ML070370023

Requests for Additional Information Regarding the Review of the License Renewal Application for James A. FitzPatrick Nuclear Power Plant

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, 2007Mr. Peter T. DietrichSite Vice President James A. FitzPatrick Nuclear Power PlantEntergy Nuclear Operation, Inc.

P.O. Box 110 Lycoming, NY 13093-0110

SUBJECT:

REQUESTS FOR ADDITIONAL INFORMATION REGARDING THE REVIEWOF 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 applicationpursuant 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 yourresponse 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 RegulationDocket No. 50-333

Enclosure:

As statedcc w/encl: See next page

ML070370023 OFFICELA:DLRPM:RLRB:DLRBC:RLRB:DLRNAMEIKingNBLe RAuluck DATE02/22/0702/22/0702/23/07 Letter to Peter Dietrich, from N B Le dated, February 23, 2007

SUBJECT:

REQUESTS FOR ADDITIONAL INFORMATION REGARDING THE REVIEWOF THE LICENSE RENEWAL APPLICATION FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT (TAC NO. MD2666)HARD COPYDLR R/FE-MAIL: JFairRWeisman AMurphy RPettis GGalletti DShum GBagchi SSmith (srs3)

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ECobey RLaufer JBoska RMathew REQUESTS FOR ADDITIONAL INFORMATIONJAMES A. FITZPATRICK NUCLEAR POWER PLANTLICENSE RENEWAL APPLICATIONSECTIONS 4.3.1, 4.7.1, and 4.7.2Section 4.3.1 Class 1 FatigueBackgroundThe staff conducted an on site Time-Limited Aging Analyses (TLAA) follow-up audit on January 8 th and 9 th, 2007 and has identified the following areas where additional information isrequired. 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 4.3.1-1Part A: Licensing renewal application (LRA) Table 4.3-2 gives the current design basis allowable cyclesand 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 calculationswere 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 SIA's updated current design basisallowable 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 cyclesperformed 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 SIA's 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 beenrecorded 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 basedon 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 definethe 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 aplant 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 YearCycle Projection" value of "0" for transient category 13, "Reactor Overpressure."(b). A "Current Design Basis Cycles, Allowable" value of "2" and an "Updated 60 YearCycle 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 YearCycle 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 YearCycle Projection" value of "0" for transient category 18, "Sudden Start of PumpColdRecirculation." (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 (100F/hr to 375F);" 22, "Vessel Flooding (375F to 330F in 10 min.);"and 23, "Cooldown (100F/hr to 100F)."(f). A "Current Design Basis Cycles, Allowable" value of "1" and an "Updated 60 YearCycle 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 YearCycle Projection" value of "34" for transient category 25, "Unbolt."Part B: Page 19 of GE Design Calculation EAS-149-1286 / DRF B13-01391 discusses GE's evaluationof 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 valuefor 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 excessiveand the temperature gradients for the transients are not defined.(ii). For the scram event that occurred on November 4, 1984, a T of 297F and a T of+437F occurred on the same day, when did T events occur and what were the actualtemperature 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 reactorSCRAMS. 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 provideyour evaluations.(ii). In LRA Table 4.3-2, Entergy projects that the number of SCRAM events occurring through60 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 12unidentified transients was approximately 330F. The staff requests the following additionalinformation:(a). Please define these unidentified transients and list the pressure-temperature datafor these transients. (b). Please define the pressure-temperature (P-T) data that were used for the limitingSCRAM event used in SIA's updated 60-year cumulative usage factor calculations. (c). Justify how these 12 unidentified transients are characterized based on the analyzedP-T limit data used in SIA's 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 ValvesBackgroundIn 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 loadsand 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 alsoindicates 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 evaluationsconsider 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 4.7.1-1Provide the basis for the referenced cycles in LRA Section 4.7.1 and elaborate on your rationalfor concluding that the LRA cycles are less than those referenced in the UFSAR Table 16.2-7.RAI 4.7.1-2As stated in UFSAR table 16.2-7, fatigue evaluation and interpretation of results was required toqualify 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 4.7.1-3UFSAR 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 4.7.1-4LRA Section 4.7.1 states that, "the number of cycles suggested by the UFSAR is greater thanthe 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 RecirculationValves. 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 BreakBackgroundThe staff reviewed LRA Section 4.7.2 and per our discussions with the applicant during thestaff's and applicant's 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 discussedclearly 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 inUFSAR Section 16.3.2 and the staff's 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 ofthe "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 theongoing 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. TaylorChief Executive Officer Entergy Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213Mr. John T. HerronSr. VP and Chief Operating Officer Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. Peter T. DietrichSite Vice President Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power PlantP.O. Box 110 Lycoming, NY 13093Mr. Kevin J. MulliganGeneral Manager, Plant Operations Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power PlantP.O. Box 110 Lycoming, NY 13093Mr. Oscar LimpiasVice President Engineering Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. Christopher SchwarzVice President, Operations Support Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. John F. McCannDirector, Licensing Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Resident Inspector's OfficeJames A. FitzPatrick Nuclear Power Plant U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093Ms. Charlene D. FaisonManager, Licensing Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. Michael J. ColombDirector of Oversight Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. David WallaceDirector, Nuclear Safety Assurance Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093Mr. James CostedioManager, Regulatory Compliance Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093Assistant General CounselEntergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. Charles Donaldson, EsquireAssistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 FitzPatrick Nuclear Power Plant cc:

Regional Administrator, Region IU.S. Nuclear Regulatory Commission

475 Allendale Road King of Prussia, PA 19406Mr. Steven LymanOswego County Administrator 46 East Bridge Street Oswego, NY 13126Mr. Peter R. Smith, PresidentNew York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399Mr. Paul EddyNew York State Dept. of Public Service 3 Empire State Plaza Albany, NY 12223-1350SupervisorTown of Scriba Route 8, Box 382 Oswego, NY 13126Mr. James H. SniezekBWR SRC Consultant 5486 Nithsdale Drive Salisbury, MD 21801-2490Mr. Michael D. LysterBWR SRC Consultant 5931 Barclay Lane Naples, FL 34110-7306Mr. Garrett D. Edwards814 Waverly Road Kennett Square, PA 19348Mr. Rick PlasseProject Manager, License Renewal Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093 Mr. James RossNuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708