ML082671120
| ML082671120 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 09/16/2008 |
| From: | Laughlin G Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML082671120 (12) | |
Text
Constellation Point Nuclear Station Energy P.O. Box 63NY 13093 Lycoming, Nine Mile September 16, 2008 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION: Document Control Desk
SUBJECT:
Nine Mile Point Nuclear Station Unit No. 1; Docket No. 50-220 Extension of Permanent Relief from Inservice Inspection Requirements of 10 CFR 50.55a(g) for the Volumetric Examination of Reactor Pressure Vessel Shell Circumferential Welds for the License Renewal Period of Extended Operation
REFERENCES:
(a) Letter from S. S. Bajwa (NRC) to J. H. Mueller (NMPC) dated April 7, 1999, Alternatives for Examination of Reactor Pressure Vessel Shell Welds, Nine Mile Point Nuclear Station, Unit I (TAC No. MA4383)
(b) Letter from J. A. Spina (NMPNS) to Document Control Desk (NRC) dated July 14, 2005, Recovery of Nine Mile Point License Renewal Application Quality (TAC Nos. MC3272 and MC3273)
(c) EPRI Report TR-105697, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), dated September 1995 (d) Letter from G. C. Lainas (NRC) to C. Terry (BWRVIP) dated July 28, 1998, Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. M93925)
(e) Letter from J. R. Strosnider (NRC) to C. Terry (BWRVIP) dated March 7, 2000, Supplement to Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. MA3395)
(f) NRC Generic Letter 98-05, Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds, dated November 10, 1998 (g) BWRVIP-74-A, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal, June 2003
Document Control Desk September 16, 2008 Page 2 By letter dated April 7, 1999 (Reference a), the NRC approved an alternative to performing reactor pressure vessel (RPV) shell circumferential weld examinations for Nine Mile Point Unit 1 (NMP1) pursuant to the provisions of 10 CFR 50.55a(a)(3) and 10 CFR 50.55a(g). The alternative allowed permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV shell circumferential welds (i.e., American Society of Mechanical Engineers Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item BI.11, Circumferential Shell Welds) for the duration of the original NMP1 operating license (i.e., until August 22, 2009).
On October 31, 2006, the NRC issued the Renewed Operating License for NMP 1, with an expiration date of August 22, 2029. In order to allow continued use of the subject alternative for the 20-year period of extended operation (i.e., from August 23, 2009 to August 22, 2029), Nine Mile Point Nuclear Station, LLC (NMPNS) requests NRC approval of the attached 10 CFR 50.55a Request Number IISI-001A. This request would allow permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV shell circumferential welds for the license renewal period of extended operation. Submittal of this request satisfies a commitment contained in Section 4.2.3 of the amended NMP1 License Renewal Application (Reference b).
The technical basis providing justification for this request is contained in report BWRVIP-05 submitted by the Boiling Water Reactor Vessel and Internals Project (BWRVIP) to the NRC (Reference c). The NRC evaluated this report and responses to Requests for Additional Information, and issued safety evaluations to the BWRVIP (References d and e). The requested alternative is consistent with the guidance provided in NRC Generic Letter 98-05 (Reference f), as supplemented by report BWRVIP-74-A (Reference g).
NMPNS requests approval of this request for alternative by August 22, 2009. This letter contains no new regulatory commitments.
Should you have questions regarding the information in this submittal, please contact T. F. Syrell, Licensing Director, at (315) 349-5219.
