ML19324B768

Forwards Response to Generic Ltr 88-20, Individual Plant Exams for Severe Accident Vulnerabilities. New Analyses Incorporating Util Prior Experience Will Be Developed in Responding to Generic Ltr as Listed

ML19324B768
Person / Time
Site:	Peach Bottom, Limerick, 05000000
Issue date:	10/31/1989
From:	Helwig D PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-88-20, NUDOCS 8911080185
Download: ML19324B768 (8)

Text

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. 10 CPR 50.54(f) j GL 88-20  ;

PHILADELPHI A ELECTRIC COMPANY l

NUCLEAR GROUP HEADQUARTERS 955 65 CHESTERBROOK BLVD.

WAYNE PA 19087 5691 '

1 (215) 640 6650 I 5 )

DAVID R. HELWIG vics enssipaMt mucLean aanveces ,

October 3

.1, 1989 Docket Nos. 50-277  :

50-278 )

50-352 50-353 License Nos. DPR-44 DPR-56 NPP-39 NPF-85 1 I

U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Peach Bottom Atomic Power Station, Units 2 and 3 l Limerick Generating Station, Units 1 and 2

[ Response to NRC Generic Letter 88-20, " Individual Plant Examinations for Severe Accident Vulnerabilities" '

L Gentlemen:

NRC Generic Letter (GL) 88-20, " Individual Plant Examination for Severe Accident Vulnerabilities," dated' November 23, l 1998, and Supplement 1 dated August 29, 1989, required Philadelphia i Electric Company (PECo)'to submit a response, as specified in 10 CFR l 50.54(f), desc'ribing the proposed programs for completing Individual l Plant Examinations (IPEs) of our nuclear stations. GL,88-20 states ,

that our program description should: J

1. identify the method and approach selected for performing the IPE,

2. describe the method to be used, if it has not been l previously submitted for staff review (the description may l be referenced), and l

3. identify the milestones and schedules for performing the IPE and submitting the results to the NRC.

8911000185 891031 PDR ADOCK 05000277

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t-00f' P PNU ' \

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Dodument Control De-k Octcber 31, 1989 PCgs 2 Our responses for. Peach Bottom Atomic Power Station (PBAPS) i Units 2 and 3 and Limerick Generating Station (LGS) Units 1 and 2, provided in^ Attachments 1 and 2, describe our plans and schedules .

for completing the IPEs. .)

i If you have any questions, or require additional I informat.lon, please contact us. ,

Very truly yours, f . i

(. Lke.

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Attachments  ;

cc: W. T. Russell, Administrator, Region I, USNRC .

T. P. Johnson, USNRC Senior Resident Inspector, PBAPS T. J. Kenny, USNRC Senior Resident Inspector, LGS l'

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COMMDNWEALTH CF PENNSYLVANIA :

• ss.

COUNTY OF CHESTER  :

D. R. Helwig, being first duly sworn, deposes and sayst  ;

That he is Vice President of Philadelphia Electric Company, that he has read the response to Generic Letter No. 88-20, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

D 1 l

C) spa Vice Presi e Subscribed and sworn to before me this3l day

< -of O M 1989.

l det&ac>/2 %h Notary Public l

l NOTARIAL SEAL CATHERINE A. MENDEZ. Notary Public Tredyffrin Two Chester County My Commission Exetres Sect. 4.1993 l'

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ATTACHMENT 1 ,

PEACH BOT 20M ATOMIC POWER STATION, UNITS 2 AND 3 RESPONSE TO GENERIC LETTER 88-20, J

" INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCIDENT VULNERABILITIES" >

j Peach Botton. Atomic Power Station (PBAPS) was chosen as a reference plant for the Reactor Safety Study, WASH-1400, and the recently l published NUREG-1150, entitled, " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants." NUREG-1150 provides '

a state-of-the-art understanding of severe accident risk and also provides an update of the risk calculated in WASH-1400. Both NRC studies concluded that the risk of a severe accident at PBAPS is extremely low. In addition to the major probabilistic studies cited, PBAPS has served as a reference plant for other LIC-sponsored studies as well as the Industry Degraded Core Rulemaking (IDCOR) program.

