga (NRC) l-t l
10CFR.50.49 Rule 3.
VERIFICATION DOCUMENTS:
L 1.
Letter from G. Zech (NRC) 4/15/85 L
to C. Reed (CECO) l-:
2.
Letter from J. J. Harrison (NRC) 8/21/85 to C. Reed (CECO) 3.
Letter from J. J. Harrison (NRC) 8/7/86 to C. Peed (CECO) 4 Letter from J. J. Harrison (NRC) 1/30/87 to C. Reed (CECO)
l PLANT tl0N STATION UNITS 1 AWD 2 DOCKET NOS.
50-295 8.50 304 PROJECT MANAGER C. P. Patel USI NO. A-26 TITLE Reactor Vessel. Pressure. Transient Protectico..
MPA NO.
B 04 TAC N05, 06705/08342
)
ISSUES SUMMApV:
)
i 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 FWRs were requested to provide an overpressure prevention system that could be used whenever the i
plants were in startup or shutdown conditions. The issue effected all operating 3
and future plants, and the staff established MPA P-04 for implementing the solution at operating PWRs.
Since 1972.there have been numerous reported incidents of pressure transients in PWPs where technical specification pressure and temperature limits have been exceeded. The majority of these events occurred while the reactors were in a j
solid-water condition during startup or shutdown and at relatively low reactor vessel temperatures. Since the reactor vessels have less toughness at lower temperatures, they are more susceptible to brittle fracture under these condi.
tions than at r.ormal 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 taken to minimize the number of future transients and reduce their severity.
s 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 generic letter provides guidance regarding review of pressure-temperature limits and indicates that licensees may have to revise low.
temperature-overpressure protection setpoints.
1MPLEMENTATIONAN'D. STATUS.
SUMMARY
(PLANTSPECIFM:
)
By letter dated August 11, 1976, the staff reauested Coninmonwealth Edison Company (CECO) to begin efforts to design and install plant systems to mitigate the consequence of pressure transients at low temperatures. CECO was also requested to examine operating procedures and to change administrative
. procedures to guard against initiating overpressure events.
By letters dated September 2,1977, February 26, 1979 and September 26, 1979, CECO proposed to modify the actuation circuitry of the existing air operated pressurizer relief valves to provide a low pressure set point at 435 psig during startup and shutdown conditions.
By letter dated April 28, 1980, the staff approved the technical specifications which included overpressure I
mitigating system and other administrative controls used to limit overpressurization events.
l
_m.,
- ,p i;
F
REFERENCES:
Zion Station Units 1 and 2 i
A-26 1
1.
RE001RD1ENT DOCUMENTS j
i TITLE NUDOCS NO.
DATE
]
" Reactor Vessel Pressure Transient Protection for PWRs."
9/78 i
NRC Letters.to Licenstes Informino Licensees of Staff' Concerns Regarding l
Overpressure Low-Temperature
~
-Conditions in PWRs Auoust 1976 i
Generic Letter 88-11, "NRC 7/12/P8 l
Position on Radiation Embrittlement of Reactor Vessel Materials and
[
lts impact on Plant Operatione j
Standard Review Plan Section 5.2
)
i 2.
1HPLEMENTAT10N DOCUMENTS:
TITLE NUDOCS.WO.
DATE NRC Letter (Schwenecer) to CECO April 28, 1980 (Peoples)
" Pressure mitigating System July 1977 Transient Analysis Results" 4
prepared by Westinghouse for the Westinghouse User's group on renetor coolant system over-pressurization.
Ceco letter (Bolger) to NRC (Schewencer)
September 2, 1977 CECO letter (Reed) to NRC (Denton)
February 26, 1979 4
CECO letter (Peoples) to NRC (Denton)
September 26, 1979 3.
VERIFICATION. DOCUMENTS:
TITLE NUDOCS_NO.
DATE Letter from E. Jordan (NRC to December 16, 1986 C. Reed (CECO)
~
f l
r
- ~
m m._.