Very truly yours, GJL/DEV
Attachment:
Nine Mile Point Nuclear Station, Unit 1, Fourth Inservice Inspection Interval - 10 CFR 50.55a Request Number IlSI-001A cc: S. J. Collins, NRC R. V. Guzman, NRC Resident Inspector, NRC
ATTACHMENT NINE MILE POINT NUCLEAR STATION, UNIT 1 FOURTH INSERVICE INSPECTION INTERVAL 10 CFR 50.55a REQUEST NUMBER 1ISI-001A Nine Mile Point Nuclear Station, LLC September 16, 2008
Nine Mile Point Nuclear Station, Unit 1 Fourth Inservice Inspection Interval 10 CFR 50.55a Request Number IlSI-001A Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
A. COMPONENT IDENTIFICATION System: Reactor Pressure Vessel Class: Quality Group A, ASME Code Class 1 Component
Description:
Volumetric Examination of all Pressure Retaining Reactor Pressure Vessel Shell Circumferential Welds Components Affected:
Circumferential Welds Description Code Code Item Caeory Number, RVWD-100 Circumferential Shell Weld B-A B1.11 RVWD-101 Circumferential Shell Weld B-A B1.11 RVWD-137 Circumferential Shell Weld B-A B1.11 RVWD-1 38 Bottom Head to Shell Weld B-A B1.11 B. APPLICABLE CODE REQUIREMENTS The applicable ASME Code, Section Xl, for the Nine Mile Point Nuclear Station (NMPNS), Unit 1 (NMP1), Fourth 10-Year Interval, In-service Inspection Program is the 2001 Edition through 2003 Addenda. The fourth 10-year interval will begin on August 23, 2009, concurrent with the NMP1 license renewal period of extended operation.
In accordance with the provisions of 10 CFR 50.55a, Codes and Standards, paragraph 10 CFR 50.55a(a)(3)(i),
Nine Mile Point Nuclear Station, LLC, (NMPNS) requests permanent relief for the NMP1 license renewal period of extended operation, from the requirements of ASME Code, Section Xl, Sub article IWB-2500, Table IWB-2500-1, Volumetric Examination of Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, Examination Item Number B1.11, Circumferential Shell Welds. See Figure 1 for weld locations.
C. REASON FOR REQUEST FOR RELIEF The technical basis providing justification for the permanent elimination of the examination requirement of the RPV shell circumference welds is contained in BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations", (Reference 1). In the report, the Boiling Water Reactor (BWR) Vessel and Internals Project (BWRVIP) concluded that the probabilities of failure for BWR RPV circumferential welds are orders of magnitude lower than that of the longitudinal welds. The NRC staff conducted an independent risk-informed, probabilistic fracture mechanics assessment (PFMA) of the analysis contained in BWRVIP-05 (Reference 1), and the results are documented in the NRC's final safety evaluation of the BWRVIP-05 report (Reference 2). This assessment concluded that the probability of failure of the BWR RPV circumferential welds is orders of magnitude lower than that of the axial shell welds and the added risk caused by not inspecting the circumferential welds is negligible.
Additionally, the NRC assessment demonstrated that inspection of BWR RPV circumferential welds does not measurably affect the probability of failure. Therefore, NMPNS has determined that the proposed alternative described below provides an acceptable level of quality and safety and satisfies the requirements of 10 CFR 50.55a(a)(3)(i).
Nine Mile Point Nuclear Station, Unit I Fourth Inservice Inspection Interval 10 CFR 50.55a Request Number IlSI-001A D. BASIS FOR RELIEF AND ALTERNATIVE EXAMINATIONS Proposed Alternative In accordance with 10 CFR 50.55a(a)(3)(i), and consistent with information contained in NRC Generic Letter 98-05, (Reference 4) and in the NRC safety evaluation for BWRVIP-74-A (Reference 10) NMPNS will implement the following alternate provisions for the subject weld examinations.
The failure frequency for ASME Code Section Xl, Table IWB-2500-1 Examination Category, B-A, Item No. B13.11, "Reactor Pressure Vessel Shell Circumferential Welds," is sufficiently low to justify their elimination from the in -
service inspection (ISI) requirement of 10 CFR 50.55a (g) based on the NRC Safety Evaluation. (Reference 2)
The ISI examination requirements of the ASME Code Section XI, Table IWB-2500-1 Examination Category B-A, Item No. B11.12, "Reactor Pressure Vessel Shell Longitudinal Welds," shall be performed, to the extent possible, and shall include inspection of the RPV Shell Circumferential Welds only at the intersection of these welds with the longitudinal welds, or approximately 2 to 3 percent of the RPV shell circumferential welds. The proposed alternative for volumetric examination of the RPV shell welds includes performing an examination, from the external outside diameter (OD) surface or where access is practical from the internal inside diameter (ID) surface of the RPV to the maximum extent possible. The examination of the remaining accessible portions of the RPV circumferential shell welds will be permanently deferred for the life of the original license and the license renewal period of extended operation.
