ML20044A554

From kanterella
Jump to navigation Jump to search
Forwards Response to Generic Ltr 90-04 Requesting Info on Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions
ML20044A554
Person / Time
Site: Waterford Entergy icon.png
Issue date: 06/26/1990
From: Burski R
ENTERGY OPERATIONS, INC.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TASK-***, TASK-OR GL-90-04, GL-90-4, W3P90-1144, NUDOCS 9006290202
Download: ML20044A554 (55)
v • d • e


Text

e

w. -

gEntergy

..e ,,, o,,,,wo =, io..

L -r_-:- 3#2;[i 70m O ations ==

b 504 73M774.

Raymond F. Butski tAv or SW g- f. h Ji .ty'_A y A1JS W3P90 1144 A4.05

~

~

June 26, 1990 9^

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. .20555

SUBJECT:

Waterford 3 SES  !

e Docket No.' 50-382 License No. NPF 38 Request for Information On the Status of Licensee Implementation

]f!

of Generic Safety Issues Resolved With Imposition of Requirements  !

or Corrective Actions (Generic Letter 90-04)

Gentlemen: '

In the subject letter you requested a review and reporting of the ' status of GSI j resolution and implementation at Waterford 3. An enclosure was provided which identified a table of GSIs included in the request.

In response to your letter we have conducted a review of the appropriate records to determine the status of GSIs at Waterford 3. The results of our review have been included on the attached table in accordance with the guidelines presented i

'in Enclosure - 1 of the subject letter. For your convenience, we have also provided an additional' enclosure with more extensive status information.

'Should you need. additional information or wish to discuss this matter further, please do not hesitate to contact George Wilson at (504) 595-2837.

Very Truly Yours, t

FB/PLC cmb

' Attachment cc w/ attach: R.D. Martin, NRC Region IV s F.J . liebdon, .NRC-NRR D.L. Wigginton, NRC NRR E.L. Blake W.M. Stevenson R.B. McGehee NRC Resident Inspectors Office

b 9006290202 900626 '

i

~ PDRx ADOCK 03000382 k

P; PDC o '.

0_

FACILITY NAME: WATERFORD 3 SES l DOCRET ND.: 50-382 ,

LICENSEE: LP&L

'l' STATUS OF LICENSEE IWLEMNTATION OF GEIERIC SAFETY ISSUES RESOLVED WITH IWOSITION OF REQUIREM NTS OR CORRECTIVE ACTIONS TITLE APPLICABILITV STATUS

  • CMENTS GSI/(W A No.}

40 (8065) Safety Concerns Associated With All BWRs ,fa ^

Pipe Breaks In The OMR Scram System 41 (8058) BWR S. cram Olscharge Volume Systems All BWRs ,fA 43 (8107) Reliability Of Alt Systems All Plants I Refuel 4 Design Change

' 3/91 51 (L913) Improving (the Re11 ability of All Plants I nefuel 4 see subetttal Open-Cytile Service Water Systems 3/91 dated 1/29/90 l-i 67.3.3 (A017) Improved Accident Monitoring All Plants c 12/12/89 see subetttal

, dated 7/7/87 75** (8076) Iten 1.1 - Post-Trip Review All Plants c 7/17/89 See sER dated (Program DeSCflption and 6/7/89 Procedure) 75 (8085) Item 1.2 - Post-Trip Review - All Plants c 5/11/88 See SER dated Data and Information Capability

~

7/5/88 l

  • Please follow attached gulfwce for completing-this column.  ;
    • The 16 items listed for GSI 75 all relate to actions derived from the generic lapilcations of  ?

Ites numbers correspond to Generic Letter'83-28 action item numbers. j-Salen ATWS events.

m m1

~

~ - - -

. = ..

T 3 O

GSI/(IPA No.) TITLE ' APPLICABILITY STATUS

  • COBOIENTS 75 (9077) Item 2.1 - Equipment Classi- All Plants C 11/15/85 See SER dated' fication and Vendor Interface 8/26/87 (Reactor Trip System Components) 75 (8086) Item 2.2.1 - Equipment Classiff- All Plants C 11/15/85 See SER dated cation for Safety-Related Components 3/30/88 75 (LOO 3) Item 2.2.2 - Vendor Interface All Plants E 9/90 Response due CL 90-03 for Safety-Related Components 75 (B078) Items 3.1.1 & 3.1.2 - Post - All Plants N/C See SER dated Maintenance Testing (Reactor 9/22/88 Trip System Components) 75 (8079) Item 3.1.3 - Post-Maintenance All Plants N/C See SER dated Testing-Changes to Test Sequire- 5/15/86 ments (Reactor Trip System Components) 75 (8087) Items 3.2.1 & 3.2.2 - Post- All Plants ' C 11/15/85 See SER' dated Malatenance Testing (All Other ' 9/22/88 Safety-Related Components) 75 (8008) Item 3.2.3 -- Post-Maintenance All Plants N/C See SER dated Testing-Changes to Test Require- 5/15/86

- ments (All Other Safety-Related Components) 75 (8080) Item 4.1 - Reactor Trip' System' All Plants N/C See SER dated

' 10/20/87 i'

Reliability (Vender-Related ,

Modifications) swa -.+.m.--_.s- -----.-__w -

ma- sews------=--am

-w- ---sa -- ~.. ww ,

, w _.ne----- w-.d. s:.ww.- - D-

,a-

ii

-j; 5

?

APPLICABILITY COIGENTS z

~

GSI/(MPA No.) TITLE . STATUS

  • i!

75 (B081) Items 4.2.1 & 4.2.2 - Reactor- All PWRs c 9/2/87 see sEn dated Trip System Reliabillty- 10/20/87:

Maintenance and Testing .

.. 4

' ~-

(Preventative Maintenance and -

~ Surveillance Program for Reactor Trip Breakers)-

75 (B082) Ites 4.3 - Reactor Trip System All W and 88M . nya Reliability - Design Modifications Plants (Automatic Actuation of Shunt Trip Attachment for Westinghouse and 88W ,

Plants) ,

-t Item 4.3 - Reactor Trip System All W & 88N "

75 (8090) n/A Reliability - Tech Spec Changes Plants (Automatic Actuation of Shunt Trip Attachment For Westinghouse and 88W Plants) 75 (8091) Item 4.4 - Reactor Trip System All 88N Plants ufg Reliability (Improvements in Maintenance and Test Procedures for 88N Plants) t 4

~

I

. [

b li

vy x w. = ~ . g.. - -

m= ; =; , . ; ;z pg. .

g

?

1 GSI/(9FA No.) TITLE APPLICABILITV STATUS

  • CO M NTS 75 (8092) Item 4.5.1 - Reactor Trip System All Plants N/C See SER dated-Reliability-Diverse Trip Features 9/2s/88 (System Functional Testing) 75 (8093) Items 4.5.2 & 4.5.3.- Reactor Trip All Plants N/C See SERs dated System Reliability - Test Alterna- s/26/87 and tives and Interva15 (System S/24/s9 Functional Testing) 86 (8084) long Range Plan for Dealing All SWRs N/A I with Stress Corrosion i Cracking in SWR Piping 93 (8098) Steam Binding of Auxiliary A11 Ptsts c 7/s6 see submittal Feedwater Pumps dated s/22/s6 i Response 1R8613-03 99 (L817) RCS/RHR Suction Line Valve All PWRs I Refuel 4 see sabetttal .!

Interlock en PIRs 3/91 dated S/21/90 ,

124 Auxiliary Feedwater System AND-1&2, Ranche' N/A' -

Re11abt11ty Seco Prairie Island 1&2 Crystal River-3 t Ft. Calhoun A-13 (8017) Snubt,er Operability Assurance - All Plants -N/c  !

Hydraulic Snubbers I

( ,

i a

i

~

. . _ _ . : =>

J-

p. -

~

.,  :=_.

--S- s a.

1.  : ~.,

7.

! GSI/(IFA No.) TITLE APPLICA81LITY STATUS

  • C0f0ENTS -

Snubber Operability Assurance -

A-13 (8022) All Plants N/c Mechanical Snubbers A-16 (0012) Steam Effects on SWR Core Oyster Creek N/A Spray Olstribution &per-1 A-35 (8023) Adequacy of Offsite. Power .All Plants N/C See SER NUREG 0787'!

Systems dated 7/81 (8.2.4) !

, 8-10 8ehavior of BWR Mark III All BWR Mark III N/A M '

Containments Plants 8-36 Develop Design, Testing and All Plants with N/A j!

Maintenance Criteria for OL Applications -l Atmosphere Cleanup System- After 4/1/80 .l Air Filtration and Adsorption i!

Units for Engineered Safety:

Features Systems and for  :

Normal Ventilation Systems B-63 (8045) Isolation of Low Pressure- All Plants N/C See SER NUREC 0787:

Systems Connected to the Reactor dated 7/81 (3.9.6).- .

Coolant System Pressure Boundary t

4 1

,y sI i-

f I 4 L

L ADDITIONAL ENCLOSURE TO W3P90-1144 GENERIC LETTER 90-04 l  ;

IMPLEMENTATION & STATUS

SUMMARY

GENERIC SAFETY ISSUES FOR WATERFORD 3 SES

ISSUES

SUMMARY

GSI No. Al (MPA No. B 107) TITLE: Reliability of Air Systems This issue arose from staff concerns related to the Three Mile Island accident I and subsequeit air operated equipment failures at other plants. Some of these equipment failures are described in Information Notice (IN) 87-28 and IN 87 28, Supplement 1. ,

The staff's generic Safety Evaluation Report, NUREG-1275, V.2, was provided to all licensees and applicants by IN 87-28, Supplement 1. Generic Letter 88-14 identified requested corrective actions. These actions consisted of three types of verification and a discussion of a program for maintaining air i quality. The three types of verification included: (1) test verification of air quality, (2) verification of adequate maintenance practices, emergency procedures, and training, and (3) verification of design and failure modes. ,

Responses concerning implementation of these actions were to be submitted within 180 days with allowances made for implementation of actions requiring ,

outages to complete. *

References:

1. NRC Information Notice No. 87 28, " Air Systems Problems at U.S. Light Water Reactors," June 22, 1987.
2. NRC Information Notice No. 87-28, Supplement 1, December 28, 1987.
3. . NUREG 1275, " Operating Experience Feedback Report - Air Systems Problems," U.S. Nuclear ReSulatory Commission, Vol. 2, December 1987.
4. NRC Letter to All.lloiders of Operating Licenses or Construction Permits for Nuclear Power Plants, " Instrument Air Supply Systems-Problems '

Atfecting Safety-Related Equipment (Generic Letter 88-14)," August 8, i 1988. 3 j

l.

