ML110670379
| ML110670379 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/10/2011 |
| From: | Dan Doyle License Renewal Projects Branch 1 |
| To: | Gambhir S Energy Northwest |
| Doyle D, NRR/DLR, 415-3748 | |
| References | |
| TAC ME3121 | |
| Download: ML110670379 (6) | |
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~O lh***i' Mr. S. K. Gambhir, Vice President, Engineering Columbia Generating Station Energy Northwest MD PE04 P. O. Box 968 Richland, WA 99352
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE COLUMBIA GENERATING STATION LICENSE RENEWAL APPLICATION SEVERE ACCIDENT MITIGATION ALTERNATIVE REVIEW (TAC NO. ME3121)
Dear Mr. Gambhir:
By letter dated January 19,2010, Energy Northwest submitted an application to the U.S.
Nuclear Regulatory Commission (NRC or the staff) to renew Operating License NPF-21 for Columbia Generating Station pursuant to Title 10 of the Code of Federal Regulations Part 54.
The NRC staff is reviewing the information contained in the license renewal application and the associated Environmental Report. The staff has identified, in the enclosure, areas where additional information is needed to complete the Severe Accident Mitigation Alternatives review.
Further requests for additional information may be issued in the future.
Items in the enclosure were discussed with Mr. Abbas Mostala. A mutually agreeable date for the response is within 60 days from the date of this letter. If you have any questions, please contact me at 301-415-3748 or bye-mail at daniel.doyle@nrc.gov.
Sincerely, Daniel Doyle, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosure:
As stated cc w/encl: Listserv
Request for Additional Information Regarding the Analysis of Severe Accident Mitigation Alternatives for the Columbia Generating Station License Renewal Review
Background:
The U.S. Nuclear Regulatory Commission (NRC) issued previous requests for additional information (RAI) related to the Columbia Generating Station (CGS) Severe Accident Mitigation Alternatives (SAMA) review to Energy Northwest by letter dated July 1, 2010 (ADAMS Accession Number ML101760421). Energy Northwest provided a partial response to the RAls by letter dated September 17, 2010 (ADAMS Accession Number ML102660151). The NRC issued two subsequent RAI letters to provide clarification on the Energy Northwest partial response (ADAMS Accession Numbers ML102870984 and ML103330246). Energy Northwest provided a second RAI response letter dated January 28,2011 (ADAMS Accession Number ML110330395).
The purpose of these RAls is to provide clarification on Energy Northwest's response dated January 28, 2011. All of the RAls in this letter refer to the January 28 RAI response letter.
Requests:
RA11:
Table 8-4 does not provide an analysis of SAMA CC-21, which was screened as Criterion C. If modeled similar to SAM A CP-01, SAMA CC-21 (procedure change) would be cost-beneficial.
Clarify the disposition of this SAMA.
RAI2:
Tables A-10, A-12, and A-14 provide a large early release frequency (LERF) importance analysis for internal, fire, and seismic events, respectively, and associated SAMA assessment.
Tables A-6 through A-7 show that release category Mil is generally a much more significant contributor to population dose/economic impact than the LERF (H/E) release category, with release category HII also being a significant contributor. Clarify how releases categories Mil and HII are considered in the LERF importance analysis.
RAI3:
The Level 1 and Level 2 seismic basic events importance lists (Tables A-13 and A-14) identify, in addition to the two initiating events, only a few basic events, and those identified appeared to be flag events, split fractions, or success terms. Neither seismically-induced failures nor random failures appear to be addressed in this importance analysis. Clarify how the seismic importance lists were developed. In the response, specifically discuss how the seismic probabilistic safety assessment (PSA) model treats both seismically-induced failures and random failures. If random failures are not included in the seismic analysis, explain how this model incompleteness impacts the SAMA evaluation.
ENCLOSURE
- 2 RAI4:
Table A-1 presents a total fire core damage frequency (CDF) of 3.6E-6/yr on the Rev. 6.2 Model column header, but the contributing fire sequences under that column header sum to 3.92E-6/yr. Environmental Report Table E.3-1, on the other hand, presents a total fire CDF of 7AE-6/yr and Table E.4-5 presents release categories that appear to support (Le., frequencies when summed equals 7 AE-6/yr) that total. Clarify these discrepancies.
RAI5:
The truncation limits for internal events, fire and seismic models used in the quantification of Revision 6.2 Level 1 and Level 2 CDFs range from 5 x 10.14 to 1 X 10.8* In response to an NRC staff RAJ (September 17, 2010), Energy Northwest explained that in general a four-order difference between the calculated total and truncation limit was maintained, except in a few cases where a lesser difference was appropriate. In a telephone clarification, Energy Northwest further explained that the expression "appropriate" referred to cases in which the calculated CDF appeared to converge using a lower truncation limit. Clarify if the following statement is applicable for both the Revision 6.2 and 7.1 PSA models: "In general a four-order-of-magnitude difference between the calculated total and truncation limit was maintained, except in a few cases where a lower truncation limit resulted in convergence between the calculated CDF and truncation limit."
RAI6:
The fire events listed in Table A-1 are almost entirely different from the fire events listed in Table E.3-7 of the Environmental Report. It appears that the Table A-1 fire events are identified by initiating event category rather than fire compartment (although the Table A-1 column header uses the term "Fire Compartment"). Clarify the difference between the fire events listed in the Environmental Report and table A-1 of the RAI response.
