ML17292B638
| ML17292B638 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 04/19/1999 |
| From: | Jack Cushing NRC (Affiliation Not Assigned) |
| To: | Parrish J WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| GL-88-20, TAC-M83695, NUDOCS 9904220048 | |
| Download: ML17292B638 (10) | |
Text
April 19, 1999 Mr. J. V. Parrish Chief Executive Officer Washington Public Power Supply System P.O. Box 968 (Mail Drop 1023)
Richland, Washington 99352-0968 It
SUBJECT:
SUPPLEMENTAL REQUEST FOR ADDITIONALINFORMATION(RAI)
REGARDING GENERIC LETTER 88-20 ATWNP-2 (TAC NO. M83695)
Dear Mr. Parrish:
The NRC staff has reviewed your response dated March 30, 1998, to our request for additional information regarding Generic Letter 88-20, Supplement 4, "Individual Plant Examination for External Events." As a result of the review, the staff has determined'that additional information
,is needed to complete the review. The information needed is detailed in the enclosure.
The enclosed request was discussed with Mr. Inserra of your staff on April 12, 1999. A mutually agreeable target date of June.18, 1999 was established for responding to the RAI. If circumstances result in the need to revise the target date, please call me at the earliest opportunity at (301) 415-1424.
Sincerely, ORIG.
S.IGN<D PY Jack Gushing, I ro]ect Manager, Section 2 Project Directorate IV 8 Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosure:
Request for Additional Information cc w/encl: See next page DISTRIBUTION:
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 2055&0001 April 19, 1999 Mr. J. V. Parrish Chief Executive Officer Washington Public Power Supply System P.O. Box 968 (Mail Drop 1023)
Richland, Washington 99352-0968
SUBJECT:
SUPPLEMENTAL REQUEST FOR ADDITIONALINFORMATION(RAI)
REGARDING GENERIC LETTER 88-20 ATWNP-2 (TAC NO. M83695)
Dear Mr. Parrish:
The NRC staff has reviewed your response dated March 30, 1998, to our request for additional information regarding Generic Letter 88-20, Supplement 4, "Individual Plant Examination for External Events." As a result of the review, the staff has determined that additional information is needed to complete the review. The information needed is detailed in the enclosure.
The enclosed request was discussed with Mr. Inserra of your staff on April 12, 1999. A mutually agreeable target date of June 18, 1999 was established for responding to the RAI. If circumstances result in the need to revise the target date, please call me at the earliest opportunity at (301) 415-1424.
Sincerely, ack Gushing, Project Manager, Section 2 Project Directorate IV 8 Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosure:
Request for Additional Information cc w/encl: See next page
Nuclear Project No. 2 0 cc w/encl:
Mr. Greg O. Smith (Mail Drop 927M)
Vice President, Generation Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352-0968 Mr. Albert E. Mouncer (Mail Drop 1396)
Chief Counsel Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 Ms. Deborah J. Ross, Chairman Energy Facility Site Evaluation Council P. O. Box 43172 Olympia, Washington 98504-3172 Mr. D. W. Coleman (Mail Drop PE20)
Regulatory Affairs Manager Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 Mr. Paul Inserra (Mail Drop PE20)
Manager, Licensing Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower 8 Pavilion 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 Chairman Benton County Board of Commissioners P.O. Box 69 Prosser, Washington 99350-0190 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 69 Richland, Washington 99352-0968 Mr. Rodney L. Webring (Mail Drop PE08)
Vice President, Operations Support/PIO Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Perry D. Robinson, Esq.