IMPLEMENTATION AND STATUS

SUMMARY

By submittal dated February 21, 1989 the licensee for the Waterford 3 SES facility responded to the design and operations verification review of the instrument air system-imposed by Generic Letter 88-14.

The licensee has completed all outstanding items with the exception of minor FSAR updates and a potential problem concerning the air accumulators associated with valves SI 602A and B. This item was identified in the  !

licensee's response and subsequently discussed in NRC Inspection Report 50 382/89 08, Inspector Followup Item (IFI) 8908-03. The licensee has scheduled resolution of this item'by the March 1991 refueling outage and in accordance with Generic Letter 88 14 will notify the staff when all requirements have been implemented by 04/01/91.

-IMPLEMENTATION DOCUMENTS:

TITLE EU DAIE CEJ 001'004 07 08/04/89 CE-002 032 00 09/28/89 01-004-000t 12 04/17/89 l OP 003 016 04 09/21/89 OP-500 008 04 09/05/89 j 1 .

.OP 901 038 04 07/15/89- .j OP 902-002- 03 08/04/89

.OP 902 004 03 08/04/89 OP-902 005- 03 08/04/89 ,

OP 002 006 03 08/04/89

.0P 902-007 03 08/04/89

.0P-902 008 04 08/04/89

-OP 903J032. 07 09/05/89

,OP 903-033' 08- 08/19/89

=

WA 01033500 WA 01033543 i

VERIFICATION DOCUMENTS TITLE NUDOCS NO. DAIE W3P89-0028. -02/21/89

'R.F. Burski (LP&L) to USNRC W3P89-3016 04/28/89 R.F.=Burski (LP&L) to USNRC Inspection Report 50-382/88-31, Section 3.E 02/09/89 Inspeccion Report 50-382/89-08, IFI8908-03 05/16/89

ISSUES

SUMMARY

GSI No. 11 (MPA No. L 913) TITLE: Imorovine the Reliability of 4

H Ooen Ovele Service Water Systems This issue arose from operating experience and studies related to Bulletin i 81-03 which led the NRC to question the compliance of the service water systems with the requirements of GDC 44, 45, 46 and Appendix B to 10 CFR Part .

50.

The resolution of GSI No. 51, along with implementation of AEOD and Region II recommendations, affected all plants and addressed the following actions:

(1) reduce flow ~ blockage problems from biofouling, (2) conduct a heat transfer testing program on safety-related heat exchangers in open cycle systems, y (3) establish a routine inspection and maintenance program for open-cycle  ;

system piping and components, (4) confirm that the service water system will -l perform its intended function in accordance with the licensing basis for the plant;.and (5) confirm the adequacy of relevant maintenance practices, l operating and emergency procedures, and training.

Generic Letter 89-13 requested licensees to advise the staff whether they have established programs to implement the above five actions resulting from the resolution of GSI No. 51, or equally effective alternative courses of action.

The Generic Letter also requested licensees to confirm to the staff that all recommended actions or equivalent alternatives have been implemented.

Re ference s : <

1. NRC Bulletin No. 81-03, " Flow Blockage of Cooling Water to Safety System i Components by Corbicula sp. (Asiatic Clam) and Mytilus sp. (Mussel)," M April 10, 1981.
2. NRC Letter to All Holders of Operating Licenses or Construction Permits .

for Nuc1 car Power Plants, " Service Water System Problems Affecting )

Safety-Related Equipment (Generic Letter 89-13)," July 18, 1989.

E

l. ,

I IMPLEMENTATION AND STATUS

SUMMARY

By submittals dated July 7,1981 and January 23, 1982 the licensee for the i Waterford 3 SES facility responded to IEB 81 03, Inspection Report

1. 50-382/83-36 documents NRC closure and Inspection Report 50-382/86-11 l documents an inspection of the actions committed to in response to IEB 81-03.

No violations or deviations were identified.

The licensee responded to Generic Letter 89-13 by submittal dated January 29, 1990 and is currently performing several implementation activities which are scheduled for completion by 05/31/91. In accordance with Generic Letter l 13 licensees are to inform the NRC within 30 days, following implementation of activities. By letter dated February 9, 1990 the staff requested that LP&L l

reference Generic Letter 89-13 and TAC No. 74081 in that response.

l

.(

1

^

L;-

IMPLEMENTATION DOCUMENTS PROCEDURE E DAIE CE 001-009 01 08/02/88 VERIFICATION DOCUMENIS TITLE NUDOCS NO. DAIE 5 W3P81 1578 07/07/81 L.V. Maurin (LP&L) to K.V. Seyfrit (NRC)

W3P82-3056 11/23/82 L.V. Maurin (LP&L) to J.T. Collins (NRC)

Inspection Report 50 382/83-36, Fection 4 01/20/84 Inspection Report 50-382/86 '.1, Section 8 07/16/86 W3P90 0207 01/29/90 R.F. Burski (LP&L) to USNRC Letter 02/09/90 F.J. Hebdon (NRC) to J.G. Dewease (LP&L) l l

l l

l l

d ISSUES

SUMMARY

GSI No. 67.3.3 (MPA A-017) TITLE: Imoroved Accident Monitoring This issue addresses compliance with Regulatory Guide 1.97. NUREG 0737,

" Clarification of TMI Action Plan Requirements," was issued in 1980, followed by Supplement 1 -(issued as Generic Letter 82 23) in December 1982. Supplement I requested proposed schedules for implementing the provisions of Revision 2 to Regulatory Guide 1.97. In addition, licensees and applicants were requested to submit details, for staff review, of how they would comply with the provisions of Regulatory Guide 1.97, Rev. 2, and to identify any exceptions to'or deviations from these provisions.

Based on licensee responses to Supplement 1, confirmatory orders were issued to operating plants in 1985. For license applications still under review, implementation would be addressed as part of the licensing process.

References:

1. Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," U.S. Nuclear Regulatory Commission, December-1980,
2. NRC Letter to Licensees of Operating Reactors, Applicants for Operating i Licenses, and Holders of Construction Permits,_" Supplement 1 to NUREG 0737 - Requirements for Emergency Response Capability (Generic

. Letter No. 82-33)," December 17, 1982.

IMPLEMENTATION A E STATUS

SUMMARY

By letter dated July 6,1983, as supplemented by letter- dated November 6, 1984, the licensee provided a response to Generic Letter 82-33 that included exceptions to the' provisions of RG 1.97, Revision 2. The exceptions were reviewed by the staff as indicated in SER NUREG 0787 Supplement 8 dated December 1984.and were found acceptable with the exception of those concerning i reactor coolant system pressure and containment isolation valve position indication.

By letter dated November 21, 1984 the licensee provided a commitment to install Category I instrumentation for the RCS pressure range instrumentation consistent with RG 1.97 Revision 2. By letter dated November 29, 1904 the licensee committed to install direct, continuous position indication for the control room. The staff found these commitments acceptable and concluded'in SER NUREG-0787_ Supplement 8 dated December 1984 that the licensee conforms to the requirements of NUREG-0737, Supplement 1, as required by Generic Letter 82-33.

The licensee's commitment to install continuous, direct position indication in the Control Room for containment isolation valves for instrument line penetrations 53 and 65 was completed prior to startup following the first refueling outage 1/87.

I LI The commitment to Category.1'RCS Pressure Instrumentation _was completed during the third refueling outage 12/89.  ;

3 IMPLEMENTATION DOCUMENTS TITLE REY DAIE Station Modification-631 Condition Identification / Work'

-Authorization 027990 Work Authorization 990002861 Design Change 3080 VERIFICATION DOCUMENTS TITLE NUDOCS NO. DAIE W3P83 1194 04/15/83_ l

'L.V. Maurin (LP&L) to G.W.- Knighton (NRC) 1 W3183 0177 07/06/83 F.J . Drummond '(LP&L) . co H.R. Denton (NRC)

W3P84-3029 11/06/84.

.K.Wi Cook (LP&L)-to G.W. Knighton (NRC)

W3P84 3246 11/21/84 K.W. Cook (LP&L) to G.W. Knighton (NRC)

  • W3P84-3318 11/29/84 K.W. Cook (LP&L) to G.W. Knighton' (NRC)

SER NUREG-0787 Supplement 8 Section 7.5.2 12/00/84 OL 50 382/NPF-38 (License-Condition) 2.C.1 03/16/85-W3P86-1670 08/27/86 d K.W. Cook (LP&L) to G.W. Knighton (NRC) 1 Letter J.H. Wilson (NRC) to J.G. Dewease (LP&L) 02/17/87 1 W3P87 1125 07/07/87

~K.W. Cook (LP&L) to USNRC

' Letter J.H. Wilson (NRC) to J.C. Dewease (LP&L) 08/20/87 l

L.

c ISSUES

SUMMARY

GSI No. 11 (17 Individual MAPS) TITLE: Generh Imolications of ATVS I Events at the Salem Nuclegg Plant i

This issue arose from staff concerns resulting from analysis of events that i occurred at the Salem Nuclear Power Plant on February 22 and 25, 1983. The  ;

analysis of the events revealed that a total loss of automatic scram  ;

capabi'ity (an anticipated transient without scram, or ATWS event) had  ;

occurred each time. The relatively mild transients, coupled with the rapid  !

manual shutdown of the reactor by the operators both times, turned these potentially serious events into little more than routine reaceor shutdowns. ,

However, the implications of these events vis a vis scram system reliability '

were considered to be extremely safety significant.