.. +
..,.m
~ PLANT ZION. STATION L' NITS 1. AWD 2 '
DOCKET NOS. 50 295.&.50-304 PROJECT MANAGER C..p..Patel USI NO. A-36 TITLE:
Contral of Heavy-Loads.ph'ase 1&Il MPA NO. C-10&l5 TAC NOS. 08092/.02093.& 52285/52280 l
liSUES.$UWARY:
This 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 11, respectively, of the resolution of this issue at operating plants.
In nuclear power plants, heavy loads may be handled in several plant areas, if these loads were to drop in certain_ locations in the plant, they may impact spent fuel, fuel in the core, or equipment that may be required to achieve safe shutdown 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 chances to ensure the-safe handling of heavy loads. The guidelines proposed in NUREG-0612 include definition of safe load paths, use of load bandling procedures, trainino of crane 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 1-(six month response) ano Phase 11 (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 B30.9 7.
Design of cranes to ANSI B30.2 or CMAA-70 Phase 11 responses were to address Sections 5.1.2 thru 5.1.6 of NUREG-0612 which covered the need for electrical interlocks / mechanical stops. or alternatively, single-failure-proof cranes or load drop analyses in the spent fuel pool area (PUR), 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 11 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 11 repert.
Based on the improvements in heavy inads handling obtained from implementation of NUREG-0612 (Phase 1), 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 11 was considered completed.
e
-+
-,~,
--e--en-,,.,
, -, -,a, e
e r-,
-h y
F i
E
REFERENCES:
Zion Station Units 1 & 2 i
A-36 j
t i
IMPL EFENTAT10W. AND STATUS
SUMMARY
(PLANT SPECIFICJ:
The Commonwealth Edison responded by several letters ranging from hovember 7, i
L 1980 thru July 23, 1983. The staff issued a Safety Evaluation on Maren 22, i
1984, which concluded that the guidelines of NUREG-0612, Section 5.1-1 and 5.3.
were satisfied and Phase 1 of this issue for Zion Station was acceptable.
Based on the review of Phase I responses, the staff concluded that the implementation of Phase I had provided sufficient protection such that the risk associated with potential heavy load drops was acceptably small. No I
n further action was required to reduce the risk associated with the handling of heavy loads.
However, the licensees were encouraged to implement any actions j
they might have considered appropriate.
TITLE NUDOCS DATE 1.
REOUIREMENT DOCUMENTS:
Letter, Darrell G. Eisenhut, NRC, to all licensees, applicants for Ots 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-0612" 06/28/85 2.
IMPLEMENTATION. DOCUMENTS tetters from CECO to NRC November 7, 1980 January 2, 1981 April 10, 1981 July 7, 1982 October 12, 1982 July 27, 1983.
Letter from S. Varga (NRC) to D. Farrar (CECO)
March 22, 1984 3.
VERIFICATION DOCUMENTS:
N/A
L -
t 0
PLANT' Zion Station Units 1 & 2 DOCKET N0(S).
50-295 and 50-304 PROJECT MANAGER C. P. Patel UST NO. A-44 TITLE Station Blackout i
68,630/68631 ISSUES
SUMMARY
This US1 was resolved in June 1988 with the publication of a new rule (10 CFR 50.63) and Regulatory Guide 1.155.
1 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
)
showed that station blackout could be an important risk contributor and i
operating experience has indicated that the reliability of ac power systems might be less than originally anticipated.
For these reasons station blackout was designated as a U51 in 1980. A proposed rule was published for comment on March 21, 1986.
A final rule, 10 CFR 50.63 " Loss of All Alternating Current Power " was published on June 21, 1988 and became effective on July 21, 1988.
Regulatory Guide 1.155 was issued at the same time as the rule and references an industry guidance document NUMARC-8700.
In order to comply with the A-44 resolution, licensees will be required to:
maintain onsite emergency ac power supply reliability above a minimum level develop procedures and training for recovery from a station blackout
[
determine the duration of a station blackout that the plant should be able to withstand use an alternate qualified ac power source, if available,-to cope with a station blackout evaluate the plant's actual capability to withstand and recover from a i
station blackout 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 their coping analysis within 270 days from issuance of an operating license or the effective date of the rule, whichever is later.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
Licensee responded on April 17, 1989. The response is under review.
