ML20217K831

Summary of 970924 Meeting W/Wcnoc in Rockville,Md to Discuss Results of Evaluations & Proposed Actions in Response to GL 96-06, Assurance of Equipment Operability & Ci During DBA Conditions. Handout & Attendance List Encl

ML20217K831
Person / Time
Site:	Wolf Creek
Issue date:	10/15/1997
From:	Thomas K NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
GL-96-06, GL-96-6, TAC-M96887, NUDOCS 9710290265
Download: ML20217K831 (20)

Text

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I gt p" *i UNITED STATES g- } NUCLEAR REGULATORY COMMISSION o WASHINGTON, D.C. 306 4-0001

%q October 15, 1997 LICENSEE: Wolf Creek Nuclear Operating Corporation FACILITY: Wolf Creek Generating Station i

SUBJECT:

RESPONSE TO GENERIC LETTER 96-06, " ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS" (TAC NO. M96887)

Backaround; On September 24, 1997, members of the NRC staff met with representatives from Wolf Creek Nuclear Operating Corporation (the licensee) to discuss the results of their evaluations and proposed actions in response to Generic Letter 96 06.

This was a followup to the meeting held on June 25. 1997. The meeting was i held in the NRC' offices in Rockville, Maryland.

Summary:

The focus of this meeting was the licensee's activities in resolving the thermally induced overpressurization of isolated pipe inside containment issue. The licensee identified 33 sections of pipe that may be susceptible to thermal overpressurization, In determining the stress allowables in these isolated piping sections, the licensee determined that the Wolf Creek licensing basis included the use of the ASME Code Section Ill, Appendix F.

with an allowable stress for faulted conditions fer Class 2 and 3 piping to be 2.4 S3 .

Of the 33 sections evaluated, 9 locations were eliminated 4 because of configuration and 5 by changing the operating procedures associated with these tsystems. For the remaining 24 locations, the licensee used the following methodology for analyzing the isolated sections:

1. For heat input, various size reactor coolant system line breaks from 2 inches up to the large break LOCA were used and various size main steam line breaks were used,

2. Using the containment pressure and temperature profile for the various line breaks, the time dependent pressure rise inside the pipe section was calculated.

3. The stress in the piping due to internal pressure was then calculated.

4. The calculated stress was compared to the stress allowables in the gf piping and valve components.

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'For the large break LOCA case, the licensee reported the following preliminary-results:

1. Seven locations passed criteria with no modification required.

2. Fifteen locations passed criteria when credit is taken for existing insulation.

3, Two locations will require field modifications.

For the small break LOCA case, only 7 of the 24 locations have been analyzed.

The areliminary results show that for smaller, uninsulated pipe, the small breat containment temperature and pressure profile may be more limiting than the large break-LOCA. Enhanced insulation or the modifications required as a result of the large break LOCA analysis would allow the stress criteria to be met.

The licensee still has to complete the evaluations for the remaining small break LOCAs and steam line breaks. It is estimated that about 2 months will be required to complete these evaluations.

The licensee had identified three modifications that would reduce the piping

-stress (1) a change in-the electrical logic to assure access to an existing-relief valve, (2) the installation-of a thermal relief valve, and (3) enhanced insulation on the residual heat removal-(RHR) suction cooling lines.

During the meeting, the licensee stated that there was a high probability-that during Refuel IX. changes 1 and 2 would be implemented. After the meetir.g.

the-licensee notified the staff that after reviewing the impact on safety of the modifications, only modification 3. above, had a high probability of implementation during Refuel-IX. The licensee's commitment is to have all evaluations and modifications completed by the end of Refuel X, currently scheduled for the spring of 1999. The licensee has performed an operability

< evaluation for all sections-of the piping in question.

The licensee has done some ' risk evaluations of the 24 locations and estimated the impact on core damage frequency (CDF). The following results were obtained:

1. None of the 24 sections perform an-accident mitigation function.

2. The postulated failure of 22 of the 2/ tections has no impact on CDF.

3. The postulated failure of the RHR suction cooling line sections would have an insignificant impact on CDF (<0.016% increase).