The study of these events resulted in the issuance of NUREG 1000 and Generic Letter 83-28. The Generic Letter contained a number of items and sub-items addressing those aspoets of GSI 75 which have been resolved, each requesting specified actions of all or identified categories of licensees and applicants. ,

It should be noted that two aspects of GSI 75 have not yet been fully resolved and thus are not included herein. One-of these was not addressed in GL 83-28 and involves possible revisions to Reg, Guide 1,33, "QA Program Requirements 4 (Operations)" to contain more detailed guidance for operational QA programs.

The second relates to Items 4.2.3 and 4.2.4 of CL 83-28 which address life testing and replacement of reactor trip system components. The staff is currently reassessing the methods for establishing reactor trip reliability and may issue a future generic communication on these items.

The 16 sub issues of GSI 75, described below, consist of items and sub items from GL 83-28 in accordance with' how they were grouped into Hulti-plant Actions (MPAs) by the staff for tracking purposes. Each sub issue relates to a single MPA and may contain more than one sub-item from GL 83-28.

Re fe rences :

1. NRC Letter to All Licensees of Operating xeactors, Applicants for Operating Licenses, and Holders of Construction Permits, " Required Actions Based on Generic Implications of Salem ATUS Events (Generic Letter 83 28)," July 8,1983.
2. NUREG 1000, Volume 1, " Generic Implications of ATUS Events at the Salem Nuclear Power Plant," U.S. Nuclear Regulatory Commission, April 1983.

'3. NUREG 1000, Volume 2 August 1983.

IMPLEMENTATION AND STATUS

SUMMARY

By submittal dated May 30, 1984 the licensee for Waterford 3 SES provided a consolidated view of LP&L's compliance concerning Generic Letter 83-28. This submittal appears as Appendix C in SER NUREG-0787 Supplement 8 dated December 1984. The SER notes that a condition will be placed in the operating license

~

,, .: .a (OL NPF 38 2.C.13) requiring the licensee to submit responses and implement the requirements of Generic Letter 83 28 on a schedule consistent with that given in the May 30, 1984 submittal.

A separate Implementation and Status Summary has been generated for each of the following 13 sub. issues of CSI 75 app 1'. cable to the Waterford 3 SES facility.

CSI/(MPA NO.) CSI/IMPA NO.)

75 (B076) 75 (B079) 75 (8085) 75 (B087) 75 (B077) 75 (b088) 75 (B086) 75 (B080) 75 (LOO 3) 75 (B081) 75-(B078) 75 (B092) 75 (B093)

VERIFICATION DOCUMENTS TITLE HUDOCS NO. D&IE SER NUREC 0787 Supplement 8 Section 7.2.9 12/00/84 and Appendix 0 W3P84 1528 05/30/84 K.W. Cook (LP&L) to C.W. Knighton (NRC)

OL $0 382/NPF 38 (License condition) 2.C.13 03/16/85 6

r i

i .

l t

ISSUES

SUMMARY

(MPA No. B 076) TITLE: Item 1.1 - Pont Trio Review (Pronram Descriotion and Procedure) 1 The resolution of this ites, applicable to all plants, requests that licensees and applicants describe their programs for ensuring that unscheduled reactor '

shutdowns are analyzed and a determination made that the plant can be restarted safely. .

As a minimum,.each licensee is requested to describe: (1) the criteria for  !

determining the acceptability of restart, (2) the responsibilities and authorities of personnel who perform the review and analysis, (3) the necessary qualificotdons and trainlug for the responsible perscnnel, (4) the

. sources of plant ivformation nervasary to conduct the rev1 w and analysis. (5) the u.ethoda and critoria for comparing the event information with known or 6 expect (d plant behavior, (6) the criteria for determining the need for an .

indepondent m.essment of an event and guidelines on the preservation of '

physical evidence to support independent analysis of the event, and (7) the systematic safety assessment procedures compiled from (1) to (6) which are used in conduct.ing the evaluation ty the staff. l R;E1.EliEElbTION AND STAT'JS

SUMMARY

In Safety Evaluation Report for Waterford 3 Response to Generic Letter 83 28 t 1.1 (Tost Trip Review) dated October 10, 1985, the staff concluded that the licensees response dated November 4, 1983 and February 2, 1984 did not meet the guidelines for post trip review in the following areas:

1. The criteria for determining the acceptaSility of the restart.
2. The criteria for determining the need for independent assessment of the event following a trip.

By review of licensee submittals dated October 3, 1985 and June 2, 1989 the staff stated the following in a letter to the licensee dated June 7,1989: .

"We find the LP&L response to Generic Letter 83 21 Item 1.1 to be acceptable.

This closes the action under TAC 57698. We have also recorded your implementation date as July 17, 1989 as noted in the June 2,1989 LP&L ,

letter."

i i

t- ,

IMPLEMENTATION DOCIMENTS

{

PROCEDURE M  ;

E D&II ,

i OP 100 012 Post Trip Reviev 01 05/31/89  :

I L VERIFICATION DOCUMENTS  !

r HEE NUDOCS NO. Dgg

[

, W3P83 3911 Section 1.1 11/04/83  :

'K.V. Cook (LP&L) to D.C. F.isenhut (NRC) ,

W3P84 0297 02/06/84 It.V. Cook (LP&L) to Mr. C.W. Knighton (NRC)  !

Letter E.E.R. 10/03/85  !

Ceorge W. Knighton (NRC) to Mr. R.S. Leddick (LP&L) ,

W3P85 2697 .

10/15/85 j K.V. Cook (LP&L) to Mr. G.W. Knighton (NRC) t W3P89 3029 06/02/89 R.F. Burski (LP&L) to USNRC Document Control Desk l Letter Frederick J Hebdon (NRC) .to 06/07/89 Mr. J. Dewease i

e,

)

4 l

l I

E. ,

h '

  • O .. ,

C i- ISSUES fiUMMARY:

(MPA No. B 085) TITLE: Item 1.2 Post Trio Review - Data and Information L Cacability Item 1.2 requests that licensees and applicants have the capability to record, recall, and display data and information to permit diagnosing the causes of unscheduled reactor shutdowns and the proper functioning of safety related equipment during these events using systematit safety assessment procedures.

The data and information are to be displayed in a form that is user friendly and reflects human factors considerations. It further requests licensees and applicents to prepare and submit a report which describes and justifies the adequacy of their equipment for diagnosing an unschr.duled reactor shutdown.

Submittais are to te reviewed by tha staff to determine vbether asiequate data and information vill be available to support the systematic tasessment of i unscheduled reactor shutdowns.  !

I IMPLEMENTATION AND STATUS _ EUMMARY:  !

A Technical Evaluation Report (TER) on the status of Itcu 1.2, " Data and information capabilities" was provided to the licensee by letter dated April 6, 1988. Based on review of licensees response dated November 4, 1983,  !

Tebruary 6, 1984 and October 15, 1985, the TER found the post trip review data ,

and information capabilities to be adequate except for the following three i

areas:

1. Parameters recorded
2. Time history recorders performance characteristics
3. Data retention capability By letter dated May 11 1988 the licensee provided additional information on the open items. In a letter to the-licensee dated July 5, 1988 the staff stated the following: "On the basis of our previous review in the TER and the LP&L response dated May 11, 1988, we find that Waterford 3 has met the requirements of Salem ATWO Item 1.2 and this action is closed."

p

_ .,m.,m-x v ~- . , , _ , ,,

~

'l h t'

?

IMPLEMENTATION.. DOCUMENTS PROCEDURE IIILE ggy. DAIG OP.100 012 - Post Trip Review 01 05/31/89 l VERIFICATION DOCUMENTS IIILE NUDOCS NO. DAIE l W3P83 3911 section 1.2 11/04/83  ;

K.W.. i Cook,(LP&L) to D.C. Eirenhut (NRC) -

W3P84 0 88 _ _

02/06/84 l 7-

'K.W. Cook (LP&L) to G.W. Knighton (NRC)  !

. .I W3P85 260'/ ~

10/15/85 ]

K.W. Cock (LP&1.) to G.W. Knighton (NRC) l

. Letter T.E.R. 04/06/88 .]

' D.L. Wigginton (NRC) to J.G. Dewease (LP&L) i W3P88 0989 05/11/88-R.F. Burski-(LP&L) to USNI:C >

Letter D.L. Wigginton (NRC) to

. 07/05/88 J.G. Dewease (LP&L) l .

l I

4,

[

t

+

!I F >

t e , , --,- ,,,,- ,- -.av e ---e,- w - , . , - , . . , -,- - , , - - , ,, - - - -

E

.e ISSUES

SUMMARY

(MPA No. B 077) TITLE: Item 2.1 - Eculement classiffeation and vendor Interface (Remeter Trio System comoonents)

Item 2.1 addresses components whose functioning is required to trip the reactor and requests all licensees and applicants to describe their program to assure that all such components are identified as " safety-related" in documents, procedures and information handling systems used to control safety related activities in the plant. In addition, the item requesto that a vendor. interface program be established, implemented and maintained for such components to ensure "that relevant vendor information is completa, current and controlled throughout the plant lifetime, that it is appropriately referenced or incorporated in plant instructions and procedures, and that it include periodic communication with the vendor. The licensees' submittals are to be i reviewed by the staff to determine their adequacy.

IMPLEMENTATION AND STATUS SUMMAlW!

A letter dated August 26, 1987 transmitted the staffs safety evaluation and TERs for Generic Letter 83 28 Item 2.1 Parts 1 and 2 (TAC No. 57699).

The staff and its contractors by review of the licensee's response dated  :

November 4, 1983, February 16, 1984, May 11, 1984 and November 15, 1985 concluded that a program exists for identifying, classifyin5 and treating components that are required for performance of the reactor trip function as

  • safety related. The staff also concluded that a vendor interface program exists with the NSSS vendor (CE) for components that are required for performance of the reactor trip function. Thus, LP&L has satisfied the requirements of Parts 1 and 2 to Item 2.1 of Generic Letter 83 28.

