ML20012E937
| ML20012E937 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 03/22/1990 |
| From: | John Hickman Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-A-44, REF-GTECI-A-47, REF-GTECI-A-48, REF-GTECI-CO, REF-GTECI-EL, REF-GTECI-SF, REF-GTECI-SY, TASK-A-36, TASK-A-44, TASK-A-47, TASK-A-48, TASK-OR GL-89-21, NUDOCS 9004090121 | |
| Download: ML20012E937 (53) | |
Text
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NUCLEAR REGULATORY COMMISSION n
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wAsWNO TON, D. C. 70665 I
March 22, 1990 l
Docket No. 50-461 s
MEMORANDUM FOR:
File FROM:
John B. Hickman, Project Manager I
Project Directorate !!!-2 Division of Reactor Projects - III, IV, V and Special Projects i
SUBJECT:
STATUS OF IMPLEMENTATION OF UNRESOLVED SAFETY ISSUES AT CLINTON POWER STATION j
Thecurrentimplementationstatusofunresolvedsafetyissues(USIs)atthe l
Clinton facility is set forth in the enclosures to this memorandum.
Enclosure I contains a copy of the information provided by the licensee in their response to Generic Letter 89-21, i
In' addition Enclosure 2 contains a status sumary for each USI applicable to t
this f acility. This status sumary is based upon the licensee's response to the Generic Letter discussions with the licensee, and my review of available NRCrecordsandinformation. Appropriate NRR technical branches have also reviewed the USI status sumary and this memo. is a copy of the staff's data base printout for this facility.
It reflects the staff's assessment of USI implementation for all 27 USIs.
It is based on revi u 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 is as follows:
A-36:
Control of Heavy Loads Near Spent Fuel Additional procedural controls to be implemented prior to the next refueling outage (9/90). This is acceptable.
A-44:
Station Blackout Staff SER is due the fourth quarter 1990. Licensee procedure changes to be completed 5/90. This is acceptable.
A-47:
Safety Implication of Control Systems Licensee response to Generic Letter 89-19 due 4/90. This is acceptable.
I 9004090121 900322 ADOCK 0500g 1
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i 9
I File March 22, 1990 l
A-48:
Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment Licensee developing hydrogen control E0P for 12/90.
All equipment is installed with procedures in place. Will submit final analysis 6 months ef ter the generic SER is issued. This is acceptable.
/s/
John B. Hickman, Project Manager Project Directorate 111-2 Division of Reactor Projects - III, IV, V and Special Projects
Enclosures:
As stated cc w/ enclosures:
K. Eccleston DISTR' ItlT 0N Mln"RM NRC & Local PDRs PD32 Rdg. File JZwolinski LLuther PShemanski
~0GC EJordan RWessman c
Q o Min'A) y 9
PD32:FDh 32:PM PD13:PD JHICKMAN RWESSMAN JCP.AIG 5/so/90
$/Lt/90 1 47/90 a
l j
o U.601563 i
L30 89(11 27).LP'
')
80.120 til/N0lB POWER COMPANY IP cuNToN Ponta st AtioN. P.o. son ets, cLatON. wNois emt November 27, 1989 j
Docket No. 50 461 Nuclear Regulatory Commission Document control Desk r
Washington, D.C.
20555
Subject:
Clinton Power Station l
Response to Request for Information Concerning Status -of Implementation of Unresolved Safety Issue (USI) Reauirements (Cenerie latter 89-21)
Dear Sir:
t Please find enclosed the results of the Illinois Power (IP) review and reporting of the status of implementation of Unresolved Safety Issues for which a final technical resolution has been achieved.
i Sincerely yours, j
Nl D. L. Holtra her Acting Manager -
Licensing and Safety SFB/krm Enclosure cc:
Regional Administrator, Region III, USNRC NRC Clinton Licensing Project Manager NRC Resident Office l
Illinois Department of Nuclear Safety l-l 1
l J*!2Ci^^'; 891227 j
FI3R ADodg050004hsg M
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F po i
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Enclosure to C-601563 Fage 1 of 6 UNRESOLVED SAFETY ISSUES FOR UNICH A FINAL TFf"JafICAL RESOIETION MAS nFN ANYKVED USI/MPA NUMBER TITLE REF. D0CiMENT AFFLIGARILITY STATUS /BAIE*/IEEARKS A-1 Water N-r SECY 84-119 All NC/SEE Appendix C Secties NUREC-0927, Rev. 1 C.5 evaluated as se NUREC-0993, Rev. 1 additional action NUREG-0737 Ites I.A.2.3 required.
SRP revisions A-2/
Asymmetric Blowdown NUREG-0609.
FWK NA MPA D-10 Loads on Reactor CL 84-04, CDC-4 Primary Coolant Systems A-3 Westinghouse Steam NUREG-0844 U-FUR NA Generator Tube Integrity SECY 86-97 SECY 88-272 CL 85-02 (No requirements)
A-4 CE Steam Generator NUREG-0844, SECY 86-97 CE-FWR NA Tube Integrity SECY 88-272 I
GL 85-02 (No requirements)
A-5 B6W Steam Generator NUREG-0844, SECY 86-97 B&W-PWR NA Tube Integrity SECY 88-272 j
GL 85-02 (No requirements)
E Mark I Containment NUREG-0408 Mark I-BUR NA A-6 Short-Term Program
- C - Complete NC - No Changes Necessary NA - Not Applicable I - Incomplete E - Evaluating Actions Required WSI3:SFB10 e
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_J Enclosure to 0-601563 Fage 2 of 6 USI/MPA NUMBER TITLE REF. DOCUMENT AFFLICABILITY STAIUS/MIE*/EBIARES A-7/D-01 Mark I Inng-Tern NUREG-0661 ~
Mark I-3UR m
Program NUREG-0661 Supp. 1
~
CL 79-57 A-8 Mark II Containment NUREG-0808 Mark II-BUR NA Fool Dynamic Imads NUREG-0487, Supp. 1/2 NUREG-0802 SRP 6.2.1.1C GDC 16 A-9 Anticipated Transients NUREG-0460, Vol. 4 All C (10/85)/
Without Scram 10CFR50.62 SSER *6, Section 15.2.1 A-10/
BWR Feedwater Nozzle NUREG-0619 BUR C (11/86)/USAR Section MPA B-25 Cracking Istter from DG Eisenhut 5.3.3.1.4.5.
dated 11/13/80 NRG Istter GL 81-11 March 1, 1988 Docket #50-461.
ISI Manuel App. 4 transmitted by IFC**
1etter U-600733 dated 11/7/86.
A-11 Reactor Vessel NUREG-0744. Rev. 1 All C (7/82)/SSER 1 Sections Material Toughness 10CFR50.60/82-26 5.3.1, 2 and 3 A-12 Fracture Toughness of NUREG-0577 Rev. 1 FUR NA Steam Generator and SRP Revision 5.3.4 Reactor Coolant Pump Supports A-17 Systems Interactions Ltr: DeYoung to All NC/GL 89-18 did not licensees - 9/72 repire actions of IF.
NUREG-1174, NUREG-1229, NUREG/CR-3922, NUREC/CR-4261, NUREG/
USI3:SFB10 W
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Emeleeure to 0-601363 Fage 3 et 6 USI/NPA 18 UMBER TITLE REF. DOCLEENT AFF1.1CARI11TT IIAIRE/381E*]EageM S CK-4470, GL 89-18 (No requirements)
A-24/
Quellfication of Class NUREC-0584, Rev. 1 All-C (9/86)/558K 5, Sections MFA B-60 IE Safety-Related SRP 3.11 3.11.3.2, 3.11.5.
Equipment inCFR50.49 55 M 6, Sectione 3.11, CL 82-09 GL 84-24, 7.7.3.
GL 85-15 SSM 7, 3.10.1.
A-26/
Reactor Vessel DOR 14tters to FMR I/due 4/99 NPA B-04 Pressure Transient Licensees 8/76 IFC letter U-601317 Frotection WUREG-0224 deced 12/6/88 NURFE-0371 respeedig to CL 48-11 SRF 5.2 identifies mood for CL 88-11 Tech. Spec. change.
i A-31 Residual Neat Removal NUREG-0606 All OLs After C (2/82)/58K Section Shutdoun Requirements RC 1.113, 01/79 5.4.2, (Table C-1).
RC 1.139 A-31 resolved SRF 5.4.7 by Inc in 5/78 by isome of SRF 5ectica 5.4.7.
SSR was isoned in 1982:
1mpleaseted prier to licensing.
A-36/
Control of Heavy leads NUREG-0612 All I/(procedere changes C-10. C-15 Near Spent Fuel SRF 9.1.5 due 9/90). SSER 5 GL 81-07, GL 83-42, Section 9.1.5.
GL 85-11 Addittemel procedural Intter free DG Eisenhut controls to be dated 12/22/80 implemented prior to oest refweltag.
WS13:SFB10
Enclosure to U-601563 Fage 4 of 6 USI/MPA leUftBER TITIJ:
REE. DDCEEEENT AFFLICARILITT ITATH1/BRI1*1EERARES A-39 Determination of SRV Fool NUREC-0802 BE C (1/86)/
Dynamic Imeds and Pressure NUREcs-073.0783,0802 SER, section 6.2.1.8 Transients NUREC-0661 SSER 5, section 6.2.1 SRF 6.2.1.1.C A-40 Seismic Design Criteria SRF Revisions, NUREC/
All NC/SSEE 7, 5setier, CR-4776. NUREC/CR-0054, 3.10.1 NUREC/m-3480, NUREC/
CR-1582. NUREC/CR-1161 NUREC-1233 NUREC-4776 NUKEC/CR-3805 NUREG/CR-5347 NUREG/CR-3509 A-42/
F1pe Cracks in Boiling NUREC-0313. Rev. 1 BUR 1/ Tech. Spec. change MFA B-05 Water Reactors NUREG-0313. Rev. 2 reystred CL 81-03, CL 88-01 (dee 2/90).
SSER 1, Section 5.2.3.
A-43 Containment Emergency NUREG-0510, All WC/USI A-43 and CL 85-22 Sump Ferformence NUREC-0869. Rev. I did met contain NUK'$-0897, RC 1.82 regairements for IFC.
(Rev. 0), SRP 6.2.2 GL E5-22 (No requirements)
A-44 Station Blackout RG 1.155 All I/due 5/90, IFC letter NUREG-1032 U-601427 deced 4/16/89 NUREG-1109 provided the results of 10CFR50.63 the IPC review on Sao.
All actions scheduled to be completed by this dote are complete. Additional procedure changee will be made by 5/90.