The procedures for these examinations shall be qualified such that flaws relevant to the RPV integrity can be reliably detected and sized, and the personnel implementing these procedures shall be qualified in the use of these procedures. Qualification and examination will be completed in accordance with the 2001 Edition through 2003 Addenda of ASME Section Xl, Appendix VIII as modified by the Performance Demonstration Initiative (PDI) and 10 CFR 50.55(a), "Codes and Standards."
Basis for Relief The technical basis providing justification for the permanent elimination of the examination requirement of the RPV shell circumference welds is contained in a report (BWRVI P-05, "BWR Reactor Pressure Vessel Shell Weld
- Inspection Recommendations" - Reference 1), that was transmitted to the NRC in September 1995 and supplemented by letters dated June 24 and October 29, 1996, May 16, June 4, June 13 and December 18, 1997, and January 13, 1998. The NRC staff conducted an independent risk-informed assessment of the analysis contained in BWRVIP-05 as documented in the final safety evaluation of the BWRVIP-05 report (Reference 2) and the supplement to final safety evaluation (Reference 3). This assessment concluded that the probability of failure of the BWR RPV circumferential welds is orders of magnitude lower than that of the axial shell welds and the added risk caused by not inspecting the circumferential welds is negligible. Additionally, the NRC assessment demonstrated that inspection of BWR RPV circumferential welds does not measurably affect the probability of failure.
The NRC issued Generic Letter 98-05, (Reference 4), permitting BWR licensees to request permanent relief from the in-service inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV shell circumferential welds, ASME Section XI, Table IWB-2500-1, Examination Category B-A, Item B1.11. The NRC stated in that BWR licensees may request permanent relief for the remaining current license period by demonstrating that:
(1) At the expiration of their license the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the NRC staff's July 28, 1998, safety evaluation (Criterion 1),
and (2) Licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the NRC staffs July 28, 1998, safety evaluation (Criterion 2).
Nine Mile Point Nuclear Station, Unit 1 Fourth Inservice Inspection Interval 10 CFR 50.55a Request Number IlSI-001A For the original operating license period, the NRC authorized the alternative allowing permanent relief from the ISI requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV circumferential welds by letter dated April 7, 1999 (Reference 13).
This request also demonstrates that the safety criteria specified in BWRVIP-74-A (Reference 9) and the associated October 18, 2001 safety'evaluation (Reference 10) will continue to be met for the license renewal period of extended operation.
BWRVIP-74-A (Reference 9) provides generic guidelines intended to present the appropriate inspection and flaw evaluation recommendations to assure safety function integrity of the RPV components during both the current operating term and the license renewal term. The NRC staffs review of BWRVIP-74 was provided by safety evaluation (SE) dated October 18, 2001 (Reference 10), which concluded that Appendix E of the July 28, 1998 SE for BWRVIP-05 conservatively evaluated BWR RPVs to 64 effective full power years (EFPY), which is 10 EFPY greater than what is realistically expected for the end of an additional 20-year license renewal period. Therefore, the staffs analysis provided a technical basis for relief from the current ISI requirements of the ASME Code Section Xl for volumetric examination of the circumferential welds as they may apply for the license renewal period. The October 18, 2001 SE further stated that to obtain relief, each licensee will have to demonstrate that:
(1) At the end of the renewal period, the circumferential welds will satisfy the limiting conditional failure probabilities for circumferential welds in Appendix E of the NRC staffs July 28, 1998 SE for BWRVIP-05, and (2) They have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the NRC staffs July 28, 1998 SE for BWRVIP-05.
Criterion (1) - ConditionalFailureProbability Demonstrate that at the expiration of the license (initial and renewed), the RPV shell circumferential welds will continue to satisfy the limiting conditional failure probability for RPV shell circumferential welds that is established in the July 28, 1998 Safety Evaluation.