F I

5 1

l ,

I

~

o-t IMPLEMENTATION DOCUMENTS ,

IIILE BIE DAIE Nuclear Operations Procedure NOP.010 01 07/16/87 i

" Control of Vendor Information" l Nuclear Operations Support & Assessment 06 10/17/89 Procedure N0 SAP.103  ;

" Operations' Assessment & Information i Dissemination" ,

, -Nuclear Operations Support & Assessment 02 04/04/89 l L

Procedure NOSA1 203 .

" Operation of the Nuclear Network System"  ;

See Itta 2.2.1 for Implementation of Reactor Trip Safety Related Components Verification Documents  ;

i 2ER'iFICA'rION_,DQQi[HIS IIILE IBl.DD.3_EL. DAIE W3P83 3911 Section 2.1 11/04/83 i K.W. Cook (LP&L) to D.G. Eisenhut (NRC) I I

-W3P84 0396 02/16/84 K.W. Cook (LP&L) to G.W. Knighton (NRC) ,

i W3P83 1341 . 05/11/84 K.W. Cook (LP&L) to 0.W. Knighton (NRC) 5 W3P85 3158 11/15/85

  • h K.W. Cook (LP&L) to G.W. Knighton (NRC) i i

Letter J.H. Wilson (NRC) to J.G. Dewease 08/26/87  :

I 0

l'

. j L

ISSUES

SUMMARY

I l'

(MPA No. B 086) TITLE: Item 2.2.1 Eauinment Classification for D lafety Related Comoonents Item 2.2.1 addresses all other safety related components and requests all licensees and applicants to describe their program used to clasify such components. The classification program is necessary to ensure that all such l components are identified as " safety related" in documents, procedures and information handling systems used to control safety related activities in the plant, and must include periodic communication with the vendor. The staff is .

to review the licensees' submittals to determine their adequacy.

This MPA originally addressed vendor ititerface prograns for safety related components in addition to component c1Assification, as identified in GL E3-

28. The original vendor interface program guidelines were modified and superseded by way of GL 90 03 on March 20, 1990. A new MPA was established to track implementation of the revised guidelines. They are discussed separately below.

Additional

Reference:

1. NRG Lecter to All Power Reactor Licenrecs and Applicants, "Melaxation of Staff Position in Generic Letter 33 28. Item 2.2 Part 2 ' Vendor Interface for Safety Reisted Components' (Generic Letter No. 90 03)," ,

March 20, 1990.

IMPLEMENTATION AND STATUS SUHMARY:

By letter dated November 4, 1983 and November 15, 1985 the licensee responded to Generic Letter 83 28 Item 2.2.1 on programs for equipment classification for all safety related components. The staffs safety evaluation and contractor Technical Evaluation Report were provided to the licensee in a letter dated March 30, 1988. The SER and TER concluded that the licensee's response meets the staff requirements and is acceptable, therefore the action under Generic Letter 83 28 Item 2.2.1 for the Waterford SES Unit No. 3 is complete.

t l

l I

1

ei l-IMPLEMENTATION DOCUMENTS PROCEDURE REY DAIE NOP 006- 02 01/05/90 NOP 012 00 09/01/89 l:- UNT 001-002- 12 05/05/89 1 UNT 005 002 09 09/07/88 t

UNT 005-015 01 09/05/89 UNT 007 003 .06 05/15/88-t UNT 007-010 03 03/31/88 UNT 007-014 04 09/08/89 UNT 007 021 07 10/14/89

- UNT 007 022: 04 11/11/89 UNT 007 028 00 05/16/89 MD 001 022 '00 05/08/87 NOEP+001 01 06/01/89 NOEP 103 00 12/23/87 j

- N0ECP 004 00 04/03/89

- . 4 N0ECI-152 01 07/25/89  :

l PE 002 005 11 07/05/88 QAP 250 09 02/19/90 VERIFICATION DOCUMENTS IIILE NUDOCS NO. DAIE W3P83 3911 Section 2.2.1 11/04/83 K.W.-Cook:(LP&L) to D.G. Eisenhut (NRC)

W3P85-3158 Att. 1 11/15/85 K.W.' Cook (LP&L) to G.W. Knighton (NRC)

Letter D.L. Wigginton (NRC) to 03/30/88 J.G. Dewcase (LP&L) i D

, r -, --. .----, -+- - , , . . , < . ,,w~, ...,#

ISSUES

SUMMARY

(MPA No. L-003) TITLE: I*em 2.2.2 - Vendor Interface for Safetv-Related Components The original needs for vendor interface programs for safety related components were specified in CL 83 23 and licensee implementation was being tracked via MPA No. B 086, together with equipment classification. GL 90 03 was issued on March 20, 1990 which relaxes and supersedes the original vendor interface  ;

program guidance based upon industry initiatives and experience. The revised interface program with the NSSS vendor covers all safety-related components within the NSSS scope of supply and is to conform with the Vendor Equipment Technical Information Program (VETIP) as described in the Nuclear Utility Task Action Committee Report, INPO 84 010 issued in March 1984. A program of periodic contact with non NSSS vendors of other key safecy related components is also specified.

Additionni Referenees:

1. NRC Letter to All Power Reactor Licensees and Applicants, " Relaxation of Staff Position in Generic Letter 83 28, Iter 2.2 Part 2 ' Vendor Interface for Safety Related Components' (Generic Lotter No. 90 03),'

March 20, 1990.

2. INPO 84 010, " Vendor Equipment Technical Information Program," Nuclear Utility Task Action Committee, March 1984.

IMPLEMENTATION AND STATUS

SUMMARY

The licensee for Waterford SES Unit 3 responded to Item 2.2 (Part 2) of Generic Letter 83 28 by submittals dated November 4, 1985 and November 15, 1985. In these submittals the licensee described their current vendor interface programs which included implementation of the NUTAC/VETIP program.

However, the staff concluded the following in a letter to the licensee dated August 8, 1988, "The licensee's response is not complete. They need to identify and describe formal programs of interface with their NSSS, diesel generator, and safety related electrical switchgear vendors. They also should establish or confirm the existence of informal programs of contact with the vendors of other key safety related equipment."

By submittal dated October 14, 1988 the licensee responded to the staff request for additional information describing their " Key Vendor Contact Program." The licensee has recently received Generic Letter 90-03 and will provide a rasponse on or before September 30, 1990. The staffs evaluation will follow at 3 later date.

In a related item, Potential Inforcement Finding (50 382/8719 01) resulted in Violation (50 382/8823 01), " failure to process vendor information properly."

Corrective action included implementation of the licensee's " Key Vendor Contact Program." This item was closed via Inspection Report 50-382/89 41 dated March 12, 1990.

a IMPLEMENTATION DOCUMENTS PROCEDURE ILE DAIE  !

NOSAP 103 06 10/17/89 NOSAI 203 02 04/04/89 l NOECP-106 00 04/21/89 N0 SAP 201 00 03/04/89 NOAP 029 01 10/03/88 N0ECP 004 00 04/03/89 ~

NOEI 153 02 11/11/88 NOP 006 02 01/05/90 NOP 010 01 07/16/87 4 QAP-000 04 12/16/88 QAP 203 02 10/07/88 >

P QAP-252 00 05/12/89 NSP 602 01 11/30/88 N0ECI-155 00 07/31/89 VERIFICATION DOCUMENTS TITLE NUDOCS NO. DAIE W3P83 3911 Section 2.2;2 11/04/83 K.W. Cook (LP&L) to D.G. Eisenhut (NRC)

W3P85 3158 Act. 2 11/15/85 K.W. Cook (LP&L) to G.W. Knighton (NRC) .

Letter D.L. Wigginton (NRC) to J.G. Dewease 08/08/88 1

W3P88-1940 10/14/88 R.F. Burski (LP&L) to USNRC "NUTAC/ Vendor Equipment Technical Information 03/00/84 Program" INPO 84 010 .

Inspection Report 50 382/87-19, Potential 05/12/88 Enforcement Finding 8719 01 -

b s

4' 9

(

Inspect. ion Report 50 382/88 23' 09/29/88 ,

Violation 8823 01 Inspection Report 50 382/89 41 03/12/90 i

i t

.- - .,- , ,r ,, , -

i ISSUES

SUMMARY

l i

(MPA No. B 078) TITLE: Items 3.1.1 and 3.1.2 Post Maintenance Testine (Reactor Trio System Comoonents)  ;

Items 3.1.1 and 3.1.2 concern post maintenance testing procedures and. vendor recommendations for reactor trip system components. Licensees and applicants are to review their test and maintenance procedures and Technical Specifications to assure that they require post maintenance operability testing of safety related components in the reactor trip system and that such  ;

testing demonstrates that the equipment is capable of performing its safety functions prior to returning it to service. Licensees and applicants are also- !

to review applicable vendor and engineering recommendations to ensure that any-appropriate test guidance.is included in the test and maintenance procedures .

or in the Technical Specifications, where required. The results of these reviews are to be submitted for staff evaluation. >

IMPLEMENTATION AND STATUS

SUMMARY

-By submittals dated November 4, 1983 and March 5, 1984 t the licensee provided their response to Generic Letter 83 28 Items 3.1.1 and 3.1.2 Post Mr.intenance Testing (Reactor Trip System Components).

The staf f Safety Evaluation for items 3.1.1 and 3.1.2 appears in a letter to the licensee dated September 22, 1988. The SER concluded that the Waterford 3 SES response meets the review guidelines and is therefore acceptable, ,

L

+

r

F ,

i ,

s,., , ,

=4 q

i '

'l a- '

l

/ IMPLEMENTATION DOCUMENTS l 1

PROCEDURE M D&IE i MD 001 014 03 06/30/89  ;

i UNT 001 002 12 05/05/89 J

' UNT 001 003 12 12/29/89 yfJLIllCATION DOCUMENTS j t

IIIll NUDOCS NO. DAIE ,

W3P83 3911 Section'3,1  ;

, 11/04/83

- K.W. Cook (LP&L) to D.C. Eisenhut (NRC)

W3P84 0601 03/05/84 .'

