JAFP-07-0079, License Renewal Application, Amendment 12
| ML071770168 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/20/2007 |
| From: | Peter Dietrich Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| JAFP-07-0079 SIR-07-084-NPS, Rev 1 | |
| Download: ML071770168 (50) | |
Text
Entergy Nuclear Northeast a
Entergy Nuclear Operations, Inc.
James A. Fitzpatrick NPP P.O. Box 110 Lycoming, NY 13093 Tel 315 349 6024 Fax 315 349 6480 Pete Dietrich Site Vice President - JAF June 20, 2007 JAFP-07-0079 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
REFERENCES:
- 1. Letter, Entergy to USNRC, "James A. FitzPatrick Nuclear Power Plant, Docket No. 50-333, License No. DPR-59, License Renewal Application," JAFP-06-0109, dated July 31, 2006
- 2. Letter, Entergy to USNRC, "License Renewal Application, Amendment 8," JAFP-07-0047, (TAC No. MD2666) dated April 6, 2007
- 3. Letter, Entergy to USNRC, "License Renewal Application, Amendment 10," JAFP-07-0053, (TAC No. MD2666) dated April 24, 2007
- 4. Letter, Entergy to USNRC, "License Renewal Application, Amendment 6," JAFP-07-0021, (TAC No. MD2666) dated February 12, 2007
- 5. Letter, Entergy to USNRC, "License Renewal Application, Amendment 9," JAFP-07-0048, (TAC No. MD2666) dated April 6, 2007
- 6. Letter, Entergy to USNRC, "License Renewal Application, Amendment 11," JAFP-07-0067, (TAC No. MD2666) dated May 17, 2007
SUBJECT:
Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333, License No. DPR-59 License Renewal Application. Amendment 12
Dear Sir or Madam:
On July 31, 2006, Entergy Nuclear Operations, Inc. submitted the License Renewal Application (LRA) for the James A. FitzPatrick Nuclear Power Plant (JAFNPP) as indicated by Reference 1. contains a response to questions concerning LRA section 4.3.1, Class 1 Fatigue, previously committed to in Reference 2. Based on the information provided in Attachment 1 Entergy determined a need to update the 60 year cycle projections, due to the identification of additional startup/shutdown transients not previously considered which are detailed in. These updated projections will be included in the future fatigue management activities described in LRA Commitment #20 documented in Reference 5. Attachment 2 contains clarifications to previous RAIs provided to the NRC in References 3, 4, 5, and 6, as requested by the NRC license renewal staff.
Should you have any questions concerning this submittal, please contact Mr. Jim Costedio at (315) 349-6358.
KA(_16ZA
June 20, 2007 JAFP-07-0079 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 26 day of June, 2007.
SITE VICE PRESIDENT PD/cf Attachments 1 and 2 cc:
Mr. N.B. (Tommy) Le, Senior Project Manager License Renewal Branch B Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-11-F1 Washington, DC 20555 Mr. Samuel J. Collins, Administrator Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 NRC Resident Inspector U. S. Nuclear Regulatory Commission James A. FitzPatrick Nuclear Power Plant P.O. Box 136 Lycoming, NY 13093 Mr. John P. Boska, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8-C2 Washington, DC 20555 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 1 0 th Floor Albany, NY 12223 Mr. Peter R. Smith, President NYSERDA 17 Columbia Circle Albany, NY 12203-6399
JAFP-07-0079 Docket No. 50-333 James A. FitzPatrick Nuclear Power Plant License Renewal Application - Amendment 12 RAI section 4.3.1, Class 1 Fatigue SIR-07-084-NPS, Rev. 1
Structural Integrity Associates, Inc.
6855 S. Havana Street Suite 350 Centennial, CO 80112-3868 Phone:
303-792-0077 Fax:
303-792-2158 themann@structintcamn June 8, 2007 SIR-07-084-NPS, Rev. 1 Mr. Kenneth Phy Entergy Nuclear Northeast James A. Fitzpatrick Nuclear Power Plant 268 Lake Road East P.O. Box 110 Lycoming, NY 13093
SUBJECT:
SI Response to NRC Requests for Additional Information at Fitzpatrick
References:
- 1. James A. FitzPatrick Nuclear Power Plant License Renewal Application, submitted July 31, 2006, SI File No. FITZ-08Q-210.
- 2. USNRC Letter, Requests for Additional Information Regarding the Review of the License Renewal Application for James A. FitzPatrick Nuclear Power Plant (TAC No.
MD2666), February 23, 2007, S1 File No. FITZ-08Q-2 10.
- 3. General Electric Report No. EAS-149-1286, DRF B 13-01391, "Reactor Pressure Vessel Fatigue Evaluation for the James A. FitzPatrick Nuclear Power Plant," January 1987, SI File No. FITZ-08Q-201.
- 4. Structural Integrity Associates Report No. SIR-02-045, Revision 1, "Updated Fatigue Analysis for James A. Fitzpatrick Nuclear Power Plant Reactor Pressure Vessel Components," September 2002, SI File No. W-NYPA-78Q-401.
Dear Ken:
Structural Integrity Associates, Inc. (SI) is pleased to provide responses to U.S. Nuclear Regulatory Commission (NRC) Requests for Additional Information (RAIs) related to Section 4.3.1 of the License Renewal Application (LRA) for the James A. Fitzpatrick Nuclear Power Plant (JAFNPP). Our proposed responses are included as Attachment I to this letter.
Entergy Nuclear Northeast (ENN) is in the process of getting the LRA for JAFNPP approved by the NRC. Section 4.3.1 of the LRA addresses the effects fatigue may have on ASME Class I structures, systems and components. Prior to License Renewal activities, an evaluation was performed for JAFNPP by General Electric to update the original RPV analysis to determine fatigue usage factors resulting from plant operating data through 1986 [3]. In 1992, the fatigue Austin, TX Centennial, CO Charlotte, NC Stonington, CT Silver Spring, MD Sunrise, FL Uniontown, OH Whittier, CA Ontario, CANADA 512-533-9191 303-792-0077 704-597.5554 860-536-3982 301-445-8200 954-572-2902 330-899-9753 562-944-8210 905-829-9817
Mr. Kenneth Phy/Entergy Nuclear Northeast SIR-07-084-NPS, Rev. I June 8, 2007 Page 2 of 3 usage factors were revised to include the effects of Power Uprate and an updated feedwater nozzle fracture mechanics evaluation.
The NRC staff conducted an on-site Time-Limited Aging Analysis (TLAA) audit on January 8th and 9h, 2007 and has identified areas where additional information is required. The questions are divided into three parts; Parts A, B & C. The attachment to this letter summarizes the NRC staff questions and Sl's proposed responses.
Please note that during the course of our review, an error was identified in a supporting calculation used to develop the Reference [41 report. SI has determined that the number of Scram events after 11.4 years of plant operation in the Reference [3] GE report was not correctly translated into the Reference [4] S1 report. This resulted in an inaccurate estimate of current cycle counts and projected 60-year transient cycles. In accordance with the SI Quality Assurance Program, Corrective Action Report (CAR-07-06) and Nonconformance Report (NCR 07-05) were initiated and transmitted to ENN.
Following further review of plant data, SI was able to re-classify the twelve missing transients as Startup and Shutdown events, identified an additional startup/shutdown transient not previously considered, and re-classified six Scram events as startup/shutdown events. With the incorporation of these 19 events, JAFNPP remains within the current allowable cycle limit in the UFSAR, and is expected to remain within the current design basis analyzed number of cycles for the current licensed operating period. However, this will require a revision to the JAFNPP cycle counting procedure and LRA Table 4.3-2 to incorporate the updated 60-year cycle projections for the affected events.
Since JAFNPP has already committed to the NRC to update existing fatigue evaluations to address environmentally assisted fatigue, we recommend that this information be addressed as a part of that effort.
V Structural Integrity Associates, Inc.