From the June 25. 1997, meeting, the licensee felt-there were three open issues:

1. Containment cooler pressure measurements,

J 21 - Analysis of Train A waterhammer forces,: and 3..- Use of emergency condition stress limits for waterhammer..

To: resolve these-issues, the_ licensee reported they- will install instrumentation to measure containment cooler pressures during the performance of the surveillance test that simulatss a' loss-of-off-site power.- They are currently analyzing Train' A and the results to date show that the Train B analysis remains-bounding. Concerning use of emergency condition stress

-limits, the -licensee has revised its- surveillance testing proceduret '.o -

eliminate the possibility of a waterhammer event during testing. With this-revision,:- the licensee concluded that the frequency of a waterhammer event would be very low and use of the emergency condition stress. limits was justified.

The technical content of the licensee's presentation.was well- received by the staff.

E Enclosed is_the handout provided at the meeting by the-licensee and' the list  !

of participants in the meeting.

Original Signed By Kristine M. Thomas, Project Manager Project Directorate IV Division of Reactor Projects Office of Nuclear Reactor Regulation

-Docket No. 50-482 :ISTR::BUT:'ON: (w/encis. I and 2)*

O etisetE1' e4

Enclosures:

1. -

Handout- PUBLIC

2. Attendees -PDIV-2 Reading- '

J. Stone W. Johnson, RIV P. Gwynn, RIV DISTRIBUTION:'(w/ encl.L2)

S. Collins (SJC1)

F.=Miraglia (FJM) 8t. Zimmerman (RPZ)

E. Adensam (EGA1)

W. Bateman (WHB)

E.-Peyton (ESP)

T. Marsh (LBM)

D. Wessman (RHW)-

-DOCUMENT NAME: 924 MEET.WC *See Previous Concurrence Sheet OFC - PDIV-2/PM PDIV-2/LA NRR:SPLB , NRR:EMEB*

NAME KTt N ye EPN TMarsf DWessman DATE /O /tc/97 i Q \S/97 /0 //b /97 N 10/3/97 0FFICIAL-RECORD COPY

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cc w/encls Jay S11 berg. Esq. Chief Operating Officer Shaw. Pittman. Potts & Trowbridge Wolf Creck Nuclear Operating Corporation 2300 N Street, NW P. O. Box 411 l Washington D.C. 20037 Burlington Kansas 66839  :

Re Su L U.kionalAdministrator.RegionIV-Nuclear Regulatory Commission WoServisorLicerisinbperatingCorporation f Creek Nuclear  :

! 611 Ryan Plaza Drive. Suite 1000 P.O. Box 411 4-Arlington,1 Texas 76011 .

Burlington, Kansa; 66839 Senior Resident inspector U.S. Nuclear Regulatory Commission

- U.S.-Nuclear Regulatory Commission Resident inspectors Office L P. O. Box 311 8201 NRC Road  ;

4 Burlington, Kansas 66839 Steedman, Missouri 65077 1032 Chief Engineer Mr. Otto L. Maynard i utilities Division President and Chief Executive Officer Kansas Corporation Commission Wolf Creek Nuclear Operating Corporation

. 1500 SW Arrowhead Road Post Office Box 411 Topeka, Kansas 66604 4027 Burlington Kansas 66839

Office of the Governor State of Kansas Topeka,I:nsas 66612 i

Attorney General Judicial Center 301 S.W. 10th 2nd Floor Topeka, Kansas 66612

. County Clerk Coffey County Courthouse Burlington, Kansas 66839 Vick L. Cooper. Chief Radiation Control Program Kansas Department _of Health and Environment Bureau of Air and Radiation Forbes Field Building 283 Topeka, Kansas 66620 e

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, .l Wolf Creek Nuclear Operating Corporation 6

L Generic Letter 96-06 Meeting September 24,1997 ;E

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Agenda

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• Introduction

• Summary

• Wolf Creek License Basis

• Preliminary Results of Completed Analyses

• Remaining Analyses

• Risk Evaluation

• Modification Plans

• OtherIssues

introduction

• Licensee

• 3RC

• Purpose of Meeting:

-Follow-up of Wolf Creek /NRC meeting on 6/25/97

-Provide findings and status L-- __. . . - j

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Summary w - - -

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• USAR. allows the use of ASME Appendix F for elastic analysis (Paragraph 3.9(B).1.4.2 and .