The review is expected to be completed by December 31, 1990.
The licensee is planning to make the modifications in year 1990. Certain procedure changes will be implemented one year af ter the notification provided by NRR in accordance with 10 CFR 50.63(c)(3).
Based on the above, this item should be implemented latest by December 31, 1992.
w
-,-w n
s c
w
c f
r R[ffRENCES:
Zion Station' Units 1 & 2 A 44 1.
REQUIREMENT.00CUMENTS TITL E WUDOCS.NO.
DATE l
10 CFR 50.63, " Loss of All Alterneting Current Power" 06/21/88 Regulatory Guide 1.155, j.
" Station Blackout" 08/88 b
2.
IMPLENEWTATION DOCUMENTS:
TITLE NUDOCS.NO.
DATE L
Letter from M. Richter (Ceco) 8904240429 4/17/89 to T. Murley (NRC) k 3.
VERIFICATION DOCUMENTS:
I TlILE NUDOCS.WO.
DATE b
b
)
s
l PLANT ZION STATION UNITS 1 AND 2 DOCKET NOS. 50-295.6 50 304 PROJECT MANAGER C..P. Patel USI NO. A-46 TITLE Seismic Qualification of Ecuipment in Operating Plauts.
MPA NO. B;105 TAC NOS. 69492/69493 ISSUES
SUMMARY
USI A-46 was resolved with the issuance of GL B7-02 on February 19, 1987, which endorsed the approach of using the seismic and test experience data proposed by the Seismic Qualification Utility Group (SQUG) and Electric Power Research 3
Institute (EPRI). This approach was endorsed by the Senior Seismic Review and Advisory Panel (SSRAP) and approved by the NRC staff.
The scope of the review was narrowed to equipment required to bring each
. affected plant to hot shutdown and maintain it there for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The review includes a walkthrough of each plant which is recuired to inspect eouip-ment.
Evaluation of equipment will include:
(a) adequacy of equipment
^,
anchorage; (b) functional capability of essential relays; (c) outliers and deficiencies (i.e., equipment with non-standard configurations); and (d) seismic systems interation.
As an outgrowth of the Systematic Evaluation Program (SEP), the need was identified for reassessing design criteria and methods for the seismic quali-fication of mechanical equipment and electrical ecuipment. Therefore, the seismic cualification of the eouipment in operating plants must be reassessed to ensure the ability to bring the plant to a safe shutdown condition when subject to a seismic event. The objective of this issue was to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at operating plants in lieu of attempting to backfit current design criteria for new plants.
Generic letter 87-02 with associated guidance, required all affected utilities to evaluate the seismic adequacy of their plants. The specific requirements and approach for implementation are being developed jointly by SOUG and the staff on a generic basis before individual member utilities proceed with l
plant-specific implementation.
IMPLEMENTATION AND STATUS.
SUMMARY
(PLANT.SPECIEIC):
For.All Plants:
The Generic implementation Procedure (GIP), Revision 0, was submitted by SOUG on June 3, 1988. The staff issued a Generic Safety Evaluation (SE) on July 29, 1988 endorsing much of the GlP but with about 70 open items to be resolved. By letter dated October 7, 1988, the licensee committed, continnent upon the final agreerent reached between the staff and SQUG by mid 1089, to-perform plant walkdown in 1990 through 1991 for Unit 2, and in 1991 through 1992 for Unit 1 and common equipment. After a series of meetings. SOUG rubnitted Revision I to the GIP on December 23, 1988.
Supplemental information was submitted by SQUG on March 17. 1989.
The staff has prepared a supplemental SE for GIP, Revision I and has submitted it to the CRGR for review. The target date for issuance of the supplemental SE is November 1989.
An additional supplement is scheduled for June 1990 and overall closeout of implementation projected for 1993.
4 l
-a
i i
REFERENCES:
Zion Station Units 1 & 2 A-46 1.