Response
In order to demonstrate that the circumferential welds satisfy the July 28, 1998 NRC safety evaluation limiting conditional failure probabilities, a comparison of the chemistry values and the predicted fluence at the end of the original license period can be made. Note that the NMP1 current license period is equivalent to 28 EFPY.
However, for the purpose of the inspection relief for the initial 40-year license, NMP1 used values for 32 EFPY to compare against the NRC 32 EFPY values presented in the SER. In addition, failure probabilities are also calculated for NMP1 at 46 EFPY, which corresponds to the end of the license renewal period of extended operation.. For the license renewal period, it is more appropriate to compare the change in failure probabilities since the NRC analysis did not consider the effects of the license renewal period (added fluence and crack growth). In this evaluation, the change in risk is the governing factor determined by the difference between the probability of failure at the end of the license renewal period (46 EFPY) and the end of the original license period (28 EFPY). The NMP1 request for relief for the original license period is given in Reference 11, and the NRC's authorization is documented in Reference 13.
For the original license period, Table 1 illustrates that NMP1 has conservatism in comparison to NRC Final Evaluation of BWRVIP-05 Limiting Plant Specific Analysis (comparing 32 EFPY values). The chemistry factor, adjustment for reference temperature (A RTNDT), and mean RTNDT, are calculated consistent with the guidelines of NRC Regulatory Guide 1.99, Rev. 2 (Reference 16). The data presented for NMP1 in the BWRVIP response to the NRC Request for Additional Information (RAI) on BWRVIP-05 is also shown in Table 1. The maximum Cu%
and Ni% variability from the most current data available is also bounded. The fluence values in Table 1 for 28 EFPY and 46 EFPY (from Reference 15) bound the highest fluence beltline circumferential weld, and were calculated using methods that are in accordance with Regulatory Guide 1.190 (Reference 17) and have been previously reviewed and approved by the NRC (References 5 and 6).
Nine Mile Point Nuclear Station, Unit 1 Fourth Inservice Inspection Interval 10 CFR 50.55a Request Number IISI-001A Table 1: Comparison of Input Parameters for NRC Staff Assessment and BWRVIP Methodology Nine Mile Point I NRC Staff Assessment (Circumferential Weld) for 32 EFPY(Ref. 2) Nine Mile Point 1 Parameter (Circumferential Weld) (Axial Weld)
Description Safety Safety Using BWRVIP Methodology Evaluation Evaluation Using BWRVIP "VIP" "CEOG" Methodology 28 EFPY** 32 EFPY* 46 EFPY** 32 EFPY 32 EFPY 28 EFPY** 46 EFPY**
Fluence, n/crrn 1.1-6 xlT1 2.21 x1018 1T-67TxlT 2 xl0 2 x10T" 1.65 xl0 18 2.49 x10 1T Initial RTNDT °F -50 -50 -50 0 0 -50 -50 Chemistry Factor 99.9 112 99.9 151.7 172.2 97.6 97.6 Cu% 0.214 0.22 0.214 0.13 0.183 0.214 0.214 Ni% 0.076 0.20 0.076 0.71 0.704 0.046 0.046 A RTNOT'F 44.7 66.5 52.7 86.4 98.1 51.2 60.8 Mean ART F -5.3 16.5 2.7 86.4 98.1 1.2 10.8 From Reference 11. Note that the weld heat number 1248 chemistry used for the 32 EFPY calculations was revised for the 28 EFPY and 46 EFPY calculations based on the resolution of the NMP1 surveillance capsule weld identity as discussed in Reference 14.
- From Reference 7, 14, and 15.