Mr. Kenneth Phy/Entergy Nuclear Northeast SIR-07-084-NPS, Rev. 1 June 8, 2007 Page 3 of 3 Please feel free to contact me with any questions you have.
Prepared by:
Reviewed by:
Terry J. Herrmann, P.E.
Senior Consulting Engineer Date: 6/8/2007 Date: 6/8/2007 Gary L. Stevens, P.E Senior Associate Approved by:
Terry J. Herrmann, P.E.
Senior Consulting Engineer Date:
6/8/2007 cc:
FITZ-08Q-401 ENN-JAFNPP:
R. Plasse L. Leiter W. Drews J. Costedio M. Durr J. Abisamra Entergy License Renewal Services (ANO-1):
M. Stroud S. Batch A. Cox ENN-WPO:
R. Penny A. Unsal D. Burch K. Tom R. Casella V
Structural Integrity Associates, Inc.
.9 REFERENCES
- 1. Reactor Thermal Cycles, GE Drawing No. 729E762, Revision 0, JAFNPP Drawing 11825-5.15-1A, September 15 1967, SI File No. W-NYPA-78Q-205.
- 2. Structural Integrity Associates Report No. SIR-02-045, Revision 1, "Updated Fatigue Analysis for James A. Fitzpatrick Nuclear Power Plant Reactor Pressure Vessel Components," September 2002, SI File No. W-NYPA-78Q-401.
- 4. JAFNPP Updated Final Safety Analysis Report, Chapter 4, Table 4.2-3 "Reactor Pressure Vessel Thermal Cyclic/Transient Limits", Rev 5/05, SI File No. FITZ-08Q-206.
- 5. New York Power Authority Memorandum JTS-95-0684, "Response to DER 95-1499, Discrepancy Between GE Fatigue Report and Reactor Analyst Procedure for Tracking Thermal Cycles", March 15, 1986, SI File FITZ-08Q-212.
- 6. Entergy Nuclear Northeast Document No. JAF-SE-03-002, "Updated Reactor Pressure Vessel Fatigue Analysis," December 2003, SI File No. FITZ-08Q-205.
- 7. Combustion Engineering Report No. CENC-1 159, "Analytical Report for PASNY Reactor Vessel for FitzPatrick Station," August 1971, SI File No. W-NYPA-78Q-204.
- 8. General Electric Report No. EAS-149-1286, DRF B 13-0139 1, "Reactor Pressure Vessel Fatigue Evaluation for the James A. FitzPatrick Nuclear Power Plant," January 1987, SI File No. FITZ-08Q-201.
- 9. GE Nuclear Energy Certified Design Specification No. 25A5024, Revision 0, "Reactor Vessel - Power Uprate," July 1991, SI File No. W-NYPA-78Q-203.
- 10. Structural Integrity Associates Calculation, Revision 1, "Updated Cycle Counts and 60 Year Projections," SI File No. W-NYPA-78Q-301.
- 11. Entergy Nuclear Northeast Engineering Report No. JAF-RPT-05-LRD04, Revision 1, "TLAA - Mechanical Fatigue," July 2006, SI File No. FITZ-08Q-204.
- 12. Structural Integrity Associates Corrective Action Report 07-06, "Calculation NYPA-78Q-301 did not Account for 12 Unidentified Transient Events", dated 4/4/2007, SI File No. FITZ-08Q-109.
- 13. J. A. FitzPatrick computer logs and strip charts from GE evaluation EAS-149-1286, SI File No. FITZ-08Q-203.
- 14. Shift Supervisor Operating Logs from J. A. FitzPatrick, SI File No. FITZ-08Q-208.
Attachment I to SIR-07-084-NPS, Rev. I Page 1 of 37 C
Structural Integrity Associates, Inc.
- 17. Nozzle Thermal Cycles, GE Drawing No. 135B9990, Revision 1, Sheets 1 through 8, JAFNPP Drawings 11825-5.01-50A through 57A, May 22 1967, SI File No. W-NYPA-78Q-206.
- 18. General Electric Report No. NEDC-32068, "Reactor Pressure Vessel Power Uprate Stress Report Reconciliation for the Fitzpatrick Power Plant", March 1992, NYPA File No. S-90-00978-38A, SI File No. NYPA-58Q-214P.
- 19. Structural Integrity Associates Calculation, Revision 1, "Fatigue Analysis of Feedwater Nozzle," SI File No. W-NYPA-78Q-305.
- 20. Structural Integrity Associates Calculation, Revision 1, "Fatigue Analysis of Recirculation Inlet Nozzle," SI File No. W-NYPA-78Q-308.
- 21. Structural Integrity Associates Calculation, Revision 2, "Fatigue Analysis of Control Rod Drive Nozzle," SI File No. W-NYPA-78Q-311.
- 22. Structural Integrity Associates Calculation, Revision 0, "Revised Fatigue Calculation for the RPV Closure Region Bolts," SI File No. W-NYPA-78Q-302.
Attachment I to SIR-07-084-NPS, Rev. I Page 2 of 37 Structural Integrity Associates, Inc.
RAI 4.3.1-1. PartA 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:
RAI 4.3.1-1. Part A (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.
Response
Whereas the original design basis allowable transients were established by GE [1 ], the current design basis allowable transient cycle values are those established by SIA [2]. Table A-I shows both the original and current design basis values.
Attachment I to SIR-07-084-NPS, Rev. I Page 3 of 37 Structural Integrity Associates, Inc.
Table A-i: Transient Event Cycle Limits Event Design Basis Transient Original Design Basis Revised Design Basis Number Analyzed No. of Analyzed No. of Cycles (GE) Ill Cycles (SIA) 121 I
Bolt-up (70'F) 123 36 2
Design Hydro Test (1250 psig, 130 36 100-F) 3 Startup (100IF/hr to 5460 F) 120 233 4
Turbine Roll and Increase to Rated 120 221 Power 5
Daily Reduction to 75% Power 10000 7566 6
Weekly Reduction to 50% Power 2000 1685 7, 8 Rod Worth Test (Sequence 400 357 Exchange)
Loss of Feedwater Heater 9
Turbine Trip at 25% Power 10 7
10 Feedwater Heater Bypass 70 34 SCRAMs 11 Loss of FW Pumps, MSIVs Close 10 12 12 Turbine Generator Trip, FW on, 40 12 MSIVs stay. open 13 Reactor Overpressure 1
1 14 Single Relief or Safety Valve 2
2 Blowdown 15 All Other Scrams 147 64 16 Normal Operation 17 Improper Start of Cold 5
5 Recirculation Loop 18 Sudden Start of Pump - Cold 5
5 Recirculation Shutdowns (events 19-23) 118 233 19 Reduction to 0% Power 20 Hot Standby 21 Cooldown (to 375°F @I 00°F/hr) 22 Vessel Flooding (375°F to 330'F in 4
10 minutes) 23 Cooldown (330'F to I 00°F @
(a_)
24 Hydrostatic Test (1563 psig, I00F) 3 1
25 Unbolt (l00F) 123 35 26 Refueling (70'F)
Attachment I to SIR-07-084-NPS, Rev. I Page 4 of 37 V
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RAI 4.3.1-1. Part A (ii)
Clarify what regulatory process was used to allow SIA's updated current design basis allowable cycle values as the current design basis for the JAFNPP design basis transients.
Response
The JAFNPP licensing basis was updated in accordance with 10CFR50.59 per JAF-SE-03-002, "Updated Reactor Pressure Vessel Fatigue Analysis" [6]. The review was performed considering the original CE analysis [7] and subsequent GE analyses [8, 9] and SIA analyses [2, 10].
The revised number of allowable cycles was included in UFSAR Section 4.2, Table 4.2-3 "Reactor Pressure Vessel Thermal Cyclic/Transient Limits" [4]. Per procedure, JAFNPP monitors transients affecting reactor vessel fatigue usage as required by Section 5.5.5 "Component Cyclic or Transient Limit" of the Technical Specifications [3]. to SIR-07-084-NPS, Rev. I Page 5 of 37 j
Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part A (iiiN 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 SIA's updated current design basis allowable cycle assessment is acceptable to use as the current design basis for JAFNPP.