-USI3:SFB10
Emeleeure to C-601M3 Fage 5 et 6 USI/hPA nut 1BER TITLE REF. DOCMEMY AFFLICARILITY STAIDE/ BRIE
- m A-45 Shutdown Decay Meat SECY 88-260 All 1/due 9/92.
Removal Requirements NUREC-1289 IFC letter 5-601549, deced l
NUREC/CR-5230 10/27/89 (A-45 mow SECT 84-260 imeluded in Individual (No requirements)
Flant M== tion CL 88-20 Fregram).
I A-46 Seismic Qualification NUREC-1030 All NC/558K 7, Section 3.10.1 of Equipment in NUREG-1211/
Operating Flants CL 87-02, CL 87-03 A-47 Safety Implication of NUREC-1217 All E/(due 3/2/90)
Control Systems NUKEC-1218 CL 89-19 A-48 Hydrogen Control Neasures 10CHt50.44 All, except FURS I/dme 12/90 for Smergency and Effects of Hydrogen SECT 89-122 with large dry Operating Frecedures (50F)
Burns on Safety Equipment containments cosylettom. All equiposet to installed. IF will submit documentation h erating ceaptience with1m l
6 mes. of issuance l
ef the General Hydregon Goetrel Program SSR.
IPC is developing Nydrogen Centrol 30F. Nydrogen mitigatlee procedures are in place.
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Inclosure to 5-691363 Page 6 of 6 USI/MPA NUMBER HII&
REr_ DOCIEIBff AFFUCARII.1TY IIA 2WUBRIE*1ERIARES A-49 Fressurized Thermal RCs 1.154, 1.99 fur IEA Shock SECT 82-465 SECY 83-288 SECY 81-687 10CRt30.61/G: 88-11 SER/SSER - Safety Evaluation Report for Clinton Power Station - IUUREC-0853
- USI3:SFB10
PLANT Citnton DOCKETN0(5).
!0 461 PROJECT MANACER John B. Hickman TECHNICAL CONTACT A. Serkir U$1 NO. A-1 TITLE Water Hamer MPA NO. W/A TAC N05.
!$$UE$ SUW.ARYt ThisUnresolvedSafetyIssue(US1)watresolvedinMarch1984 with the publication of NUREG 0927,
- Evaluation of Water Hamer in Nuclear Fower Plants
- Technical findings Relevar.t to Unresolved Safety Issue A-1."
Also on March 15, 1984, the EDO sent the Comissioners SECY 84-119 titled, Resolution of
'Jnresolved Safety Issue A 1, Water Hamer."
In SECY 84-119, the staff concluded that the frequency and severity of water hamer occurrerces 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)properdesignoffeed-water valves and control systemst and (c) increased operator awareness and trainirg. Therefore the resolution of U$1 A 1 did not involve any hardware or designchangesonexIstingplants.
ItdidinvolveStandardReviewPlan(SRP) changes (forwardfits)andacon.prehensivesetofguidelinesandcriteriato evaluate and upgrade utility training programs (per TM1 Task Action Plan item 1.A.2.3).
In addition, the assumption was w,ade that for PWRs 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 10 failure by water hamer that opened a direct pathway to the environment.
IMPLEMENTATIONANDSTATUS
SUMMARY
(PLANTSPECIFIC):
No changes necessary. SER Appendix C Section C.5 evaluated as no additional action required for plant operation, r
l l
i
REFERENCES:
Clinton A-1 1.
RE001REMENT DOCUMENT $t TITLE NUDOCS NO.
DATE Letter from Denton to Utilities.
84031503'40 03/05/84
'hotice of issuance and Availtbility NUREG 0927 Rev. 1 Safety issue A l' 2.
IMPLEPENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE NUREG 0927 ' Evaluation of Water 83C6C60413 05/31/83 Hammer in Nuclear Fower Plants.
Technical Findings Relevant to Unresolved Safety issue A-l' NUREG-0993 Rev. 1 8306C60418 March 1984
'Re9ulatory Analysis for for U51 A-1, Water Hammer" SRP Sectionst 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 Hamer" SER related to the operation NUREG 0853 February 1982 of Clinton Power Station (C.5) 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS MO.
DATE
1 0-I j
PLANT Citaton.......~...........
DOCKETN0(S).
$d6},;;,,;;,;;,.,;.,,,
j J )3,kyn........
TECHNICAL CONTACT Jgi,Rg h n :
PP.0 JECT MANAGER g
TITLE Asynenetric Blowdowgdg. f a.RCS.............-
i USl NO. Ad.
l' MPA NO. D 10.....
TAC N05.
3,gne...............m............-........
li&UES.5VMMARY:
This USl was resolved in January 1981 with the publication of NUREG-0609, I
"Asymetric Blowdown Loads on PWR Primary Systems.'
i in October 1975, the NRC notified each operating PWR licensee of a potential safety problem concerning the fact that asynenetric LOCA loads had not been considered in the design of any PWR piping system.
In June 1976 the NRC i
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 January 1978 with a request for additional information to all PWR licensees.
NUREG 0609 provided guidance for these analyses, for operating PWRs, Multi Plant Action (MPA) Item 0-10 was established by NRC's Division of Licensing for implementation purposes, i
r i
During the course of the work on US! 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 new technical work, it was demon-l strated that pipes would leak before break and that new fracture mechanics techniques for the analyzing of pipin failures assured adequate protection against failures in primary system pi ing in PWRs (Generic Letter 84-04). This v
was reflected in a revision of Genera DesignCriteria(GDC)-4(AppendixAto 10 CFR Part 50) published in the Fgdert) Rtgister in final form on April 11, 1986, and in a subsequent revision fo*G5C-4 puti11shed in the FLderal Roginter i
on July 23,1986, in addition, it has also been satisfactortTy ceTn6nidra33 in the course of the A-2 effort that there is a very low likelihood of simultaneous j
pipe loading with both LOCA and safety shutdown earthouake (SSE) loads, i
Therefore, the last revision of GDC 4 represented the final technical action of 4
NRC regarding the issue of asynenetric blowdown loads issue in PWRs primary 1
coolant main loop piping.
IWPLEMEWT ATIOW. AWD.ST&TUS.SUWARY. ( PL AWT-SPE CIFIC):
N/A, PWR only I
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4
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g(Qt[Nll:
Clinton I
A-2 1.
- amommn,
)
IHLI WP9915.50a MIL 1
Generic letter devaluation of i
Primary Systems for Asynenetric LOCA Loads" 01/20/78 Task Action Plan A-2, Asynenetric Blowdown Loads on Reactor Primary Coolant System ' NUREG-0371 Test
(
Action Plans for Generic Activities 11/78 l
- Asynnetric Blowdown Loads on PWR
\\
Primary Systems," NUREG-0609 US NRC NP.R 01/81 l
GDC-4, " Environmental and Dynamic
{
Effects Design Basis
- j GL 84-04,
- Safety Evaluation of I
Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main loops."
2.
IMPLEV.EWTATION DOCUWEWTS:
t IIILL NV9.0CSW9:
MI!
N/A, PWR only i
3.
VERIEICAT10W.000UMEWIS:
nts uUo0Cs.u0.
Eng.
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i 4
L PLANT Citsten.~.... -. _...........
DOCKETN0($).60461.---..............
PROJECT MANAGER J. Mickasa.....-...
TECHNICAL CONTACT I. Surphy............
U$1 NO. A.3..A.4..and.A.5 TITLE ltaan. Generator. Tube lateerity --... -.......
MFA NO..-~.-...---....
T AC W OS. W o me................... -. -............
15$00$.$UMMARY:
U$1s A 3 4 and 5, were resolved in Septed er 1988 with the publication of NUREG08d4,NRCIntegratedProgramfortheResolutionofUnresolvedSafety issues A-3, A 4, and A 5 Regarding Steam Generator Tube Integrity.' U$1s A-3, A-4, and A 5 did not result in new generic requirements for Industry in view of the small potential for reducing risk.
Steam generator tube integrity was designated an unresolved safety issue in 1978 after it became apparent that steam generator tubes were sub.iect to widespread segradation, frequent leaks, and occasional ruptures (i.e., gross failures). U$1 Task Action Plans A-3, A 4, and A 5 were established to evaluate the safety significance of these problems for Westinghouse, Coe ustion Engineering, and Babcock & Wilcox steam generators, respectively. These studies were later combined into a single effort because PWR vendors were all experiencing many of the same problems.
NUREG-0844 provides a generic risk assessment that indicates that risk from steam generator tube rupture ($GTR) 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.
KUREG-0844 also identifies a number of staff-recomended actions that can further improve the effectiveness of licensee programs in ensuring the integrity of steam generator tubes and in mitigating the consequences of a $GTR. 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-recomended actions; however, such recomended actions do not constitute HRC requirements. The staff's assessment of licensee responses to Generic Letter 85-02 was provided to the Comission in SECY 86-97.
g.PjEMENHTJ0W. AWD STATUS.S@jKARY.(PLANT SPEC},FJ}:
N/A, PWR only
j O
8
-R[FIPENES:
Clinton
~~
A 3, 4, 5 1
1.
Rjg1RJggi. DOCUMENT $i Tj,Tg NyDogjh QATE I
J NUREG 0844, 'NRC Integrated Septod er 1988 i
Program for the Resolution of Unresolved Safety issues A 3, A 4, and A 5 Regarding Steam Generator Tube Integrity" I
Generic Letter 85-02 04/17/85 SECY-86 97, Steam Generator U51 Program - Utility Responses to Staff Recomendations in Generic Letter 85-02 03/04/86 1
2.
IMPQKENTATION. DOCUMENTS:
lilg NUD0@,,%
0AT,E,
)
N/A, PWR only l
l 1
l
)
l 3,
1EJJ,Fj$ATION. DOCUMENT),.
i TITLE WyDOC,}.WO.
- DATJ, l
l 1
l l
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FLANT G11ston....-....-........._
DOCKET N0(S), 50 461... -... - mau PROJECT MANAGER k Wjikmes........_. TECHNICAL CONTACT k h{rfgk........
U51 NO. Ru TITLE Mark.1.Costainm_ent.Short. Term Prggrger.. ;..,,;,. u;u MFA NO wu TAC N05. gp_nh. -......... u.
...n.....-..
. m = ;;., =
lilVES $YNNARYt This 051 was resolved in December 1977 with the public6 tion of NUREG 0400,
- Mark 1 Containment Short-Term Program Safety Evaluation Report.'
