ML20137L570

Requests for Addl Info Re Pilgrim Nuclear Power Station to Credit Containment Overpressure in Net Positive Suction Head Analysis for Emergency Core Cooling Pumps

ML20137L570
Person / Time
Site:	Pilgrim
Issue date:	03/31/1997
From:	Wang A NRC (Affiliation Not Assigned)
To:	Boulette E BOSTON EDISON CO.
References
TAC-M97789, NUDOCS 9704070265
Download: ML20137L570 (9)

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g p I j - ' March 31'l'1997 CAVd4'Qj { i hNi O 7 J'- Mr.I.Th$asBolilette}'Ph'.De';* x NDI ~

Senior.Vice President - Nuclear-r . ,3 i.. N

' Boston Edison Company .< ,s .

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, Pilgrim Nuclear Power. Station

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RFD #1 Rocky Hill Road 1/fi me ,

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SUBJECT:

REQUEST.FOR ADDITIONAliINFORMATION ON THE PILGRIM NUCLEAR. POWER j STATION TO CREDITS. CONTAINMENT OVERPRESSURE IN THE NET POSITIVE.

SUCTION HEAD ANALYSIS FOR THE EMERGENCY-CORE COOLING PUMPS (TAC NO.

.M97789)

Dear Mr. Boulette:

) -Enclosed is a requ'est'for. additional information-(RAI) regarding your subject submittal dated January 20, 1997. The RAI provides comments and' i questions regarding equipment qualification.in support of your amendment request. The staff requests.that the response-be provided as soon as possible-as this is a restart issue.

o Sincerely, (Original Sigr.ed By)

Alan B. Wang, Project Manager i- Project Directorate I-3

, Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation.

Docket No. 50-293

Enclosure:

Request for Additional Informationt cc w/ enc 1: See next page i Distribution:',

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%.....,[ . March 31, 1997 Mr. E. Thomas Boulette, Ph.D Senior Vice President - Nuclear Boston Edison Company Pilgrim Nuclear Power Station RFD #1 Rocky Hill Road Plymouth, MA 02360

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON THE PILGRIM NUCLEAR POWER STATION TO CREDIT CONTAINMENT OVERPRESSURE IN THE NET POSITIVE SUCTION HEAD ANALYSIS FOR THE EMERGENCY CORE COOLING PUMPS (TAC NO.

M97789)

Dear Mr. Boulette:

Enclosed is a request for additional information (RAI) regarding your subject submittal dated January 20, 1997. The RAI provides comments and questions regarding equipment qualification in support of your amendment request. The staff requests that the response be provided as soon as possible as this is a restart issue.

Sincerely, t

Alan B. Wang, roject Manager  !

Project Directorate I-3 i Division of Reactor Projects - I/II l Office of Nuclear Reactor Regulation  !

l Docket No. 50-293 j

Enclosure:

Request for Additional Information cc w/ enc 1: See next page i

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E. Thomas Boulette Pilgrim Nuclear Power Station

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Mr. Leon J. Olivier' Mr. Jeffery Keene.

Vice President _of Nuclear.' Licensing Division Manager- l Operations & Station Director Boston Edison Company '

Pilgrim Nuclear Power Station 600 Rocky Hill Road IFD #1 Rocky Hill Road Plymouth, MA 02360-5599 Plymouth, MA 02360 Ms. Nancy Desmond Resident Inspector Manager, Reg. Affairs Dept. ,

U. S. Nuclear Regulatory Commission Pilgrim Nuclear Power Station q Pilgrim Nuclear Power Station RFD #1 Rocky Hill Road .i Post Office Box 867 Plymouth, MA 02360 i Plymouth, MA 02360 I Mr. David F. Tarantino Chairman, Board of Selectmen Nuclear Information Manager

-11 Lincoln Street Pilgrim Nuclear Power Station

-Plymouth, MA 02360 RFD #1, Rocky Hill Road Plymouth, MA 02360 Chairman, Duxbury Board of Selectmen Town Hall Ms. Kathleen M. O'Toole 878 Tremont Street Secretary of Public Safety Duxbury, MA 02332 -

Executive Office of Public Safety One Ashburton Place .