K.W. Cook (LP&L) to G.W. Knighton (NRC)  !

Lettet D.L. Wigginton (NP.C) to 09/22/88 j J.G. Devease (LP&L) j i

}

v a

-t t

f i

7 l

t F

L 3

..-.._.-,.,.---.,.ra ... - - , ,_ , , ,.

r1

-)

t

. , ' i

)

1' ISSUES

SUMMARY

(MPA No. B 079) TITLE: ItJa 3.1.3 - Post Maintenance Testing Channes u Test Reauirements (Reactor Trio System Components l Item 3.1.3 requests identification of any applicable post maintenance test-requirements in existing Technical Specifications for reactor trip system l i- components which can be demonstrated to degrade rather than enhance safety. - >

1.icensees and applicants are to perform the required reviews and notify the staff '

of their findings. Appropriate changes to these test requirements, with  !

supporting justification, are to be submitted for staff approval.

l 1MPLEMENTATION AND STATUS

SUMMARY

! i By submittals dated Now;ber 4,1983 and March 5,1984, the licensee provided their response to Cenerie Letter 83 28 Item 3.1.3 Post. Maintenance Testing -

Changes to Test Requiremonto (Reactor Trip System Components).

i The staff r;afety evaluation and technical evaluation for Item 3.1.3 appears in

~

a let.ter to the licensee dated May 15, 1986. The SER and TER concluded, baled on the licensee's submittals documenting that no post maintenance test '

requireaeats were found in t.he Appendix A Technical Specifications for the

'laterford Steam Electric Station Unit 3, that could degrade rather than enhance 3 safety. tesolution of Item 3.1.3 is acceptable and now considered closed.

-m

.r,,

,r-

, , -- i-e

.c IMPLEMENTATION DOCUMENTS NONE <

Vf11FICATION DOCUMENTS IIILE MIE W3P83 3911 Section 3.1 11/04/83 K.W. Cook (LP&L) to D.C. Eisenhut (NRC)

W3P84 0601 .

.. 03/05/84 K.W. Cook (LP&L)! to C.W.LKnighton.(NRC)

Letter J.H. Wilson ~-(NRC) to 05/15/86 C.W. Muench (LP&L) 1.

. 't' k

i I

l' -l j -}

I '

, .o .*

t ISSUES SLMMARY:

(MPA No. B 087) TITLE: Items 3.2.1 and 3.2.2 Post-Maintenance Tearine (All Other Saferv Related Comoonents)

Items 3.2.1 and 3.2.2 concern post maintenance testin5 procedures and vendor recommendations for all safety related components other than the reactor trip system components. Licensees and applicants are to review their test and maintenance procedures and Technical Specifications to assure that they require post maintenance operability testing of all safety related components (non reactor trip system components) and that such testing demonstrates that -

the equipment is capable of performing its safety functions prior to returning it to service. Licensees and applicants are also to review applicable vendor and engineering recommendations to ensure that any appropriate guidance is included in the test and maintenance procedures or in the Technical Specifications, where required. The results of these reviews are to be submitted for staff evaluation.

IMPLEMENTATION AND STATUS

SUMMARY

By submittal letters dated November 4, 1983 and November 15, 1985 the licensee responded to Generic Letter 83 28 Items 3.2.1 st.d 3.2.2 Post Maintenance Testing (All Other Safety Related Components).

-The staf issued Safety Evaluation Report dated September 22, 1988 concluded that post maintenance testing of all safety related components at Waterford Steam Electric Station Unit 3 meets the review guidelines and is acceptable.

i I

l

-(

t..

7 e

IMPLIMENTATION DOCUMENTS I l

PROCEDURE E E  :

t

, MD 001 014 03 06/30/89 [

UNT 001 002 12 05/05/89- -!

UNT 001 003' 12 12/29/89 i VERIFICATION DOCUMENT $ t

!I IIIM; ElDOCS NO. M  !

.i W3P83 3911 Section.3.2. 11/04/83 i K.W. Cook (LP&L) to D.G. Eisenhut (NRC) i l

W3P85 3158 Attachment 3 11/15/85 l K.W. Cook (LP&L)'to G.W. Knighton (NRC) i h

Letter.D.L. Wigginton (NRC) to. 09/22/88  :

J.G. Dewease (LP6L) I f

f

.h

'i

{

j i

f

<q

S

?

=.

-i 3

e f

1 j

4, J

a o  :

1 L  !

ISSUES

SUMMARY

! TITLE:

(MPA No. B 088) Item 3.2.3 Post-Maintenance Testine - chaneen r to Test Reauirements (All Other Safety.Related Comoonents) i Item 3.2.3 requests identification of any applicable post maintenance test i c requirements in existing Technical Specifications for safety related components  !

which can be demonstrated to degrade rather than enhance safety. Licensees and l applicants are to perform the required reviews and notify the staff of their j findings. Appropriate changes to these test requirements, with supporting l justification, are to be submitted for staff approval. j IMPLEMENTATION AND STATUS

SUMMARY

By submittal letter dated November 15, 1985 the licensee responded to Generic Letter Item 3.2.3 Post Maintenance Testing Changes to Test Requirements (All Other Safety Related Components).  !

The staff issued safety evaluation and technical evaluation dated May 15, 1986

  • concluded, (based on the licensee's statement that no Technical Specifications for post. maintenance testing at Waterford 3 SES were found to degrade rather than enhance safety) that the licensee's response is acceptable and Item 3.2.2 is now ,

considered closed. l

-i 1

I  !

m cgp!r -

-i

m. ] sk f.;'

T sj  :

-c- (

1 Qa ,t' -

1 l

'l IMP 12 MENTATION DOCUMENTS j 1

NONE-  ;

p ,

a-m

)

' 8 s

- i

,p

. VERIFICATION DOCUMENTS ,

3 y b  :

s gl , , .

J

..IIILE NUDOCS NO. jgIE j

n. ' '

1- -

W3P85 3158 Attachment 3 11/15/85- )

3; % K.W. Cook (LP&L) to G.W. Knighton (NRC) m.

Y Letter J.H. Wilson (NRC) to 05/15/86 i ,

G.W..Muench (LP&L) {

w .

,2 ,

1 J I' (

f

.' d -

i h

I

'G'i j

i ,iI!

- fI i rj f

3; ,

y lG' e g: , 1 4

i 4

.p.%

7 t

. 4

?

5

I e

'5 N ', f

?

f -

t

?

'I\'

s i

I

?

6 i j -.  !

4

, , ,,,-, <, . , . , . . - , , , .- . + . . . , ,.m....or., _.,, . , a ... . - . . , . . .,.__,.,..~,y_ . < , -

h 4

ISSUES SUHMARY!

(MPA No. B 080) TITLE: Item 4.1 Reactor Trio System Reliability (Vendor Related Modifications)

The resolution of this item, applicable to all plants, requests that each licensee review all vendor recommended reactor trip breaker modifications to verify that either: (1) each modification has been implemented, or (2) a written evaluation of the technical reasons for not implementing a modification exists. Submittals were to be made by all licensees / applicants.

For those plants that were licensed at the time, the submittals were to be reviewed by the cognizant regions and Safety Evaluations were issued by NRR.

For plants licensed since 1983, Item 4.1 was to be included as part of the licensing review and the results reported in the licensing SER or in one of the supplements.

IMPLEMENTATION AND STATUS

SUMMARY

By submittal letters dated November 11, 1983, February 16, 1904 and Nover;ber JO,15BA che licar,see responded to Generic Letter 83 28, Item 4.1, Reactor Trip System Reliability (Vendot +Related Modifications).

The staff issued safety evaluation dated October 20, 1987 concluded that the-Waterford 3 SES position on Item 4.1 meets the staff requirements and is acceptable.

The staff inspectors performed a followup inspection on Generic Letter 83 28 Item 4.1 in accordance with Temporary. Instruction 2515/91 (IR 50 382/P9 01, Section 3.3.3 dated March 21, 1989).

The NRC inspectors found that Waterford 3 uses General Electric AK.2 25 type reactor trip breakers and that the General Electric Company (GE) has not j issued any modification recommendations. However, GE has noted that the bearing grease may start to solidify after about 7 years, possibly affecting breaker response time. The licensee returns the reactor trip breakers to CE for refurbishment on a 5 year interval. The preventive maintenance program checks breaker response times and these times are in the trending program.

1 l

,y.- - g y .-- .,

v.  :<

.i:

h~ ,'~.

IMPLEMENTATION DOCUIENTS Task Card Number 016329 016330

[ 016331

-016332-016334 016335-016336-016337

'016333-

_ VERIFICATION DOCUMENTS IIILE DAIE W3P83 3911.'Section 4 1'.

11/04/83 K.W. Cook.-(LP&L) to D.0; Eisenhut (NRC)

W3P84 0396 02/16/84

.K.W. Cook (LP&L) to C.W. Knighton (NRC)

W3764 3344 11/30/84 <

K.W. Cook (LP&L) to G.W. Knighton (NRC)

'l Lotter 10/20/87 .I

' J.H. Wilson (NRC)' to J .G. Dewease (LP&L)

Inspection Report 50 382/89 01,.Section 3.3.3 03/21/89 l

u

.I 1 i

l l

.1

.- i ISSUES

SUMMARY

(MPA No. B 081) TITLE: Items 4.2.1 and 4.2.2 - Reactor Trio Systea Reliability - Maintenance and Testine (Preventive Maintenance and Surveillance Program for Reactor Trio Breakers)

Item 4.2.1 addresses development of a planned program of periodic maintenance, including lubrication, housekeeping and other items recommended by the equipment suppliers. Item 4.2.2 addresses development and implementation of a program for trending of parameters which affect breaker operation and are measured durin5 testing in order to predict performance degradation. All PVR licensees and applicants were to provide descriptions of their programs for staff review.