)
The objectives of the Mark I short-tere program weret (a)toexaminethe containment systes of each BWR f acility with a Mark I contairment design to J
verify that it would maintain its integrity and functional capability when subjected to the most probable hydrodynamic loads induced by a postulated j
design-basis LOCA, and (b) to verify that licensed Mark I BWR facilities'could cont 1nue to operate safely, without undue risk to the public health and safety i
until such time as a methodical, comprehensive long tem program is conducted.
I The NRC staff used a safety factor of at least two to failure for the weakest structural or mechanical component in the Mark I containment system in judging
)
that containment integrity and functions would be assured under most probable I
design basis LOCA-induced hydrodynamic loads.
As indicated in NUREG-0408, the staff required full implementation of the l
calculation of the hydrodynamic loads and structural analysis as an interim measure until complete implementation of the long term program had been I
achieved.
In NUREG-0408 the staff concluded that the objectives of the Short-Term Program had been satisfied, thus documenting the basis for resolving this safety issue. This issue is considered complete for all affected BWRs.
[
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l 1MPLEWENTATI E A g $TATUS.SutmARY.(PLANT SPECIF1,C,):
l
- N/A, Fark I only I
i I
i t
i
t i
$[l[&Elill:
Clinton A6 1.
REQVIREMEWT.DOCUELWTS:
l 111k[
EUDOCS NO, BT[
j i
- Mark 1 Containment 10/77 Short Term Program Safety l
Evaluation Report * (See Table 12 i
for letters to BWR licensees requestingaction) j i
2.
1MP([MENTAT10W00CVMENTS:
TITLE
[yp.pij;WO, DAI[
N/A, Mark 1 only t
I t
3.
((,RJ{}C1.710W.00CUMEWT5 TITLE N90pCS.NO.
DAT[
l I
i f
h I
i t
a
i PLANT Qi njen............... -.
DOCKETN0($). 50 461...;; u : m,,
PROJECT MANAGER h,dligtmas........;
TECHNICAL C0K1ACT hfydritk t
U$1 NO. A 7.. -.
TITLE M.),r,t;l;L oc c. T e rs: P ree rgm,; u,n;u : 1..
7, MPA MO.
TAC WOS. ha. u;, _..... : a;;;;;n.,un : --..... : _ :.
l
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1$$UES.$Ul94ARY t This U$1 was resolved in August 1982 with the publication of Supplement 1 to NUREG-0061, Safety Evaluation Report, Mark 1 Containment Long-Term Program
- and Standard Review Plan Section 6.2.1.1.C.
For operating BWRs, MPA D-01 was established for implementation purposes.
The focus of this U$1 was the suppression pool hydrodynamic loads, associated with a postulated LOCA, which had not explicitly been included in the original Mark I containment design. The issue was identified during large scale testing of a Mark 111 containnent design. The staff addressed this issue in NUREG 0661, published in July 1980, and in $Upplement 1 to NUREG 0661, published in August 1982.
The objective of the long term program (LTP) was to establish the design basis loads that are appropriate for the anticipated Itfe of each Mark i BWR facility and to restore the originally intended design safety margins for each Mark I containment system. The principal thrust o# the LTP was the development of l
generic methods for defining suppression pool hydrodynamic loadings and the associated structural assessment techniques for the Mark I configuration. On the basis of experimental and analytical programs conducted by the Mark 1 I
Owners Group, it was determined that the hydrodynamic load definition pro-j cedures, with some modifications defined in NUREG 0661, provided a conservative estimate of these loading conditions. Thus, the requirements associated with i
this U$1 were concerned with the structural assessment of Mark I containments and related structures to the hydrodynamic loads defined by the staff in the LTp.
l In January 1981, the staff issued ' Orders For Modification of License and Grant I
of Extension of Exemptions" to each licensee of a Mark I plant. The orders i
1 required the licensees to assess the suppression pool hydrodynamic loads in accordance with General Electric docusents and NUREG 0661 on a defined i
I schedule. For some plants, the implementation schedule was extended by a subsequent order.
IMPLElsEWT&T10W. AWD. STATUS.$UMGRY.(PL AWL $pECI FIC h l
4 ft/A, Mark I only i
i
o M il:
Clinton A-7 1.
SE003REMENT-DOCWREWTS:
I lilLE guppeg_vgt gAtt
- $afety Evaluation NUREG-0661,k 1 Containsent Long Report, Mar Term Program" 07/80 NUREG-0661, Supplement 1 08/82 Orders for Modification to License for Applicable Licensees 1981 P.
INPLIP[WTATIOW. DOCUMENTS:
11L[
- UDOCSW92
- gA3, N/A, Mark I only 3.
VERITICAT10W-00CUMEUTS:
TITLE NUDOCS WO.
GATE l
l I
i
[
PLANT Clieten..................--.,
DOCKET N0(S). 50 461............. ~,
PROJECT MANAGER J...Hickmas...-.... - TECHNICAL CONTACT h.hdr4ck.. -....
USI NO. A.S. --
TITL E Ras k.11. Costa t enes t-tool Dynami c-Leade.... -.-.......
MFA NO.. - -.
T AC N0 5. None............-................ - - --...... -. - --.
1$500s.SumARY:
j 6.
USl NO. AJ TITLE: Mark.11.Coptstasest Pool.Dysamic-Leeds l
\\
This USI was resolved in August 1961 with the publication of NUREG-0808,
- Mark
[
11 Containment Program Lead Evaluation and Acceptance Criteria,* and Standard Review Plan (SRP) Section 6.2.1.10. The requironent is that the 11 BWRs having j
the Mark 11 containment shall seet the requirements of GDC 16.
As stated in NUREG-0808, the original design of the Mark 11 containment system l
considered only those loads normally associated with design basis accidents j
that were known at the tire. These included pressure and temperature loads associated with a LOCA, seismic loads, dead loads..iet impingement loads, hydrostatic loads due to water in the suppression chamber, overload pressure I
test loads, and construction loads. However, since the establishment of the i
original design criteria, additional loading conditions were identified that
}
must be considered for the pressure-suppression containment-system design.
l t
in the course of performing large scale testing of an advanced design pressure-suppression containment (Mark 111), and during implant testing of Mark I containments, new suppression pool hydrodynamic loads were identified that had not been included explicitly in the original Mark 11 containment design basis.
These additional loads result from dynamic effects of drywell air and steam t
i being rapidly forced into the suppression pool during a postulated LOCA and t
from suppression-pool response to safety / relief valve (SRV) operations these t
are generally associated with plant transient operating conditions. Because i
these new hydrodynamic loads had not been considered, the NRC staf f determined that a detailed reevaluation of the Mark 11 containment system was required.
l The issuance of NUREG 0B08, NUREG-0B02, " Safety Relief Valve Quencher Loads:
l i
l Evaluation for BWR Mark II and III Containments," and NUREG-0487
' Mark 11 i
Containment Lead Plant Program Load Evaluation and Acceptance Crlteria,"
docurented acceptable methods for calculating the hydrodymanic loads associated with plant transient conditions. Specifically, the loads referenced in these i
NRC staff reports, as modified by the acceptance criterl., constituted the resolution of USI A-8.
SRP Section 6.2.1 has been modified to reflect the applicability of these reports to Mark 11 containment evaluations.
l l
Implementation 15 believed to be complete for all Mark 11 BWRS.
As part of the l
licensing process, the staff required that the applicants utilire the new I
calculation methodology defined in the reference docurents before a full power t
license was issued.
EEMENT AT ION. AND.ST ATUS
SUMMARY
. ( PL ANT. Sg ! F I C ) :
tl/A, Mark 11 only
~------U
i 1
l c
RJi[RE([}:
Clinton AB r
1.
M001REMENT.00CggT}:
l 1
IIIki W991LE EAIE
{
GDC 16, Containment Design t....
NUREG-DBOB ' Mark 11 Containment Program Load Evaluation and Acceptance Criteria" August 1981 l
I Standard Review Plant 6.2.1.1.c.
" Pressure Suppression Type BWR Containments
- Revision 1-4 NUREG-0487, " Mark 11 Containment November 1978 i
lead Plant Program Load Evaluation j
and Acceptance Criteria" i
n.
Supplement 1 September 1980 i
b.
Supplement 2 fetiruary 1901 j
NURG-0802 " Safety Relief Yalve October 1982 Quencher Loads: Evalu6 tion for BWR Mark 11 and 111 Containments' i
2.
1MPL EMEWTAT10W.00CLM{NTjt l
TlTJ,[,
NUDQC$.NO.
OATE N/A, Marl 11 only i
i 3.
y,ER1rlCAT10W.00CUMEWTS TITLE WUD0CS.NO.
9,AT1 l
l
o l
Pt. ANT Clinton DOCKET 110($).
50 461 i'
TR00ECT MANAGER J. Hickman T[CHNICAt. CONTACT J. Knuck US! NO. A9 TITLE ATVS per 10 CrR $0.62 MPA 140.
TAC NOS. None 1$$UES $UMMARY:
This U$1 was resolved in June 1984 with the publication of a final rule (10 CFR 50.62)toreovireimprovementsinplantstoreducethelikelihoodoffailureof the reactor protection system (RPS) to shut down the reactor following snticipated transients and to mitigate the consequences of an anticipated trantientwithoutscram(ATWS) event.
The rule includes the following design-related requirements:
50(2(C)(1),
diverseandindependentauxiliaryfeedwaterinitiationandturbinetrip(C)(3) for all PWRst 50.6P(C)(2), diverse scram systens for CE and B6W reactorst 50.6?
titernate red injectier. (ARI) for tWRst 50.62(C)(4); standby liquid control system (SLCS) for BWRst and 50.62(C)($), automatic trip of recirculation pumps under conditions indicative of an ATWS for BWRs.
Information requirements and an implementation schedule are also specified.
IMPLEPENTATION AND STATUS
SUMMARY
(PLAllT SPECIFIC):
No changes necessary SER suppl. 7 accepted the licenste's SLCS.
SER suppl. 6 accepted the licensee's ATWS recirculation pump trip design.
The licenste's ARI was accepted by letter 5/18/87.
e I
f ggl[ ggt Clinton l
A9 1.
RiaWitDEWT.DOClistuts:
f l
E M
- US l
i MUREG-0460, and supplements.
03/80 i
' Anticipated Transients Without j
$ cram for Light Water Reactors' l
Federal Register Notice f
49FR20045(10CFR$0.62) 06/26/84 i
l I
l h
2.
IMPLEW.[WT ATICW 40 CUM [WT$
f l}T,M, EUDOC$ WO.
DATE Clinton $ER sup, 6 NUREG-0853 7/86 f
I Clinton SER sup, 7 9/86 l
$ER on ARI B705200148 5/18/87 l
l i
i 3.