Response
During the original design of JAFNPP, GE, as supplier of the Reactor Pressure Vessel (RPV) and associated ASME Section III Class 1 components, developed thermal and pressure cycle diagrams for the RPV and nozzles [1, 17]. These thermal, pressure, and flow rate cycle diagrams established the transient events to be evaluated for cyclic operating conditions. The number of transients was originally established based on an estimate of what would be experienced by the RPV during a 40-year operating period. The original design basis allowable cycles were considered to be conservative bounding values for design.
Combustion Engineering (CE), the RPV fabricator, used the cycles defined by GE as input to tcalculate fatigue usage for the RPV in accordance with the ASME Boiler and Pressure Vessel Code. The CE analysis [7] was the original design basis fatigue calculation for the JAFNPP RPV.
In 1987, GE performed an updated fatigue evaluation to assess the effect of actual cyclic duty from the first 11.4 years of plant operation on RPV components [8] using temperature, pressure and flow data obtained from plant records. These components included the closure studs, the recirculation inlet nozzle, the feedwater nozzle and the control rod drive nozzle. The analyzed number of cycles used in this evaluation was an estimate made by GE based on plant operating data through 1986. Components were selected for evaluation considering the severity of the stresses as well as their sensitivity to changes in reactor conditions based on past GE fatigue analysis experience. Components which were originally exempted from fatigue evaluation in the original design basis fatigue calculation were reevaluated and those fatigue exemptions were confirmed to remain valid.
A review of actual plant history in 2002 revealed that seven thermal cycle events reached as high as 92% of the original design basis cycle limits. Based on the limited remaining thermal cycle margin for these events, SIA updated both the actual and projected thermal cycle count for JAFNPP [2, 10]. SIA used the same methodology as GE had previously used to update the actual and 40-year projected thermal cycle counts, except that additional plant data that was available was factored into revised projections for 40 and 60 years.
For the 2002 update, thermal cycle transient data collected from plant instrumentation and operator logs was obtained. The data was reviewed and compared to the design basis thermal cycle diagrams. Design inputs for this evaluation primarily consisted of:
- 1. The thermal cycle monitoring reports for the time period from January 1st, 1989 to June 30th 2001. Data for the time period between 1986 (i.e., end date of previous fatigue update) and I/1/89 was not available at the time of the 2002 update.
Attachment I to SIR-07-084-NPS, Rev. I Page 6 of 37 V
Structural Integrity Associates, Inc.
- 2. Additional plant process computer data and strip charts retrieved from microfiche or provided on CD-ROM by JAFNPP personnel.
- 3.
Design basis thermal cycle diagrams used to categorize the actual thermal cycles captured in the data.
- 4. RPV Power Uprate Design Specification, since this specification altered many of the transient pressure and temperature profiles.
The 2002 evaluation categorized each thermal transient experienced at JAFNPP based on a comparison to the design thermal cycle diagram [1]. For any transient event where detailed data was unavailable, conservative assumptions were applied to assign each transient event to a bounding design basis event. For example, the GE evaluation [8] identified that there were 4 Loss of Feedwater Pump events (Event 11) through Year 11.4, which was incremented to 5 events in the 2002 SI evaluation [2, 10]. A subsequent JAFNPP review [5] identified that the actual count through Year 19.5 (June 1994) should be 2 events. The Loss of Feedwater Pump transient would result in significantly higher temperature and pressure changes in the RPV bottom head than other events that could reasonably be used (e.g., Shutdown and Startup) if more detailed data had been available.
By reviewing actual plant operating conditions, SIA determined the actual numbers of accrued event cycles. These numbers were used to project the numbers of cycles expected through sixty years of operation [10]. The 60-year projections were used to revise the number of cycles from those originally estimated by GE [1] and included in the CE analytical report [7] and the 1987 GE evaluation [8]. The projected number of cycles for 60 years was set equal to (but never less than) the results from the straight-line projections.
SIA's updated current design basis allowable cycle assessment is acceptable to use as the current design basis for JAFNPP because: (1) associated fatigue calculations demonstrate that CUF values will remain less than allowable, and (2) the Fatigue Monitoring Program ensures that actual numbers of cycles do not exceed the numbers used in the CUF calculations. The Fatigue Monitoring Program also requires that appropriate corrective action be taken prior to any analyzed number of cycles being exceeded.
Attachment I to SIR-07-084-NPS, Rev. I Page 7 of 37 Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part A (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.
Response
The requested information is provided in Table A-2. The Year 30.5 cycle count data was the most recent data available when the 60-year transient projections were calculated [ 11]. The cycles as of Year 30.5 from Reference [1111 represent an update to the SIA work from 2002 [10],
which was done as of 26.5 years of operation.
Attachment I to SIR-07-084-NPS, Rev. I Page 8 of 37 Structural Integrity Associates, Inc.
Table A-2: Number of Cycles as of Year 30.5 Compared to Year 60 Projections Event Design Transient Year 30.5 Cycle Updated Year 60 Number Count [11]
Cycle Projection [ 11 I
Bolt-up (70°F) 18 35 2
Design Hydro Test (1250 psig, 18 35 1 00-F) 3 Startup (100 °F/hr to 546*F) 113 216 4
Turbine Roll and Increase to Rated 107 204 Power 5
Daily Reduction to 75% Power 3,608 6,674 6
Weekly Reduction to 50% Power 814 1,526 7, 8 Rod Worth Test (Sequence 165 310 Exchange)
Loss of Feedwater Heater 9
Turbine Trip at 25% Power 5
7 10 Feedwater Heater Bypass 27 32 SCRAMs 11 Loss of FW Pumps, MSIVs Close 5
10 12 Turbine Generator Trip, FW on, 10 12 MSIVs stay open 13 Reactor Overpressure 0
00) 14 Single Relief or Safety Valve 1
1 Blowdown 15 All Other Scrams 57 62 17 Improper Start of Cold 0
00)
Recirculation Loop 18 Sudden Start of Pump - Cold 0
0 Recirculation Shutdowns (events 19-23) 126 244 19 Reduction to 0% Power 20 Hot Standby 21 Cooldown (to 3751F @ 000F/hr) 22 Vessel Flooding (3750F to 330OF in
- j 10 minutes) 23 Cooldown (3300 F to 100I F @
I 100-F/hr) 24 Hydrostatic Test (1563 psig, 100IF) 1 1
25 Unbolt (1 00°F) 17 34 26 Refueling (70'F)
Note: 1. Although zero events are anticipated, the number of design cycles remains unchanged from the reference [2] analysis. to SIR-07-084-NPS, Rev. I Page 9 of 37 V
Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part A (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.
Response
The 60-year projections in LRA Table 4.3-2 were developed based on the same approach used in both the 1987 GE evaluation and the 2002 SIA evaluation. Straight-line projections of historical data and actual transient events recorded through June 30, 2005 (or 30.5 years of plant operation) were utilized, as documented in Reference [11].
There are two instances where more than one transient event is grouped in Table 4.3-2; the Rod Worth Test/Sequence Exchange (Events 7 & 8) and the Shutdown (Events 19-23). Since the RPV does not experience any changes in pressure or temperature for Events 7 & 8, Event 7 does not contribute any fatigue usage to the RPV. The Sequence Exchange (Event 8) does involve a change in feedwater conditions, so this event is used in the evaluation of the feedwater nozzles.
Events 19 through 23 are sequential stages of a normal plant shutdown. A shutdown and cooldown to 100°F entails one each of transient Events 19 through 23. Each Shutdown event is therefore considered to result in one cycle of Events 19 through 23. In the case of the grouped shutdown transients (Events 19-23), the maximum stress intensity for the entire pressure and temperature range is used to calculate fatigue usage. Therefore, all plant shutdown events to cold ambient conditions were assumed to cause each of Events 19 through 23.