As shown in Table 1, the impact of irradiation results in a lower plant specific mean RTNDT for the NMP1 circumferential weld material as compared to that for any of the NRC's plant-specific analyses which were performed for the Combustion Engineering (CE) fabricated RPVs with the highest adjusted reference temperatures. Comparison of the NMP1 specific data and the data used in the NRC Final Safety Evaluation indicates that the combined effects of the Ni% and Cu% on the Chemistry Factor, which is by itself bounded by the NRC Independent Assessment, and the initial RTNDT. Therefore, the limiting plant-specific conditional probability of failure P(FIE), determined by the NRC, bounds the NMP1 case through the projected end of the original license period.
Thus, for the original license period, the BWRVIP specific results relative to NMP1 as presented in BWRVIP-05 and subsequent RAI responses are consistent with those in the NRC Independent Assessment. Both analyses conclude that the failure probability associated with the circumferential welds is extremely small, and is orders of magnitude less than that for axial welds. Therefore, the NMP1 circumferential weld satisfies, at the end of the original license period, the limiting conditional failure probability for circumferential welds stated in the NRC's July 28, 1998, Safety Evaluation. Note that the discussion above is applicable for the original license period.
For the license renewal period, an NMP1 specific probabilistic fracture mechanics (PFM) evaluation was performed with the VIPER Program (Reference 8) using the data under the column "Using BWRVIP Methodology" in Table 1 for 28 EFPY and 46 EFPY (end of license renewal period). This evaluation was performed using the VIPER probabilistic fracture mechanics program that was developed as part of the BWRVIP-05 (Reference 1) effort. The same low temperature over pressure (LTOP) event parameters (Temperature = 881F, Pressure = 1150 psi) used in the BWRVIP-05 effort-were used in this NMP1 specific calculation. Using the BWRVIP methodology the conditional probability of failure for the NMP1 circumferential weld was found to be less than 1x10.7 for 28 EFPY and 46 EFPY (no failures predicted in 107 trials for both 28 EFPY and 46 EFPY). The BWRVIP frequency of over-pressurization was determined to be 1x10-3/yr. This gives a total probability of failure for NMP1 of less than 2.5x1 0-12/yr for the circumferential welds for 28 EFPY (40 years) and 46 EFPY (60 years) of operation. Note that the failure probabilities are reported to be the same for 28 EFPY and 46 EFPY since there were no failures in 107 Monte Carlo trial simulations. The 46 EFPY includes higher fluence and considers crack growth for 18 EFPY beyond the original license period (28 EFPY).
For NMP1 axial welds with the data shown in Table 1 under the column "Using BWRVIP Methodology," the total probability is <2.5x10l&/yr for both 28 EFPY and 46 EFPY, as no failures were predicted for either the original license period or the license renewal period.
Nine Mile Point Nuclear Station, Unit I Fourth Inservice Inspection Interval 10 CFR 50.55a Request Number I1SI-001A The fact that no failures occurred through the initial license period and license renewal period shows that the reliability of the NMP1 RPV is extremely high. Most importantly, the results show that the increase in failure probability due to the license renewal period is essentially negligible. In addition, the reliability of the circumferential welds is likely much higher than calculated because of lower stress (axial stress is one half the hoop stress) and lower chemistry factors.
These calculations have been performed conservatively assuming a constant fluence at all weld locations equal to the peak fluence. For example, the maximum fluence anywhere in the beltline circumferential weld is assumed to exist throughout the circumferential weld. The peak fluence at any axial weld is assumed to exist at all axial weld locations. In reality, the fluence varies both circumferentially and axially. If analysis were performed considering these fluence variations, the resulting probability of failures would be lower than calculated using the peak fluence at all weld locations.
Thus, the BWRVIP-05 NMP1 specific results as determined using the BWRVIP-05 methodology and subsequent BWRVIP responses to NRC RAIs are consistent with those in the NRC Independent Assessment. Both analyses conclude that the failure probability associated with circumferential welds is extremely small. In addition, due to the NMP1 specific conditions, the failure probability for the axial welds is also extremely small. Most importantly, the increase in failure probability due to operation during the license renewal period is extremely small since no failures were predicted even through the license renewal period. Thus, it is concluded that the NMP1 circumferential weld satisfies, at the end of the license renewal period, the limiting conditional failure probability for circumferential welds in the NRC staffs July 28, 1998 safety evaluation.