.gfyFEA3gW.00tuMEWTS:
l Ing wuo0cs.g pan
~
t l
l-I l
(
i l
-o i
I PtANT Clinten DOCKETN0($).
00 41 ikDJECT MANAGO!
J. Hickman -
TECHNICAL CONTACT K. Wichman -
USI NO. A-10 TITLE BWR reedwater Nor:1e tracking,,_,,,, _ u j
t MPA NO. B 25 TAC NOS. None -
l 6
ISSU[$ $UMMARY:
s This issue was resolnd in November 1980 with the publication of NUREG 0619.
- BWR ftidwater Nor:1e and Control Rod Drive Returr Line Norrie Cracking.' MFA
[
B 25 was establisted by NRC's Division of Licensing for implementation purposes.
Inspections of operating EWRs conducted u
[
feedwater rcarles of 20 rt+ctor vessels. p to April 1970 revealed cracks in the It was determined that cracting was due to high cycle fatigue caused by fluctuations in water teriperature within the itsst1 in the r.ortle regior, j
Py letter dated Noventer 13,etter stated that NUREG 0019 provided the resolu-1900, a copy of NUREG 0619. The s
[
tien of the staff's generic technical activity US! A 10, which resulted from i
the ir.scryict discovery of cracting in feedwater rortles and control rod drive rcturn lir.e r. caries. NUREG-0619 cescribes the technical issues, General Electric ard staff studies and analyses, and the staff's positions and require-merts. Licerisees were required to respond, pursuant to 10 CFR 50.54(f), that I
they wculd meet in.plettntation dates it dicated in NUREG 0619.
[
Cereric letter 81 11 was subsequently issued to provide technical clarification f
to the Nonrber 13, 1900 letter, to clarify that it had been sent to FWR littrsees for inferrc.ation only, and that no response was required from PWR lictnstes.
[
IPPLEPENTATICh_ AND STATU$ $lEMARY.iplANT SPECIFIC):
f Iraplemented pre-licensing.
{
i t
i t
.2
REFERENCI):
Clinton i
A-10
[
1.
jgU)}{,MJNT DOCLMENT$t TITli NUDOCS N_0_.
GAT [
Letter from D. Eisenhut f
tran5r.itting NUREG 0019,
'DWR reedwater Nortle and Control F.ot Drive Return Line hor:1e Cracking,'
i resolutien of A 10 to litenstes 11/13/60 l
t Generic Letter 81 11, 'BWR reedwater Ner:1e and Control Pot trive Return Line Nctrie Cracking (NUREG-0019)'
02/20/81 2.
Il!.PLEhlHTAT!j,N DOCUMENTS:
TITLE NUDOLS NO.
DATE
$6fety Evaluatier. Rtrert NUREG-0853 2/82 related to the operation of i
Clinton Power Stetter i
(!rction3.9.3.1) f i
r I
3.
YEklF1 CATION DOCUMENTS:
t TITLE NUDOCS_NO.
DATE I
I I
f i
PLANT Clinten DOCKETo0($).
50 461 PROJECT RWAGER Jchn B. Hie ben TECHNICAL CONTACT B. Elliott U$1 NO. A 11 TITLE Reactor Yessel Materials Toughness MPA NO.
TAC NOS. None
{
f
!$5DE$ StMMARY:
i This U$1 was resolved in October 1982 with the publication of NVREG 0744,
Pressure Yessel Material Fracture Toughness .
NUREG 0744 was issued by Generic Letter 82 26 and provided only a rethodology to satisfy the require-I ments of 10 CFR Part 50, Appendix G.
No licensee response to Generic Letter 82 26 was required.
l Fecause of the remote possibility that nuclear reactor pressure vessels desi ntd to the ASME Boiler and Pressure Yessel Code would fail, the design of i
nuclear facilities does not provide protection against reacter vessel failure.
l Frevention of reactor vessel failure depends primarily on maintaining the j
reactor vessel material fracture toughness at levels that will resist brittle i
fracture durir,g plant operation. At service times and operating conditions typical of current operating plants, reacter vessel fracture toughness prettrties provide adequate margins of safety against vessel feilure; however, as plants accumulate more and more service tire, neutron irradiation reduces the material fracture toughness and initial safety margins.
l i
Aspendix G to 10 CFR part 50 requires that the Charpy upper shelf energy i
tiroughcut the life of the vessel be no less than 50 ft-lb unless it is de:rerstrated that lower values will provide margins of safety against failure i
equivalent to these provided by Appendix G of the A$ME code. U$1 A 11 was l
initiated to address the staff's concern that some vessels were projected to have beltline materials with Charpy upper shelf eterpy less than 50 f t-lb.
j NUREG 0744 provides a method for evaluating reactor vessel materials when their Charpy upper shelf energy is predicted to f all below 50 ft-1b.
Plants will use 1
the prtscribed nethod when analysis of irradiation damage predicts that the charpy upper shelf energy is below 50 f t-lb, j
IITLEMENTATICH AND STATUS SUMKARY (PLANT $PECIFIC):
i Implemented prior to licensing.
I 1
k[f tk[PiC[$t Clinton A 11 1.
RIOUlttHENT DOCUMENTS:
TITLE NUDOCS No.
DATE i
l NUREG 0744 Revision 1, 'Fressure 10/82 Yessel Meterial Fracture Toughness
- Generic Letter 82 26, ' Pressure Yessel Material Fracture Toughness" 11/12/62 1
2.
IMPlfP[t;TATION DOCUMENTS:
f TITLE NUDOCS NO.
DATE Safety Evaluation Peport NUREG-0653 7/82 l
related to the operation of CPS, Supplement 1 l
(Section5.3.1.1) i i
3.
VERIFICATION DOCUMENTS:
[
TITLE NUDOCa NO.
DATE l
i t
f
PLANT Clinton.
DOCK [TN0(5). 50_461 PROJLC1 MANAGER 0. Hickren TECHNICAL CONTACT R. Ot+nsen (RES) 051 NO. A-12 TITLE Potential of Low Fracture Toughness and Lane 11er Tearing in pWR $G and.RCP. Support $ _.
f MPA NO.
TAC NOS. Lone.
1$$UES $tW.ARY:
I This USI was resolved in October 1983 with the publication of NUREG 0577,
" Potential of Low fracture Toughness and Lamellar Tearing in PWR Steam j
Generator and Reactor Coolant Pump Supports.' The resolution contained no backfit requirenentst it only applied to plants with a new construction permit i
issued after October 1983. Standard Review Plan Section 5.3.4 was issued at l
the sane ties this US! was resolved.
[
t The concern in this U$1, as the title indicates, was the potenti61 of Icw fracture tougtness of scoe r.attrials selected for fabrication of stear.
generator (SG) and reactor coolant pun:p (RCP) supports in operating PWRs.
Larellar tearing has also of concern. Tracture toughness is a measure of a rateritl's resistanct to fracture in the presence of a previously existing crack. Generally, a r.attrial is considtred to have adequate fracture toughness if it can withstand loading to its design limit in the presence of detectable r
flews under stated conditicns of stress and teciperature.
l The m.odificatiers to address this US1 could involve n.aintainin n.inimum tit.ptrature around the supports abcst its fracture tr6nsition erperature, or r
tetel replactment of existirg $G and RCP supports with supports febricated of r.attriel grade which has a hi her Charpy upper shelf energy and a lower 1
transition tenperature. Anal sis perforred for the resolution of this U$1 dettre.ined that, even with th failure of the SG and RCP supports, the amount l
cf incremental reltast of radioactivity wculd not be sufficiently high enough tt,'ustify any r.odification in ternis of incrtasing the toughness of these supports. This conclusion is based on a value-irpact analysis docunented in l
/ttperdix C of NUREG-0577.
l Il.plERENTATI(>N AhD 51 ATUS.$UW.ARQLANT SPECIFldt f
N/A, PWR only
o REftktNCES:
Citnton A 12 1.
[j$p)Ifpf g DOClWENT$:
TITlf WLIDOCS WO.
DATE NUREG 0577. Rev. 1
'rotential 10/83 of Low Fracture Toughness and L6ne11ar Tearing in PL'R Steam Generator erid Feactor Coolant Putp $vrports' 2.
IP.T'l EWINT AT10k. DOC UMENTS:
TITLE NUDOCs.WO.
DATE N/A, PWR only 3.
VERIFICATION DOClMENTS:
TITLE NilDOCS WO.
DATE l
i 1
l l
i c
ILANT Clinton DOCKETN0(S).
!0 4f1 l
PROJECT MANAGER J. Hickman TECHNICAL CONTACT D. Thatcher U$l NO. A.17 TITLE systems Interactions in Nuclear Power Plants MPA N0.
TAC NOS. None ISSUE $ SlMMARY:
Generic Letter (GL) 8918, dated Se stenber 6,1989 was sent to all power reacterlicenseesandconstitutestseresolutionofUS!A17. The generic letter did not require any licensee actions.
GL8918hadtwoenclosureswhich(a)outlinedthebasesfortheresolutionof USl 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 t9e Individual Plant Examination (1PE)forSevereAccidentYulnerabilities.
Specifically the staff expectedthatinsightsconcerningwaterintrusionandfloodingfromInternal sources, as described in the appndix to NUREG.1174, would be considered in the IPE program. Also considered in the resolution of this US1 was the expectation that licensees would continue to review information on events at operating nuclear power plants in accordance with the requirements of-Ttil Task Action Planitem1.C.5(NUREG.0737).
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
No action required.
1972 letter on this subject was not received by this licensee since it pre. dated the Construction Permit.
y 1
REFERENCES:
Clinton A 17 1.
!gul}4$NTDOCLM[NT$:
TITLI NUD005 NO.
M Ger.eric Letter 89-18 09/06/89 NUREG 1174 ' Evaluation of May 1969 Systees Interactions in Nuclear rower Plants
- AUREG-1229 *Re9ulatory Analysis August 1989 for Resolution of U$1 A-17*
liVREG/CR-3922
- Survey and January 1985 i
Evaluation of System Interacti(n Events and Ecurces*
liVREG/CR-4261 'Assessa+nt of June 1986 Systen: Interactien Experience in Nucletr rower riants' i
i:UREG/CR-4470 *$urvey and August 1986 Evaluation of Vital Instrunentation and Control Power Surply Ever.ts" t:RC Letters to Licersees 9/72 Inforn.ir.g Lictnstes of Staff Concerns regarding rottntit.1 reilure of !!on-Categery I l
Equipn. tnt t
2.
IMPLEl'EllTATION DOCUMENTS:
i TITLE NUDOCS NO.
