ML063320461

Request for Permanent Relief from Inservice Inspection Requirements of 10 CFR 50.55a(g) for the Volumetric Examination of Circumferential Reactor Pressure Vessel Welds

ML063320461
Person / Time
Site:	Nine Mile Point
Issue date:	11/16/2006
From:	Harland G Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML063320461 (13)

Text

0 Constellation Energy Nine Mile Point Nuclear Station November 16, 2006 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION: Document Control Desk

SUBJECT:

Nine Mile Point Nuclear Station Unit No. 2; Docket No. 50-410 Request for Permanent Relief from Inservice Inspection Requirements of 10 CFR 50.55a(g) for the Volumetric Examination of Circumferential Reactor Pressure Vessel Welds

REFERENCES:

(a) EPRI Report TR-105697, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), dated September 1995 (b) Letter from NRC to the BWRVIP, Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. M93925) dated July 28, 1998 (c) Letter from NRC to the BWRVIP, Supplement to Final Safety Evaluation of the BWR Vessel and. Internals Project BWRVIP-05 Report (TAC No. MA3395) dated March 7, 2000 (d) NRC Generic Letter (GL) 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 Pursuant to 10 CFR 50.55a(a)(3), Nine Mile Point Nuclear Station, LLC (NMPNS) requests NRC approval of proposed Request for Alternative Number 21SI-001 for use at Nine Mile Point Unit 2 (NMP2). The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety. Request Number 21SI-001 (Attachment 1) requests permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of reactor pressure vessel (RPV) shell circumferential welds (i.e., ASME Section XI, Table IWB-2500-1, Examination Category B-A, Item BI.l1, Circumferential Shell Welds). This request is for the remaining term of the NMP2 Renewed Operating License.

,4c47

Document Control Desk November 16, 2006 Page 2 The technical basis providing justification for relief from the examination requirements of RPV shell circumferential welds is contained in a report submitted from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) to the NRC (Reference a). The NRC evaluated this report and responses to Requests for Additional Information, and issued safety evaluations to the BWRVIP (References b and c).

Additionally, NRC Generic Letter 98-05 (Reference d), provides guidance to BWR licensees regarding requests for permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV shell circumferential welds (i.e., ASME Section XI, Table IWB-2500-1, Examination Category B-A, Item B1.11, Circumferential Shell Welds).

NMPNS requests approval of this proposed request for alternative by November 30, 2007 to support preparations for the next NMP2 refueling outage, which is currently scheduled for the spring of 2008.

Should you have questions regarding the information in this submittal, please contact M. H. Miller, Licensing Director, at (315) 349-1510.

Very truly yours, Gary Harland Acting Manager Engineering Services GH/DEV/kms

Attachment:

(1) Nine Mile Point Nuclear Station, Unit 2 - 10 CFR 50.55a Request Number 21SI-001 cc: S. J. Collins, NRC T. G. Colbum, NRC Resident Inspector, NRC

ATTACHMENT (1)

NINE MILE POINT NUCLEAR STATION, UNIT 2 10 CFR 50.55a REQUEST NUMBER 21SI-001 Nine Mile Point Nuclear Station, LLC November 16, 2006

Nine Mile Point Nuclear Station, Unit 2 10 CFR 50.55a Request Number 21SI-001 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

1. ASME Code Components Affected System: Reactor Coolant System Code Class: ASME Code Class 1 Component

Description:

Volumetric Examination of all Pressure Retaining Reactor Pressure Vessel Shell Circumferential Welds Components Affected:

2RPV-AA Bottom Head Radial Plate to Shell 1 B-A B1.11 2RPV-AB Shell ito Shell 2 B-A Bl l 2RPV-AC Shell 2 to Shell 3 B-A B1.11 2RPV-AD Shell 3 to Shell 4 B-A B1.11

2. Applicable Code Edition and Addenda The applicable ASME Code, Section Xl, for the Nine Mile Point Nuclear Station (NMPNS), Unit 2, Second 10-Year Interval, In-Service Inspection Program is the 1989 Edition with no addenda.

3. Applicable Code Requirement In accordance with the provisions of 10 CFR 50.55a, "Codes and Standards," paragraph 10 CFR 50.55a(a)(3), Constellation Energy Group (CEG) requests permanent relief for the remaining license period and the license renewal period of extended operation for Nine Mile Point Nuclear Station (NMPNS), Unit 2, from the requirement of ASME Code Section Xl, Subarticle 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.