In 2002, SIA used various JAFNPP reports [13] to determine the actual cycles recorded during the first 26.5 years of operation since this is the latest data that was available at the time. The number of cycles the plant would incur in 60 years of operation was projected based on a straight-line extrapolation of the rate of cycles observed between Year 3 and Year 26.5 (i.e., the previous 23.5 years of plant operation). The data for the first 3 years of operation included a large number of events associated with initial operation and is not considered representative of future performance. Based on established maintenance rule programs and continuing advances in technology, the future rate of transient occurrence is not expected to exceed the rate of transient occurrence observed during the past 23.5 years of operation.
The 60-year number of cycles for four transients, Event 9 (Turbine Trip at 25% Power), Event 10 (Feedwater Heater Bypass), Event 12 (Turbine Generator Trip with MSIVs Open and Feedwater On) and Event 15 (All Other Scrams), were projected based on the rate of cycles occurring over 13.5 years of operation (Year 13.0 to Year 26.5). Projections for these events did not use data from the first 13 years of operation because there was a significant reduction in the accumulation of these events after Year 13. The projections for these four events based on only the last 13.5 years of operation are considered realistic and more representative for future plant operation.
Attachment I to SIR-07-084-NPS, Rev. I Page 10 of 37 Structural Integrity Associates, Inc.
In preparing the JAFNPP LRA, ENN compiled four additional years of thermal cycle data that were collected since the plant 60-year thermal cycle limits were established by SIA in 2002 [11].
The most recent semi-annual report for the component cyclic or transient limit monitoring program at the time the LRA was being prepared included a record of actual transient events through June 30, 2005, or 30.5 years of operation. Reference [11] performed revised projections to determine whether the thermal cycle limits established by SIA in 2002 remained valid for 60 years of operation, based on the updated Year 30.5 thermal cycle data. The same methods used to project 60-year counts in 2002 were applied to the Year 30.5 data. The results of this analysis are shown in the final column of Table A-2 and Table 4.3-2 of the JAFNPP LRA.
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RAI 4.3.1-1. Part A (vi)
Explain how the cycles were recorded prior to 1988, when JAFNPP did not implement a plant computer to track transient events.
Response
The method used to record transient events has evolved and improved since initial plant operation due to improved procedures and widespread availability of computer technology.
When JAFNPP entered commercial operation, operator log entries were the primary means used to determine when transient events occurred. Periodically, engineering personnel reviewed operator logs, strip charts and plant process computer printouts retained in plant records. A plant procedure [16] was implemented in the 1980s to provide a more structured approach to accounting for the information collected from operator logs, strip charts and computer printouts.
This procedure listed each transient event, the number of transient cycles experienced during the reporting period, the total number of cycles accumulated to date and the projected number of cycles to Year 40. The information collected from this procedure is maintained in plant records. to SIR-07-084-NPS, Rev. I Page 12 of 37 Structural integrity Associates, Inc.
RAI 4.3.1-1. Part A (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."
Response
Refer to Tables A-I and A-2. The revised design basis number of analyzed cycles for Event 13 is one cycle, as shown in Table A-1. To date, none of these events have occurred and none are expected to occur for 60 years, which is reflected in Table A-2. Since the projected number of cycles through 60 years of operation does not exceed the number assumed in the fatigue analyses, the value is acceptable. The JAFNPP Fatigue Monitoring Program assures that the analyzed number of cycles is not exceeded.
Attachment I to SIR-07-084-NPS, Rev. I Page 13 of 37 S Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part A (vii)
(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."
Response
Refer to Tables A-I and A-2. The revised design basis number of analyzed cycles for Event 14 is two cycles, as shown in Table A-I. To date, one of these events have occurred and none are expected to occur in the future, which is reflected in Table A-2. Since the projected number of cycles through 60 years of operation does not exceed the number assumed in the fatigue analyses, the value is acceptable. The JAFNPP Fatigue Monitoring Program assures that the analyzed number of cycles is not exceeded. to SIR-07-084-NPS, Rev. 1 Page 14 of 37 Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part A (vii)
(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."
Response
Refer to Tables A-1 and A-2. The revised design basis number of analyzed cycles for Event 17 is five cycles, as shown in Table A-1. To date, none of these events have occurred and none are expected to occur for 60 years, which is reflected in Table A-2. Since the projected number of cycles through 60 years of operation does not exceed the number assumed in the fatigue analyses, the value is acceptable. The JAFNPP Fatigue Monitoring Program assures that the analyzed number of cycles is not exceeded. to SIR-07-084-NPS, Rev. I Page 15 of 37 V
Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part A (vii)
(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."
Response
Refer to Tables A-I and A-2. The revised design basis number of analyzed cycles for Event 18 is five cycles, as shown in Table A-1. To date, none of these events have occurred and none are expected to occur for 60 years, which is reflected in Table A-2. Since the projected number of cycles through 60 years of operation does not exceed the number assumed in the fatigue analyses, the value is acceptable. The JAFNPP Fatigue Monitoring Program assures that the analyzed number of cycles is not exceeded.
Attachment I to SIR-07-084-NPS, Rev. I Page 16 of 37 Structural integrity Associates, Inc.
RAI 4.3.1-1. Part A (vii)
(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 (100°F/hr to 3750F);" 22, "Vessel Flooding (375 0F to 330OF in 10 min.);" and 23, "Cooldown (100°F/hr to 100°F)."
Response
Refer to Tables A-I and A-2. The revised design basis number of analyzed cycles for Events 19-23 is 233 cycles, as shown in Table A-I. To date, 126 of these events have occurred and 244 were previously projected to occur for 60 years, which is reflected in Table A-2.
However, as discussed in the answer to Part C below, a review of plant records was performed to classify the twelve unidentified events mentioned in the 1987 GE evaluation [8]. This review was able to identify that these twelve previously unidentified events, as well as one additional event, should be classified as shutdown and startup events. This results in a revised projection of 270 cycles for Events 19-23 for 60 years, as shown in Table C-2 below. The 40-year projected number of cycles for Events 19-23 is 186 cycles. Since the projected number of cycles through 40 years of operation does not exceed the number assumed in the fatigue analyses, the value is acceptable for the remaining operating period. However, since the projected number of cycles through 60 years of operation exceeds the number assumed in the fatigue analyses, corrective action may be required if the trend continues and accumulated numbers of cycles approach the analyzed value of 233. The JAFNPP Fatigue Monitoring Program assures that corrective action will be taken prior to the accumulation of more than 233 cycles.
Attachment I to SIR-07-084-NPS, Rev. I Page 17 of 37 Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part A (vii)
(f). A "Current Design Basis Cycles, Allowable" value of "V' and an "Updated 60 Year Cycle Projection" value of "1" for transient category 24, "Hydrostatic Test (1563 psig)."
Response
Refer to Tables A-I and A-2. The revised design basis number of analyzed cycles for Event 24 is one cycle, as shown in Table A-i. To date, one of these events have occurred and none are expected to occur in the future, since these events are no longer required by the ASME Code.
This is reflected in Table A-2. Since the projected number of cycles through 60 years of operation does not exceed the number assumed in the fatigue analyses, the value is acceptable.
The JAFNPP Fatigue Monitoring Program assures that the analyzed number of cycles is not exceeded. to SIR-07-084-NPS, Rev. I Page 18 of 37 Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part A (vii)
(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."