RAI7:
Additional Comment #2 discussed in the January 19, 2011, conference call (ADAMS Accession Number ML110400510) does not appear to have been addressed. The Phase I screening for SAMAs AC/DC-OS, CB-02, CB-05, CC-13, and FR-02 need to be re-evaluated based on the total risk reduction benefit and associated implementation cost.
RAJ 8:
Comment #2 discussed in the January 19, 2011, conference call (ADAMS Accession Number ML110400510) does not appear to have been entirely addressed. Explain the reason for the increase in fire population dose risk for SAMAs CW-02, CW-03, and CW-04 (Analysis Cases 18 and 19 in Table B-3) and the increase in internal events CDF and population dose-risk for SAMA AC/DC-30R (Analysis Case 45 in Table B-2).
RAI9:
The calculated total for the internal, fire, and seismic events listed for the release categories presented in Tables A-3, A-4, and A-5 (5.61 E-06/yr, 1.02E-OS/yr, and 4.31 E-06/yr respectively)
- 3 are not the same as the total CDFs given for internal, fire, and seismic events in Table A-1 (7.4E-6/yr, 1.4E-6/yr, and 4.9E-6/yr respectively). Explain these differences. Also, the percentage contributions presented in Tables A-3, A-4, and A-5 total to much less than 100%
for each table (e.g., totals to 75% in the case of the internal events release categories).
RA110:
Table A-1 (seismic) shows that the CDF for a couple of the seismic damage states (i.e., S2P2, S20P2) was completely eliminated in PSA Rev. 7.1. Explain.
RA111:
Section 2.2 provides a sensitivity analysis of the assumed 0.3 hot short probability (if CPTs were known to be present for the circuits; otherwise, 0.6) for three selected SAMAs that address fire events. The basis for selecting the three SAMAs is the risk reduction worth significance of the hot shorts they address and that they address numerous important functions. Clarify Energy Northwest's basis for believing that the sensitivity analysis results for these three SAMAs bound the effect for other fire SAMAs. In the response, specifically address the potential for multiple hot shorts in series and whether the factor of 2 impact determined for SAMA FR-07b is bounding for the fire SAMAs. Alternatively, specifically assess the impact of using a 0.6 hot short probability (or 0.3 if these circuits are known to be protected by CPTs) on the analysis results for fire-related SAMAs FR-OB, FR-09R, FR-12R, and FR-11 R.
Also, the hot short probability assumption could result in an underestimate of the estimated risk reduction for SAMAs identified principally to address internal events if the SAMA addresses cutsets that contain hot shorts. Assess the impact of using a 0.6 hot short probability (or 0.3 if these circuits are known to be protected by CPTs) on the analysis results for non-fire-related SAMAs AC/DC-15, AC/DC-23, AC/DC-27, CC-02, CP-01, CW-02, CW-07, CC-24R, FW-05R, and OT-09R, which have significant fire risk reduction contribution to the total estimated benefit.
RA112:
Table 2-3 notes that the "Late" time category (Le., greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) is not used in PSA model Rev. 7.1. Clarify that all Level 2 sequences are mapped into "early" or "intermediate" release categories. If not, assess the impact of this incompleteness on the results of the sensitivity study.
March 10,2011 Mr. S. K. Gambhir, Vice President, Engineering Columbia Generating Station Energy Northwest MD PE04 P. O. Box 968 Richland, WA 99352
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE COLUMBIA GENERATING STATION LICENSE RENEWAL APPLICATION SEVERE ACCIDENT MITIGATION ALTERNATIVE REVIEW (TAC NO. ME3121)
Dear Mr. Gambhir:
By letter dated January 19, 2010, Energy Northwest submitted an application to the U.S.
Nuclear Regulatory Commission (NRC or the staff) to renew Operating License NPF-21 for Columbia Generating Station pursuant to Title 10 of the Code of Federal Regulations Part 54.
The NRC staff is reviewing the information contained in the license renewal application and the associated Environmental Report. The staff has identified, in the enclosure, areas where additional information is needed to complete the Severe Accident Mitigation Alternatives review.
Further requests for additional information may be issued in the future.
Items in the enclosure were discussed with Mr. Abbas Mostala. A mutually agreeable date for the response is within 60 days from the date of this letter. If you have any questions, please contact me at 301-415-3748 or bye-mail at daniel.doyle@nrc.gov.
Sincerely, IRA!
Daniel Doyle, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosure:
As stated cc w/encl: Listserv ADAMS Accession No ML110670379 I OFFICE LA:DLR PM:RPB1 :DLR BC:RPB1 :DLR PM:RPB1 :DLR
. NAME SFigueroa
- DDoyle BPham DDoyle DATE 03/09/2011 03/09/2011 03/09/2011 03/10/2011 OFFICIAL RECORD COpy
Letter to S. Gambhir from D. Doyle dated March 10, 2011
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE COLUMBIA GENERATING STATION LICENSE RENEWAL APPLICATION SEVERE ACCIDENT MITIGATION ALTERNATIVE REVIEW (TAC NO. ME3121)
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