Office of the Commissioner Boston, MA 02108 Massachusetts Department of Environmental Protection Mr. Peter LaPorte, Director One Winter Street Attn: James Muckerheide Boston, MA 02108 Massachusetts Emergency Management Agency Office of the Attorney General 400 Worcester Road One Ashburton Place P.O. Box 1496 20th Floor Framingham, MA 01701-0317 Boston, MA 02108 Chairman, Citizens Urging Mr. Robert M. Hallisey, Director Responsible Energy Radiation Control Program P.O. Box 2621 Massachusetts Department of Duxbury, MA 02331 Public Health 305 South Street Citizens at Risk Boston, MA 02130 P.O. Box 3803 Plymouth, MA 02361 Regional Administrator, Region I U. S. Nuclear Regulatory Commission W.S. Stowe, Esquire 475 Allendsle Road Boston Edison Company King of Prussia, PA 19406 800 Boylston St., 36th Floor Boston, MA 02199 Ms. Jane Fleming 8 Oceanwood Drive

. Duxbury, MA 0233

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Chairman  !

Nuclear Matters Committee  ;

Town Hall 11 Lincoln Street Plymouth, MA 02360 Mr. William D. Meinert-Nuclear Engineer Massachusetts Municipal Wholesale -

Electric Company P.O. Box 426 Ludlow, MA 01056-0426 ,

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1 REQUEST ."Ok ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST DATED JANUARY 20, 1997 .

SUPPORTING THE PILGRIN NUCLEAR POWER STATION NPSH ANALYSIS l 2 1. Page 32 of Safety Evaluation No. 2983 indicates that the EQ analysis differs from the final safety analysis report (FSAR) analysis in two  !

respects: (1) the EQ analysis takes credit for passive heat sinks in the drywell, wetwell, and suppression pool which tend to moderate the drywell i temperature response; and (2) the EQ analysis takes credit for a 720 vs. l 300 drywell spray flow which provides more cooling to the drywell airspace and reduced temperature for equipment located in the drywell.

Page 32 also indicates that a review and update of all environmental qualification data files was performed to verify environmental qualification at the new 75F qualification envelope for drywell temperature. It is not clear if the 75F qualification envelope for drywell temperature is based on the EQ or FSAR analysis described above.

a. Provide clarification.

b. Clarify what is meant by " equipment qualification was verified."

(1) Is equipment qualified to the requirements of paragraph (e) of 10 CFR 50.497 If not provide justification and describe how equipment has been

, qualified.

(a) For equipment qualified using methods permitted by paragraph (k) of 10 CFR 50.49 (i.e., D0R guidelines -- no aging required), describe and provide justification for the process used to assure each item of electric equipment important to safety covered by 10 CFR 50.49 will meet its specified performance requirements when it is subjected to the conditions predicted to be present when it must perform its safety function up to the end of its qualified life (Ref. Paragraph (j) of 10 CFR50.49).

(2) Do EQ test profiles (for accident and post-accident) for each piece of equipment envelop the new EQ accident and post-accident profiles? Or, if not, identify how and provide justification for each case where the EQ test profile does not envelop the new accident and post-accident profiles.

c. Identify how and provide justification for each case where the EQ test profile does not envelop the new accident profile (accident and post-accident profiles) based on the FSAR analysis for establishing accident profile.

2. Page 35 of Safety Evaluation No. 2983 indicates equipment qualification was verified for resulting post-LOCA building ambient temperature profiles without loss of offsite power.

ENCLOSURE

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a. Explain how this verification covers equipment qualification for ,

post-LOCA building ambient temperature profiles with (versus without) '

loss of offsite power and with single active failure with and without l loss of offsite power. Also, explain how this verification covers i equipment qualification for 1.0CA building ambient temperature profiles  !

with and without loss of offsite power and with single active failure.

Similarly, explain how steam line break (i.e., other accidents besides LOCA) are covered.

b. Clarify what is meant by " equipment qualification was verified."  :

(1) Is equipment qualified to the requirements of paragraph (e) of 10 CFR 50.497 If not provide justification and describe how equipment has been qualified.

