ML063060257
| ML063060257 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/29/2006 |
| From: | Hernandez S NRC/NRR/ADRO/DLR/REBB |
| To: | Kansler M Entergy Nuclear Operations |
| Hernandez S, NRR/DLR/REBB, 415-4049 | |
| References | |
| TAC MD2667 | |
| Download: ML063060257 (11) | |
Text
November 29, 2006 Mr. Michael R. Kansler President Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING SEVERE ACCIDENT MITIGATION ALTERNATIVES FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT (TAC NO. MD2667)
Dear Mr. Kansler:
The U.S. Nuclear Regulatory Commission staff has reviewed the Severe Accident Mitigation Alternatives analysis submitted by Entergy Nuclear Operations, Inc., in support of its application for license renewal for the James A. FitzPatrick Nuclear Power Plant, and has identified areas where additional information is needed to complete its review. Enclosed are the staffs request for additional information.
We request that you provide your responses to these questions within 60 days of the date of this letter, in order to support the license renewal review schedule. If you have any questions, please contact me at 301-415-4049 or via email at shq@nrc.gov.
Sincerely,
/RA/
Samuel Hernandez Environmental Project Manager Environmental Branch B Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-333
Enclosure:
As stated cc w/encl: See next page
ML063060257 Document name: C:\\FileNet\\ML063060257.wpd OFFICE LA:DLR GS:DLR:REBB PM:DLR:REBB NAME S. Figueroa J. Muir S. Hernandez DATE 11/2/06 11/6/06 11/29/06
Letter to M. Kansler, from Samuel Hernandez, dated November 29, 2006
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING SEVERE ACCIDENT MITIGATION ALTERNATIVES FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT (TAC NO. MD2667)
DISTRIBUTION:
Email:
F. Gillespie / P.T. Kuo (RidsNrrDlr)
M. Rubin (RidsNrrDraApla)
R. Franovich (RidsNrrDlrRebb)
E. Benner (RidsNrrDlrReba)
B. Palla R. Schaaf S. Hernandez J. Muir R. Emch B. Mcdowell, LLNL (mcdowell5@llnl.gov)
N. Le J. Boska E. Cobey, RI G. Hunegs, RI DLR/REBB DLR/REBA
Request for Additional Information Regarding the Analysis of Severe Accident Mitigation Alternatives for the James A. Fitzpatrick Nuclear Power Plant 1.
Provide the following information regarding the development of the James A. Fitzpatrick Nuclear Power Plant (JAFNPP) Probabilistic Safety Analysis (PSA) used for the Scientific Apparatus Makers Association (SAMA) analysis, i.e., Revision 2, October 2004:
1.
Section E.1.4 discusses the differences between the Individual Plant Examination (IPE) and Revision 2, but does not discuss Revision 1. Provide additional detail about Revision 1, including:
1.
the date of Revision 1 2.
a summary of the changes between Revision 0 and Revision 1, and 3.
the total Core Damage Frequency (CDF) and the major contributors to CDF.
2.
Section E.1.4 states that the results of the peer review process are included in Section 5 of the Individual Plant Examination for Internal Events, Revision 2, October 2004.
Provide a copy of Section 5.
3.
The contributions to CDF (Section E.1) are very different from those in the IPE. Provide the reason for the changes in the major contributors to the CDF between each version of the PSA (i.e., the IPE, Revision 1 and Revision 2).
4.
It is stated that the PSA represents the plant operating configuration and design changes as of December 2003, and component failure and unavailability data as of December 2002. Identify any changes to the plant (physical and procedural modifications) since December 2003 that could have a significant impact on the results of the PSA and/or the SAMA analyses. Provide a qualitative assessment of their impact on the PSA and their potential impact on the results of the SAMA evaluation.
5.
There are inconsistencies in the CDF values reported in the ER, specifically:
The baseline CDF and the total release frequency used in the SAMA analysis is indicated to be 2.74x10-6 (page 4-36, and Table E.1-10, respectively), however, the point estimate CDF appears to be 3.11x10-6 (Table E.1-1) and the mean CDF is given as 3.70x10-6 (Table E.1-3).
The text below Table E.1-3 states that the ratio of the 95th percentile to the mean is 3.83; however, the ratio is actually 2.84.
Rectify these inconsistencies. If the mean CDF is different from the CDF used in the SAMA analyses, include an explanation of why the mean CDF was not used.
6.