REQUIREMENTS DOCUMENTS:
i TITLE NUDOCS NO Date 1.
REOUIREMEt!T DOCUMENTS TITLE NUDOCS NO.
DATE Generic Letter 87-02, "Verifi-cation of Seismic Acequacy of J
Mechanical ard Electric Equipment in Operating Reactors" 02/19/87 t!UREG-1211. " Regulatory Analysis for Resolution of Unresolved Safety
!ssues A-46..."
02/87 NUREG-1030, " Seismic Qualification l
of Equipment in Operating Plants, Unresolved Safety Issue A-46" 02/87 i
Letter attached with " Generic Safety Evaluation Report on 500G GIP,) Revision 0,"fromL.Shao (NRC toNeilSmith(SQUG) 07/29/88 2.
111PLEME!! TAT 10N DOCUMENTS:
TITLE NUDOCS NO.
DATE
" Generic Implementation Procedure (GIP for Seismic Verification of Nuclear Plant Equipment," Pcvision 0 06/88
" Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment," Revision 1 12/88 LetterfromG.E.Trzyna(CECO) to T. Murley 88010120202 October 7, 1988 6
L e
l FLANT Zion Station Units 1 & 2 DOCKETN0(S).
50-295 and 50-304 PROJECT MANAGER C. P. Patel TECHNICAL CONTACT J. Pauck i
USI NO. A-47 TITLE Safety Implication of Control Systems in LWR L
Nuclear Power Plants MPA NO.
TAC NOS.
ISSUES SUt4ARY:
USI A-47 was resolved September 20, 1989, with the publication of Generic L
Letter (GL)69-19.
The generic letter states:
The staff has concluded that all PWR plants should provide automatic steam generator overfill protection, all BWR plants should provide automatic reactor vessel overfill protection, 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 automatic overfill protection is available to mitigate main i
feedwater overfeed events during reactor power operation.
- Also, the system design and setpoints should be selected with the objective of minimizing inadvertent trips of the nairi feedwater system during plant startup, normal operation, and protection system surveillance. The Technical Specifications recommenda-tions are consistent with the criteria and the risk considera-tions of the Commission Interim Policy Statement on Technical Specification Improvement.
In addition, the staff recommends that all BWR recipients reassess and modify, if needed, their operating procedures 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."
Also, page 2 of the generic letter provides for additional actions for CE and B&W plants. The generic letter provides amplifying guionnce for licensees.
The generic letter requires that licensees provide NRC with their schedule and commitments within 180 days of the letter's date.
The implementation schedule for actions on which commitments are made should be prinr to startup after the first refueling outage, but no later than the second refueling outape, beginning 9 months after receipt of the letter, IMPLEMEllTATION AND STATUS SuttMARY (PLAtlT SPECIFIC):
Response due to NRC by March 1990 9
O i
REFERENCES:
Zion Station Units 1 1. 2.
i l-A-47
]
1.
RE0VIREMENT DOCUMENTS TITLE NUDOCS NO.
DATE i
Generic Letter 89-19 09/20/89
" Request for Action Related to Resolution of USI A-47" NUREG-1217 " Evaluation of Safety June 1989 implications of Control Systems in LWR Nuclear Power Plants" NUREG-1218 " Regulatory Analysis July 1989 for Resolution of USI A-47" e
2.
lHPLEMENTATIOM 00CllMENTS:
TITLE NUDOCS t:0.
DATE 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE p
I L
e
?
- PLANT Zion Station Units 1 &'2 DOCKETN0(S).
50-295 and 50-304 PROJECT MANAGER C. P. Patel USI NO. A-49 TITLE Pressurized Thermal' Shock NPA NO.
A TAC NOS.
60777/60778 ISSUES
SUMMARY
The final rule (10 CFR 50.61) on pressurized thurmal shock (PTS) was approved by the Commission in July 1985.
Regulatory Guide 1.154, " Format and Content of Plant-Specific Pressurited 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 demonstrate that PTS is not a concern for the
- facility.
Neutron-irradiation of reactor pressure vessel weld and plate materials decreases the fruture toughness of the materials.