Response
Refer to Tables A-I and A-2. The revised design basis number of analyzed cycles for Event 25 is 35 cycles, as shown in Table A-I. To date, 17 of these events have occurred and 34 are projected to occur for 60 years, which is reflected in Table A-2. Since the projected number of cycles through 60 years of operation does not exceed the number assumed in the fatigue analyses, the value is acceptable. The JAFNPP Fatigue Monitoring Program assures that the analyzed number of cycles is not exceeded. to SIR-07-084-NPS, Rev. I Page 19 of 37 Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part B Page 19 of GE Design Calculation EAS-149-1286 / DRF B13-01391 discusses GE's 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 (AT) for six of these events had exceeded the AT value for this transient. The staff noted that the bases provided on page 19 forj ustifying 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:
RAI 4.3.1-1. Part B (i)
Explain why the six transients specified in GE calculation can be grouped into "Shutdown" transient for the plant when the AT values for these six events were determined to be excessive and the temperature gradients for the transients are not defined.
Response
Each of the six transients listed on page 19 of the 1987 GE report was reviewed against the record copy of the original reactor pressure strip charts and computer logs [0], as well as station Scram reports [15], to identify the time frame and rate of temperature change for each transient.
Whereas the AT for these events exceeded the AT for the Rapid Heatup and Rapid Cooldown events, the rate of change of temperature for the actual events was significantly less than the Rapid Heatup and Rapid Cooldown events. In fact, per the assessment provided below for each event, the rate of temperature change remained well within the limits of the Shutdown and Startup events. Therefore, based on the review provided below, these six events have been re-classified.
The information obtained for each of the six transients is summarized as follows:
1-9-83 Scram The Scram event occurred on 1-9-83 due to a failure of a relay that caused the 345kV breakers to open. The event continued until the plant reached atmospheric pressure in the RPV. The time it took for the change from 430"F to 100F (-330'F) was greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The rate of temperature change was at all times less than the design basis transient rate of I 000F/hr for the shutdown transient.
11-4-84 Scram The Scram event occurred on 11-4-84 with the plant at 30% power during startupfrom a planned maintenance outage when the condensate bypass controller failed with the reactor at low power.
The High Pressure Coolant Injection system started and injected to the reactor to maintain level.
Shutdown continued until zero pressure. The shutdown from 397°F to I 00°F (-297°F) took approximately 12.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The temperature change in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period remained less than 100F. Review of the event strip charts showed that the maximum rate of temperature change for the remainder of the shutdown event was approximately 50F/hr. Following the shutdown, the plant entered a startup to a pressure of approximately 940 psig. The startup event from I 00°F to 537°F began on 11-5-84 around 6:30 PM and took approximately 19.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
Attachment I to SIR-07-084-NPS, Rev. 1 Page 20 of 37 V
Structural Integrity Associates, Inc.
7-19-85 Scram The Scram occurred on 7-19-85 due to a turbine trip concurrent with a loss of the 10300 and 10500 buses while transferring house loads that led to a loss of condenser vacuum when the "A" Circulating Water pump tripped and air admission valves to 3 condenser water boxes opened.
The recovery event began on 7-20-85 with the reactor taking approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> to increase from 3647F to 540°F (+176*F). The highest rate of temperature increase occurred in the first three hours with temperature increasing from 364°F to 539*F at a rate of less than 60*F/hr.
7-26-85 Scram The Scram event occurred on 7-26-85 during turbine control valve testing when a Scram signal was received on two Reactor Protection System channels, triggering a full Scram signal. The recovery began on 7-27-85 and continued through 7-28-85. The time to increase 168'F was greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. For much of the time, the pressure held relatively constant. The highest rate of temperature increase occurred between 3640 F to 411 0F, which occurred over a time period of 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> with a rate of slightly more than 30°F/hr.
8-20-85 Scram The Scram event occurred on 8-20-85 when the "B" inboard Main Steam Isolation Valve (MSIV) closed while testing Main Steam line radiation monitors. The event occurred due to a latent failure of one of the solenoid valve coils designed to maintain the MSIV open during the surveillance test. The recovery began on 8-23-85 and continued through 8-25-85. The times were not consistently noted on the charts and the computer logs were illegible during much of that time. The highest observed rate of temperature increase occurred between 470'F to 518"F, which occurred over a time period of 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> with a rate of slightly more than 30°F/hr.
Based on the above review, the six transient events listed in Table 5 of the GE evaluation were re-classified as Shutdown and Startup events.
Attachment I to SIR-07-084-NPS, Rev. I Page 21 of 37 Structural integrity Associates, Inc.
RAI 4.3.1-1. Part B (ii)
For the scram event that occurred on November 4, 1984, a AT of -2970F and a AT of +4370F occurred on the same day, when did AT events occur and what were the actual temperature gradients associated with these events.
Response
Based on a review of all available plant records for this event, the Scram event occurred on 11 84 with the plant at 30% power during startup from a planned maintenance outage. The event occurred when the condensate bypass controller failed with the reactor at low power. The High Pressure Coolant Injection system started and injected to the reactor to maintain level. Shutdown continued until zero pressure. The shutdown from 397°F to 100°F (-297'F) took approximately 12.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The temperature change in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period remained less than 1 00'F. Review of the event strip charts showed that the maximum rate of temperature change for the remainder of the shutdown event was approximately 50°F/hr. Following the shutdown, the plant entered a startup to a pressure of approximately 940 psig. The startup event from 100lF to 537°F began on 11-5-84 around 6:30 PM and took approximately 19.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
Attachment I to SIR-07-084-NPS, Rev. I Page 22 of 37 S Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part B (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.
Response
Based on the response to RAI 4.3.1-1, Part B (i) above, the number of recorded occurrences for some of the transient categories in LRA Table 4.3-2 will change. The changes are described in the response to RAI 4.3.1-1, Part C below.
Attachment I to SIR-07-084-NPS, Rev. I Page 23 of 37 Structural itegrity Associates, Inc.
RAI 4.3.1-1. Part C RAI 4.3.1-1. 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.
Response
The operational transients and occurrences identified in the GE report were compared to the data supplied to GE by JAFNPP in 1986. This review verified that the number of operational transients, including the number of unidentified operational transients, is consistent with the input provided to GE.
In order to determine if the unidentified operational transients could be more accurately classified, a review of operating logs [14] for the time period in question was performed. From this data, all of the twelve unidentified operational transients were able to be located, identified and categorized. One additional shutdown event on 6/17/76 was identified during the review and some dates were modified slightly to coincide with the beginning of the events. The resulting classification of these events is provided in Table C-I.
Based on the information provided in Table C-I, the twelve unidentified events and the one additional event were all classified as Shutdown (Event 19-23) and Startup (Event 3) events.
Attachment I to SIR-07-084-NPS, Rev. I Page 24 of 37 Structural Integrity Associates, Inc.
Table C-1: Classification of Unidentified Transients for JAFNPP Event Description Classification Date 3/18/75 Shutdown due to "D" Emergency Diesel Generator failure Shutdown and inoperable "A" EDG. Plant shut down to cold condition.
& Startup 5/22/75 Shutdown following Turbine Trip/Reactor Scram as part of Shutdown startup test program. Plant shut down to cold condition.
& Startup 5/28/75 Shutdown to repair #1 main steam stop valve. Plant shut Shutdown down to cold condition.
& Startup 6/2/75 Reactor Scram and MSIVs closed during test. Continued in Shutdown (1 Hot Standby until startup commenced later that same day.
& Startup 6/11/75 Reactor Scram on MSIV isolation due to loose PCIS Shutdown 0) connection. Continued in Hot Standby until startup
& Startup commenced later that same day.
6/17/75 Reactor Scram on Low RPV water level. Plant shut down to Shutdown cold condition.
& Startup 12-25-75 Shutdown to hot standby to repair a main turbine hydraulic Shutdown ()
control system (EHC) oil leak from 40% power. Continued Shut in Hot Standby until startup commenced 12-26-75.
6-4-76 Turbine Trip/Reactor Scram due to Moisture Separator Shutdown ()
Reheater drain tank high level. Continued in Hot Standby
& Startup until startup commenced later that same day.
6-17-76 Reactor Scram during surveillance test. Continued in Hot Shutdown l Standby until startup commenced later that same day.