Event IE-T1, loss of offsite power, has a risk reduction worth (RRW) of 2.316 and frequency of 3.5x10-2 (Table E.1-2). Thus, the contribution of this event alone to CDF is about 57%. Given that station blackout (SBO) is 43% of CDF (Table E.1-1), the loss of offsite power is a major contributor to other accident types shown in Table E.1-1.
However, all of the Phase II SAMAs listed for the event IE-T1 address provisions related to SBO, and do not address the contribution of this event to other accident types. Identify the other major accident types where the event IE-T1 is a contributor, and the SAMAs that address these accidents.
7.
Table E.1-2 identifies numerous Phase I SAMAs that have been implemented. For each of the Phase I SAMAs noted, confirm whether credit has been taken for the improvement in the current PSA.
8.
On page E.1-85, it is indicated that the accident sequence quantification truncation limit was reduced from 10-9 to 10-11/yr in response to peer review recommendations. Provide the truncation value used in each PSA revision, and where available, the change in CDF due to this reduction.
2.
Provide the following information relative to the Level 2 analysis:
1.
Describe the process and assumptions used to group the numerous source terms for internal initiators into a much smaller number of source term groups.
2.
The JAFNPP source terms (Table E.1-11) appear amenable to grouping into three release time categories, e.g., 0-8, 8-24, and 24+ hours. However, the SAMA analysis uses only two, i.e., <24 and >24 hrs. Explain the rationale for this coarse grouping, and how the use of just two release time categories affects the results of the population dose and risk reduction estimates. Discuss whether this provides a conservative or non-conservative bias, and the magnitude of this bias.
3.
Interfacing-systems loss-of-coolant accident (ISLOCA) events contribute 28% of the large early release, and are assigned to the Early release mode in Table E.1-15.
Provide the estimated population dose and offsite economic cost risk values for an ISLOCA event and a comparison of these values to those for the Early release mode.
4.
It is stated on pg. E.1-60 that releases are integrated over a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> period following reactor pressure vessel (RPV) failure or event initiation (if no RPV failure occurs).
However, information in Table E.1-11 indicates that late releases do not start until about 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and continue until about 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />, and early releases continue until about 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. Thus, it appears that the later portions of the releases may not be included in the total integrated release. Explain the overall accident time frame for the assessment, and justify that the release fractions reported in Table E.1-11 include the majority of the fission products released for each release category.
3.
Provide the following information with regard to the treatment and inclusion of external events in the SAMA analysis, :
1.
The individual plant examination of external events (IPEEE) fire CDF value is 2.56x10-5 /y (page 4-46). However, the emergency response (ER) states that a more realistic value may be closer to one third this value, or 8.53x10-6 /y, based on the conservative assumptions listed on pages 4-46 and 4-47. These assumptions are generic and qualitative in nature. Provide a quantitative analysis of the conservative assumptions that justifies the factor of three reduction.
2.
In the ER, the potential for risk reduction in external events is considered in the context of an upper bound assessment in which the internal event benefits are increased by a factor of 16 to account for the combined effect of external events and analysis uncertainties. The impact of external events should be reflected in the baseline evaluation independent of the uncertainty assessment. Provide a revised baseline evaluation (using a 7 percent discount rate) that accounts for risk reduction in both internal and external events, and an alternative case using the 3 percent discount rate.
Reflect any corrections in the multiplier that may have resulted from addressing RAI 3a.
3.
Provide an assessment of the impact on the revised baseline evaluation if SAMA benefits are increased to account for uncertainties in the analysis. Reflect any corrections in the multiplier that may have resulted from addressing RAI 1e.
4.
In Section E.1.3.2, Entergy states that a number of fire-related improvements were identified and that these improvements have been implemented. Despite these improvements, the fire zone CDF is a factor of three to nine greater than the current internal events CDF, depending on the level of conservatism assumed in the IPEEE fire analysis. SAMA candidates identified based on internal risk contributors will not necessarily address the fire risk. Describe any further efforts made to determine if any SAMA candidates exist to address fire risk contributors beyond those already identified in the IPEEE, and explain why the fire CDFs cannot be further reduced in a cost-effective manner.
5.
NUREG-1742 states that there is a fire-induced seismic vulnerability due to failure of the hydrogen line in the turbine building. Given that the fire analysis shows a contribution to risk from a fire in the turbine building, provide details on actions taken to reduce the risk due to a seismic-induced fire, and whether a SAMA to further reduce the risk from this event is cost-beneficial.
4.
Provide the following information concerning the MACCS2 analyses:
1.
The meteorological data used was obtained from the Nine Mile Point/JAFNPP meteorological monitoring system and regional National Weather Services stations.