The fracture toughness sensitivity
- r uiation-induced change is increased by the presence of certain
]
materials suc., n copper. Decreased fracture toughness makes it more likely 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-precursor events have happened that could have potentially threatened vessel 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
- se<ere pressurized overcooling event cannot be ruled out.
- IMPLEMENTATION AND STATUS
SUMMARY
(PLANT-SPECIFIC):
By letter dated January 17, 1986, the Commonwealth Edison Company (CECO)-submitted information on the material properties and the fast neutron fluence of the reactor pressure vessel in compliance-with the requirements of 10 CFR 50.61.
In Safety Evaluation dated August 14, 1986 the staff concluded that Zion Unit I did not meet-the fracture toughness requiren..'nts of 10 CFR 50.61 neither for 32 effective full power years of operation or to the end of the current license.
Therefore, the. staff requested that Ceco submit plans for flux reduction program as required
- by paragraph 10 CFR 50.61(b)(3).
CECO submitted another submittal on December 29, 1986.
The staff again did not support CECO's position in a Safety Evaluation dated May 7, 1987. A third submittal was made by CECO on September 18, 1987. The staff issued a Safety Evaluation on February 26, 1988, reconfirming its previous conclusion on August 4, 1986 and again on May 7, 1987.
Zion Unit 2 meets the screening criteria for 10 CFR 50 61 On December 28,
- 1989, the staff has proposed (54 FR 52946) to change PTS Rule.
Zion Unit I will be in compliance with the revised rule when it becomes effective.
This issue is considered open for Zion Unit 1.
QD f
+
REFERENCES:
Zion Station Units 1 1 2 t
A 1._
REOUIREMENT DOCUMENTS TITLE NUDOCS NO.
DATE
-10 CFR 50.61, " Fracture Toughness 7/85 8
Requirements for Protection Against D
Pressurized Thermal Shock Requirements" Reg. Guide 1.154, " Format and Content 1/89 of Plant-Specific Pressurized Thermal l
Shock Safety Analysis Reports for PWRs" SECY 82-465, " Pressurized Thermal l,
Shock"-
11/23/82 SECY 83-288,=" Proposed Pressurized Thermal Shock Rule" 07/15/83 Regulatory Guide 1.154 '
" Format and Content of Plant-
. Specific Pressurized Thermal Shock Safety. Analysis Reports for Pressurized Water Reactors" 02/87 Generic Letter 88-11. "NRC Position i
on Radiation Embrittlement of Reactor Vessel Materials'and Its impact on Plant Operations" 7/12/88 2.
IMPLEMENTATION-DOCUMENTS:
.)
TITLE NUDOCS NO.
DATE.
]
LetterfromG. Alexander-(CECO)
'1/17/86 i
to H. Denton (HRC)
LetterfremS.Varga(NRC)to 8/14/86 D.-L..Farrer(CECO)
Letter from Ceco to NRC 12/29/86 Let'ter from D. Muller (NRC) to 5/7/87 i
,D. L. Farrer (CECO) i Letter frem L. De1 George (Ceco) to 9/18/87 NRC LetterfremD. Muller (NRC)to 2/26/88 L. Butterfield-(Ceco)
LetterfromD.Itu11er(HRC)to 10/6/88 H. Bliss (Ceco) 3..-
VERIFICATION DOCUMENTS:
TITLE NUDOCS N0.
DATE
ENCLOSURE 3 CATEGORIES FOR USI STATUS (1) Where an item is not applicable to the facility, "NA" is entered in the status-column.
fl.
(2) Where an iten is applicable to the facility, but no thanges were necessary, "NC" is entered in the status column. Also, "NC" is entered if the USI i
2
~' -
was implemented )rior to licensing, as no changes were necessary after issuance of-an 0..
No implementation dates are entered for items that are "NC."
(3)-Whereanitemisapplicable.tothefacilityandchangesarecomplete, "C" is entered in the status column and the date implementation was
-completed is entered.
(4) Where an item is applicable to the facility and is not fully i piemented, "1" is entered in the status column and the projected implementation date is entered.