IMPLEMENTATION AND STATUS

SUMMARY

By submittal letters dated November 4, 1983, March 8, 1984, November 30, 1984 and September 2, 1987 the licensee provided their response to Generic Letter 83+28 Items 4.2.1 and 4.2.2 Reactor Trip System Reliability + Maintenance and Testing (Preventive Maintenance and Surveillance Program for Reactor Trip Breakers).

The staif safety evaluation dated October 20, 1987 concluded the following for the Waterford 3 SES facility.

"We find that the licensee's extended maintauance interval (performed annually) meets the otaff requiremants ano is acceptable. This acceptance is based on GE's recommendation that maintensnce on RTBs located in mild environments shov.id be performed annually. The vendor recommendation tnat RTBs located in harsh environments or experiencing severe load conditions be maintained more frequently is not applicable to these RTBs because of-their location in a mild environment and reduced service duty at Waterford 3. We

-find the licensee's position on Item 4.2.1 meets the staff requirements and is acceptable.

The licensee will trend undervoltage trip attachment dropout voltage, trip force, and breaker response time. They do not trend breaker insulation resistance. The licensee will perform any appropriate preventive or corrective maintenance if the analysis of the trend data indicates need for such action. We find that the licensee's position on Item 4.2.2 is acceptable."

s ,  ;;

IMPLEMENTATION DOCUMENTS q TITLE REY DAIE OP-903 006 03 05/02/89 J ME 004 155 08 04/12/90 j TASK CARD NUMBERS 010829 .

010836 010819 010823 -

-010826  ;

010828 010831 010833 010834 l

IE Bulletiin 79 09 - Failures of GE Type AK 2

-Circuit Breakers.in Safety Related Systems ..

.[

Availability Data Program Infor Bulletin 81 02 '

and Supplement 1 dated November 5, 1981 3 e  :}

IE Information Notivo 85 58 07/17/85 y LGeneral Electric Service Advise Letter (SAL) i

175 dated April.2, 1979 and Supplement dated April 15, 1983 j CE ADP.Information Bulletin 83 07 06/15/83' l

'{gg1FICATION' DOCUMENTS

.d

6 TITLE NUDOCS No. DAIE .v. A W3P83 3911 Section 4.2 11/04/83-K.W... Cook (LP&L) to D.G. Eisenhut (NRC)  !

W3P84 0609 . 03/08/84 K.W. Cook (LP&L) to G.W. Knighton (NRC) .

.W3P84 3344 11/30/84 4 K.W. Cook (LP&L) to G.W. Knighton (NRC) ,

i W3P87 2052 09/02/87 K.W.' Cook (LP&L) to USNRC Letter J.H. Wilson -(NRC) to J.G. Dewease (LP&L) 10/20/87 p

's I

1 P

f

ISSUES

SUMMARY

(MPA No. B 092) TITLE: Item 4.5.1 - Reactor Trio System Reliability -

Diverse Trio Fectures (System Functional Testine)

Item 4.5.1 requests that licensees perform on line functional testing of the reactor trip system, including independent testing of the diverse trip features. The diverse trip features to be tested include the breaker undervoltage and shunt trip features on Westinghouse, B&W and CE plants; the circuitry used for power interruption with the silicon controlled rectifiers on B&W plants; and the scram pilot valves and backup scram valves (including all initiating circuitry) on CE plants.

Licensees were requested to confirm that the required on line functional surveillance testing is being performed for the diverse trip features of the plant.

Some licensees do not test backup scram valves on line, because such testing would result in a reactor scram. In such cases the NRC allows scram valves to be tested during each refueling outage to avoid unnecessary reactor scrams and challenges to the reactor protection system. Conformance with this item is verified by follow up inspections.

IMPLEMENTATION AND STATUS S'.NMARY:

By submittal letters dated Nove9her 4,1983 and February 6,1984 the licetsae l respondes to Generic Letter 83 26 Item 4.5.1 Reactor Trip System Rolf ebility - '

Diverse Trip Features (System Functional Testing). 1 The staff concluded the following in its safety evaluation dated September 28, 1988.

"The licensee stated that on line functional testing of the reactor protection syv em will be performed usfng operating procedure op.903 006, " Manual Reactor Trip Test." We have reviewed this procedure as part of the plant records and the tests of the undervoltage and shunt features for all eight reactor trip breakers are included. The tests are performed in accordance with the procedures and as required by the Technical Specifications every 18 months.

Based on our review, we conclude that the on line reliability verification testing of the reactor protection system by Waterford 3 meets the review guidelines and is acceptable."

I t

'('

. .O  !

,- .e  ;

IMPLEMENTATION DOCUMENTS 1

PROCEDURE E E' I OP 903 006 03 05/02/89  :

VERIFICATION DOCUMENTS IIILE NUDOCS NO. D&IE W3P83 3911, Section 4.5.1 11/04/83-  ;

K.W. Cook (LP&L)1to D.G. Eisenhut (NRC)  !

l W3P84 0299 02/06/84 I K.V. Cook'(LP&L) to G.W. Knighton (NRC) l Letter D.L. Wigginton;(NRC) to 09/28/88 J.G. Dewease (LP&L) l L

h

't i

p .!

l

[

.i t

I

-i r

i

.I f

b u

< N$ - - - , - - , . a - . , , - -u ,_ - - , --.-+ we-, - - ~.~-n, ..,,,,-----n.-,--- - - - , -

, ,aw . - , .

4 ISSUES SlMMARY:

(MPA No. B 093) TITLE: Items 4.5.2 & 4.5.3 - Reactor Trio System Rellability - Test Alternatives and Yntervals ff,y,gtem Functional Testine)

Item 4.5.2 requests licensees and applicants to certify whether their plants are designed to permit on line functional testing of the reactor trip system (RTS). For plants not designed to permit such testing, licensees are requested to commit to design modifications which would permit such testing and provide an implementation schedule, or to provide justification for not implementing on line testing capability. The staff will consider alternatives to on line testing where special circumstances exist and where the objective of high reliability can be met by other means.

1 Item 4.5.3 requests licensees and applicants to confirm that on line '

functional testing of the RTS is being performed and that existing test  !

intervals required by their Technical Specifications are adequate for I achieving high RTS reliability. All four vendors submitted topical reports 4 which presented analyses demonstrating that current test intervals provide {

high reliability. Based on staff review of the Owner's Group topical reports, '

the contractors' indesendent analyses, and the generic safety evaluation findings in NUREG 0460, the staff concluded that the existing intervals, as I recommended in the topical reports, for on line functional testing are I consistent with achieving high RTC availability at (1; operating reactors.

Licensees and applict.nts are to Jubmit a der.criptiur. of how they ate irrplementing the provisions of their Owner's Group topical report.

Additional Refennees: .

1. Topical Report WCAP 10271, " Evaluation af the Surveillance Frequencies and Out of Service Tices for the Rewtes Protection Systeins," 1985,
2. Topical Report WCAP 10271, Supplement 1.
3. NECD 30844, "BWR Owner's Group Response to NRC Generic Letter 63 28, Item 4.5.3," January 1985.

4, NECD 30851P, " Technical Specification Improvement Analyses for BWR Reactor Protection System," May 1985.

5. CE NPSD 277, " Reactor Protection System Test Interval Evaluation, Task 486," December 1984.
6. BAW-10167, " Justification for Increasing the Reactor Trip System On Line Test Interval," May 1986.
7. BAW 10167, Supplement 1, February 1988.

l IMPLEMENTATION AND STATUS

SUMMARY

By submittal letters dated November 4, 1983 and August 12, 1987 the licensee responded to Generic Letter 83 28 Item 4.5.2.

Acceptance of Waterford 3 SES response to Item 4.5.2 appears in the staff TER enclosure to letter dated August 26, 1987.

By submittal letters dated November 4, 1983, March 8, 1984, December 7, 1984 and January 30, 1985 the licensee responded to Generic Letter 83 28, Item 4.5.3.

The staff issued safety evaluation and technical evaluation provided to the licensee as enclosures to letter dated May 24, 1989 concluded the following:

"On the basis that Waterford 3 data and information is consistent with CE Owners Group reports, we find your submittals acceptable. This completes the LP&L efforts under item 4.5.3."

IMPLEMENTATION DOCUMENTS IIILE DAIE Combustion Engineering, "RP/EFAS Extended i Task Interval Evaluation" CEN 327 May 1986 VERIFICATION DOCU @lTIS IIILL NUDOCS NO. DAIE L

W3P83 3911, Sections 4.5.2 and 4.5.3 11/04/83 K.W. Cook (LP&L) r,c D.G. Eisenhut (NRC) 93P84 0610 G3/08/84

/. . W. Cook (LP&L) to G.W. k'nighton (NRC)

W3P84 3381 12/07/84 l K.W. Cook (LP&L) to G.W. Knighton (NRC)

W3P85 0245 01/30/85 K.W. Cook (LP&L) to G.W. Knighton (NRC)

W3P87 1136 08/12/87 K.W. Cook (LP&L) to USNRC Letter J .H. Wilson (NRC) to J.G. Dewease (LP&L) 08/26/87 Letter F.J. Hebdon (NRC) to J.G. Dewease (LP&L) 05/24/89 i

i i-ISSUES SUHMA3X; GSI-No. 11 (MPA No -B 098) TITLE: Steam Binding of Auxiliary Feedwater Pumos W

Thu issue concerns the potential disabling of auxiliary feedwater pumps by j steam binding caused by back-leakage of main feedwater past the isolation i check valves. IE Bulletin 85-01, issued October 29, 1985, requested that j certain licensees implement procedures for monitoring the auxiliary.feedwater  ;

piping temperatures for indications of possible back leakage and for restoring  !

the pumps to operable status if steam binding were to occur. I Generic Letter 88 03, issued February 17, 1988, stated that the plants that  :

received Bulletin 85 01 should continue following the Bulletin's i recommendations, and requested that these recommendations be followed on all  !