Criterion(2) - Limiting the Frequency of Cold Over-pressureEvents Demonstrate licensees have implemented operatortrainingand establishedprocedures that limit the frequency of cold over-pressure events to the amount specified in the NRC staff's July 28, 1998, safety evaluation.
Response
The NRC indicated that the potential for, and consequences of, non-design basis events not discussed in the BWRVIP-05 report should be addressed. In particular, the NRC stated that non-design basis, low temperature over-pressure transients (LTOP) transients, should be considered. The NRC further went on to describe several types of events that could be precursors to an LTOP event. The BWRVIP provided a response to this issue concluding that Condensate and Control Rod Drive (CRD) pumps could cause such a condition leading to an LTOP event. This was summarized in the NRC Safety Evaluation for BWRVIP-05 (Reference 2).
NMP1 has in place procedures which monitor and control reactor pressure, temperature, and water inventory during all aspects of cold shutdown minimizing the likelihood of an RPV LTOP event. Additionally, these procedures are reinforced through the NMP1 reactor operator training program. The procedural controls and training provisions that will be used for the license renewal period of extended operation will be the same as those used for the original operating license period for NMP1 (see Reference 11 and 13).
The RPV leakage pressure test procedures used at NMP1 have sufficient procedural guidance to prevent LTOP events. The leakage test is performed at the conclusion of each refueling outage. These pressure tests are infrequently-performed, complex tasks, and the test procedures are controlled as Special Plant Evolutions. As such, a requirement is included in the procedures for an extensive pre-job briefing to be conducted with all essential personnel including Operations management. The briefing details the anticipated testing evolution with special emphasis on conservative decision making, plant safety awareness, lessons learned from similar in-house or industry operating experiences, the importance of open communications and finally the process in which the test would be aborted if plant systems responded in an adverse manner. Vessel pressure and temperature are required to be monitored throughout the tests to ensure compliance with the plant Technical Specification pressure-temperature curve. Also, the procedures require the designation of a "Principal Test Engineer" for the duration of the test who is a single point of accountability, responsible for the coordination of testing from initiation to closure, and maintaining operations and plant management cognizant of the test status.
Nine Mile Point Nuclear Station, Unit 1 Fourth.Inservice Inspection Interval 10 CFR 50.55a Request Number 11SI-001A With regard to inadvertent system injection resulting in an LTOP condition, the NMP1 high pressure make-up system, ,(i.e., the High Pressure Coolant Injection (HPCI)) and the normal Feedwater system are interconnected.
The HPCI system is a mode of operation of the Condensate and Feedwater systems rather than an independent, stand alone'system. The HPCI system utilizes two condensate pumps, two feedwater booster pumps, two motor-driven feedwater pumps, and an integrated control system. As such, the HPCI system contains only instrumentation and control components as its own dedicated equipment. HPCI initiation is prompted by the Reactor Protection System under the following conditions: (1) a turbine trip, or (2) low reactor water level. During shutdown of the unit, the associated booster and feedwater pumps in the system are secured in accordance with operating procedures. Equipment malfunction or inappropriate operational action would be necessary to cause inadvertent system operation.
During normal cold shutdown conditions, with the RPV head installed, RPV level and pressure are controlled with the CRD System, Condensate Feedwater System, and Reactor Water Cleanup (RWCU) systems using a "feed and bleed" process. The RPV is not taken solid during these times, and plant procedures require opening of the head vent valves after the reactor has been depressurized to approximately 15 psig.
The Liquid Poison System is another high pressure water source to the RPV; however, there are no means of automatic system activation. System injection requires an operator to manually reposition a key-locked control switch to start the system from the Control Room. The system may also be operated from a remote local test station. The only injection path to the RPV is through two explosive actuated injection valves that are interlocked with the key-locked switch in the Control Room. The injection rate for each pump is approximately 30 gpm, which would give the operator sufficient time to control reactor pressure. Local testing of the pumps uses demineralized water from a test tank and is a closed test loop.