DATE
[
i
~
3.
VERIFICATION 00CtHENTS:
r TITLE NUDOC NO.
DATE r
i
o PLAT:T Clinton DOCKET N0($). LO,461 PROJECT MANAGER J, Hi e b.a n TECHNICAL.C0hTACT h $heranski,;;
US! N0. A-24 TITLE (unlificationofClass.1EE,qu,ip, ment 7
MFA NO.
TAC NOS. None
- 7...;;
iss0Es $UMMARY:
This 051 was resolved in July 1981 with the publication of NUREG 0588, lated Revision 1
" Interim Staff Position on Environrental Qualification of Safety Re Electrical Equipment.' Part I of the report is the criginal NUREG 0588 that was issued for co'.sents that report, in conjunction with the Division of 0 prating Reactor (DOR) Guidelines, was endorsed by a Corrission Memorandum and Ordtr as the interim position on this subject until
- final
- positions were esteblisted in rule mahing. On January 21,1983 the Comissien attended 10 CFR E0.49 (the rule), effective february 22, 1963, to codify existing qv611fic6tien s.ethods in national standards, regulatory guides, and certain NFC publicatiotis, ircluding NUREG 05BB.
[
The ru1t is based on the 00R 6cidelines and NUREG 0588. These provide on (a) how to establish environrentel service concitions, (b) how to sekvidence
[
ect ri.ethods which are considered approp(c) ate for qualifying the equipe,ent in ri different areas of the plant, and such other areas as margin, aging, ard dceurentation, h0 REG-0588 does not acdress all trees of qualifications it does suppleinent, in selected areas, the provisions of the 1971 and 1974 versions of l
1EEE Star.dard 323. The rule recognin s previous qualific6 tion efforts cce.pleted as a result of Cor.enission Menorandum and Order CL180 21 ard also reflects different versions IEEE 323, dependent on the date of the construction i
pert.itSefetyEvaluationReport(SER). Therefore, plant specific requirerents stay very in accordance with the rule.
In sum ary, the resolution of A-24 is er, bodied in 10 CFR 50.49. A reasure of whether each licenstt has implerented the resolution of A 24 may therefore be fcurd in tht determin6 tion of compliance with 10 CFR 50.49. This was addressed by 72 SERs for operating plants. issued shcrtly after publication of the rule and subsequently in crerating license reviews pursuant to Standard Review plan l
Secticn 3.11. This was further addressed by the first-round environmental l
qualification inspections conducted by the NRC.
MLEFENTATION,AND, STATUS $UmARY_(plANTSPECIFlC):
i Implemer,ted prior to licensing.
l i
+
l 1
=
a O
r 3GIRENCES:
Citeton A.24 t
1.
LILUIRElit,NT,DOClMENTS:
l t
E NltDOCS NO.
E D0R'GuidelinesforEvaluating Environmental Qualificati0ti e' Class IE Electrical Equipent in Operatirig reactors
- NUREG 0!SB, *Interisi Staff Position on Envirorsental Qualification of Saftty F. elated Electrical Equipent" 12/79 Cowission Fen 4orandoni tr>d Order, l
CLI.[0 21, on D0R Cuidelines and NUREG-0088 05/23/60 h0 REG.0!B8, revision 1 07/81 l
10 CFR E0.49 (40 FR 2730 2733) 01/21/83 Sithdard and Review Plan 3.11 I
Er.virenrantal Qualification of Mechanical ar.d Electrical Equipment 07/81 l
2.
IMPLEVELTATION DOClNINTS:
f TITEE NUDOCS.N0 DATE
[
2 Safety Esaluaticr. Report NUREG-0853 2/82 relat(c to the operation of and Clinton Power Statien Supplenent 2,5 & 6 5/83,1/86,7/66 (Section3.11)
Letters fron. IP C002190200 2/14/66 4/04/o6 Clinton F$AR 3.
YFRTFICATION. DOCUMENTS:
f TITLE NUDOCS NO.
DATE i
l i
r i
j i
f FLANT Clinton DOCKET N0(S).
50-461 PROJECT MANAGER J. Hickman TECHNICAL CONTACT Chu Liang t
J.i USI NO. A-26 TITLE Reactor Vessel Pressure Transient Protection l
MPA NO.
TAC N05. 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 R* view Plan Section 5.2.
The licensees of all operating PWRs were requestcd.o provide an overpressure prevention system that could be used whenever ine plants were in startup or 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 1
exceeded.
The majority of these events occurred while the reactors were in a soliti-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-
' twas than at normal operating temperatures.
In light of the frequency of the h
reported transients and the associated potential for vessel damage, the NRC l
staff concluded that measures should be taken to minimize the number of future l
tre.nsients and reduce their severity.
L Generic Letter 88-11, *NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its impact on Plant Operations " was published July 12, Lp 1988. ThislgenericletterprovidesguidanceregardIngreviewofpressure-temperature limits and indicates that licensees may have to revise'10w-temperature-overpressure protection setpoints.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
- N/A - PWR Only Licensee evaluated and responded to GL 88-11. Response included conmitnent to submit appropriate TS change request.
NRC confirmed adequate response.
a
o j.
REFERENCES:
Clinton A-26 t
'1.
REQUIREMENT DOCUMENTS:
TITLE.
NUDOCS NO.
DATE NUREG-0224
" Reactor Vessel Pressure Transient Protection P
for PWRs."
9/78 l
NRC Letters to Licensees Informing Licensees of Staff Concerns Regarding Overpressure Low-Temperature Conditions in PWRs August 1976 Generic Letter 88-11 "NRC 7/12/88 Position on Radiation Embrittlenant of Reactor Vessel Materials and Its impact on Plant Operations" l
Stendard Review Plan Section 5.2 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Clinton Power Station 0812140178 Response to Generic Letter 88-11 12/06/88 Response to NRC GL 88-11 5/08/89 3.
VERiflCATION DOCUMENTS:
TITLE NUDOCS NO.
DATE 1
o o
j PLANT Clinton DOCKETN0(S).
!0-4fl.
0 PROJECT MANAGER J. Hickman.
TECHNICAL CONTACT R. Jones..
US1 NO. A-31 TITLE RHR Shutdown.Re,girements j
MPA NO.
TAC NOS. None....
i ISSU_ES.SUbiMARY:
This USI was M 601ved in May 1978 with the publication of Standard Review Plan (SRP) Sect h 5 4 7.
Only those plants expected to receive an operating license.stier January 1,1979 were affected by this resolution. The USI involvhd evteHishnent of crMoria for the design and operation of systems necessary it, take a power reacttir from norn.a1 operating conditions to cold shutdown.
SRP Section S.4.7 stated that, for purposes of implementation plants would be divided into three classes: ClassIwouldrequirefullcomp1Iancefor Construction Perr.it (CP) or preliminary Design Approval (PDA) applications which were docketed on or after January 1, 1978. Class 2 required a partial implementation for all plants for which CP or PDA applications were docketed be: fore January 1,1978, and for which an Operating License (OL) issuance was exptcted on or after January 1, 1979. Class 3 affected all operating reactors arc' all other plants for which issuance of the OL was expected before January 1, 1979. The extent to which Class 3 plants would require implementation was besed on the combined staff review of related plant features.
In general, the outcone of these evaluations were that only plants receiving an OL after January 1, 1979 were effected by this USI resolution, and there were no backfits to crerating plants that had received an cperating license before January 1,1979.
IMPLEVENTAT10htAND. STATUS
SUMMARY
J g NT SPECIFIC):
Implen+nted prior to licensing.
p.
o o
+
i-
REFERENCES:
Clinton
~~*-
A.31 1.
REptilkEllENT DOCUMENTS:
TITLE NUDOCS NO.
DATE NUREG 0800 " Standard Review Plan,"
5/78 SRP Section 5.4.7 NUREG-0606 " Unresolved Safety Issues Sunnary" Regulatoiy Guide 1.139. " Guidance for Resicual Heat Removal" Regulatory Guide 1.113 2.
IMPLEFEliTATION DOCUMENTS:
TITLE NUDOCS NO,,
DATE Safety Evaluation Report NUREG-0853 2/82 related to the operation of Clinton Power Station (Section5.4.2) 3.
VERlf] CATION. DOCUMENTS:
TITLE NUDOCS NO.
DATE 1
)
)
l PLANT Clinton DOCKETN0(S), 50-461 3
FRNECT mat!AGER John B. Hickman TECHNICAL CONTACT J. Wermiel USI NO. A-36 TITLE Control of Heavy Loads. Phases I & II MPA NO.
C-10. C-15 TAC NOS.
ISSUES
SUMMARY
i This USI was resolved in July 1980 with the publication of NUREG-0612,)" Control i
of Heavy Loads at Nuclear Power Plants," and Standard Review Plan (SRP Section j
9.1.5.
The staff established MPAs C-10 and C-15 for the implementation of a
Phases I and II, 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 changes to ensure the safe handling of heavy loads. The guidelines proposed in !!UREG 0612 include definition of safe load paths, use of load handling procedures, training 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 I (six month 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
!!14.6. 6.
6.
Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9 l
7.
Design of cranes to ANSI B30.2 or CHAA-70 Phas'e 11 responses were to address Sections 5.1.2 thru 5.I.6 of NUREG-0612 which covered the need for electrical interlocks /mechanicel stops, or alternatively, single-failure-proof cranes or load drop analyses in the spent fuelpoolarea(PWR),containmentbuilding(PWR),reactorbuilding(BWR),other H
areas and the :ipecific guidelines for single-failure-proof handling systems.
I 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.
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.
1
i While not a requirement, NRC encouraged the implemer',ation of any actions identified in Phase !! regarding the handling of heavy loads that were considered appropriate.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
Review completed in $$ER 5. Section 9.1.5.
Licensee comitments were determined to be acceptable.
Implementation is incomplete. Additional procedural controls are to be implemented by 9/90.
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REFERENCES:
Clinton i
A-36 i
1.
RE001REMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE l
Letter, Darrell G. Eisenhut, NRC, to all licensees, ap)1icants for OLs and holders of C*s transmitting flVREG-0612 and staff positions 12/22/80 i
Generic Letter 85-11 Hugh L.
Thompson,ilRC,toalllicenseesfor Operating Reactors " Completion i
of Phase II of ' Control of Heavy Loads at Nuclear Power Plants' NUREG-0612" 06/28/85 2.
IMPLEMEllTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE SSER 5 related to the NUREG-0853 1/86 operation of Clinton Supp. No. 6 PowerStation(Section 9.1.5)
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3.
VER1rlCATION DOCUMENTS:
TITLE NUDOCS tl0.