Identify where the monitoring stations are located relative to JAFNPP.
2.
The baseline evacuation speed (2.0 meters/s) is an average of the maximum and minimum speeds, and is said to be conservative. Explain why the average value is considered conservative. Explain why this value is not the same as for Nine Mile Point, which used 1.8 meters/second.
3.
Entergy assumed 100% evacuation within the emergency planning zone (EPZ), which is non-conservative. NUREG-1150 assumed a 99.5% evacuation within the EPZ, and previous SAMA analyses (including Nine Mile Point) have assumed 95% evacuation.
Address the potential impact on the off-site exposure risk and averted public exposure cost if 5% of the population fails to evacuate the EPZ.
4.
The MAACS2 analysis is based on a core inventory scaled by power, with an increase of 25% for long half-life nuclides. Clarify whether the 25% increase of the long half-life nuclides is from the generic burnup data in MAACS2, and whether the scaling is considered a correction for the JAFNPP specific burnup. Confirm that the adjusted core inventory adequately reflects the fuel enrichment and burnup expected at JAFNPP.
5.
Provide the full reference identifications for the following:
1.
National Hydrology Dataset 2.
U.S. Department of Agriculture, 2002 3.
Census of Agriculture, 2002 4.
New York Statistical Information System 5.
Northern New York Travel and Tourism Center 6.
MACCS2 version used 7.
Evacuation travel time estimate study 5.
Provide the following with regard to the SAMA identification and screening process:
1.
SAMA 61 was evaluated by eliminating failures of both DC battery chargers, i.e.,
71BC-1A and 71BC-1B. These events do not appear in Table E.1-2. Provide the risk reduction worth and the probability of failure of these events.
2.
Table E.1-2 indicates that Phase I SAMAs to improve procedures and training have been implemented to address event NR-LOSP-7HR. In spite of these improvements, this event is the highest risk reduction worth ranked non-initiator event. No Phase II SAMAs were recommended for this event. The risk reduction worth of this event can be reduced by reducing the probability of the event itself, or by implementation of SAMAs for other events in the accident sequences that include the event NR-LOSP-7HR (e.g., if emergency diesels were a "perfect" back-up, the RRW of NR-LOSP-7HR would be zero). Hence:
i.
Given that the human error contribution to the probability of the event has been addressed, identify and evaluate hardware-based SAMAs that could facilitate recovery of offsite power (e.g., automatic switching gear, redundant switching gear, or other hardware that facilitates operator recovery of offsite power, given the grid is available within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />).
ii.
Identify other events in the accident sequences involving power recovery that would reduce the risk reduction worth of NR-LOSP-7R, and identify and evaluate SAMAs for these events. Such events may already be identified in the Phase I and/or Phase II SAMAs, but not attributed and/or evaluated for NR-LOSP-7HR (e.g., SAMAs 26 through 36). Identify and evaluate potential SAMAs that might lower the importance of this event.
3.
Table E.1-2 indicates that SAMA 57 was evaluated to address event NVP-XHE-FO-LVENT (operator fails to initiate local containment vent). This SAMA, controlling containment venting within a narrow pressure band, would be subject to the same failure to vent human error as in the basic event. Conversion of the containment vent system to a passive design would appear to be more effective in reducing the risk from this event. Provide an evaluation of the costs and benefits of converting the vent system to a passive design.
4.
SAMA 53 was evaluated based on reducing CDF due to reactor protection system (RPS) failure (event C). However, event C, which has a RRW of 1.057, also has a large Risk Achievement Worth. The staff estimates that an order-of-magnitude increase in this event would increase CDF by 48%. Provide an assessment of SAMA 53 in the context of preventing an increase in likelihood of event C. Also, consider other SAMAs that could prevent a degradation of the RPS over time.
5.
Table E.1-2 indicates that SAMA 49 was considered to address event FXT-ENG-FR-76P1. This SAMA involves the addition of an entire new system. The addition of a redundant diesel fire pump would appear to be more cost effective.
Provide an evaluation of the costs and benefits of adding a redundant diesel fire pump, in lieu of SAMA 49.
6.
Table E.1-2 identifies a procedure change for operator action to be taken within 5 minutes of fire pump starting to ascertain whether flooding is occurring in the relay room (event IE-RRFLOOD). Explain why 5 minutes was used and whether corrective action is required within 5 minutes. Provide an assessment of the potential impact if no action is taken for some longer period. Justify why no additional SAMAs were identified to address internal flooding events.