Procedural controls are in place to respond to an unexplained rise in reactor pressure which could result from a spurious activation of an injection source. Actions specified include determination and isolation of the pressure source, verification of reactor head vents and/or MSIVs open and, as necessary, relieving reactor pressure using available plant equipment (e.g., electromagnetic relief Valves, reactor water cleanup system and reactor bottom drain).
During normal cold shutdown conditions, reactor water level and temperature are maintained within established ranges in accordance with operating procedures. Procedures governing the conduct of operations require that the Control Room operators frequently monitor for indications and alarms to detect problems and abnormalities as early as possible. An Operations procedure also requireslthat the control room supervisor be notified immediately of any change or abnormality in plant indications and controls. Furthermore, reactor water level and temperature operating bands and changes thereto are established under the direction of the Shift Manager. Therefore, any deviations in reactor water level or temperature from a specified band will be identified and corrected. Finally, plant conditions and on-going activities are discussed during each shift turnover. This ensures that on-coming operators are cognizant of activities that could adversely affect reactor level, pressure, or temperature.
Plant specific procedures have been developed to provide operator guidance regarding compliance with the plant Technical Specifications and RPV pressure-temperature curve-limits. Additionally, operators receive training on RPV brittle fracture and the relationship of these pressure-temperature curve limits.
During plant outages, NMP1 work control processes ensure that the outage schedule and changes to the schedule receive a thorough shutdown risk assessment review to ensure defense-in-depth is maintained. Work is coordinated through the Work Execution Center which provides an additional level of Operations oversight. In the Control Room, the Shift Manager is required, by procedure, to maintain cognizance of any activity that could potentially affect reactor safety during refueling outages. Expected plant responses and contingency actions to address unexpected conditions that may be encountered are required to be evaluated as stated in the administrative controls for risk management and management of outages.
As discussed above, NMP1 has implemented procedural controls and training to minimize the probability of an.
LTOP event. Accordingly, the above information and.the supporting technical documentation contained in the BWRVIP-05 report and NRC Safety Evaluation provide a basis for excluding RPV circumferential welds from the ISI 001A-6 of ISI 001A-9
Nine Mile Point Nuclear Station, Unit I Fourth Inservice Inspection Interval 10 CFR 50.55a Request Number IISI-001A augmented examination requirements of 10 CFR 50.55a(g) and ASME Section XI.
Summary In summary, the NMP1 specific chemistry, and adjusted reference temperature (ART) were compared against the NRC's July 28, 1998, safety evaluation values for 32 EFPY. The NMP1 values were found to be bounded demonstrating that the NRC SE conclusions regarding failure probability have been satisfied. An NMP1 specific probabilistic fracture mechanics evaluation was performed to determine the probability of failure when subjected to an LTOP event during the license renewal period. In additiorn, it was confirmed that NMP1 has taken steps to reduce the potential for LTOP events through procedural controls and personnel training. An evaluation to identify the sources for increased pressure was also performed and found that the probability of a cold overpressure transient is considered to be less than or equal to that used in the NRC evaluation.
In effect the criterion in RG 1.174 regarding defense-in-depth, and safety margins are maintained and USNRC safety goals are not exceeded.
NMPNS has concluded that permanent deferral of the examination of the RPV circumferential shell welds for the license renewal period of extended operation and the reduced examination coverage of the circumferential welds is justified and presents an acceptable level of quality and safety to satisfy the requirements in accordance with 10 CFR 50.55a (a)(3)(i).
E. IMPLEMENTATION SCHEDULE Pursuant to 10 CFR 50.55a(a)(3)(i), NMPNS requests permanent relief for the license renewal period of extended operation. NMPNS has demonstrated that the criteria specified in GL 98-05 (Reference 4) are met for the original operating license period, (NRC SE, dated 4/7/99, Reference 13), and that the criteria of BWRVIP-74-A (Reference 9) are met for the license renewal period of extended operation. Therefore, the requested duration of the proposed alternative is justified.