(5) Where a USI resolution was recently issued and the licersee's evaluating their response, "E" is entered in the status column and the projected response date is entered, if-known.
s
M~~F
(: '
Q 1;
s s.
,;Mn; 4:
N,. y m-P2;e N.
1
~ 2..C, 4 -
,z.
. >.- s v L :.ty$.;.e r..si r.v.e i..e.,
. r+: ;
.. ~...
FOR lhM!T FR;' :R :ECT "AsA!!R3 vt n Pr',
.; a
.::,e
.ie:"E tr:*R. !F*. *VE Ne'M.c*
. ~.' i d.:N. i ' *:L ".*.: N'. '.~.*..N*t.: "**"tNi
- * ' r*.' " * *. *.E N T
.~
a.a
-. ~,
Y.?}ER
':At!-
siATUS it::LANT.NA"It!!:N;
.s
'd c.,-
,48!:s e...x... a a! n :.
.a:
r..:,
s.;t:
.r...,,..
s
. + ^
- se
- n. a3wr. t.l a t>..k.**
N s
- c..h'
%.a
~
aa -
- . c 94
-n..
i i a
%.t p.e. ng p s v..:n,
?"9..J NT
- v.:i. *.*.:
2 s.
. n. s?
-: +s 4 r.:g te.rA test ;t
- E /
N*.
'. N."'.'Yw\\ -
n... :..
s
.a
<w
..y......,
..PAi U
A ;;
.:E ii!;M. !!NEFA!:S it'3E ;N!!5R 1! !
N/A I PLANil lNLY A ild !!!AM !!NE:ATOR .'!E -
/
N. A
'!!W FLAN'i NLY
.c
- liEiF!!Y A Ci MPr-! Ir:RT i!E9 ?RC3 RAM 5A
.9r ! $W: :NLY A.c,..: x s ie..t..rs..= :R..:.A>.
/
t <:
.y.. :. rsi.v.
....m 4
.n
- 0?
MARK !! *:NTAINMENT CL:P.A"l:
/ /
'v A Na II :sR C1LY L5.o
,,.1...R,,.,R...,,,
.. : ic.
.:3 A J,i,5
,34/ 0!91 l'
- !CD L'R 1/9/19 "En !/: /!9
. A.or.
ti * ;;;? r0'. r.t. om?.'. c..F:ewn.t h'J
. d 4... 4 1..
...aay a
a e ra
- g. sNi.y e
.d t*
g...
- A, g <It:e.n..
4, R.i AL-i
.;e
.:."O LTR O'..:*.'.' 7. 0 a' 4 : 4
.E J
vs":.NE.::
A-!:. FRACTL'RE i. GENES! lF Si!?9
!/
N/A
- F AFTER !! CNLY
.Er R=,T..
un. :r.sa.a4.am..n*hT
..i.
..s *
'Vif :'.?FORi3 A-;'
315'E*! !NTERACTl N N:
'E-NOFICt;EMENTS A-:4 UALIFICATIONOFCLA!!IE' 04/1.'/24 0
.*E*: LIR
!Nif 1/i!
5 AFETV-RE'.?!!D E3UiFM.ENT.
A-04 El?!:P VE35EL ?Ri!!.?E !RAN!!EN! 09/:6/79 C CIC: LIR SER 4/*S!!0 - LT;FS
~.....w.
A :;-
Rh; i JT!:WN : E0' :RE*ENT3
! !. N/A NEJ FLANTS ONLY. IRF.
J
- a. t r.ette.nL e.r, we.A /.I,.a..asa.s N.r a
.nn..rNT.*i.,/
?.
e.
e..rr.o ?.R ? a7/ 8+
- r Jo e
v,.i.. <,e. 4
.tn
.r m.
r e. 6.:
A *:
/
~ N/A
!W; ONLY VW.VE i:CL OYNAM!C LDAOS AN!