Table 3.9(B)-7).

• Large break LOCA analysis complete and required modifications identified.

• Analysis ongoing for small break LOCA and steam line breaks. Additional modifications may be necessary.

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Wolf Creek License Basis l

The AS ME Boiler and Pressure Vessel code

of record for Wolf Creek is the 1974 l edition, up to and including Winter 1974 l addenda (Paragraph 2.1.1 of S

SUPPS .

l Piping Design Specification M-200?.

j Appendix F is cited in Subsection :SB j (Class 1 components) in paragraph NB- .

! 3235 and in :SB-3656 of the 1974 edition of j the Code.

Wolf Creek License Basis Section 3.9(B).1.4. 2 of the Wolf Creek USAR states that the stress allowables of Appendix F are used for elastically analyzed code components for faulted conditions for Seismic l

Category 1 items other than the NSSS.

Table 3.9(B)-7 of the Wolf Creek USAR gives l the allowable stress for faulted conditions for l Class 2 and 3 piping to be 2.4 S3.

l

Preliminary Results of Completed Analyses A total of 33 locations were evaluated.

9 locations were eliminated:

- Four locations were eliminated from consideration due to configuration.

- Five locations will be eliminated with operating procedure changes.

3 Preliminary Results of Completed Analyses

- Analyses methodology for remaining 24 locations:

- Use various size line breaks.

- Calculate time dependent pressure rise inside pipe section.

- Calculate stress in piping due to internal l pressure.

- Compare stress to allowables in piping and in valve components.

I

Preliminary Results of Completed Analyses Results using large break LOCA for

! containment pressure and temperature profile:

- 7 locations passed criteria with no modifications i

necessary.

- 15 locations passed criteria when existing.

insulation is credited (will require modifications).

- 2 locations require field modifications.

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Preliminary Results of Completed Analyses

• Results using small break LOCA for containment pressure and temperature profile (7 of 24 locations completed):

- Small break LOCA pressure and temperature

< profile appears more limiting for some locations.

- Results indicate that enhanced insulation or large break LOCA modifications would allow stress criteria to be met.

Remaining Analyses

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- Complete evaluation of small break LOCA.

- Complete evaluation of steam line breaks.

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i i Risk Evaluation i

l The 24 piping sections were evaluated for impact on Core Damage Frequency (CDF):

- None of the 24 sections perform an accident mitigation function.

- Postulated failure of 22 of the 24 sections has no

! impact on CDF.

- Postulated failure of valves in the RHR suction cooling line sections would have an insignificant impact on CDF (<0.016%).

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Modification Plans

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Modifications planned for Refuel IX*:  ;

- One location will have an electrical logic modification to assure access to an existing thermal 1 relief valve.

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- One location (B31.1 piping) will have a thermal relief valve installed.

- Insulation qualification or enhancement for the RHR suction cooling line sections.

Contingent upon ability to complete design, procure materials, and complete field work within the current outage schedule.

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Modification Plans i

Future Modifications (by Refuel X):

- Insulation qualification or enhancement depending on results of remaining analyses and industry disposition of this issue.

- Any of the RefuelIX modifications which could not be completed.

Other Issues g m -

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- Open Issues from 6/25/97 meeting:

- Containment Cooler pressure measurements.

- Analysis of Train A waterhammer forces.

- Use of emergency condition stress limits for waterhammer.

Questions and Answers 1 i

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Enclosure 2 ARENDANCE LIST MEETING WITH WOLF CREEK NUCLEAR OPERATING CORPORATION SEPTEMBER 24, 1997 Wolf Creek Nuclear Operatino Corocration Terry Garrett Bob Osterrieder Terry Damashek Bill Selbe Altran Corooration Thomas Esselman El Kurt Cozens '

NRC Jim Stone Coretta Saadu John Fair William Bateman George Hubbard Kamal Manoly Dick Wessman Jim Tatu:n

. Bill Long Tad Marsh

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