DATE l.
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PLANT Clinton DOCKETN0(S).
LO-461 l
PROJECT MANAGER J. Hickman TECHNICAL CONTACT J. kudrick USl NO. A-39 TITLE Determination of SRV Pool Dynamic Loads and Temperature. Limits.
MPA NO.
TAC NOS. None..
f ISSUES
SUMMARY
ThisUSIwasresolvedwiththepublicationofStandardReviewPlan(SRP)
Section 6.P.I.1.C. in October 1982. In addition, NUREGs 0763, 0783 and 0802 were issued for Mark I, Mark II, and Mark 111 containments, respectively.
EWR plants are equipped with safety / relief valves (SRVs) to protect the reactor from overpressurization.
Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially subirerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool produces high-pessure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment inside containment.
NUREG-0802 presents the results of the staff's evaluation of SRV loads. The tvaluation, however, is limited to the quencher devices used in Mark 11 and 111 contairitnent s.
With respect to Mark I containtnents, the SRV acceptance criteria are presented in NUREG-0661, "Sefety Evaluation Report, Mark 1 Containment and Loris-Terra Prograr," end are dealt with as part of USI A-7.
SRP Section 6.2.1.1.C addresses the applicable review criteria, since all Mark l
11 end 111 containment designs are understood to have completed their operating license (OL) reviews subsequent to resolution of this USI and reflection of the resolution in the SRP.
i IPPLEMENTAT10hANDSTATUS
SUMMARY
.JPLANT. SPECIFIC):
Impleraented prior to licensing.
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RtFERENCES:
Clinton A-39 1.-
RE001REMENT DOCUMENTS:
l 9
l TITLE NUDOCS NO.
DATE SRP 6.2.1.1.C, Pressure Suppression Type BWR Containnents i
NUREG-0802, " Safety / Relief Yalve Quencher Loads:
Evaluation for EWR Mark 11 and III Containments, i
Generic Technical Attivity A-39" 1982 NUREG-0661, " Safety Evaluation Report -
7/80 Mark I Long Term Program" 2.
IMPLEMENTATION DOCUMENTS:
TJ!_l,E NUDOCS NO.
DATE Safety Evaluatin Report NUREG-0853 2/82 related to the operatiori cf and Clinton Power Station supp. 4 & 5 2/85,1/86 (Section6.2.).8) i 3.
VERIFICATION DOCUMENTS:
+
TITLE NUDOCS N0.
DATE
- The appliceble SRP revision number would depend on the date of the evaluation for each specific plant.
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l DOCKETN0(S).
50-461 PLANT Clint t.n 3,__.___
Ph00ECT MANAGER J. Hickman TECHNICAL CONTACT H. Ashar USI NO.
A TITLE Seismic Design Criteria...
MPA NO.
TAC NOS.
_None...
ISSUES.
SUMMARY
i The staff has resolved USI A-40 as docunented in NUREG/CR-5347, "Reconnenda-tions for Resolution of Public Connents 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, "Seist.ic Qualification of Equipment in Operating Plants " the staff concluded that tanks in plants that weresubjecttolicensingrevIewbythestaffafter1984hadbeenreviewedto currect requirenants and found accepteble.
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 (Collaway 1/2, l
Wolf Creek, Shearon Harris 1, and Watts Bar 1/2) are being handled on a plant-specific basis.
USI A-40 originated in 1977. Thebasicobjectiveswere(a)tostudythe seismic design criteria, (b) to quentify the conservatism associated with the I
criteria, and (c) to recomend modificetions to the Standard Review Plan (SRP) if changes are justified. Lawrence Livermore National Laboratory (LLNL) completed the study and published its findings in NUREC/CR-1161, " Recommended Revisions to USNRC - Seismic Design Criteria," dated May 1980. The report recomended specificchangestotheStandardReviewPlan(SRP). NRC staff reviewed the l
report and developed some other changes that would reflect the present state of l
seismic design practices. The resulting SRP changes were issued for public l
corrent in June 1988, and the final SRP changes are to be published in October 1989.
TtA major SRP changes consist of (a) clarification of development of site specific spectra, (b) justification for use of singic synthetic tine-history by pcwer spectral density function, (c) location end reductions of input ground motion fer soil structure interaction, and (d) design of above-ground vertical I
tanks.
Except for item (d), these items do not constitute any additional i
requirerr+nts for current licenses and applications, and thus, no backfitting is l
being required for these items.
However the revised provisions could be used formarginstudiesandreevaluationsorIndividualplantexaminationfor externalevents(IPEEE).
TheparticipantutilitiesintheSeismicQualificationUtilityGroup(SQUG) agreed to implen,ent the changed criteria for flexible vertical tanks for their plants.
For the four plants where this issue has to be resolved on an indi-vidualbasisa10CFR50.54(f) request-for-informationletterhasbeensentto the affected utilities.
If the inforniation received indicates that large above-ground vertical tanks do not meet the new criteria, plant-specific i
backfitt will be considered.
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IMPLEMENTATION.ANDSTATUS
SUMMARY
,(PLANTSPECIFICl:
l N/A, No action required.
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REFERENCES:
Clinton A-40 t
1.
REQUIREMENT _ DOCUMENTS:
TITLE NUOOCS.NO.
DATE f
Regulatory Analysis for NUREG-1233 Sept. 1989 l
USI A-40 Recommendations for Resolution NUREG/CR-5347 June 1989 cf Public Contents on USI A-40 Standard Review Plan NUREG-0800 To be issued i
Sections 2.5.2, 3.7.1, 3.7.2,3.7.3(Revision 2) i Response of Seismic NUREG/CR-4776 Feb. 1987.
Category I Tanks to Earthquake Excitation i
Engineering Characteri-NUREG/CR-3805 Feb.-Aug. 1986 ution of Ground Motion, Vols. 3,4,5 Proceedings of the NUREG/CR-0054 June 1986 Workshop on Scil-Structure Interaction Value Impact Assessment NUREG/CR-3480 Aug. 1984 for Seismic Design Criteria Seismic Hazard Analysis NUREG/CR-1582 Oct. 1981 Application of Methodology, Results and Sensitivity Studies, Vol. 4
{
Recommended Revision to NUREG/CR-1161 May 1980 Nuclear Regulatory Commission Seismic Design Criteria Power S)ectral Density Functions NUREG/CR-3509 June 1988 Compati/>1e with NRC R.G. 1.60 Response Spectra 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Request for Information Letters Docket Nos.
May 1989 to Owner's of Callaway 1&2, Wolf 483, 486, 482, Creek 1, Shearon Marris 1 Watts 400, 390, 391 i
Bar 1&2 3.
VEPIFICAT10h DOCUMENTS:
TITLE NUDOCS.NO.
DATE
l PLANT Clinton DOCKETN0(S).
50-461 PROJECT MANAGER J. Hickman TECHNICAL CONTACT W. Koo I
USI NO. A-42 TITLE Pipe Cracks in Boiling Water Reactors MPA NO.
TAC NOS. 69130 ISSUES
SUMMARY
This USl was resolved in February 1981 with the publication of NUREG-0313 i
Revision 1."TechnicalReportonMaterialSelectionandProcessingGuide1Ines for BWR Coolant Pressure Boundary Piping.' That NUREG document was issued to all holders of BWR operating licenses or construction permits and to all applicants for BWR operating licenses. The staff established NPA B-05 for implementation of the resolution at operating plants.
Pipes have cracked in the heat-affected zones of welds in primary system piping in BWRs since mid-1960. These cracks have occurred mainly in Type 304 stainless steel, which is the type used in most operating BWRs. The major problem is recognized to be intergranular stress corrosion cracking (IGSCC) of austenitic l
stainless steel components that have been made susceptible to this failure by being " sensitized ' either by post-weld heat treatment or by sensitization of a narrow heat affected zone near welds, l
" Safe ends" that have been highly sensitized by furnace heat treatment while attached to vessels during fabrication were found to be susceptible to IGSCC in the late 1960s. Most of the furnace-sensitized safe ends in older plants have been removed or clad with a protective material, and only a few BWRs still have furnace-sensitized safe ends in use. Most of these, however, are in smaller diameter lines.
Cracks reported before 1975 occurred primarily in 4-inch-diameter recirculation loop bypass lines and in 10-inch-diameter core spray lines.
Cracking is most often detected during inservice inspections using ultrasonic test techniques.
Some piping cracks have been discovered as a result of primary coolant leaks.
NUREG-0313. Revision 1 provided the HRC staff's revised acceptable methods for reducing the IGSCC susceptibility of BWR code class 1, 2, and 3 pressure boundary piping of sizes identified above and safe ends.
In addition, it provided the requirements for augmented inservice inspection of piping with nonconforming materials.
1
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As a result of further IGSCC degradations in larger piping, the staff provided licensees with additional requirements in several NRC communications (i.e.,
Du11etins 82-03, 83-2, and 84-11). The long-term resolution of IGSCC in BWR piping (including the scope of A-42) was provided in NUREG-0313, Revision 2 which was transmitted to all holders of BWR operating licenses via Generic Letter 88-01.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
No changes required NUREG-0313 rev. I was implemented prior to licensing. Licensee response to GL 88-01 is under staff review.
kT e',
.k i
REFERENCES:
Clinton L
A-42 E'
l 1..
REQUIREMENT DOCUMENTS:
i TITLE NUDOCS NO.
DATE i
NUREG-0313. Revision 1, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,"
07/80 k'
Generic Letter 81-03, "Implemen-tation of NUREG-0313, Rev. I for Selection and Processing Guidelines
'for BWR Coolant Pressure Boundary L
Piping (Generic Task A-42)"
2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE t
Safety Evaluation Report NUREG-0853 2/82 related to the operation of CPS (Section5.2.3) t Response to GL 88-01 8808050333 7/?9/88 Response to RAI, GL 88-01 8909280020 9/21/89 3.
VERIFICATION COCUMENTS:
i TITLE NUDOCS NO.
DATE h.
L a
\\
DOCKETN0(S). 5f:,411-PLANT Clinton r
PROJECT MANAGER J. Hickman.
TECHNICAL CONTACT A. Serkir. _
i USI NO. A-43 TITLE Containment.Emerg,eycy,$ g ; Performance c
MPA NO.
TAC N05. None JJSUES
SUMMARY
- 19. USI NO. A-43 TITLE: Containnwnt Emergency. Sump;ferformance The resolution of this USI was presented to the Consnission in October 1985 in SECY-85-349.
NUREG-0897, Revision 1. " Containment Ener presents the results of the staff's technical findings.gency Sump Performance,"
These findings estab-lished a need to revise current licensing guidance on these matters.