- (2) Do EQ test profiles (for accident and post-accident) for each piece of equipment envelop the new EQ accident and post-accident profiles? If not, identify how and provide justification for each case where the EQ test profile does not envelop the new accident and post-accident profiles.

3. Page 31 of Safety Evaluation No. 2983 states: " Equipment at PNPS I requiring qualification to meet the requirements of 10 CFR 50.49 are  !

listed in the Environmental Qualification Master List (Ref. 46)." Page 33 of Safety Evaluation No. 2983 states: "The containment electrical penetrations that support the operation of active equipment are listed on the EQ Master List (Ref. 46). Containment electrical penetrations not 1!sted on the EQ Master List contain cabling that is not required to  !

function electrically in a post-accident environment but may continue to operate. Although functionally passive, penetrations not listed on the EQ Master List must remain leaktight to ensure containment integrity."

These statements appear to indicate that equipment considered functionally passive (e.g., non-safety electrical penetration) do not have to meet the requirements of 10 CFR 50.49 but must remain leaktight '

to ensure containment integrity (i.e., to ensure the requirements of paragraph (b)(1)(iii) of 10 CFR 50.49 are met). Provide clarification.

Identify other electrical equipment which have been determined to not have to meet 10 CFR 50.49 requirements because they are considered functionally passive.  ;

4. Define the original plant accident requirements profile which had a peak temperature of 330 *F to which electrical penetrations were qualified.

5. Provide the results of a linear slopes comparison analysis which utilizes Arrhenius methodology (similar to that shown in attachment 7 of General Electric Proprietary Document NEDC-32123P, Report PIR-CPD92045 Service Life Estimate for the Epoxy Sealant in the Penetration Assemblies at the Browns Ferry Nuclear Plant Unit 3, dated August 20,1992) which compares

the 18929.01 equivalent hours obtained for the test profile of 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at 352 *F, plus 23.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at 309 'F, plus 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> at 135 *F to equivalent hours at the original plant accident requirements profile, plus the equivalent hours for 40 years at 150 *F.

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6. Electric Power Research Institute's (EPRI's), Nuclear Power Plant Equipment Qualification Reference Manual indicates that the Arrhenius method has been employed to relate accident test temperatures to postulated accident temperatures. If the Arrhenius model and activation energy value are applicable to the test and

. accident temperatures, then the model may arguably be used in various ways to draw correlations between the accumulated thermal ,

damage occurring during various phases of LOCA testing. This approach has been used principally to support long-term operability -

in post-LOCA environments when it is desirable to have a test  !

duration be shorter than the actual required operability time. For l example, the test temperature plateau dropped to 212 *F at 5 days ,

into the 30-day test. The required post-LOCA temperature dropped to l 190 *F after 5 days and remained constant for an' additional 175 i days. Thus, although the test temperature envelopes the required  !

post-LOCA temperature, it lasts only 25 days and not 175 days. It is a common practice to argue that the higher test temperature (212

• F) can be viewed as an accelerated version of the actual post-LOCA temperature (190 *F). After using Arrhenius methods to determine equivalent degradation for 25 days at 212 *F and 175 days at 190 *F, if it turns out the equivalent degradation for 25 days at 212 *F is greater than 175 days at 190 *F, it can be argued that the test is conservative with respect to the actual post-LOCA conditions.

Another example (provided by PNPS in response to an NRC Request for 1

Additional Information) uses a device at PNPS that is required to be i operable for a period of 33 days (30 days plus a 10-percent margin) in a temperature environment of 150 *F maximum. Post-LOCA testing was conducted for a period of 20 days at 200 *F. The Arrhenius equation is used to rietermine the equivalent time at the required temperature. Since the available time-temperature curve from the

! vendor is based on testing, the equivalency and margin utilized in .

qualification are determined utilizing Arrhenius techniques. The j use of this method provides a means to quantitatively evaluate i

, variable accident conditions to determine equipment thermal  !

degradation. A Degradation Equivalency Analysis uses the Arrhenius j Methodology to show that the degradation of the equipment experienced due to test conditions is equal to or greater then the 3 i

degradation the equipment would experience from PNPS conditions.