PWR's.

I References; i

1. IE Bulletin 85 01, " Steam Binding of Auxiliary Feedwater Pumps," U.S. ,

Nuclear Regulatory Commission, October 29, 1985. -j

2. NRC Letters to All Licensees, Applicants for Operating Licenses, and Holders of Constructions Permits for Pressurized Water Reactors,

" Resolution of Leneric Safety Issue 95, ' Steam Binding of Auxiliary Feedwater Pumps' (Generic Letter 88 03)," February 17, 1988

' IMPLEMENTATION AND STATUS

SUMMARY

1 i

By submittal dated February 26, 1986 the licensee for the Waterford 3 SES Unit 3 facility responded to IE Bulletin 85-01 coneerning generic problems involving the inoperability of auxiliary feedeter pumps as a result of steam  !

binding. (NOTE: The system described in IEB . -01 is called the Emergency Feedwater System (EFW) at the licensee's facility.)

During a subsequent NRC inspection conducted on June 1-30, 1986, a deviation '

of the licensee's commitments in response to IEB 85 01 was identified. The licensee responded to Deviation (50 382/8613 03) by letter dated August 22, 1986. The NRC closed the Deviation via Inspection Report 50 382/86-16. NRC Inspection Report 50-382/86-17 closed IEB 85 01 stating that the licensee's program meets the rec;uirements of the bulletin and Temporary Instruction 2515/69.

By letter dated April 5, 1988 the licensee responded to Generic Letter 88-03.

The staff concluded in letter dated June 20, 1988 that the licensee's response waa satisfactory and no further action was necessary.

L

i o- .

l IMPLEMENTATION DOCUMENTS Il.ILE ~ BJN DATI 01 004 000 12 09/13/88' OP 903 001 11 03/23/90 3 Station Modification No. 531 09/21/85 'I and CIWA 020497-VERIFICATION DOCUMENTS  ;

TITLE NUDOCS NO. DATE W3P86-0036 02/26/86' i K.W. Cook (LP&L) to R.'D. Martin (NRC) (

Inspection Report 50-382/86 13, 07/24/86 Deviation 8613-03  ;

W3P86-1949 08/22/86 K.W. Cook (LP&L) to R.D. Martin (NRC) ~f Inspection Report 50-382/86 16 10/16/86 Inspection Report 50 382/86 17 10/24/86 W3P88-0049= 04/05/88 l- _R.F. Burski (LP&L) to USNRC. l Letter D.L. Wigginton (NRC) to J.G. Dewease (LP&L) 06/20/88 L 1 i

f

\i

't m.

I e

p i

ISSUES

SUMMARY

t GSI No. 22 (HPA No. L 817) TITLE: RCS/RHR Suction Line Valve Interlock on PWRs ,

This issue concerns the inadvertent closing of RHR suction valves when the RHR system is in use.

I Interlocks are provided on these valves to ensure that a double barrier (i.e.,

two closed valves) is maintained between the RCS and RHR systems when the plant,is at normal operating conditions. However, the loss of one instrument -l bus or disturbance of one logic channel will result in the automatic closure of one of the RHR suction line isolation valves. Such closure gives rise to the potential for RHR pump damage and loss of decay heat removal capability if j the RHR pump is not interlocked with the RHR suction valves, j i

The scope of this issue was broadened in June 1986 to include the less i frequent but higher risk mode of failure associated with mid loop operation.

Generic Letter 87 12 addressed this concern.  !

Generic Letter 88 17 superseded GL 87-12 and requested responses regarding

licensee plans with respect to operation on shutdown cooling. This letter requested expeditious licensee actions in the areas of: (1) training of operators before entering a reduced inventory condition, (2) implementation of procedures and administrative controls related to decay heat removal, (3) _

temperature and level indications, and (4) alternate means of adding water to the RCS. Further, GL 8817 identified a number of programmed enhancements to l be developed-in the following six areas: (1) instrumentation .(2) procedures, 3 (3) equipment, (4) analyses, (5) Technical Specifications, and (6) RCS s.

perturbations. l

References:

1

1. NRC Letter to All Licensees of Operating PWRs and Holders of l Construction Permits for PWRs, " Loss of Residual Heat Removal (RHR) }

While the Reactor Coolant System (RCS) is Partially Filled (Generic i Letter 87 12)," July 9,1987.

2. NRC Letter to All Holders of Operating Licenses or Construction Permits for Pressurized Water Reactors (PWRs), " Loss of Decay Heat Removal (Generic Letter No. 88 17), 10 CFR 50.54(f)," October 17, 1988.
3. NUREG/CR 5015, " Improved Reliability of Residual Heat Removal Capability in PWRs as Related to Resolution of Generic Issue 99," U.S. Nuclear Regulatory Commission, Hay 1988.

M

i

-IMPLEMENTATION AND STATUS

SUMMARY

By submittal dated December 23, 1988 as supplemented by submittal dated February 1,1989, the licensee- for 'iaterford 3 SES responded to Generic Letter 88 17.

The staff noted several observations concerning the licensee's response dated December 23, 1988 in letter dated March 8, 1989 and expressed the staffs-intent to audit the expeditious actions and programmed enhancements taken by

~

the licensee at a 1ater date.

Inspection Report 50-382/89 13 dated June 15, 1989 documents the NRC's inspection of the licensee's expeditious actions and the provisions taken with j regard to observations in the NRC letter dated March 8,1989, IR 89-13 made  ;

several observations but concluded that the licensee's expeditious actions j appear responsive to Generic Letter 88 17. Also stated was the NRC's intent 1 to inspect the licensee's long range programmed enhancements during a i subsequent inspection,

'l l

The licensee has completed all actions required by Generic Letter 8817 with j the exception of the following: (1) Technical Specification change to remove  !

auto closure interlock on shutdown cooling suction valves. Scheduled for  !

completion during the March 1991 refueling outage. (2) Install the new SDC l 1evel indication system (see submittal dated May 21, 1990), Scheduled for I completion during the March 1991 refueling outage.  ;

i I

IMPLEMENTATION DOCUMENTS f i

TITLE BEE DAIE _

OP-001 003. 09 09/23/89 j OP-901 046 05 07/10/89 )

OP-903-072 05 09/23/89 OP 009-008 08 09/19/89

.0P-009-005 10 03/19/90 RF 003 002 01 09/08/89  !

MM-008 001 03 9/18/89 MD 001 026 02 09/15/89 i RF-003-001 01 09/18/89 ,

LESSON PIAN NO. M011-881-01,2  !

LESSON PLAN NO, A010 000-00 3

df, a;>

a 1

< ,4

?

4 } 1.4
;;j;.e5 5L M , VERIFICATION DOCUMENTS IS IIILE NUDOCS No. DAIE' i ,

W3P88-3091 12/23/88-R.F. Burski~(LP&L) to.USNRC W3P89 0101 .

02/01/89 R.F. Burski (LP&L) to USNRC Letter 03/08/89 J.A. Calvo'(NRC) to J.G. Dewease (LP&L)

Inspection Report 50-382/89-13 06/15/89 Letter D.L. Wigginton (NRC) to LP&L 05/18/90 W3P90 1314 05/21/90 R.F. Burski (LP&L) to USNRC

'a ,'

'i l

j a'

. l

)'

, > 1 9,4 1

l

,1 l

1 4 1

i 1 1

L

,?' e i

ISSUES

SUMMARY

GSI No. A:11 (MPA No. B 017) TITLE: Snubber Ooerability Assurance - l Hydraulic Snubbers  !

This_ issue concerns operability of hydraulic snubbers which is required to

~

.f assure that the structural integrity of the reactor coolant system is maintained during and following a seismic or other event initiating dynamic loads. Operating experience in the 1970's indicated the need for changes,  ;

clarifications and improvements in snubber Technical Specifications. These  !

changes provided for: (1) precluding use of an arbitrary snubber capacity as a limit for inservice test requirements, (2) elimination of the requirement that seal material be approved by NRC, (3) implementation of a monitoring program to assure anubber reliability, (4) development and implementation of clearly defined inservice test requirements, and (5) permissible in place inservice testing.  !

By letter dated November 20, 1980, the NRC requested that all power reactor  !

licensees (except Systematic Evaluation Program (SEP) licensees) incorporate .l the above changes in plant specific Technical Specifications. A similar request was sent to SEP licensees on March 23, 1981. Also, revisions to the Standard Technical Specifications (W, GE, CE and BW)' incorporated the >

appropriate Technical Specifications to address these changes for NTOLs.

References:

1. NRC Letter to All Power Reactor Licensees (Except SEP Licensees),

" Technical Specification Revisions for Snubber Surveillance," November 20, 1980,

2. NRC. Letter to All SEP Power Reactor Licensees, (Except SEP Licensees).

" Technical Specification Revisions for Snubber Surveillance," March 23, 1981.

IMPLEMENTATION AND STATUS

SUMMARY

The Standard Technical Specification for CE had been revised to include the appropriate Technical Specifications for Hydraulic Snubbers Operability Assurance.

The licensee reviewed Generic Letter 84-13 which revised the snubber technical specification attached to Generic Letter issued 11/20/80 and determined that their " Draft" technical specification had already been changed to the format described in Generic Letter 84-13. l The licensee's Technical Specification issued 03/16/85 included the l I

appropriate requirements.

l o

l l

l l

w_-

I. * 'S 4 '(

j IMPLEMENTATION DOCUMENTS PROCEDURE .EEY- DAIE  ;

TS 3.7.8 12/27/85 TS 4.7.8 03/16/85 PE 005 014- 05 11/18/88 j 11-PE 005 012 03 12/19/87  :

ISIC&S 1.10 03 10/02/87 :. ?

~!

VERTFICATION DOCUMENTS  !