RG 1,82 Revision 0 arid Standard Review Plan Section 6.2.2, "Cor.tainment Heat Removal Systerns" were revistd to reflect this new guidance.
No licensee actions were t
required.
Initially an issue existed concerning the availability of ade tion cco1Ing water following e loss-of-coolant accident (LOCA)quate recircula-when long-term recirculatic.n of coolitig water from the PWR containn.ent sump, or the BWR resic* cal heat renovel system (RHR) suction intake, must be initiated and inainteiried to prevent core I:41t.
The technical concerns evaluated under US! A-43 were:
(e) post-LOCAadverse cor.ditions resulting front potential vortex forrnation and air ingestion and subsequent pur.p failure, (b) blockage of surnp screens with LOCA generated ir f ulation dt:bris causing inadequate net positive suction head (NPSH) on purps, ard (c) RHR and containn+nt spray pun;ps inoperability due to possible air, debris, or particulate ingestion on punp seal and bearing systems.
This revised guidance applies only to future construction perniits, preliminary design approvals, final design 'ipprovels, standardized designs, and applica-tions for licenses to manufacture. The staff performed a regulatory analysis to determint if this new guidance should be applied to operating plants. The i
results of this analysis were reported in t'l' REG-0869 Revision 1, "USI A-43 i
Regulatory Analysis," issued in October 1985. The staff concluded that the regulatory analysis does not support any new generic requirenants for present licensees to perform debris assessn.ents, l
IMPLEMENTATION.AND STATUS
SUMMARY
(PLANT SPECIFIC):
1.
N/A, No action required.
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REFERENC_E_S:
Clinton A-43 1.
g,0llJgMENT. DOCUMENTS TITLE NUDOCS NO.
DATE NUREG-0869, Rev. 1. "USl 10/85
. A-43 Regulatory Analysis" NVREG-0897 Rev. 1 "Conta 10/85 Dergency $ ump Performance"inment GL 85-22, " Potential for Loss 12/03/85 of Post-LOCA Recirculation Casability Due to Insulation c
Otaris Blockage" 2.
IMPLEMENTATIONDOCUMENLS:
E TITLE NUDOCS NO.
DATE N/A, No action required 3.
VERIFICATION.DOCUMENS:
TITLE NUDOCS.NO.
DATE l
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i PLANT Clinton DOCKETN0(S).
50-461 PROJECT MANAGER J. Hickman TECHNICAL CONTACT P. Gill USI NO. A-44 TITLE Station Blackout MPA NO.
TAC NOS.
68529 ISSUES
SUMMARY
This USI was resolved in June 1988 with the publication of a new r.ule (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 i
unavailability of the redandant onsite emergency ac power systems. WASH-1400 showed that station blackout could be an important risk contributor and 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 USI in 1980. A proposed rule was published for comment on March 21, 1986.
A final rule, 10 CFR 50.63, 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 station blackout i
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 operating license or the effective date of the rule, whichever is later.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
Incomplete Licensee responded to requirements in a letter dated April 16, 1989.
Staff SER is expected to be completed on December 31, 1990. The licensee has two years after the SER is issued to complete implementation.
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REFERENCES:
Clinton l
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A-44 L
1.
RE0u1REMENT. DOCUMENTS:
TITLE NUDOCS NO.
DATE 10 CFR 50.63, " Loss of All Alternating Current Power" 06/21/88 i
" Station Blackout" 08/88 2.
JfM,P_LEkENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Conformance of the HPCS Div.
8904060208 3/31/89 t
to the NUKARC 87-00 AAC Criteria CPS Response to the Station 8904250413 4/16/89 Blackout Rule i
3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS.NO.
DATE i
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PLANT Clinton DOCKETN0(S). 50-461 PROJECT MANAGER J. Hickman TECHNICAL CONTACT R. Jones USI NO. A-45 TITLE Shutdown Decay Heat Removal Requirements MPA NO.
TAC NOS. 74396 ISSUES
SUMMARY
USI A-45 was resolved by SECY 88-260, " Shutdown Decay Heat Removal Requirements (USI-A-45) " issued September 13, 1988, without imposing any new licensing requirements other than the Individual Plant Examination (IPE), as described i
below.
At the same time the staff issued NUREG-1289, "Regulttory and Backfit 1
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 Commission is currently implementing the Severe Accident Policy (50 FR 32130) and will require all plants presently operating or under construction to undergo a systematic examination termed the IPE. The reason for this examina-l tion is to identify any plant-specific vulnerabilities to severe accidents.
The IPE analysis intends to examine and understand the plant emergency pro-cedures, design, operations, maintenance, and surveillance, in order to identify vulnerabilities. The analysis will examine both the decay heat removal systems and these systems used for other related functions. This includes CE plants without power-operated relief valves.
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.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
l No changes required, subsumed by the IPE program.
l-IPE puidance was provided in GL 88-20 and GL 88-20 supplement 1.
The licensee l
responded to GL 88-20 S1 on October 27, 1989. That response provided the plan I
and time table for parforming the IPE.
NRC staff are reviewing the response for adequacy.
e 1
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kEFEREhcis:
Clinton A-45 1.
gpulREMENTDOCUMENTS TITLE NUDOCS NO.
DATE Federal Register Notice "10 CFR Part 50, Shutdown Decay Heat Ren. oval Requirements' NUREG/CR-5230 " Shutdown Decay Heat April 1989 Removal Analysis: Plant Case Studies and Special Issues Suninary Report' KUREG-1289 " Regulatory and Backfit 11/30/88 Analysis for the Resolution of USI A-45" SECY-88-260 " Shutdown Decay Heat 09/13/88 Rer.cvel Requirenients 2.
It!PLEMEi!TATJp Q O,CUg g S:
TITLE NUDOCS N0.
DATE GL 88-20:
IPE for Severe 11/23/88 Accider.t Vulnerabilities GL 88-20 Supplement No. 1 8/29/89
-CPS Eesponse to GL 88-20 8911060034 10/27/89 Supplement 1 3.
LEP.JflCATIOil DOCUMENTS:
i TITLE NUDOCS NO.
DATE i
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xlANT Clinton DOCKET N0(S). None l
PROJECT MANAGER J. Hickman TECHNICAL CONTACT P. Y. Chen USI NO. A-46 TITLE Seismic Qualification of Equipment in Operating f
None ISSUES
SUMMARY
l USI A-46 was resolved with the issuance of GL 87-02 on February 19, 1987, which I
endorsed the approach of using the seismic and test experience data proposed by the Seismic Qualification Utility Group (SQUG) and Electric Power Research 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 required to inspect equip-ment.
Evaluation of equipment will include:
(a) adequacy of equipment anchorage; (b) functional capability of esiential relays; (c) outliers and deficiencies (i.e., equipment with non-standard configurations); and (d)seismicsystemsinteration.-
AsanoutgrowthoftheSystematicEvaluationProgram(SEP),theneedwas identified for reassessing design criteria and methods for the seismic quali-fication of mechanical equipitent and electrical equipment. Therefore, the seismic qualification of the equipment in operating 31 ants must be reassessed to ensure the ability to bring the plant to a safe sautdown 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.
l Generic Letter 87-02 with associated guidance, required all affected utilities l
to evaluate the seismic adequacy of taeir plants. The specific requirements j
and approach for implementation are being developed jointly by SQUG and the l
staff on a ger.eric basis before individual member utilities proceed with plant-specific implementation.
IMPLEMENTATIONANDSTATUS
SUMMARY
(PLANTSPECIFIC):
Complete The Seismic Qualification Utility Group (SQUG) has been developing a Generic Implementation Procedure (GIP) to facilitate its member utilities implementation of the resolution of USI A-46.
The NRC staff has been reviewing the completed portion of the GIP and associated documents and l
reports, and has issued a generic Safety Evaluation Report (SER) on GIP Revision 0.
However, there will be supplements to the GIP for resolution of open issues. The licensee plans to perform seismic verificatinn plant walkdowns using GIP after receipt of the final SER Supplement and resolution of all open issues.
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REFEREt4CES:
Clinton
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A-46 l
1.
Rig!@EMENT DOCUMENTS:
L TITLE NUDOCS.NO.
.DATE f
Generic Letter 87-02, "Verifi-cation of Seismic Adequacy of r
Mechanical-and Electric Equipment in Operating Reactors" 02/19/87 NUREG-1211. " Regulatory Analysis for Resolution of Unresolved Safety Issues A-46..."
02/87 NUREG-1030, " Seismic Qualification of Equipment in Operating Plants, Unresolved Safety Issue A-46" 02/87 Letter attached with " Generic Safety Evaluation Report on SQUG GIP,) Revision 0," from L. Shac (NRC toNeilSmith(SQUG) 07/29/88
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2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE "Ger,eric Implen.entation Procedure (GIPforSeismicVerificationof iuclear Plant Equipment," Revision 0 06/88
" Generic linplementation Precedure (GIP)forSeismicVerificationof Huclear Plant Equipment," Revision I 12/88 3.
VERlflCATI,0N DOCUMENTS:
TITLE NUDOCS NO.
DATE l
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PLANT Clinton-DOCKETN0(S). 50-461 PROJECT MANAGER J. Hi c knen..
TECHNICAL CONTACT
- 0. Mauck USI NO. A-47 TITLE Safety In. plication of Control Systems in LWR Nuclear. Power Plants MPA NO.
TAC NOS. _None ISSUES.SUMM3RJ:
USI A-47 was resolved September 20, 1989, with the publication of Generic Letter (GL) 89-19.
The generic letter states:
"The staff has concluded that all FWR plants should provide autenietic steam generator overfill protection, all BWR plants shculd provide autcrratic reactor vessel overfill prctection, and that plant procedures and technical specifications for all plents should include provisions to verify periodically the cperability of the overfill protection and to assure that autor.atic overfill prctection is available to n.itigate nain ferdwater overfeed events during teactor power operation. Also, the system design and setpoints should be selected with the (bjective of minimizing inadvertent trips of the main feedwater systeni during plant startup, normal operation, and protection system surveillance. The Technical Specifications recomnenda-tiens are consistent with the criteria ard the risk considera-tions of the Connission Interim Policy Statement on Technical Specification Improvennt.
In addition, the staff reconatends that all BWR recipients reessess and mcdify, if needed, their operating procedures and operator training to assure that the operators.can niitigate reactor vessel overfill events that may cccur via tre condensate booster pumps during reduced system pressure coeration."
Also,lants.page 2 of the generic letter provides for additional actions for CE and E8W p The generic letter provides emp11fying guidance for licensees.