& Startup 6-18-76 Reactor Scram during startup due to exceeding Ist stage Shutdown (1) pressure setpoint with Turbine Stop Valves shut. Continued
& Startup in Hot Standby until startup commenced 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> later.
6-30-76 Turbine trip/Rx Scram due to loss of 125VDC input to Shutdown ()
EHC/load unbalance circuit. Continued in Hot Standby until
& Startup startup commenced later that same day.
8-30-76 Reactor Scram while replacing Reactor Water Recirculation Shutdown ()
flow transmitter. Continued in Hot Standby until startup commenced 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> later.
& Startup 1-22-77 Reactor Scram on high neutron flux at 44% power.
Shutdown )
Continued in Hot Standby until startup commenced 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />
& Startup later.
&_Startup Note:
- 1. Classifying the event as a Shutdown (including each of Events 19 through 23) and Startup (Event 3) is conservative because the temperature and pressure changes are significantly greater than for the Scram event (Event #15). to SIR-07-084-NPS, Rev. I Page 25 of 37 Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part C (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.
Response
ENN has determined that the number of cycles used in the 2002 SIA evaluation [2, 10] to project through 60 years of operation was incorrectly determined [ 12] because the evaluation did not properly account for the twelve unidentified transients assigned to the Scram transient category from the 1987 GE analysis [8]. This resulted in an inaccurate estimate of current transient cycles and an inaccurate 60 year transient cycle projection for the Scram event (Event 15). The corrected number of events is provided in Table C-2, considering the updated classifications shown in Table C-1, and the re-classification of the six Scram events discussed in the response to RAI 4.3.1-1, Part B (i).
Attachment I to SIR-07-084-NPS, Rev. I Page 26 of 37 V
Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part C (iii)
In the GE stress report, GE also mentioned that the change in AT associated with these 12 unidentified transients was approximately 3300F. The staff requests the following additional information:
(a). Please define these unidentified transients and list the pressure-temperature data for these transients.
Response
The change in AT of 330'F was not for an unidentified transient, but rather for the Scram which occurred on 1-9-1983. The information related to this transient event was previously discussed in the response to RAI 4.3.1-1, Part B (i). Available data was reviewed and used to further investigate the twelve unidentified transients. As indicated in the response to the previous question, these events have been re-classified as shutdown and startup events. Thus, the maximum possible temperature differential has been assumed to occur for all of the previously unidentified events, thereby making these classifications conservative. to SIR-07-084-NPS, Rev. I Page 27 of 37 Structural integrity Associates, Inc.
RAI 4.3.1-1. Part C (iii)
(b). Please define the pressure-temperature (P-T) data that were used for the limiting SCRAM event used in S IA's updated 60 year cumulative usage factor calculations.
Response
The pressure-temperature data is obtained from the original thermal cycle diagrams [1, 17] for JAFNPP and was modified for power uprate, according to References [9, 18]. The evaluation used the governing stress analyses for the closure bolts as a basis to estimate fatigue usage. The remaining three components (recirculation inlet nozzle, feedwater nozzle and CRD penetrations) were reanalyzed using finite element methods. These transients are defined in the SIA evaluation [19, 20, 21). Figures C-I through C-3 identify the Scram transients used in the SIA calculations for the feedwater nozzle. Figures C4 through C-8 identify the Scram transients used in the SIA calculations for the recirculation inlet nozzle. Figures C-9 through C-13 identify the Scram transients used in the SIA calculations for the control rod drive nozzles. These transient definitions reflect the unique pressure and temperature severity of the design basis Scram event, modified for power uprate conditions, for each evaluated component. The severity is different for each component due to the different thermal regions of the vessel, e.g., the feedwater nozzle is affected by incoming feedwater flow. Consistent with the design basis, the Scram event definitions also include the subsequent return to full power conditions after the initiating Scram.
Te.p('F) -
-Pe (psW)
Boo 13201280 550 i240 S1280 600 !
- 112 1 11080120 1 040 E400 no__
3000I:Q 0
5000 10000 15000 2000 2M000 Tnin 1
)ore Figure C-I: Transient 11 for Feedwater Nozzle Attachment I to SIR-07-084-NPS, Rev. I Page 28 of 37 Structural Integrity Associates, Inc.
T- -
-F- -
s0 I
00-I
-30 970 1970 2970 3970 4970 T"- (
O
)
Transient 12 for Feedwater Nozzle Figure C-2:
[ -T@vri
(*F) -
-Pre-osir pg) 450 400 350
-1000
-950 9w0
-000 750
. 700 850-01W 450 A
400 350 300 200 150 100 50 0
1000 2000 30 4M 5000 T'm" (Seconds)
Figure C-3: Transient 14 for Feedwater Nozzle Attachment I to SIR-07-084-NPS, Rev. I Page 29 of 37 C
Structural Integrity Associates, Inc.
1-Tn F) -
-PftosmzI(ppsW 500M 10000 1
2000 2
n-s (m)cnd Figure C-4: Transient 11 for Recirculation Inlet Nozzle
[-TeMp (T) --
Psue(l)
"s5 NO0 mat speomdStfM3 w~
ftM Sf sr o n anptdet Un1 VA19
-1180 1140 1100 1060o 940 900
-W0 970 1970 2970 3970 4970 9970 6970 Figure C-5: Transient 12 for Recirculation Inlet Nozzle 7970 Attachment I to SIR-07-084-NPS, Rev. I Page 30 of 37 1
Structural Integrity Associates, Inc.
1-TietwC)
Ruswe (piow 555 50 j 548 E
53-530-520
-1400 mius spice oaons dufmg f ir 32 seconds of 1380o Ummulief Thhis is nt ainmaiiMplolduesto
-J S
1220
-1340
~-1300 I
- 12M0
~11i0*
11*0
-1140j 11080 1020 11000 0800 1040 4-92
-32 968 168 29H8 3968 TmM 49 5098 6988 709 Figure C-6: Transient 13 for Recirculation Inlet Nozzle
-*Temp
("F) --
-Pressure (piIg) 600 500 JE T110011000 8500
-750
-700 850
.500 1450 I-10 I
2no 100 0
2M00 4000 8000 80oo 10000 12000 14000 10m0 s1=000 mm ondC)
Figure C-7: Transient 14 for Recirculation Inlet Nozzle Attachment I to SIR-07-084-NPS, Rev. I Page 31 of 37 V
Structural Integrity Associates, Inc.
I-Teavn)
-ftPe~swe(psig) 555 560 545 1540 535 530 1080
-1040 II 1000 980
-15 985 1985 2985 3985 4985 5985 6985 7985 TWM ("m"do)
Figure C-8: Transient 15 for Recirculation Inlet Nozzle Temp (TF)
-Pressure (psig) j3W A._
0 5000 10000 15000 20000 25000 30000 35000 40000 45000 Tnme (sconds)
Figure C-9: Transient 11 for CRD Nozzle.
Attachment I to SIR-07-084-NPS, Rev. 1 Page 32 of 37 Structural Integrity Associates, Inc.
I-Temp (F)
-Pressure (psig)
,1596-
~540 1020 Pressurespiie o
IrsdxnT first 30 T
seconds of transient This is not seen on matin plot due to time scale..
525 -
-30 4970 9970 14970 19970 Time (seconds)
Figure C-10: Transient 12 for CRD Nozzle
-I Temp (F) I
-Pressure (p519) 555 140D E154 I
-32 4968 9968 14968 19968 Time (seconds)
Figure C-11: Transient 13 for CRD Nozzle Attachment I to SIR-07-084-NPS, Rev.. I Page 33 of 37 Structural Integry Associates, Inc.
I-TeM (F)- -Prsuspa) 600 1100 1050 I
0 2000 40(0 000 8000M 10000 1200 14000 16000 18000 20000 TlIn (sconds)
Figure C-12: Transient 14 for CRD Nozzle I-r-W (f) -
-Pr-.u. (piq) 540 535 530 5'.