Philadelphia Electric Company (PEco) has participated in many of these probabilistic studies and as a result, has gained an appreciation of severe accident behavior, understands the most likely severe accident sequences at PBAPS and has a quantitative understanding of the overall probabilities of core damage and fission product releases associated with severe accidents. '

Probabilistic Risk Assessment (PRA) familiarization training which ,

includes the results and insights from NUREG-1150, other studies of f PBAPS, and insights from the Limerick Generating Station-(LGS) PRA, has been given to over 600 members of the Nuclear Group. A PRA task force was formed and has developed a Risk and Reliability Program ,

l Plan to focus the utilization of PRA to support PBAPS. The Plan provides a documented statement of the PRA program objectives and

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planned activities at PECo. Each Nuclear Group organization is committed to supporting the Plan.

l L Even though PECo has supported and reviewed portions of the NRC-sponsored NUREG-1150, it has been determined that in order to comply with the objectives of Generic Letter 88-20 and the guidance

(' provided in NUREG-1335, " Individual Plant Examination: Submittal Guidance," PECo should develop core damage frequencies (Level 1 PRA) l and a containment performance analysis (Level 2 PRA) for PBAPS.

E Developing, reviewing, and maintaining a PECo-sponsored PRA based in part on NUREG-ll50 will provide the necessary traceability, utility involvement, and consistency between PBAPS and LGS Individual Plant Examinations (IPEs).

The PBAPS is a two unit station; each unit being a General Electric Bolling Water Reactor utilizing a Mark I containment. The containment evaluation will utilize the Modular Accident Analysis Program (MAAP) and other published computer code results to assist in understanding the Mark I containment response to a variety of severe accidents. Computer codes and published results will also be used to characterize the timing and radionuclide release of severe accident sequences at PBAPS. Issues identified regarding severe accident phenomenology will be addressed in development of the Level 2 portion of the PBAPS PRA.

PECo Rocpon2@ to GL 88-20

• Attachment 1 Page 2 '

Results from NUREG-1150 indicate that individual early fatality and latent cancer fatality risks from internally initiated events and fire initiated accidents for PBAPS are well below the NRC safety goals. Although PBAPS meets published NRC safety goals, PECo will perform the Level 1 and 2 PRAs to further develop our understanding of severe accidents and associated probabilities. This approach .

will provide a structure to logically identify the benefits and competing risks associated with possible plant or procedural changes and provide a model that can be easily updated to incorporate future operational, design, and procedural changes.

In conclusion, PECo has been actively involved in PBAPS PRA studies since 1981. New analyses incorporating this prior experience will be developed in responding to Generic Letter 88-20 as follows.

1. Level 1 PRA models will be developed to represent Unit 2 and Unit 3.

2. A Level 2 PRA will be performed to assess containment effectiveness and probability of radioactivity releases following core damage.

3. PRA development and documentation will incorporate the .

guidance of NUREG-1335.

4. The results of the IPE may also be applied to the resolution of other USIs and GSIs as appropriate. The specific safety issues that will be addressed will be '

identified in the IPE summary report.

Our plans for completing and submitting the IPE are as follows.

MAJOR ACTION DATE

! Complete Level 1 PRA requirements February 1992 Complete Level 2 PRA requirements February 1992 Investigate hardware and May 1992 procedure changes Submit summary report to the NRC September 1, 1992 l

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ATTACHMENT 2 ,

LIMERICK GENERATINC STATION, UNITS 1 AND 2 RESPONSE TO GENERIC LETTER 88-20,

" INDIVIDUAL PLANT EXAMINATION FOR SE7ERE ACCIDENT VULNERABILITIES" Philadelphia Electric Company (PECo) has satisfied the objectives of Generic Letter 88-20 for Limerick Generating Station (LGS), Units 1 and 2. An appreciation of severe accident behavior has been attained, the most likely severe accident sequences are understood, a quantitative understanding of the overall probabilities of core damage and fission produce releases has been gained, and hardware and procedures have been modified that help prevent or mitigate severe accidents.

A May 6, 1980 letter, from D. G. Eisenhut (NRC) to E. G. Bauer (PECo), requested that a Probabilistic Risk Assessment (PRA) be performed for LGS and the results submitted to the NRC. PECo complied with this request and presented the results at a meeting with the NRC on December 9, 1980. The PRA was submitted with the LGS Final Safety Analysis Report (FSAR) and the Environment Report Operating License (EROL) on March 17, 1981. Because of the schedule, PECo relied heavily upon outside experts to prepare this PRA.

During the period from 1981 to 1982, PECo extensively reviewed and revised the PRA. The revisions were also reflected in the Severe Accident Risk Assessment, submitted to the NRC in April, 1983.

Considerable NRC and NRC consultant effort was devoted to the review of the LGS PRA. The NRC review involved numerous meetings, formal l written questions, and site visits.

l In recent years, PECo has updated the Level 1 internal events portion of the LGS PRA to reflect the as-built, as-operated plant.