The use of Arrhenius methodology to support qualification of equipment for LOCA and/or longer term post-LOCA environments has not ,

been endorsed by NRC Regulatory Guide, has not been generally i accepted, by itself, to demonstrate qualification of equipment in I

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operating experience or research test results. Therefore, the use-  ;

of Arrhenius methodology, by itself - without supporting justification or technical basis, is not considered an acceptable i

approach for supporting qualification of electric equipment for LOCA  :

3 environments. Provide additional justification supporting the

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conclusion (i.e., engineering judgment) that equipment that has not

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been tested for the full time required by actual post-LOCA accident conditions remains qualified for the new higher post LOCA

environments and/or provide supporting justification that Arrhenius methodology supports qualification of equipment for longer term ,

post-LOCA environments.

, 7. Based on information presented, it appears that acceptability for

qualification was originally based on a 10% weight loss of the j penetration's epoxy sealant. Given this 10% weight loss as acceptance criteria, it is not clear how the results from linear slope comparison 4

analysis can indicate a comparison ratio change from o.55 to 2.00. It is not clear how it can be concluded (as implied by information presented in

.EQ qualification package) that both the original test (having a profile

of 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at 352 *F plus 23.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at 309 *F plus 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> at 135 *F) 2 and the additional test (having a profile of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> at 340 *F and 10 4 days at 281 *F) on the same material can produce a weight loss of 10%.

Provide clarification.

8. General Electric Proprietary Document NEDC-32123P, Report PIR-CPD92045

Service Life Estimate for the Epoxy Sealant in the Penetration Assemblies 4 at the Browns Ferry Nuclear Plant Unit 3, dated August 20, 1992,

indicates that the activation energy for the penetration's epoxy was l revised from a value of 1.13 eV to 1.16 eV based the theory of life testing and use of thermogravimetric Analysis (TGA). TGA does not

provide a valid method for establishing the activation energy for material like the epoxy used in the penetration. The use of TGA is thus

not considered acceptable. Describe how and to what extent TGA has been i

utilized for the qualification of electric penetrations and other i i

equipment installed at Pilgrim.

9. Wyle Report No. 47066-PEN-1.1, Qualification Verification Report on General Electric Electrical Penetrations No. 238X600NLG1 for use in ,

Pilgrim 1 Nuclear Power Station, Revision F, dated March 20, 1996, indicates that a test profile of 340 *F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 281 *F for 10 days (which is described in General Electric Proprietary Test Report,

! Qualification Test for F01 electrical Penetration Assemblies, dated i

November 9,1971) was used to show a test profile to plant requirement ratio of 2.00:1 for both the cast epoxy and the vulkene cable. Given that the test report dated November 9, 1971, only describes a

qualification test of the penetration's epoxy, it is not clear what test was used for the vulkene cable to demonstrate it's qualification for the test profile of 340 *F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 281 *F for 10 days. Provide clarification.

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10. Provide the results of a linear slo)es comparison analysis which' utilizes Arrhenius methodology (similar to tiat shown in attachment 7 of General Electric Proprietary Document NEDC-32123P, Report PIR-CPD92045 Service Life Estimate for the Epoxy Sealant in the Penetration Assemblies at the Browns Ferry Nuclear Plant Unit 3, dated August 20,1992) which compares the equivalent hours obtained for the test profile of 340*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 281*F for 10 days to equivalent hours at the current plant accident requirements profile plus the equivalent hours for 40 years at 150'F.

11. Wyle Report No. 47066-PEN-1.1, Qualification Verification Report on General Electric Electrical Penetrations No. 238X600NLG1 for use in Pilgrim 1 Nuclear Power Station, Revision F, dated March 20, 1996, indicates that exposure of the penetration epoxy to radiation will cause weight loss (i.e., aging). From information presented in the EQ documentation package, it is not clear how aging due to radiation effects l has been addressed and combined with aging due to thermal effects to I demonstrate qualification of the penetration. Provide clarification.

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