ML17159A047
| ML17159A047 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/27/1997 |
| From: | Poslusny C NRC (Affiliation Not Assigned) |
| To: | Byram R PENNSYLVANIA POWER & LIGHT CO. |
| Shared Package | |
| ML17159A048 | List: |
| References | |
| GL-88-20, TAC-M74478, TAC-M74479, NUDOCS 9711190046 | |
| Download: ML17159A047 (10) | |
Text
~R hE0, 0
Cy O
tO0
+~'0
++**+
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205594001 October 27, 1997
~, y qual M8-317 S @
Hr. Robert G.
Byram Senior Vice President-Generation and Chief Nuclear Officer Pennsylvania Power and Light Company 2 North Ninth Street Allentown, PA 18101
SUBJECT:
REVIEW OF THE SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2, INDIVIDUALPLANT EXAMINATION SUBMITTAL INTERNAL EVENTS (TAC NOS.
H74478 AND M74479)
Dear Hr. Byram:
Enclosed is the NRC staff's safety evaluation (SE), which addresses the Susquehanna Steam Electric Station (SSES),
Units 1 and 2, Individual Plant Examination (IPE) submittal for internal events and internal flooding.
This SE was prepared by the Office of Nuclear Regulatory Research (RES) and includes a technical evaluation report prepared by the Brookhaven National Laboratory (BNL).
A "Step 1" review was performed by the RES and its contractor.
This review examined the IPE results for their "reasonableness" relative to the design and operation of SSES.
The NRC staff employed BNL to review the front-end
The SSES IPE resulted in a total core damage frequency (CDF) of 1E-7/} eactor cycle (rc) of 15 months, including the contribution from internal flooding.
Anticipated transients without scram contribute about 78X to the total
- CDF, loss of direct current bus about 14X, flooding about 7X, and all other initiating events about 2X.
The contribution of station blackout as an accident type is negligible (7E-ll/rc).
The NRC staff agrees that "The PP8L approach to probabilistic risk analysis for Susquehanna differs in several aspects from a conventional PRA."
The NRC staff believes that there are deterministic aspects in any PRA and is concerned that the differences of the SSES IPE from a conventional PRA may have contributed to an unduly optimistic treatment of several important aspects of the analysis and, in particular, of the treatment of.common-cause failures (CCFs),
human error, and containment performance.
7iii90046 97i027 DR ADQCK 05000387,:,
I I IIIIIIIIIIIIIIIIIIIIIIIIIIII IIIIIII
~~~ @~@TE@ >~V It is stated in your submittal that the method employed in the IPE falls under the category of "other systematic examination" of Generic Letter (GL) 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities," indicating that it is not a conventional probabilistic risk assessment (PRA) apparently because it combines deterministic and probabilistic evaluations.
The deterministic evaluations are based on Pennsylvania Power and Light Company's (PP8L's)
"defense-in-depth" approach developed as part of your severe accident management policy.
0
~
R.
Byram PP&L applied two sets of criteria to identify vulnerabilities at SSES:
(a) the criteria provided in.Appendix 2 of GL 88-20, which are quantitative; and (b) the defense-in-depth criteria, which are qualitative.
Because of the low calculated
Although the NRC staff has many concerns regarding the technical integrity of the IPE's quantitative
- approach, we believe that the PP&L's approach to identify instances of plant improvements is adequate.
It appears that your staff has performed a thorough evaluation of plant performance under accident conditions on the basis of your qualitative approach.
It is stated in the submittal that "PP&L does not 'rely upon low calculated plant damage frequency as a means of ensuring adeq'uate protection against severe accidents, but rather the satisfaction, by each accident
- sequence, of a set of Defense-in-Depth criteria." It is further stated that "for each accident
- sequence, the plant failures which have caused the specific plant damage state are identified.
With this information, it can be establish whether SSES satisfies the Defense-in-Depth criteria for each accident sequence.
Hodifications to plant equipment, procedures, or training, are considered for those accident sequences which do not meet these criteria."
Employing the defense-in-depth
- criteria, PP&L identified and implemented several plant modifications and improvements to the procedures, which were credited in the IPE.
Therefore, the NRC staff does not believe that the weaknesses identified in this SE limited PP&L's ability to identify plant improvement and concludes that your defense-in-depth approach is an adequate approach for identifying instances of plant improvements, which is one objective of GL 88-20.
On the basis of the information contained in your IPE submittal, which was provided through your direct interactions with the NRC staff and by written correspondence, the NRC staff concludes that the PP&L's SSES IPE is complete with respect to the information requested by GL 88-20 (and associated NUREG-1335) and that PP&L's IPE process is adequate to meet the following two general objectives of the IPE program as stated in GL 88-20.