...A.,....,,91...o i c. :-.r:
A 40
!!!!*!C *E3!!N CRl!ERIA -
/ /
NC
!L'EELMPED if A-43 GHCRT-TE*MFROGRAM:
- -42
- !:E ERACKS IN !D1 LING WATER
-l N/A ihR CNLY
- r. Er.......;.ns-A-43
- 'NTAINMENT EMERGENCY SUMP N:
lN:0 ;NLY F5:::RMANCE 4-34 3?ATICN !LACKCUT
! / 11:2 I 3ER 10/31/90 A,43 tu.r.*.r.W3.e Ay
.g?. t..r.w.n.yh.i f
1g
- we
.n.ec.o..~.' E v..:EVtrE a*t
..m
- JL A':..: Cbl.c 8M.14 C ':
.A r..
19 / ?..?.101. 4
.. s. 3 *.* < $". n.i.t - ' a A '. t.u. q p 9
a.7; 1e.ja;$0 g r.a'J NNS.tM.tp-I9..C.
v.,.
m er o. t.r. w.e.N'i. 4' N *F :, A'i.NC. :* ANT:
u, A 47
..'.... E r: f.v..'.'..:* !" Ai t. ".Ne c r 37;. ri.
- , tran..-
- v. g c - g.e g e.v..ry. *i :
.c
<e....
t...
Ild 5 4-49 "Y:RC;EN :'NTROL NEA!U:E3 41:
N.' A N/A 3*Y :siA!N
- .r?. e e..r u.v. e r.. :N.c u.;1q 3 vb.
..F r
- ,;c.'4 t r-". ':. e r.q {
.J jedQ '
E:::+. ! ". %.' s ' Nl. e.q{EW${ Z.u}{K f*!*t/g{
?
i
' ', $ Q J.' * [ Q.7
- ? !*y r*I-**?g/KU'*
g*J s.s 4.
s
....n.
... c k
. -Pf% ?--
^)
s w
J g.m.
i 4
Pa;tNs.
- nv/a,l o. s
.s s.
4 a
g
- i. ] C.'..i n t f f '. N*.*.u,t : :'. :. '. C. ] *J 7. 4 -
s a.
t
,nt. '.ytH' l
pao...
.. u t r.a '.r.* ? u;g a*.TE :
??'
~*
.c:
...t.......'tta;ty;.*y: gaa.r
'.= tie.r.:yt 'u.:.r :g7 :.t tg.er.:. vvei.'.
- c. '. +* ::r..a v.: g*.
....t.
.. ~
.a
. n.w.::.7
%4+{
- +m,. i.:
3
. 2 F.A.NI N!*:
(.
.1..".g '.
1 9 3.,s
'da?:3 gand*:
- j.** :*
a h C. '. N' ' ( t.n..* ' Ut. t.:
t
.e.<
.a w. f..= - ' ;t. Lw.w t*i n 4.
- 1-
.St a ' e. a'.y. I. a g * : ag 1-t 'a
- 2. t. 4 3.(
4 4
5
- tgr.t. ;
- -] v;;)...3.:s.t.e p,T,: c.
n
- 1. n *.
=.::T..o...t. n..e.r *:g3..:at p.n m.9p ti..eg ga v
- t
.tren af.y
.t.
4 4 4T..::r t. '. y n
- 4. s. 4. - e.r + :e...ise;;i.;. 7.it -. t. y'..;:: 4', v /.'
y,r 4.
t a.t
. : u 4.5+. :. 9 9.y i
... wn a.e. 5 - :tg t'e;w :rg.er.gi.c. n::
+
ga
.tyn.e.eg;; a.y.r..
m 9 4'. f.. s*.t t. y 101
-fa*K ! SHCR*-il:" :EC****-
i N?A
=X *.'.R av.y
't
- e.n.-
,agey.
- a.....v; *:;v..c. :c. -
i ia
.vy.. t. yp e..y
..m o,
w :qi, *.' r..g * ;; g.v.e s't : ~. =.,3:v.'a sea v,y..r g t;q.my.y
- +
~
4:
> a t
- t. e.r.
p t.m* :; mm.
u
..e..
.. J 3 3:.
