ML070370023
| ML070370023 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 02/23/2007 |
| From: | Le N NRC/NRR/ADRO/DLR/RLRB |
| To: | Peter Dietrich Entergy Nuclear Operations |
| Le, Ngoc, NRR/DLR/RLRA, 415-1458 | |
| References | |
| TAC MD2666 | |
| Download: ML070370023 (7) | |
Text
February 23, 2007 Mr. Peter T. Dietrich Site Vice President James A. FitzPatrick Nuclear Power Plant Entergy Nuclear Operation, Inc.
P.O. Box 110 Lycoming, NY 13093-0110
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION REGARDING THE REVIEW OF THE LICENSE RENEWAL APPLICATION FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT (TAC NO. MD2666)
Dear Mr. Dietrich:
By letter dated July 31, 2006, Entergy Nuclear Operations, Inc., submitted an application pursuant to Title 10 Code of Federal Regulations Part 54, to be reviewed by the U.S. Nuclear Regulatory Commission (NRC) for renewal of the operating license for James A. FitzPatrick Nuclear Power Plant. The NRC staff performed site audits of the Aging Management Programs and related Time-Limited Aging Analyses provided in the license renewal application. The staff has identified, in the enclosure, an area where additional information is needed to complete the audit reviews.
Based on discussions with Mr. Rick Plasse of your staff, a mutually agreeable date for your response is within 30 days from the date of this letter. If you have any questions regarding this letter or if circumstances result in your need to revise the response date, please contact me at 301-415-1458 or via e-mail at nbl@nrc.gov.
Sincerely,
/RA/
N. B. (Tommy) Le, Senior Project Manager License Renewal Branch B Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-333
Enclosure:
As stated cc w/encl: See next page
ML070370023 OFFICE LA:DLR PM:RLRB:DLR BC:RLRB:DLR NAME IKing NBLe RAuluck DATE 02/22/07 02/22/07 02/23/07
Letter to Peter Dietrich, from N B Le dated, February 23, 2007
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION REGARDING THE REVIEW OF THE LICENSE RENEWAL APPLICATION FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT (TAC NO. MD2666)
HARD COPY DLR R/F E-MAIL:
JFair RWeisman AMurphy RPettis GGalletti DShum GBagchi SSmith (srs3)
SDuraiswamy YL (Renee) Li RidsNrrDlr RidsNrrDlrRlra RidsNrrDlrRlrb RidsNrrDe RidsNrrDci RidsNrrDeEemb RidsNrrDeEeeb RidsNrrDeEqva RidsNrrDss RidsNrrDnrl RidsOgcMailCenter RidsNrrAdes DLR Staff GHunegs, SRI ECobey RLaufer JBoska RMathew
REQUESTS FOR ADDITIONAL INFORMATION JAMES A. FITZPATRICK NUCLEAR POWER PLANT LICENSE RENEWAL APPLICATION SECTIONS 4.3.1, 4.7.1, and 4.7.2 Section 4.3.1 Class 1 Fatigue
Background
The staff conducted an on site Time-Limited Aging Analyses (TLAA) follow-up audit on January 8th and 9th, 2007 and has identified the following areas where additional information is required. The following requests for additional information (RAIs) was formerly documented as Question 418 of the staff aging management review database, which has been closed at the request of the U.S. Nuclear Regulatory Commission.
RAI 4.3.1-1 Part A:
Licensing renewal application (LRA) Table 4.3-2 gives the current design basis allowable cycles and updated 60-year cycle projections for the James A. FitzPatrick Nuclear Power Plant (JAFNPP) design basis transients. The cycle values in the Current Design Basis Cycles, Allowable column of the table represent the updated current design basis allowable cycles performed by Structural Integrity Associates (SIA) and the cycle values in the Updated 60 Year Cycle Projection column of the table represent 60-year cycle projections as of actual JAFNPP operations through Spring 2005. The staff requests the following additional information:
(i). The original current design basis allowable cycles for the original metal fatigue calculations were performed by General Electric Company (GE). Provide the current design basis allowable cycle values that were calculated by GE for the JAFNPP design basis transients.
(ii). Clarify what regulatory process was used to allow SIAs updated current design basis allowable cycle values as the current design basis for the JAFNPP design basis transients.
(iii). Discuss the methods used to establish the original current design basis allowable cycles performed by GE and the updated current design basis allowable cycles by SIA. Identify the differences in the methods used by GE and SIA and justify why SIAs updated current design basis allowable cycle assessment is acceptable to use as the current design basis for JAFNPP.