1.
To develop an appreciation for severe accident behavior.
If necessary, to reduce the overall probabilities of core damage and fission product releases by modifying, where appropriate, hardware and procedures that would help prevent or mitigate severe accidents.
- However, on the basis of the NRC staff's findings from the review of the IPE submittal and associated information, we are unable to conclude that the SSES IPE results are reasonable, given the SSES design, operation, and history.
The NRC staff has documented weaknesses in all three review areas, which include the front-end analysis, the HRA, and the back-end analysis.
The NRC staff is concerned that the IPE's weaknesses have influenced the ranking of the resultant accident sequences as well as the overall CDF.
Furthermore, the NRC staff is concerned that lack of explicit delineation of CCFs, all human
Ct
)
t <i-
actions credited, and several important phenomena in the back-end analysis may have resulted in an incomplete and potentially skewed IPE model, one which may.
not allow an assessment of the impact of modeling assumptions and data used on the IPE's results through sensitivity and importance analyses.
Therefore, the NRC staff has concerns about PPKL's quantitative understanding of the contribution of CCFs, human failures, and containment performance to plant risk.
The NRC staff concludes that the SSES IPE does not meet the remaining two general objectives of the IPE program:
2.
To understand the most likely severe accident sequences that
'ould occur at the plant.
3.
To gain a more'uantitative understanding of the overall probabilities of core damage and fission product releases.
Therefore, the NRC staff cannot conclude that the SSES-IPE has met the intent of GL 88-20.
If you have any questions regarding the
- SSES, Units 1 and 2, evaluation, of your IPE submittal, please contact me at (301) 415-1402, and I will arrange discussions with RES (the principal contributor to the SE).
The NRC staff will entertain, dialogue about options for a supplemental submittal that will meet the objectives noted above.
Docket Nos. 50-387/50-388
Enclosure:
As stated cc w/encl:
See next page DISTRIBUTION Sincerely,
/s/
Chester
- Poslusny, Senior Project Manager Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Docket File PUBLIC PDI-2 Reading BBoger JStolz CPoslusny CAnderson, RGN-I HO'Brien THarris (E-Hail SE)
RHernan HDrouin OGC MCunningham s
C>~ ~C~O OFFICE PDI-2 M
NAME CPoslusn
- rb PD -2 O'rien Jst 2
D DATE
%0 ~ 97 0
2.~97 9
2097 OFFICIAL RECORD COPY DOCUMENT NAME:
SU74478.GEN
I I,
R.
Byram actions credited, and several important phenomena in the back-end analysis may have resulted in an incomplete and potentially skewed IPE model, one which may not allow an assessment of the impact of modeling assumptions and data used on the IPE's results through sensitivity and importance analyses.
Therefore, the NRC staff has concerns about PP8L's quantitative understanding of the contribution of CCFs, human failures, and containment performance to plant risk.
The NRC staff concludes that the SSES IPE does not meet the remaining two general objectives of the IPE program:
2.
To understand the most likely severe accident sequences that could occur at the plant.
3.
To gain a more quantitative understanding of the overall probabilities of core damage and fission product releases.
Therefore, the NRC staff cannot conclude that the SSES IPE has met the intent of GL 88-20.
If you have any questions regarding the
- SSES, Units I and 2, evaluation of your IPE submittal, please contact me at (301) 415-1402, and I will.arrange discussions with RES (the principal contributor to the SE).
The NRC staff will entertain dialogue about options for a supplemental submittal that will meet the objectives noted above.
Sincerely, Docket Nos. 50-387/50-388
Enclosure:
As stated cc w/encl:
See next page Chester Poslu y, Senior Project Manager Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
t
rector-Bureau of Radiation Protection Pennsylvania Department of Environmental Resources P. 0.
Box 8469 Harrisburg, Pennsylvania 17105-8469 Hr. Jesse C. Tilton, III Allegheny Elec. Cooperative, Inc.
212 Locust Street P.O.
Box 1266 Harrisburg, Pennsylvania 17108-1266 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 General Manager Susquehanna Steam Electric Station Pennsylvania Power and Light Company Box 467 Berwick, Pennsylvania 18603 Hr. Herbert D. Woodeshick Special Office of the President Pennsylvania Power and Light Company Rural Route 1,
Box 1797 Berwick, Pennsylvania 18603 George T. Jones Vice-President-Nuclear Operations Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Dr. Judith Johnsrud National Energy Committee Sierra Club 433 Orlando Avenue State College, PA 16803 Chairman Board of Supervisors 738 East Third Street
- Berwick, PA 18603
0
~u i 0