F. PRECEDENTS
- Nine Mile Point Nuclear Station, Unit 2, NRC letter dated November 5, 2007 (TAC No. MD3696) 0 Dresden Nuclear Power Station, Units 2 and 3, Quad Cities Nuclear Power Station, Units 1 and 2; NRC letter dated March 23, 2005 (TAC Nos. MC2190, MC2191, MC2192, and MC2193)
G. ATTACHMENTS None H. REFERENCES
- 1. Electric Power Research Institute (EPRI) Proprietary Report TR-105697, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendation, BWRVIP -05,"' dated September 1995.
- 2. Letter, G. C. Lainas (NRC) to Carl Terry, BWRVIP Chairman, NRC Report "Final Safety.Evaluation of the BWR Vessel Internals Project BWRVIP-05 Report," (TAC No. MA93925), Division of Engineering Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, dated July 28, 1998.
- 3. Letter, J. R. Strosnider (NRC) to Carl Terry, BWRVIP Chairman, NRC Report "Supplement to Final Safety Evaluation of BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. MA3395), Division of Engineering, Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, dated March 7, 2000.
Nine Mile Point Nuclear Station, Unit I Fourth Inservice Inspection Interval 10 CFR 50.55a Request Number IlSI-001A
- 4. United States Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds," dated November 10, 1998.
- 5. Letter, P. S. Tam (NRC) to P. E. Katz (NMPNS), "Nine Mile Point Nuclear Station, Unit No. 1, Issuance of Amendment Re: Pressure-Temperature Limit Curves" (TAC Nos. MB6687), dated October 27, 2003.
- 6. Letter, P. S. Tam (NRC) to P. E. Katz (NMPNS), "Nine Mile Point Nuclear Station, Unit No. 2, "Issuance of Amendment Re: Pressure-Temperature Limit Curves" (TAC No. MC0331), dated January 27, 2004.
- 7. Structural Integrity Associates Calculation 0800297.300.RA, Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts.
- 8. Viper Computer Code, Version 1.2, Structural Integrity'Associates, January 1998.
- 9. BWR Vessel Internals Project, BWRVIP-74-A, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal, June 2003.
10, Letter, C. I. Grimes (NRC) to C. Terry, BWRVIP Chairman, "Acceptance forReferencing of EPRI Proprietary Report TR-1 13596, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines (BWRVIP-74)" and Appendix A, "Demonstration of Compliance with the Technical Information Requirements of the License Renewal Rule (10 CFR 54.21)," dated October 18, 2001.
- 11. Letter, R. B. Abbott (NMPC) to Document Control Desk (NRC), Nine Mile Point Unit 1, Proposed Alternatives for Examination of Reactor Pressure Vessel Shell Welds, dated December 10, 1998.
- 12. Structural Integrity Associates Report No. SIR-06-512, "Technical Justification for Elimination of Nine Mile Point Unit 1 Reactor Pressure Vessel Circumferential Weld Inspections for the License Renewal Term,"
dated August 21, 2008, Rev. 1.
- 13. Letter, S. S. Bajwa (NRC) to J. H. Mueller (NMPC), Alternatives for Examination of Reactor Pressure Vessel Shell Welds, Nine Mile Point Nuclear Station, Unit 1 (TAC No. MA4383), dated April 7, 1999.
- 14. Letter, R. B. Abbott (NMPC) to Document Control Desk (NRC), Request for Additional Information Regarding Reactor Pressure Vessel Structural Integrity at Nine Mile Point Nuclear Station Unit 1 (TAC No.
MA1 200), dated September 4, 1998.
- 15. Report Number NMP-405778, Neutron Transport Analysis for Nine Mile Point Unit 1, MPM Technologies, May 2006.
- 16. NRC Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Rev. 2, May 1988.
- 17. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.
Nine Mile'Point Nuclear Station, Unit I Fourth Inservice Inspection Interval 10 CFR 50.55a Request Number IISI-001A REACTOR VESSEL SHELL WELD LOCATIONS Figure 1 Ua 2
I b h 1 b*.J A
'1 I~. -I L41 C-,
0*
z CJ z
Z: z CL
-j
.z (3
La S. zLa (3
z La (3