TITLE NUDOCS NO. DAIE -r Generic Letter 84 13 " Technical 05/03/84 Specification for Snubbers"  ;

LetterL to File W3P84-1622 06/08/84 -t

.t

't

.(

-t r

-I 1

i

'[- ,

o b

IL

,i

(

'2 L

h

, jS

j t* I

, ,,. s ISSUES

SUMMARY

GSI No. 6-11 (MPA No. B 022) TITLE: Snubber Ooerability Assurance -

Mechanical Snubbgra This aspect of the issue addresses mechanical snubbers. In the mid 1970's, several deficiencies were noted in the Technical Specifications for assuring snubber reliability. Also, mechanical snubbers were not included in the Technical Specifications surveillance requirements. Many licensees used mechanical snubbers as original equipment and others requested to replace their hydraulic snubbers with mechanical ones to simplify or avoid inservice '

surveillance. The most likely failure for an unsurveilled mechanical snubber is permanent lock up which can be harmful to plant systems during normal ,

operations and during seismic events initiating dynamic loads. Therefore, changes were needed which would: (1) include mechanical snubbers in the surveillance program. (2) preclude use of an arbitrary snubber .apacity as a limit for inservice test requirements, (2) implement a monitoring program to assure snubber reliability, (4) develop and implement clearly defined test requirements, and (5) permit in place inservice testing.

By letter dated November 20, 1980, the NRC requested that all power reactor licensees (except Systematic Evaluation Program (SEP) licensees) incorporate the above changes in plant specific Technical Specifications. A similar  ;

request was sent to SEP licensees on March 23, 1981. Also, revisions to the i Standard Technical Specifications (W, GE, CE and BW) incorporated'the appropriate Technical Specifications to address these changes for NT014, i Re ference s : 7

1. NRC Letter to All Power Reactor Licensees (Except SEP Licensees),

" Technical Specification Revisions for Snubber Surveillance,"-November 20, 1980.

2. NRC Letter to All SEP Power Reactor Licenseet, "Tochnical Specification ,

Revisions for Snubber Surveillance," March 23, 1981. l IMPLEMENTATION AND STATUS

SUMMARY

The Standard Technical Specification for CE had been revised to include the appropriate Technical Specification for Mechanical Snubbers Operability-Assurance.

The licensee reviewed Generic Letter 84 13 which revised the snubber technical specification attached to Generic Letter issued 11/20/80 and determined that their " Draft" Technical Specification had already been changed to the format

. described in Generic Letter 84 13.

The licensee's Technical Specification issued 03/16/85 included the appropriate requirements.

t,

.. (

_4'.

o .

V.. :

-IMPLEMENTATION DOCUMENTS PROCEDURE ggy DAIG TS. 3.7,8 12/27/85 TS 4,7,8 03/16/85 PE-005 014 05 11/18/88 MM 007 011 03 11/17/86 MM 006 023 03 05/02/88 ISIC&S 1.10 03 10/02/87-i

' VERIFICATION DOCUMENTS TITLE NUDOCS NO- DAIE j Generic Letter 84-13'" Technical 05/03/84 Specification for Snubbers" i

Letter to File W3P84-1622

~

06/08/84 i

l.

i

'l 1

t 1

(

)

l..

se t 1

ISSUES

SUMMARY

GSI No. 6 11 (MPA No. B 023) TITLE: Adecuacy of Offsite Power Systems This issue arose from a July 1976 degraded grid voltage condition which occurred at Millstone 2 and which resulted in blown fuses in certain -

engineered safety feature equipment. As a result, the staff determined that a potential existed for supplying both safety and non safety equipment with voltages outside the design range, which could render the equipment inoperable.

Letters were sent to licensees in June 1977 which requested installation of degraded voltage relays designed to separate the safety buses from offsite power whenever the degraded voltage condition existed for more than about 10 seconds. Licensees were also requested to propose Technical Specifications with LCOs and surveillance requirements for these relays and associated instrumentation. Some licensees chose instead to institute procedures for manual-actions in the event of these degraded voltage conditions. The Regions reviewed these procedures and eventually found them to be acceptable.

Some licensees resolved this issue in conjunction with MPA B 048, " Adequacy of Station Distribution Voltage," which was initiated by the letter to all power reactor licensees (except Humboldt Bay) on August 8, 1979. MPA B-048 requested licensees to reanalyze their plants to ensure that safety related equipment was not subjected to voltages outside design limitations when the

- grid voltage was at its maximum and minimum levels. Af ter performing these analyses, licensees were then to perform a test to measure station voltages at various places in the plant to verify the accuracy of the calculations. As a result of this review, many licensees made tap changes to transformers to optimize station distribution voltages. These tap changes often affected MPA-B023 calculations and caused changes to the undervoltage relay setpoints.

The changes imposed by resolution of this issue were incorporated into licensing reviews after 1977 through Branch Technical Position PSB-1 and, subsequently,=a 1981 revision to SRP 8.3.1, Appendix A.

References!

1. NRC Letter to Northeast Nuclear Energy Company, " Millstone Nuclear Power Station. Unit Nos. 1 and 2," June 2, 1977,
2. .NRC Letter to All Power Reactor Licensees (Except Humboldt Bay),

" Adequacy of Station Electric Distribution Systems Voltages," August 8, i 1979.

3. Branch Technical Position PSB-1, " Adequacy of Station Electric Distribution Voltages," July 1981, I l

l l

l l

l l

l l

l

.i l

..(, J

, e~

IMPLEMENTATION AND STATUS

SUMMARY

In SER NUREG 0787, July 1981 section 8.2.4 the staff evaluated the Waterford 3 SES design to-the' criteria of BTP PSB 1 and concluded based on its review that ,

the offsite= power system for Waterford 3 was acceptable.

~

In a follow up inspection (Inspection Report 50 382/85 07) the NRC reviewed I the required analytical techniques and assumptions used by the licensee to optimize the maximum and minimum load conditions expected throughout the anticipated range of voltage variation of the electrical AC distribution. 1 system,-as required by BTP PSB 1 section 4 No violations or deviations were- l identified, i

IMPLEMENTATION DOCUMENTS TITLE El2d D6IE FSAR Subsection 8,2 12/18/86 i

VERIFICATION DOCUMENTS IITLE NUDOCS NO. DAIE ,

SER NUREG 0787 Section 8,2,4 07/00/81 Inspection Report 50-382/85-07 04/22/85 i

1 l

1

U' 7 J

,'.o o-

, 0 .. , -

ISSUES

SUMMARY

GSI No. B:11 (MPA No. B-045) TITLE: Isolation of Low Pressure Systems Connected to the Reactor Coolant System Pressure Boundary This issue resulted from staff concerns regarding the potential failure of ,

valves comprising the pressure isolation barrier between the reactor coolant >

system (RCS) and interfacing low-pressure systems. Such a failure could result in overpressurization and attendant rupture of low pressure piping and/or components, with a loss of coolant outside containment. The Reactor Safety Study (WASH 1400) identified the intersystem loss-of-coolant accident (ISLOCA) in PWRs as a significant contributor to risk from core melt. The study focused on two specific pressure isolation configurations consisting of two in series check valves, with or without an open motor operated valve in series. This accident scenario was designated as Event V.

Concerns regarding Event V, as well as the staff's position that valve closure integrity could be improved by testing, led to the issuance of a Ger.eric Letter entitled " LWR Primary Coolant System Pressure Isolation Valves," dated February.23, 1980, which requested a response from all licensees specifying whether their facilities contained the Event V configurations.

For the 34 facilities (32 PWRs, 2 BWRs) responding affirmatively, orders were issued on April 20, 1981 imposing certain corrective actions, including implementation of periodic testing of the identified Event V pressure isolation valves (PIVs) and Technical Specifications addressing surveillance

-and limiting conditions of operacion for these PIVs.

References:

1. WASH-1400 (NUREG 75/014), " Reactor Safety Study An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," U.S. Nuclear Regulatory Commission, October 1975.
2. NRC Letter to All LWR Licensees, " LWR Primary. Coolant System Pressure Isolation Valves," February 23, 1980.

IMPLEMENTATION AND STATUS

SUMMARY

'This concern is addressed by the licensee's Technical Specification. PIVs separating the RCS from attached low pressure systems are listed in Table 3.4-1 w.ith the maximum permissible leakage rates. The test frequency is ,

specified in TS 4.4.5.2.3.

In response to NRC FSAR Question 211.67 the licensee evaluated the RCS Boundry Valve concern in accordance with NUREG 0677, "The Probability of Intersystem LOCA: Impact,Due to Leak Testing and Operational Changes."

In SER .NUREG 0787 Section 3.9.6 dated July 1981 the staff concluded that LP&L's commitmants to periodic leak testing of pressure isolation valves between the RCS and low pressure systems will provide reasonable assurance that the design pressure of the low pressure system will not be exceeded, and thus reduce the probability of an occurrence of an intersystem LOCA.

---( _.

.,; ,, 0;

,. ,,w .

r., t g

f :' l IMPLEMENTATION DOCUMENTS 1 IIIl4 EEY DATI TS TBL.3.4.1' 03/16/85' TS 4.4.5.2.2 03/16/85 FSAR 5.2.5 12/18/86 FSAR=- TBL.S.2.11 12/18/86 OP-903-008 04 .05/20/88 l

!- VERIFICATION DOCUMEN'd i

TITLE NUDOCS NO. DAIE j t

. FSAR Question 211.6 ' 11/00/80 ,;

Amendment 13 NUREG 0787 SER SectioT 3.9.6 _07/00/81 Generic Letter'87-06 " Periodic Verification -03/13/87- i~

of Leak Tight Integrity of Pressure

-Isolation Valves" i 1 .'

[

Licensee submittal to GL 87-06 ~06/11/87  ;

Letter 1K.W. Cook (LP&L) to USNRC-'

i NUREG 0677 05/01/80' 1 ,; .

l 1

l.

h

!-n

{

l' o

?

i 1

1 1

. . _ _ _ _ - _ _ - _ _ _ _ _ - _ . __-_-__

Retrieved from "https://kanterella.com/w/index.php?title=ML20044A554&oldid=2561141"