The generic letter requires that licensees provide NRC with their schedule and consnitn4nts within 180 days of the letter's date. The implementation schedule for 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.
IMPLEMENTATION.AND. STATUS
SUMMARY
_(PLANT SPECIFIC):
HRC requirements were provided in GL 89-19. Licensee response is due by April 18, 1990.
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RIFERENCES:
Clinton I
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A-47 1.
Rl0VIREMEtiT DOCUMENTS TITLE NUDOCS NO.
DATE Generic Letter 89-19 09/20/89 f
" Request for Action Related i
to Resolution of USI A-47" NUREG-1217 " Evaluation of Safety June 1989 Inplications of-Control Systems in LWR Nuclear Power Plants" NUREG-1218
- Regulatory Analysis July 1989 for Resolution of USI A-47" 2.
IMPL EMENTAT10h_ DOCUMENTS:
TITLE NUDOCS NO.
DATE h
3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE T
9 9
t PLANT Clinton DOCKETN0(S).
50 461 PROJECT MANAGER J. Hickman TECHNICAL CONTACT J. Kudrick USI H0. A 48 TITLE Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment MPA NO.
TAC NOS.
None i
ISSUES
SUMMARY
I I
The NRC staff concluded April 19, 1989, that USI A-48 is resolved, as stated in SECY 89-122.
USI A 48 was initiated as a result of the large amount of hy(TMI) accident..drogen genera and burned within containment during the Three Mile Island This issue covers bydrogen control measures for recoverable degraded core accidents for all BWRs and those PWRs with ice condenser containments.
Extensive research in this area has led to significant revision of the Com-mission's hydrogen control regulations, given in 10 CFR 50.44, published December 2, 1981.
l 10 CFR 50.44 requires inerting of BWR Mark I and Mark !! containments as a method for hydrogen control. The BWR Mark I and Mark 11 reactor containments have operated for a number of years with an inerted atmosphere (by addition of an inert gas, such as nitrogen) which effectively precludes combustion of any hydrogen generated. USI A-48 with respect to BWR Mark I and 11 containments is not only resolved but understood to be fully implemented in the affected plants.
The rule for BWRs with Mark III containments and PWRs with ice condenser containments was published on January 25, 1985. The rule required that these plants be provided with a means for controlling the quantity of hydrogen produced, but'did not specify the control method.
In addition, the task action plan for USI A-48 provided for plant-specific reviews of lead plants for i
reactors with Mark III and ice condenser containments. Sequoyah was chosen as the lead plant for ice condenser containments and Grand Gulf for Mark III containments. Both of the lead plant licensees chose to install igniter-type systems which would burn the hydrogen before it reached threatening concentra-tions within the containment.
Final design igniter systems have been installed
[
not only in both lead plants, Sequoyah and Grand Gulf, but in all other ice l
condenser and Mark III plants as well. The staff's safety evaluations of the final analyses required to be submitted by these licensees by the rule are I
scheduled for completion in 1989.
L Large dry PWR containments were excluded from USI A-48 because they have a greater ability to accommodate the large quantities of hydrogen associated with a recoverable degraded core accident than the smaller Mark I, II, III and ice condenser containments. However, this issue has continued to be considered and, in 1989, hydrogen control for large dry PWR containments was identified as a'high-priority Generic Issue (GI) 121. The resolution of GI 121 is being I
l actively pursued in close coordination with more recent research findings.
1 L
l l
1
p Clinton A-48 1
h ISSUES
SUMMARY
(CONT.):
i The t:RC staff has concluded that USI-A-4B is resolved as stated in SECY 89-122.
If interested, the report should be consulted for further details regarding the relationship of A-48 to other ongoing hydrogen activities.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
HRC staff SER on the Owners Group Topical Report on Hydrogen Control for BWR-6 Mark III to be issued shortly, l.icensee's final analysis to be submitted 6 months after generic SER.
f i
REFERENCES:
1.
F.EOUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE
~
10 CFR 50.44, Standards for 12/81 Combustible Gas System in Light-Water-Cooled Power i
Reactors SECY-89-122 Resolution of USI A-48, " Hydrogen Control Measures and Effects of Hydrogen' Burns on Safety I
Equipment" 04/19/89 2.
IMPLEt1ENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE CPS H Control Final Analysis 8708260185 8/24/87 2
Response to NRC questions on Prelim, analysis of CPS H control 8602240308 2/21/86 2
CPS H Control, License Condi$ ion 5 Prelim, Analysis 8511070172 11/04/85 L
3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE 1
L l<
l'
l
. e o l
L PLANT Clinton.
DOCKETN0(S). 50-461 PROJEC1 MANAGER J. Hickman TECHNICAL CONTACT B. Elliott.
US) NO. A-49 TITLE Pressurized. Thermal. Shock I
HPA NO.
TAC NOS. None ISSUES
SUMMARY
The final rule (10 CFR 50.61) on pressurized thermal shock (PTS) was approved by the Comission 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 requirenents 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 irrediation of reactor pressure vessel weld and plate materials i
decreases the fracture toughness of the materials. The fracture toughness sensitivity to radiation-induced change is increased by the presence of certain c.aterials such as copper. Decreastd fracture toughness makes it more likely thet, if a severe overcooling event occurs followed by or concurrent with high v(ssel pressure, and if a small cract is present on the vessel's inner surface, thet crack could grow to a size that n.ight threaten vessel integrity.
Severe pressurized overcooling events are improbebit since they require r.ultiple failures and improper operator perfomance. However, certain precursor events have happened that could have poter.tielly threatened vessel integrity if Edditional failures had occurred and/or if the vessel had been rcre highly irradiated. Therefore, the possibility of vessel failure due to a sev(re pressurized overcooling event cannot be ruled out.
IMPLEPENTATION AND STATUS
SUMMARY
(PLANT. SPECIFIC):
i N/A, TWR only i
l l:
l
/
r 0
i
/
- s. o i
REFERENCES:
Clinton
>/
A-49 1.
REQUIREMETDOCUMENTS:
, TITLE NUDOCS NO.
DATE i
10 CFR 50.61, "Tracture Toughness 7/85 i
Requirements for Protection Against Pressurized Thermal Shock Requirenents" Reg. Guide 1.154,
- Format and Content 1/89 of Plant-Specific Pressurized Thers.al c
a Shock Safety' Analysis Reports for PWRs" SECY 82-46$, " Pressurized Thennal Shock
- 11/23/82 SECY 83-288, "Preposed Pressurized Therr.:a1 Shock Rule" 07/15/83 P.egulatory Guide 1.154
Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Repcrts for Pressurized Water Reactors" 02/87 Generic Letter 88-11, "NRC Position on Radiaticn Entrittlen.er.t of Reactor Vessel Materials and Its impact on Plant Operations" 7/12/88 2.
lHPLEEEl4TATION DOCUMENTS:
TITLE NUDOCS NO.
DATE II/A, PWR only b
3.
VERIFICATION DOCUM,E TS:
Q TITLE NUDOCS.NO.
DATE l
)
l l
~
y l cc o.
F#pt No.,
1 0C/01/90.
llS11N6 0F INCOMPLETE USl M1A-FORINFU1FROMPROJECTMANAGERS
(
ISSUE !$$UE DESChlfilVE NME IMPLEMENT IMPLEMENT LICENSEE COMEN1 STAFF COMENT NUMBER DATE S1ATUS 81 PLANT NME: CLINTON 1
A-01 NAl[R NAMER
/ /
NC y
'A02 ASYMETRIC llLOWDONN LDADS ON
//
N/A PWRDNLY l
REAC10R FRlMARY C00LANT SYSTEMS A 03 NESilN6 HOUSE STEM BENERAIDR IUBE' //
N/A NES11N6HOUSEONLY INTEER!iY A 04 CE STEAM SENERATOR TUBE INTE6RITY / /
N/A CE PLANTS ONLY A 05 EIN STEAM SENERATOR Tulle
/ /
N/A B6WPiANTSONLY j
INTE6Rl1Y 1,-06 MARK 1 SHORT TEPR PR06 RAM
/ /
N/A MK 1 SWR ONLY A 07 MARK 1LONS-TERMFR06 RAM
'l /
N/A MK1BWRDNLY A 06 MAFK 11 CONTA!NMENT POOL DYNAMIC
//
N/A MK11BWRDNLY LOADS-LONS1ERMPROERAM A 09 ATNS
/ /
NC A 10 EM FEEDWATER N0ilLE CRACKING
/ /
NC A11 REAC10RVISSELMATERIALS
//
NC l
10U6HNEES A12 FRACTURE TOU6HNEES OF STEAM
//
N/A CP AFTER B3 DNLY
~BENEF,ATORANDREACTORCOOLANT I
PUMPSUFFORTS A 17 SYSTEMSINTERAtll0N
//
NC NOREDUlREMENTS A04 OUALIFICAi!DN OF CLASS 1E
//
NC SAFETY-RELATEDEDUlPMENT 3
A 26' REACTOR VESSEL FRESSURE TRANS![NT / /
N/A BL BB-11 PWRONLY FROTECTION A31 RHRSHUTDOWNREQUIREMENTS
//
NC LICENSIN6SER
(
A 36 CONTROL OF HEAVY LOADS NEAR SPENT 09/30/90 1 PROCEDURES FUEL A39 DETERMINAil0NOFSAFETYRELIEF
//
NC VALVE P00L DYNAMit LOADS ANI-TEMPERATURELIM116 A 40 SEISMICDES!6NCRITERIA-
//
NC SHDR1 TERM FROGRAM
{
A 42 PlFE CRACKS IN B0! LING WATER
//
NC 6L BB-01 N/A REACTORS A-43 CONTAINMENTEMER6ENCYSUMP
//
NC
!NFODNLY FERFORMANCE A 44 51A110NBLACK001 12/31/92 1 SER12/31/90 A-45 SHUTDOWN DECAY HEAT REMOVAL
/ /
NC IPE SUBSUMED BY SEVERE ACC REQUIREMENTS A 46 SE!!MitDUALIFlCAil0NOF
/ /
N/A DLD PLANTS ONLY EDU!FMENTINOPERATINGPLANTS A 47 SAFETY IMPLICA110NS OF CONTROL 04/10/90 E NEW REQUIREMENTS SYSTEMS A-4E' HYDR 06EN CONTROL MEASURES AND 12/31/90 1 PENDINB SER REVIEWING 06 TOPICAL EFFECTS OF HYDR 06EN TURNS ON SAFETY EDUlFMENT A-49 PREESURllED THERMAL SHDCK
/ /
N/A FWR DNLY
.,, -