(iv). For each transient in LRA Table 4.3-2, clarify how many operational cycles have been recorded up to the time that the 60-year transient projections were calculated, as given in the Updated 60 Year Cycle Projection column of LRA Table 4.3-2.
(v). Provide a technical discussion to clarify how the 60-year projections were performed based on recorded transient data. In particular, if a particular transient category in LRA Table 4.3-2 is made up of more than one specific transient, clarify which specific transient is used to define the transient and clarify how the total number of cycles were used to derive the 60 year cycle projections.
Enclosure (vi). Explain how the cycles were recorded prior to 1988, when JAFNPP did not implement a plant computer to track transient events.
(vii). Justify why the following values in LRA Table 4.3-2 are acceptable:
(a). A Current Design Basis Cycles, Allowable value of 1" and an Updated 60 Year Cycle Projection value of 0" for transient category 13, Reactor Overpressure.
(b). A Current Design Basis Cycles, Allowable value of 2" and an Updated 60 Year Cycle Projection value of 1" for transient category 14, Single Relief Valve Blowdown.
(c). A Current Design Basis Cycles, Allowable value of 5" and an Updated 60 Year Cycle Projection value of 0" for transient category 17, Improper Start of Cold Recirculation Loop.
(d). A Current Design Basis Cycles, Allowable value of 5" and an Updated 60 Year Cycle Projection value of 0" for transient category 18, Sudden Start of Pump!Cold Recirculation.
(e). A total Current Design Basis Cycles, Allowable value of 233" and a total Updated 60 Year Cycle Projection value of 244" for Shutdowns, which comprises transient categories Nos. 19, Reduction to 0% Power; 20, Hot Standby; 21, Cooldown (100EF/hr to 375EF); 22, Vessel Flooding (375EF to 330EF in 10 min.);
and 23, Cooldown (100EF/hr to 100EF).
(f). A Current Design Basis Cycles, Allowable value of 1" and an Updated 60 Year Cycle Projection value of 1" for transient category 24, Hydrostatic Test (1563 psig).
(g). A Current Design Basis Cycles, Allowable value of 35" and an Updated 60 Year Cycle Projection value of 34" for transient category 25, Unbolt.
Part B:
Page 19 of GE Design Calculation EAS-149-1286 / DRF B13-01391 discusses GEs evaluation of 12 transients (i.e., nine reactor SCRAMS, one turbine trip, two feedwater pump trips) that had been grouped into the Shutdown transient for the plant. The report stated that the change in reactor coolant temperature (T) for six of these events had exceeded the T value for this transient. The staff noted that the bases provided on page 19 for justifying why these events can be categorized as plant heatups or cooldowns are based on qualitative analysis without using any temperature gradient data. The staff requests the following additional information:
(i). Explain why the six transients specified in GE calculation can be grouped into Shutdown transient for the plant when the T values for these six events were determined to be excessive and the temperature gradients for the transients are not defined.
(ii). For the scram event that occurred on November 4, 1984, a T of !297EF and a T of
+437EF occurred on the same day, when did T events occur and what were the actual temperature gradients associated with these events.
(iii). Clarify how your response to this part (Part B) factors into your response to Part A, particularly with respect to the number of recorded occurrences for the transient Categories in LRA Table 4.3-2.
Part C:
(i). In the GE stress report, GE characterized 12 unidentified operational transients as reactor SCRAMS. GE identified that 63 occurrences of these transients had occurred prior to 1987.
Verify the operational transients and occurrences identified in the GE stress report and provide your evaluations.
(ii). In LRA Table 4.3-2, Entergy projects that the number of SCRAM events occurring through 60 years of operation for the All Other SCRAM events will be 62. Explain how the number of cycles projected through 60 years of operation can be 62 when 63 occurrences had been recorded through 1987.
(iii). In the GE stress report, GE also mentioned that the change in T associated with these 12 unidentified transients was approximately 330EF. The staff requests the following additional information:
(a). Please define these unidentified transients and list the pressure-temperature data for these transients.
(b). Please define the pressure-temperature (P-T) data that were used for the limiting SCRAM event used in SIAs updated 60-year cumulative usage factor calculations.
(c). Justify how these 12 unidentified transients are characterized based on the analyzed P-T limit data used in SIAs updated cumulative usage factor calculations.
(iv). Clarify how your response to this part (Part C) factors into your response to Part A, particularly with respect to the recording the number of cycles for the transients defined in LRA Table 4.3-2 and using this data to project the 60-year cycles for the transients.