/
/
1040 1000 I
/
/
7
-is 4985 9M85 145 1998 Fgr (transi)
Figure C-13: Transient 15 for CRD Nozzle Attachment I to SIR-07-084-NPS, Rev. 1 Page 34 of 37 Structural integrity Associates, Inc.
RAI 4.3.1-1. Part C NO I)
(c). Justify how these 12 unidentified transients are characterized based on the analyzed P-T limit data used in SIA's updated cumulative usage factor calculations.
Response
The unidentified transients were classified as Scram events (Event 15) by GE in the 1987 fatigue update. As discussed in the response to Part C (i) above, the previously unidentified events have been re-classified as shutdown and startup events based on a review of additional JAFNPP data.
This is conservative because these events result in the maximum pressure and temperature range and the associated fatigue usage is higher than for the Scram event. Revised projections for all affected events are provided in Table C-2.
Attachment I to SIR-07-084-NPS, Rev. I Page 35 of 37 Structural Integrity Associates, Inc.
RAI 4.3.1-1. Part C (iii)
(iv). Clarify how your response to this part (Part C) factors into your response to Part A, particularly with respect to the recording the num ber 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.
Response
Due to the re-classification of events described in the responses in RAI 4.3.1-1, Part B (i) and Part C (i) above, the number of shutdown, startup and Scram transient events require modification from the values shown in JAFNPP LRA Table 4.3-2.
Note that the revised number of Scrams (Event # 15) has been changed because the SIA evaluation [10] accounted for the twelve unidentified events in the Year 3 number of Scram events, but not in the Year 26.5 number of Scram events. In addition, the re-classification of the six Scram events described in the response to Part B (i) further reduced the number of Scram events, and increased the number of startup and shutdown events.
The results of these changes are shown in Table C-2.
Attachment I to SIR-07-084-NPS, Rev. I Page 36 of 37 Structural Integrity Associates, Inc.
Table C-2: Updated Transient Event Projections for Startup, Scram & Shutdown Transients Revised Revised Revised Revised Year 60 Current Event Design Transient Year 3 No.
Year 13 Year 30.5 Projected Design Basis Number of Cycles No. of No. of No. of Analyzed No.
Cycles Cycles Cycles of Cycles (4) 3 Startup (100°F/hr to 546*F 30 ()
68 (2) 132 (2) 242 233 at 100-F/hr)
~SCRAMs 11 Loss of FW Pumps, MSIVs 0
4 5
11 12 Close 12 Turbine Generator Trip, FW on, 2
9 10 12 12 MSIVs stay open 13 Reactor Overpressure 14 Single Relief I
I I
1 2
Valve Blowdown 15 All Other Scrams 20 48 51 7y) 57 64 19-23 Shutdowns 29_(_
67 (_
_145_
270 233 Notes:
- 1. The number of cycles for Event 3 has been revised from 17 and for Event 19-23 has been revised from 16, which was used in the reference [2] analysis to include the 13 additional events listed in Table C-I.
- 2. The number of cycles has been revised to include the 13 additional events listed in Table C-1, and the 6 additional events described in the response to RAI 4.3.1-1, Part B(i). The number of Year 13 cycles has been revised from 49 and for Event 19-23 has been revised from 48. The number of Year 30.5 cycles has been revised from 113 and for Event 19-23 has been revised from 126.
- 3. The number of Year 13 cycles has been revised from 54 and the number of Year 30.5 cycles has been revised from 57 to exclude the 6 events described in the response to RAI 4.3. 1-1, Part B(i).
- 4. From Table A-I of SIA analysis.
Attachment I to SIR-07-084-NPS, Rev. I Page 37 of 37 Structural Integrity Associates, Inc.
JAFP-07-0079 Docket No. 50-333 James A. FitzPatrick Nuclear Power Plant License Renewal Application - Amendment 12 Update Previous RAI Responses:
3.5.2-4 3.6.2-1 4.2.6-1 Table 3.3.2-13 LRA Revisions associated with Class 1 Fatigue
The following information revises the previously submitted response in Attachment I of LRA Amendment 10.
RAI 3.5.2-4 Revised Response How is JAFNPP monitoring the vent pipe bellows? Has JAFNPP considered a Type B test?
Discussion: The applicant indicated that a Type B test is performed once every 10 years. The applicant will provide a supplemental response.
Revised Response:
JAF performs the Type "B" Leak Rate Test once every ten years in accordance with ST-39B and ST-39B-X201. The testing interval is in accordance with the requirements of Appendix J.
Vent Line to Torus penetration bellows consist of two sections of two-ply stainless steel bellows.
Type B LLRT testing consists of pressurizing the space between the two plies of each bellows section, and measuring leakage as inlet flow to this space. This effectively tests all of the surface area of each bellows section.
The rest of the penetration assembly, including the vent insert in the Torus shell and mounting plates connecting the bellows to the vent piping and vent insert, is carbon or stainless steel of welded construction. Type A ILRT testing includes pressurizing the assembly from the Torus airspace, and measuring leakage as inlet flow to the Containment. This effectively tests all of the surface area of the assembly except the two two-ply bellows sections. Therefore, the combination of Type A and Type B testing effectively tests the entire assembly.
As noted in the response to NRC audit question 200 (provided in JAFP 07-0048, dated April 6,2007), there is no history at JAF of exposure of this material to corrosive contamination; neither is there any history of corrosion or other degradation of the assembly.
There is no history of leakage of the bellows assemblies under Type A or Type B testing.
Exposed inner (i.e., Torus side) surfaces of the assemblies are viewed during Type B testing and during other Torus internal inspections. There is no convenient method for inspecting the unexposed portions of the assemblies, and no perceived need to do so in light of the available history.
The following information supplements the previously submitted response to RAI 3.6.2-1 in Attachment 2 of LRA Amendment 9.
Add the following to the Periodic Surveillance and Maintenance Program:
A power factor or partial discharge test will be performed in accordance with industry standards.
The initial test will be completed prior to the period of extended operation. The frequency of the test will be adjusted based on the initial test results; the test frequency shall be at least once every ten years.
Page 1 of 6 JAFP-07-0079
LRA Table 3.6.2-1 is revised as shown below (strike-outs deleted, underlined text added).
Component Type Intended Material Environment AERM AMP NUREG Table I Notes Function
-1801, Item Vol.2 Item Oil-filled cable Pressure Carbon Mineral oil Loss of Oil analysis J
system - MECH boundary
- steel, (internal) material External (passive stainless Outdoor surfaces mechanical for
Conducts bnsFa'-t*en Outdoor NPe6 Periodic J, 602 ELEC (passive electricity material-Weather, Soil Breakdown Surveillance electrical for SBO vaPieus Voltage of and recovery)
Organic Stress insulation Preventive polymers leading to Maintenance electrical failure Notes for Table 3.6.2-1 602 - Based on vendor information this transmission cable is not subject to water treeing, since it is designed for continuously wet conditions. Industry and plant operating experience has not provided any information on failures of this type of cable. The only portion of the cables exposed to the environments (outdoor weather and soil) is the okolene (black polyethylene) outer jacket, which is over the lead sheath and serves as an anti-corrosion and moisture protection. These environments do not affect the oil impregnated paper insulation. However, breakdown of insulation (reduced insulation resistance) leading to electrical failure will be managed by the Periodic Surveillance and Maintenance Program.
The following information supplements the LRA FSAR section A.2.2.1.6, Reactor Vessel Axial Weld Failure Probability, with information previously submitted in response to RAI 4.2.6-1 in Attachment I of LRA Amendment 6.
Add the following paragraph to LRA FSAR section A.2.2.1.6:
A.2.2.1.6 Reactor Vessel Axial Weld Failure Probability The BWRVIP-74 SER states it is acceptable to show that the mean RTNDT of the limiting beltline axial weld at the end of the period of extended operation is less than the value given in Table 1 of the BWRVIP-74 SER. This value supports the axial weld failure probability and is based on the assumption of essentially 100% (> 90%) inspection of the axial welds in the beltline region.