- 3
- 4. >-s e:.S
- r. :. a...g ; st,:0
- c. :. e.f a a r g 9
..e 4 33
- t. "3.:*r,ac. wm..:4 y
- u..r.- ;.ay 4'N;.
i-4
.; t
."o, wy,y l
+
h3 4.i.a
- 4;r.'<Xa*
- :. *nt.e.c. H.$
u t
y.a
- -):?.n
- i.
- $. t t..a.i;9:
f.....
i i
- *"k. N C
- C.
l is 4 83
' [) A*t.t.
', N.t *u. N.:*a.' f :...E g u. -
g,' a*
P:,apy:q
- T. n.p. y t
e
. s t.r6 ;.g 4. :
g., g;n, e-ag.mai -. q,s-n<
...wr
- P.4.0 *it!S
- "3 i
- . t a. '
- :- w:
.. <.'.r... A.a ?..t.* y -
t :: -
r r
gr.
.' t.:
..i,::~qqru.tv.e.g I
g
.. a w...
Q.*4 J".p.t.p.!"P 4.?
- y r.h P..M.:..tr
. g /1( r:.:4.*
a.
- a:.e+v..;r 6n.. : u..t.:w.:'ii -
a-a A. '.s... w ;.:..:::..::::.:.g*.:
ava v:;se.e.r.g*i m.gf..t.ryo.. a
- s t aa.f:0 r
./.>+
c..
a m.r 1 ;c v. es.i. v ay Cn.?,
.t.. : c a.g*. *e.iq ae m. a.r.ve y, c.
.r y >> >
n 3:y si.g*g c.y.y..:-.-
4.
.a 1
r.m nr
' n'...:
aa.a+; m.6
- .r:yy w.*es s 3:.4p ::r.NT ^ 7,P.'../.:'a "w
F..F 3 '. ?. f. ' ',,' * '.,'.: '
- f.n" './".'e* 4 4 v
i aaa g..o-
- u t y o + s n.r.. ::gr. ty g n.'ET 4
N/A
- W" ?.N.Y
.n c vo.r.:3$...o34 1.a
- 3. 4 a. ags.
un.
, c: ATn.'.r L!a.t a
..~.
s
.. s
- a. s-
.....= ! C.c.S cy r..e.tt. e.g i A.
j,i we c.e..::r..u..u. D. Y 3 2 6
. a s
ii:57 7ERM FR:3 RAM
- -42.
!?E *RA*X! !N POILIN5'WATIR
/
N1 -
FaR *N'.f
- t a. r.*., e.
8 a. 4 '.
. e.t;{.s.u.eq', em.ep:.E.NrY.c".m"F
/
Nr.
4
."..' : " ". Y. '-
w.
.t e.: n. w.eN?.c.
a m
m l
- .42,
- id...N ! 4r.XcJT 1a/ *si,ig a.i
. at a er
..a. a,n..e.ig,)
F g. a s. -
.an.'. po.gs.ra a y gg;?..t.e.ngyp' yr.
v.t.:"..=.ge.:y e y:.qr p s a
..e,
.c n.a..?: w:N*:.
- rm w..
< 3 4t
- :>.*'a+.iy. :.te.: ay re.
.ia 4
- a *
.a.9[8Ee aE"<E'.
s i..
a:r.e. <.+:.i o., n ;. q
- a f
. e m.: =*.e.gi
.9. ac.;
.t,y.: ::
- a. c w
a..
. i.
I 1 4*
- eIf vvJ.a.e:t.aNs. r.aN;:n.
ar.
r
- . :n -
e Nrz :r eu r ;r.ug Nt. e
.a
......... s
- y: e+:
'.):
J e
- p,r*
- y ar %';".I **1:-.".*: ;g3
/
t i, / e*
3/4 9;y *.64TAIN
- 1 u
w
)
.rf ?.a'*
F:
a.v P
- e.'
- y (". p i a *3
.N 43.
- c. 6ar: v. :.an..' c o r N *i.
4
- .m.g,..,*:rPA.
,.2 3,
- u.,.., :.:
J
...e,... </ : :
- tr,/.,i :.3 1
Q
_