Section 4.7.1 - Recirculation Isolation Valves
Background
In qualifying the Recirculation Isolation Valves for TLAA, the Licensee in LRA Section 4.7.1, Recirculation Isolation Valves, partially quotes the method of analysis and qualification requirements stated in Table 16.2-7 of the Updated Final Safety Analysis Report (UFSAR) in reference to the 28" Suction and Discharge Recirculation Valves. It also states that fatigue evaluation was not required for these valves.
The UFSAR indicates that these valves are designed to withstand the effects of cyclic loads and provides the criteria, method and requirements for the fatigue evaluations in Table 16.2-7.
Table 16.2-7 provides specific criteria with experimental and analytical methods for evaluating and qualifying the 28" Suction and Discharge Recirculation Valves. For the analytical method it specifies the ASME Boiler and Pressure Vessel Code, Nuclear VesselsSection III Article 4. In Art 4, Design, fatigue evaluation is part of the required design criteria and the UFSAR in Table 16.2-7 provides cycles with defined service of operation for fatigue evaluations. It also indicates that the results from the fatigue evaluations were plotted and showed that the flange region of the valve is adequate for the defined service. Table 16.2-7 also provides requirements with calculated values and allowables for the primary and the primary plus secondary stresses for the valve body, bonnet and bonnet joint bolts.
Referencing the recirculation isolation valves, the UFSAR states, For fatigue evaluations consider 30 cycles of normal pressurization followed by blowdown and 270 cycles of normal pressurization followed by normal depressurization.
As these valves are not ASME class valves, no specific fatigue analysis was required; however, the number of cycles suggested by the UFSAR is greater than the number of cycles allowed as part of the Fatigue Monitoring Program, so the transients suggested will not be exceeded.
Thus this TLAA will remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(I).
RAI 4.7.1-1 Provide the basis for the referenced cycles in LRA Section 4.7.1 and elaborate on your rational for concluding that the LRA cycles are less than those referenced in the UFSAR Table 16.2-7.
RAI 4.7.1-2 As stated in UFSAR table 16.2-7, fatigue evaluation and interpretation of results was required to qualify the 28" Suction and Discharge Recirculation Valves for 40 year life. Discuss how this UFSAR requirement will be maintained for a 20 year life extension.
RAI 4.7.1-3 UFSAR Table 16.2-7 also provides requirements with calculated values and allowables for the primary and the primary plus secondary stresses for the valve body, bonnet and bonnet joint bolts. In LRA space, determine the suitability of joint bolts for UFSAR defined service cycles projected to end of life extension. If not required provide justification.
RAI 4.7.1-4 LRA Section 4.7.1 states that, the number of cycles suggested by the UFSAR is greater than the number of cycles allowed as part of the Fatigue Monitoring Program, so the transients suggested will not be exceeded. Provide a reference that contains the number of cycles allowed in the Fatigue Monitoring Program for the UFSAR defined services for fatigue evaluations stated in UFSAR Table 16.2-7 for the 28" Suction and Discharge Recirculation Valves. In LRA Section 4.7.1, show that these cycles projected to end of life extension are less than the UFSAR cycles.
Section 4.7.2 Leak Before Break
Background
The staff reviewed LRA Section 4.7.2 and per our discussions with the applicant during the staffs and applicants conference call on 2/12/07, the staff requests that the applicant revise and resubmit Section 4.7.2 after addressing the following points:
- 1. Remove all references to the terms Leak Before Break, as the TLAA being discussed clearly predates the formal concept of Leak Before Break documented in Standard Review Plan Section 3.6.3.
- 2. Rename section 4.7.2 with a title that describes the primary purpose of the analysis in UFSAR Section 16.3.2 and the staffs original Safety Evaluation Report for licensing of the facility, Section 5.2.2.
- 3. Discuss the specific time limited part of the original analysis and provide a description of the plant modifications (or non-installations) which were permitted based on the original analysis including locations, or systems in the plant to which the modifications were permitted.
- 4. Conclude with an explanation of how the TLAA will be addressed in the context of the ongoing LRA review OR include a commitment to be submitted with the completed TLAA for review at least 2 years prior to entering the period of extended operation.
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 Resident Inspector's Office James A. FitzPatrick Nuclear Power Plant U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 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 Mr. Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271
FitzPatrick Nuclear Power Plant cc:
Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. Steven Lyman Oswego County Administrator 46 East Bridge Street Oswego, NY 13126 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 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 Mr. Rick Plasse Project Manager, License Renewal Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093 Mr. James Ross Nuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708