Subarticle IWB-2500 requires components specified in Table IWB-2500-1 to be examined. Table IWB-2500-1 requires volumetric examination of all RPV shell circumferential welds each inspection interval (i.e., Examination Category B-A, Item No. B1.11).

4. Reason for Request 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 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 as documented in the final safety evaluation of the BWRVIP-05 report (TAC No. M93925), (Reference 2). This assessment concluded that the File: 21SI-001 ISI 001-1 of ISI 001-10

Nine Mile Point Nuclear Station, Unit 2 10 CFR 50.55a Request Number 21SI-001 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).

5. Proposed Alternative and Basis for Use 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), Nine Mile Point Nuclear Station will implement the following alternate provisions for the subject weld examinations.

The failure frequency for ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, "Reactor Pressure Vessel Shell Circumferential Welds," is sufficiently low to justify their elimination from the inservice 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 Xl, Table IWB-2500-1, Examination Category B-A, Item No. BI.12, "Reactor Pressure Vessel Shell Longitudinal Welds," shall be performed, to the extent possible, and shall include inspection of the Reactor Pressure Vessel 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 OD surface or where access is practical from the internal ID surface of the Reactor Pressure Vessel to the maximum extent possible. The examination of the remaining accessible portions of the Reactor Pressure Vessel circumferential shell welds will be permanently deferred for the life of the current license and the license renewal extended-period of 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 1995 Edition through 1996 Addenda of ASME Section XI, 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 (BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations"), (Reference 1), that was transmitted to the USNRC 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 (TAC No.

M93925), (Reference 2) and 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 USNRC also issued Generic Letter 98-05, (Reference 4), permitting BWR licensees to request permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV shell circumferential welds, ASME Section Xl, Table IWB-2500-1, File: 21SI-001 FSI 001-2 of ISI 001-10

Nine Mile Point Nuclear Station, Unit 2 10 CFR 50.55a Request Number 21SI-001 Examination Category B-A, Item B1.11. The USNRC stated in the generic letter that Boiling Water Reactor (BWR) licensees may request permanent relief 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 staffs July 28, 1998, safety evaluation, 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.

Generic Letter 98-05, Criterion (1)

Demonstrate that at the expiration of the license, 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 30, 1998 Safety Evaluation.

Response

In order to demonstrate that the circumferential welds satisfy the July 28, 1998 NRC safety evaluation limiting condition failure probabilities, a comparison of the chemistry values and the predicted fluence at the end of the original license period was performed. NMP2 current license period is equivalent to 36 EFPY and was compared against the NRC calculation for 32 EFPY.

This comparison is conservative since fluence and crack growth as a result of four additional EFPY was added to the NMP2 32 EFPY to compare against the NRC 32 EFPY values. Failure probabilities were also calculated for NMP2 at 36 EFPY. For the license renewal extended period of operation, it was more appropriate to compare the change in failure probabilities since the NRC analysis did not consider the effect of the license renewal extended period of operation (added fluence and crack growth).

For plants with Reactor Pressure Vessels fabricated by CBI Nuclear Company (CBIN), the peak end-of-license neutron fluence for circumferential welds used in the NRC PFM analysis was 5.lxl018 n/cm2 . At NMP2, the highest fluence anticipated at the end of the current license period (36 EFPY) is 5.8x1017 n/cm2 and at the end of the license renewal extended period of operation is 8.6x1017 n/cm 2 (Reference 7). Thus, embrittlement due to fluence effects is much lower at the end of the current license period, and the NRC analysis even at the end of 32 EFPY is conservative for NMP2 in this regard. Therefore, there is conservatism in the already low circumferential weld failure probabilities as related to NMP2.

Table 1 illustrates that NMP2 has additional conservatism in comparison to the NRC's Final Evaluation of BWRVIP-05 Limiting Plant Specific Analysis and Independent Assessment Fracture Analysis limiting case for the current license term. The chemistry factor, ARTNDT, mean ART, and ART are calculated consistent with the guidelines of Regulatory Guide 1.99, Rev. 2. The data used for the evaluation based on the BWRVIP-05 methodology are also shown in Table 1.