7.
Table E.2-1 indicates that SAMAs 8, 14, and 22 were modeled by assuming that reactor building failures were completely eliminated. However, no reduction in population dose are shown. Explain this apparent contradiction.
6.
Provide the following with regard to the Phase II cost-benefit evaluations:
1.
For a number of the Phase II SAMAs listed in Table E.2-1, the information provided does not sufficiently describe the associated modifications and what is included in the cost estimate. Provide a more detailed description of the modifications for Phase II SAMAs 18, 26, 27, 28, 30, 36, 38, 52, 59, and 60.
2.
In Table E.2-1, the percent change in CDF and population dose is reported for each analysis case. However, the change in the offsite economic cost risk (OECR) is not reported. Provide the change in the OECR for each analysis case.
3.
In Table E.2-1, SAMA 25 is indicated to provide no CDF reduction. Explain why the CDF reduction would not be equivalent to that for SAMAs 11 and 17.
4.
SAMA 57, control containment venting within a narrow band of pressure, is intended to eliminate failures associated with successful venting. The benefit of this SAMA was determined by reducing the operator failure to vent by a factor of three. It is not clear that reducing the failure to vent probability is related to the actual benefit from this SAMA. Also, the cost of $400,000 appears high for what appears to be a procedure and training issue. Justify the benefit and cost for this SAMA.
5.
The ER does not provide any indication of Entergys plans regarding the five Phase II SAMAs found to be potentially cost-beneficial (Table 4-4). Describe Entergys plans regarding these SAMAs, and any other potentially cost-beneficial SAMAs that may emerge from further analyses in response to these RAIs.
6.
Several Phase II SAMAs in Table E.2-1 provide a CDF and/or offsite dose reduction, but an estimated benefit of $0, i.e., SAMAs 43, 50, 53, 55, and 56. Provide the estimated benefits for these SAMAs.
7.
For certain SAMAs considered in the ER, there may be lower-cost alternatives that could achieve much of the risk reduction at a lower cost. It this regard, discuss whether any lower-cost alternatives to those Phase II SAMAs considered in the ER, would be viable and potentially cost-beneficial. Evaluate the following SAMAs (previously found to be potentially cost-beneficial at other plants), or indicate if the particular SAMA has already been considered.
If the latter, indicate whether the SAMA has been implemented or has been determined to not be cost-beneficial at JAFNPP.
1.
Enhance dc power availability (provide cables from diesel generators or another source to directly power battery chargers).
2.
Provide alternate dc feeds (using a portable generator) to panels supplied only by the dc bus.
3.
Modify procedures and training to allow operators to cross-tie emergency ac buses under emergency conditions which require operation of critical equipment.
4.
Develop guidance/procedures for local, manual control of reactor core isolation cooling following loss of dc power.
5.
Enhance loss of service water procedure to provide more specific guidance to deal with or prevent a complete loss of the system.
6.
Manual venting of containment using either a local hand wheel or gas bottle supplies (considered for Nine Mile Point Unit 1) as a possible alternative for containment pressure control.
FitzPatrick Nuclear Power Plant cc:
Mr. Gary J. Taylor Chief Executive Officer Entergy Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Mr. John T. Herron Sr. VP and Chief Operating Officer Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Peter T. Dietrich Site Vice President Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093 Mr. Kevin J. Mulligan General Manager, Plant Operations Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093 Mr. Oscar Limpias Vice President Engineering Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Christopher Schwarz Vice President, Operations Support Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. John F. McCann Director, Licensing Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Ms. Charlene D. Faison Manager, Licensing Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Michael J. Colomb Director of Oversight Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. David Wallace Director, Nuclear Safety Assurance Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093 Mr. James Costedio Manager, Regulatory Compliance Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093 Assistant General Counsel Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector's Office James A. FitzPatrick Nuclear Power Plant U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093
FitzPatrick Nuclear Power Plant cc:
Mr. Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Peter R. Smith, President New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy New York State Dept. of Public Service 3 Empire State Plaza Albany, NY 12223-1350 Oswego County Administrator Mr. Steven Lyman 46 East Bridge Street Oswego, NY 13126 Supervisor Town of Scriba Route 8, Box 382 Oswego, NY 13126 Mr. James H. Sniezek BWR SRC Consultant 5486 Nithsdale Drive Salisbury, MD 21801-2490 Mr. Michael D. Lyster BWR SRC Consultant 5931 Barclay Lane Naples, FL 34110-7306 Mr. Garrett D. Edwards 814 Waverly Road Kennett Square, PA 19348