Due to various obstructions within the reactor vessel, JAFNPP is able to inspect approximately 88% of the axial welds in the beltline region. The NRC granted a relief request for less than 90% coverage. The projected 54 EFPY mean RTNDT value for JAFNPP is well below the limiting mean RTNDT of 114 °F. The 2% difference in the amount of inspected weld will not offset the 16.8 OF margin between the 97.2 OF mean RTNDT for JAFNPP and the 114 OF mean RTNDT used in the NRC SER for BWRVIP-74. Therefore, the axial weld failure probability will not exceed 5 x 10-6 per reactor operating year during the period of extended operation. As such, this TLAA has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).
Page 2 of 6 JAFP-07-0079
The following information revises the previously submitted response in Attachment 2 of LRA Amendment 11 for LRA Table 3.3.2-13.
Security Generator System Changes LRA Table 3.3.2-13, Security Generator System Summary of Aging Management Evaluation, is revised changing Note E to Note A as shown below. (strikeouts deleted, underlined text added)
NUREG Aging Effect Aging Notes Componen Intended Aging Magint
-1801, Table 1 Type Function Material Environment Requiring Management Vol.2 Item Management Programs Item VII.H2-3.3.1-14 E-A Sight Pressure Carbon Loss of glass boundary steel material Oil analysis 20 1
1_
1
, (AP-30)
Based on the information provided in Attachment 1 for LRA section 4.3.1, Class I Fatigue, the following revisions to the LRA have been completed.
LRA Section 4 Changes Table 4.1-1, List of JAFNPP TLAA and Resolution, is revised as shown below. (strikeouts deleted, underlined text added)
TLAA Description Resolution Option Section Analyc-; r
-alid 10 CFR 51.21 (G)(1 )i OR Analysis projected Class 1 fatigue 10 CFR 54.21 (c)(1)(ii) 4.3.1 OR Aging effect mana-ged 10 CFR 54.21 (c)(1)(iii)
Analysis projected 10- C-FR 54.21 (G)(4141ii Effects of reactor water OR environment on fatigue Analye-will be8 projected 4.3.3 life OR Aging effect managed 10 CFR 54.21(c)(1)(iii)
Page 3 of 6 JAFP-07-0079
Section 4.3.1.1, Reactor Vessel, is revised as follows. (strikeouts deleted, underlined text added)
The reactor pressure vessel was designed in accordance with ASME Section II1. Fatigue analyses were performed as required based on an allowed number of transient cycles.
An evaluation of fatigue usage factors was performed in 2002 accounting for sixty years of operation. This analysis projected that all components of the vessel would have fatigue usage factors below 1.0. Not all reactor vessel components have fatigue usage factors.
Fatigue analyses were originally performed for limiting components of the vessel, as listed in Table 4.3-1. Fatigue usage factors for other vessel components not listed in Table 4.3-1 are bounded by the most limiting location. The Fatigue Monitoring Progqram will assure that the analyzed numbers of transients used in fatique calculations are not exceeded during the period of extended operation.
Therefore, the effects of aging associated with TLAA (fatigue analyses) for reactor pressure vessel fatigue re.main. v.lid for the period" of extended operatione inccRd.nc.
i..th 10-CFR 54.21 (G)(4)(4) are managed per 10 CFR 54.21 (c)(1)(iii).
Section 4.3.1.2, Reactor Vessel Internals, is revised as follows. (strikeouts deleted, underlined text added)
A fatigue evaluation was also performed on the tie rod assemblies installed as part of the core shroud repair. The maximum CUF for the tie rod components is 0.0575 for the spring rod based on 120 startups/shutdowns. The eiFeRt proiected number of startups and shutdowns is 242 and 270 respectively/sh'-tdGw..
allowed for 60 years of operation is 2-33. Therefore, a conservative projection of the fatigue usage of the tie rods for 60 years of operation would be (2-33270/120) x 0.0575, which equals a CUF of 04-40.13.
Table 4.3-2, Projected Cycles, is revised to replace the 60 year cycle projections as follows.
Transient 3, Startup - Replace 216 with 242.
Transient 11, Loss of FW Pumps, MSIVs Close - Replace 10 with 11.
Transient 15, All Other Scrams, Replace 62 with 57.
Transients 19-23, Shutdowns, Replace 244 with 270.
Table 4.3-2, Projected Cycles, is revised adding the following footnote for transients 13, 17, and
- 18. (underlined text added) 2Although zero events are anticipated, the analyzed number of design basis cycles remains unchanged.
Page 4 of 6 JAFP-07-0079
Section 4.3.3, Effects of Reactor Water Environment on Fatigue Life, is revised as follows (strikeouts deleted, underlined text added)
The effects of environmental-assisted thermal fatigue for the limiting locations identified in NUREG-6260 have been evaluated. Depen.ding on the option chosen, Whi"h may va,;
Fby compoen.t, this T-LAA will be projected through the period. of eten*-,ded ope.ation or the'The effects of environmentally assisted fatigue will be managed per 10 CFR 54.21(c)(1)(iii). Fei t~hoceP locAa~tion_ with CU-FFe leers, than 1.0, the TL1APA. hac bheen projected through the period of extended operation pcr1OCF=R51.1()1(i)
LRA Appendix A Changes Section A.2.1.24, Periodic Surveillance and Preventive Maintenance, is revised to add the following item to the list of components inspected by the program. (underlined text added)
Internal surfaces of carbon steel components in the floor and roof drainage system Section A.2.2.2.1, Class 1 Metal Fatigue, is revised to add the following to the last sentence of this section. (strikeouts deleted, underlined text added)
Consequently, the TLAA (fatigue analyses) based on those transients wIll* "rmain" V"ald for the pero*d of *e..tende*d*
operation in accr-,-,-,d.anceP_ W*th 10- CFR 51.21 (c)(4 )(i-or are projected through the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii)7 or the aging effect is managed per 10 CFR 54.21(c)(1)(iii).
Section A.2.2.2.3, Environmental Effects on Fatigue, is revised as follows. (strikeouts deleted, underlined text added)
The effects of reactor water environment on fatigue were evaluated for license renewal.
Projected cumulative usage factors (CUFs) were calculated for the limiting locations identified in NUREG/CR-6260. For= the l""at"nA wi;.th CU*
lRes than 1.0, the TLAA has been projected through the period of eAended operation per 10.
.FR 51.21
- ()(1)(ii).
Several locations may exceed a CUF of 1.0 with consideration of environmental effects during the period of extended operation. For these locations, prior to the period of extended operation, JAFNPP will (1) refine the fatigue analysis to lower the predicted CUF to less than 1.0; (2) manage fatigue at the affected locations with an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC); or (3) repair or replace the affected locations.
Depending on the
,pti9, ch9s*e, which mnay r Y,] by comRponent, this 1 A
AP. w beh projected through the perFid of Ox)t*ded operatien oar the The effects of environmentally assisted fatigue will be managed per 10 CFR 54.21 (c)(1)(iii).
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(
LRA Appendix B Changes Section B. 1.20, Oil Analysis, is revised to clarify the description of underground oil filled cables (strikeouts deleted, underlined text added)
Attributes Affected Enhancements
- 1. Scope of Program The Oil Analysis Program guidance documents will be enhanced to periodically sample lubricating oil in tho underground oil fillod cGble, the security generator and the fire pump diesel, as well as the oil internal to underground oil filled cables.
Section B.1.27.2, Structures Monitoring, is revised to clarify the physical location of lubrite surfaces (strikeouts deleted, underlined text added)
Attributes Affected Enhancements
- 4. Detection of Aging Effects Guidance for performing periodic inspections to confirm the absence of aging effects for lubrite surfaces in the teor-w drywell radial beam seats will be added to the Structures Monitoring Program procedure.
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