ML19327B761
| ML19327B761 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/03/1989 |
| From: | Rosenblum R SOUTHERN CALIFORNIA EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-88-20, NUDOCS 8911130091 | |
| Download: ML19327B761 (5) | |
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V Southem Califomia Edison Company i
23 PARMER STREET IRVINE, CALIFORNIA 92718
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U. S. Nuclear Regulatory Commission
,9 Attention: Document Control Desk i
L Washington, D.C.
20555 l
Gentlemen:
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Subject:
Docket Nos. 50-206, 50-361 and 50 362 l
Generic Letter 88-20, " Individual Plant Examination (IPE) i for Severe Accident Vulnerabilities" i
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~ San Onofre Nuclear Generating Station Units 1, 2 and 3 i
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i Enclosed is our response to the subject Generic Letter. The Generic Letter specifically requested that we: 1) identify the method and approach for i'
performing the IPE 2) describe the method if it has not been previously submitted for staff review; and 3) identify milestones and schedules for l
perforwance of the IPE. This information is included in the enclosure.
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We plan to meet the requirements for an IPE by performing level 1 1
and evaluating containment Probabalistic Risk Assessments (PRAs)havior and release of fission products performance and the in containment be p
to the environment. One IPE will be performed for San Onofre Unit I and one for San Onofre Units 2 and 3.
We currently have significant internal
.i experience in the performance of PRAs but are increasing our staffing and
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training to fully meet the requirement for utility staff involvement. The l
increase in staffing and training and our desire to have a quality document which can continue to be useful after submittal to the NRC will require j
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AMI 8911130092 891103 POR ADOCK 05000206..
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Document Control Desk J
November 3,1989 i
that a deliberate and consistent pace be established for this work.
i Therefore, our schedule for submittal reflects full use of the three years allowed for this work.
j po If you require additional information, please contact me, f
t-i Respectfully subaitted.
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r Subscribed and sworn to before me this l
7sh day of 7br Au. 749 l
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J. B. Martin, Regional Administrator NRC Region V C. Caldwell. NRC Senior Resident inspector, San Onofre Units 1, 2 and 3 c
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ENCI45URE EBSDOMER TO GENERIC LETTER 88-28 l
IMDIVIDD&L P1 ANT REAMIMhTION FOR REYERE ACCIDENT YULMER&EILITIES Southern California Edison Company (SCE) has a
program of Probabilistic Risk Assessment (PRA) development and application for San Onofre Nuclear Generating Station (SONGS), Units 1, 2 and 3.
The program has been ongoing for several years and when complete will meet the requirements and objectives of Generic Letter 88-20.
This letter summarizes the program currently underway and provides a commitment for the completion of the program for SCE to meet the requirements of the generic letter.
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Eg's PRA Proaram SCE's PRA program involves a systen.atic examination consistent with
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state-of-the-art PRA methods and the objectives of Generic latter 88-20 and NUREG-1335, " Individual Plant Examinationt Submittal Guidance."
Through the performance of this examination, SCE is
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fulfilling a number of objectives including the following:
o Examining, in an integrated manner, the overall safety and reliability of the nuclear unit;
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o Providing assurance that the plant is designed and operated l
consistent with the highest objectives of safety and excellence; o
Identifying plant-specific vulnerabilities and evaluating potential enhancements where appropriate; o
Developing an appreciation of severe accident behavior and the I
most likely plant accident sequences; and o
Developing a quantitative understanding of the overall core damage probability and fission product release.
Generic Letter 88-20 notes that the " quality and comprehensiveness of the results derived from an IPE depend on the vigor with which the utility applies the method of examination and on the utility's commitment to the intent of the IPE".
SCE embarked on the current program based on its own internal desire to evaluate facility safety.
The program is performed by a dedicated SCE PRA group with assistance from consultants as appropriate.
Results are reviewed with and by' plant operations personnel and system engineers for consistency with design and operation.
Further, the results are,
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- used on a regular basis to evaluate the reliability and performance of the plant.
The level of involvement and detail associated with the program requires careful and consistent development.
Therefore, the program will be continued at a consistent pace to assure active and thorough involvement by all appropriate plant and engineering personnel.
The schedule provided in this enclosure for the development of the PRA program and the IPE is consistent with this ongoing level of involvement.
IPE Methodoloav The following paragraphs identify the method which will be employed in performing the IPE and indicates the contents of the reports to be submitted to the NRC.
In all aspects, the IPE will be completed in accordance with GL 88-20 and NUREG-1335.
The level 1
(" front-end") portion of the IPE will be completed consistant with current methods and approaches, such as those described in NUREG/CR-2300, PRA Procedures Guide.
Additional l
I considerations, for example the severe accident phenomenological issues considared in the generic letter, will be made.
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considerations may result in minor variations to the specific tasks described in NUREG/CR-2300.
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assessment of internal flooding will be performed upon completion of all other aspects of the front-end analysis.
This I
will allow the flooding assessment to benefit from the detailed j
plant model developed through the front-end evaluation.
The flooding analysis will include consideration of potential flooding
- sources, the potential for flood induced plant accidents or transients, the likelihood of system failure due to flooding, and i
the ability to detect and mitigate postulated flooding conditions.
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f The evaluation of Unresolved Safety Issue A-45,
" Shutdown Decay i
Heat Rehoval Requirements," will be included in the IPE analysis.
Vulnerabilities in the decay heat removal systems will be L
identified if they exist and reported with the IPE results.
i The containment evaluation ("back-end" analysis) will be completed 1
consistent with current methods and approaches, such as those described in NUREG/CR-2300, PRA Procedures Guide.
Any additional considerations will be explained in the IPE submittal.
This will include development of a plant-specific containment event tree as t
described in Section 2.2.2.5 of NUREG-1335.
The potential containment failure and mechanisms of Table 2.2 of NUREG-1335 will be treated in the containment event tree evaluation.
To the extent
- possible, an approximate approach utilizing applicable inforr.ation performed for plants similar to SONGS will,
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be utilized to perform the back-end analysis.
This approach is consistent with the generic letter requirements.
Wherever
,l' reference calculations are adopted for applicability to SONGS, an evaluation of the principal difference between SONGS and the reference facility will be performed to provide assurance that the assumptions and results are directly applicable or to identify the general impact of differences.
Based upon the completed front and back-end analyses, significant i
vulnerabilities and insights will be evaluated.
This will be done by evaluating those sequences and cutsets contributing a
significant portion of each functional failure type and category of accident sequence.
The evaluation will include consideration of relative risk or importance measures and the performance of i
sensitivity studies to enhance the insights developed.
Any 9
identified vulnerability that warrants immediate correction will i
be thoroughly evaluated and appropriate corrective action taken.
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'nformation indicating that currently planned or anticipated modifications may not be beneficial, will be evaluated i
and the project will be considered for deferral or cancellation.
IPE Submittal
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The results of the review will be documented in a two tier approach. As described in GL 88-20, the information submitted will i
include the results of the examination, and a summary of the l
insights gained.
Detailed documentation will be retained by SCE and utilir,ed in SCE's ongoing risk evaluation and safety assessment program.
The report submitted to the NRC will be consistent with the general guidance in Table 2.1 of NUREG-1335.
The schedule for performance of the IPE is as follows:
IPE Schedule The current schedule for performance of the IPE is as follows:
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SCHEDULE 1.
Completion of the Front-End Analysis 9/1/91 2.
Completion of the Back-End Analysis 3/1/92 3.
Completion of the IPE and Submittal of 9/1/92 i
Results l
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