Winston 8 Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 Mr. Bob Nichols Executive Policy Division Office of the Governor P.O. Box 43113 Olympia, Washington 98504-3113
SUPPLEMENTAL RE VEST FOR ADDITIONALINFORMATION RESPONSE TO GENERIC LETTER 88-20 "INDIVIDUALPLANT EXAMINATIONFOR EXTERNALEVENTS" WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 NP-2)
DOCKET NO. 50-397 Seismic According to your response to the previous seismic question 1, regarding the calculation of mean hazard curves, the figure provided on page 15 indicates that the calculated mean exceedance probability at 0.3g is not consistent with the probability density function, which was also provided in the same figure. Since a logarithmic scale is used for the annual exceedance probabilities, the location of the mean exceedance probability should be shifted near the 85-th percentile curve. The mean seismic hazard, if calculated correctly, would be about two times larger than the value used in the WNP 2 IPEEE analysis.
Please provide the seismic probabilistic risk assessment (PRA) results obtained with corrected and revised mean hazard curves (e.g., revised seismic core damage frequency (CDF) calculations, identification of dominant contributing sequences and seismic failures).
The following addresses your response to the previous seismic question 2, regarding the initial screening of components. The list of screened components provided in Attachment 1 indicates that storage tanks (e.g., emergency diesel generator (EDG) day tanks) and transformers were screened based on walkdown observation, generic calculation and judgment that these components are generally rugged.
In past seismic PRA studies, however, these components were often identified as weak links.
Please describe the rationale for this screening, such as unique seismic design features of the WNP-2 plant, when compared with other plants.
3.
According to your response to the previous seismic question 3, regarding the comparison of new and old floor spectra, the comparisons provided indicate that the 0.25g design basis earthquake (DBE) spectra do not envelope the 0.5g median spectra at the higher frequency range (higher than 10 Hz), particularly for the diesel generator building and turbine building.
Please explain why this optimistic assumption does not impact the component screening and the calculated plant CDF.
4.
Regarding your response to seismic question 4, structural fragilityanalysis, the EQE report provided in Attachment 3 is a very brief outline of fragility analysis, and does not provide any more details than the submittal.
To confirm the validity of the screening performed, please provide the detailed structural fragilitycalculation packages for the following components:
- Reactor building, overturning moment of biological shield wall (0.51g), and
- Turbine building, shear on column line 13 wall (0.51g).
Your IPEEE submittal and responses to previous seismic questions 5 and 6 indicated that a large number of structures and components were screened based on the generic evaluation of seismic design margin. However, the described generic screening criteria are not consistent with.the guidance cited in NUREG-1407 (Section 3.1.1.3) for seismic PRA fragilityestimation.
Further clarification is provided below.
Generic screenin criteria for flexible com onents....... The response relating to the assumed response factor, FRE, stated that a factor of 1.28 was assumed for question 5b, and also a factor of 2.0 was assumed separately for question 5c. Therefore, in the fragility calculation, a factor of 2.56 (=1.28 x 2.0) was assumed to represent the ratio of the SSE demand to the median demand.
This assumption is not consistent with the provided comparison of old and new floor spectra.
Avera e demand/code allowable ratio..... The response relating to the ratio of average demand to code allowable stated that the assumed value of 0.7 for this ratio is considered to be conservative.
This observation is not consistent with the seismic design practice in the nuclear industry, particularly for anchorages and seismic support structures.
Generic screenin criteria for ri id e ui ment......ln the response to the previous question 6 (pp.27-29), a median fragilityvalue of 0.86g was estimated.
Then, an additional capacity of 0.19g was considered to account for the horizontal resistance afforded by friction, to obtain the total fragilityof 1.05g. Addition of friction forces to the calculated capacity values has not been accepted in past fragilityanalyses.
Please identify the screened structures and components which would not have been screened ifthe generic screening criteria discussed above had not been used, and provide fragilityestimates for those components.
Please'also describe how the seismic CDF estimate, the identification of dominant contributing accident scenarios to the seismic CDF, and the identification of dominant seismic failures, would be affected by incorporation of these additional components into the seismic PRA model.
r 4 Fire There is no supplemental RAI in the fire area.
Hi h Wind Flood and Other External Events HFO 4
There is no supplemental RAI in the HFO area.
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