PECo personnel were extensively involved in performing this update.

As analyses were completed, reviews were performed by PECo station and engineering personnel. The updated PRA was also subjected to

peer review by outside PRA experts. Most recently, PECo provided the NRC with additional information, including meetings and site visits, to support preparation of an NRC position regarding the need for severe accident mitigation design alternatives at LGS.

l The LGS is a two unit station, each unit being a General Electric Boiling Water Reactor (BWR) utilizing a Mark II containment. A preliminary draft of a containment evaluation nearly identical to that requested by NUREG-1335, " Individual Plant Examination:

Submittal Guidance," has already been completed. The Modular Accident Analysis Program (MAAP) computer code results are utilized to assist in understanding the Mark II response to different severe accident challenges. The evaluation includes quantified containment event trees (CETs) and the necessary deterministic assessments.

MAAP and Meltdown Accident Response Characteristics (MARCH) computer code published radionuclide source term information are used to characterize the radionuclide release sequences. The draft containment evaluation specifically addresses the issues which have been identified in NRC-industry meetings regarding source terms, and

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PECo Response to GL 88-20 Attachment 2 Page 2 the issues that the NRC has identified in Generic Letter 88-20 and l NUREG-1335. The approach is structured to logically identify the I benefits and competing risks associated with possible plant or procedural changes. The preliminary results are that LGS does not have unusually poor containment performance and meets the published safety goal guideline for severe releases. Performing the Level 1 and Level 2 PRA updates has further developed our understanding of severe accident behavior and the overall probabilities of core damage and fission product releases. 1 We are now implementing PRA applications at LGS. A PRA task force j was formed and has developed a Risk and Reliability Program Plan to focus the utilization of PRA to support LGS. The Plan provides a documented statement of the PRA program objectives and planned activities at PECo. Each Nuclear Group organization is committed to supporting the Plan. PRA familiarization training, including the ,

l results and insights from the LGS PRA, has been given to over 600 members of PECo's Nuclear Group. Licensed operator requalification training given to LCS personnel includes training on the most likely core damage events. Changes to procedures based upon PRA insights have been made. Hardware changes have been analyzed. The Station Maintenance organization has used PRA results to prioritize its work and is beginning a Reliability Centered Maintenance project. These ,;

applications accomplish the follow-on uses of the Individual Plant l Examination (IPE) results.

I In-summary, PEco has actively used PRA since 1980 and is currently in the process of finalizing the results for LGS as follows.

1. The PRA has been updated to the current plant configuration and procedures. Completion of the PRA documentation is currently in progress; however, LGS will complete implementation of Rev. 4 of the BWR Owners Group Emergency Procedure Guidelines (EPGs) this year which must then be reflected in the PRA.

1 L 2. A new Level 2 PRA is being performed to support the

! response to Generic Letter 88-20 for LGS.

3. The submittal guidance of NUREG-1335 requires some additional analyses and/or documentation in the following specific areas.

L Level 1 requirements which must be documented include:

simplified P& ids data source identification common cause treatment human error description requantification criteria for vulnerability internal review 1

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PECo Response to GL 88-20

, Attachment 2 .;

Page 3 o

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Level 2 requirements whlk:h must be documented include:

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- description of mitigative systems including isolation 1 l procedures, severe accident operability, and manual l t-actions 1

- description of accident sequence bins j

- incorporation of EPG Rev. 4 >

- containment challenges and time of failure

- determining source term magnitudes '

- quantification of containment event trees

- sensitivity calculations for operator actions, in-vessel phenomena and ex-vessel phenomena

- internal review Additional reporting requirements, which must be t

[ documented, include:

- discussion of functional accident sequences

- identification of vulnerabilities

- identification of hardware or procedural changes to eliminate vulnerabilities

- disposition of Unresolved Safety Issue (USI) A-45

4. The results of the IPE may also be applied to the i resolution of other USIs and GSIs as appropriate. The specific safety issues that will he addressed will be

-identified in the IPE summary report.

In addition, during the final stages of licensing for LGS Unit 2, PECo agreed to have an external peer review of the IPE process and results performed prior to submitting the IPE to the NRC.

Our plans for completing and submitting the IPE are as follows.

MAJOR ACTION D_ ATE Complete Level 1 PRA requirements June 1991 Complete Level 2 PRA requirements September 1991 Investigate hardware and November 1991 procedure changes Complete draft summary report January 1992 Conduct external p3er review March 1992 Accommodate peer review comments June 1992 Submit summary report to the NRC July 1992

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