File: 21SI-001 ISI 001-3 of ISI 001-10

Nine Mile Point Nuclear Station, Unit 2 10 CFR 50.55a Request Number 21SI-001 Table 1: Comparison of Input Parameters for NRC Staff Assessment and BWRVIP Methodology Parameter Nine Mile Point Unit 2 NRC Staff Nine Mile Point Unit 2 Description (Circumferential Weld) Assessment for 32 (Bounding Axial Weld)

EFPY (Circumferential Welds)

Using BWRVIP Safety Evaluation* Using BWRVIP Methodology Methodology "VIP" 36 EFPY 54EFPY 32EFPY 36 EFPY 54 EFPY Fluence, n/cm7 5.8x10 8.6x10" 5.1x10"' 6.06x10" Initial RTNDT' -50 -50 -65 -40 -40 0F Chemistry Factor 54 54 109.5 95 95 Cu % 0.04* 0.04* 0.1 0.07* 0.07*

Ni % 0.82* 0.82* 0.99 0.89* 0.89*

A RTNDT (TF) Monte Carlo Predicted 109.5 Monte Carlo Predicted Mean ART (°F) 4.89 1 9.7 44.5 37.16 1 46.94

• Note: Cu and Ni values are maximums from different heat/lot numbers for the beltline welds and are used together to create a bounding result.

The methodology used for the RPV neutron fluence calculation is in accordance with the recommendations of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," and has been approved by the NRC in letters to NMPNS dated October 27, 2003 (Reference 5) and January 27, 2004 (Reference 6).

As shown in Table 1, the impact of irradiation results in a lower mean ART for NMP2 (4.890 F) as compared to the NRC Final Safety Evaluation (SE) for CBIN plants (44.5 OF) at the end of the original license period. Comparison of the NMP2 specific data and the data used in the NRC Final Safety Evaluation indicates that the difference is the fluence at the end of 36 EFPY and the chemistry factor. Note that for the circumferential welds, even at the 54 EFPY (end of license renewal extended period of operation), the mean ART (9.7 OF) is also bounded by the NRC SE mean ART (44.5 OF).

Based on the data presented in Table 1 (under "NRC Staff Assessment"), the NRC performed probabilistic fracture mechanics calculations using the FAVOR code. Results of the NRC evaluation showed that the conditional probability of failure was 2.0x107 for 32 EFPY for CBIN fabricated vessels. The conditional probability of failure was 1.0x10-6 for 32 EFPY using the BWRVIP-05 methodology and the limiting plant. The NRC evaluation used a frequency of an over pressure event occurring of lx1 0 3/yr. This results in a total probability of failure of 5x1 0-1 /yr. As presented in the final safety evaluation, NUREG 1560, Vol. 1, core damage frequencies (CDF) for BWR plants were reported to be approximately 10-7/yr to 10- 4/yr. Since the failure frequency for the CBIN fabricated plants (i.e., elimination of circumferential weld examinations) contributes less than the amount of change of large early release frequency (LERF) and CDF, the failure frequency for RPV circumferential welds is sufficiently low to justify elimination ofinservice inspection.

The NMP2 specific PFM evaluation was performed with the VIPER Program (Reference 8) using the data under the column "Using BWRVIP Methodology" in Table 1 for 36 EFPY and 54 EFPY (end of license renewal extended period of operation) with fluence adjusted accordingly (See Table 2). This evaluation was performed using the VIPER probabilistic fracture mechanics program developed as part of the BWRVIP-05 (Reference 1) effort. The same LTOP event File: 21SI-001 FSI 001-4 of SI 001-10

Nine Mile Point Nuclear Station, Unit 2 10 CFR 50.55a Request Number 21S-01 parameters (Temperature = 88°F, Pressure = 1150psi) used in the BWRVIP-05 effort were used in this NMP2 specific calculation. Using the BWRVIP methodology the conditional probability of failure for the NMP2 circumferential weld was found to be less than lx10 7 for 36 EFPY and 54 EFPY. The BWRVIP frequency of over-pressurization was determined to be lx 0-3/yr. This gives a total probability of failure for NMP2 of less than 2.5xl 0-12/yr for the circumferential welds for 36 EFPY (40 years) and 54 EFPY (60 years) of operation. The 54 EFPY includes higher fluence and considers crack growth for 18 EFPY beyond the original license term (36 EFPY).

For the NMP2 axial welds with the data shown in Table 1 under the column "Using BWRVIP Methodology," the total probability is <2.5x1 0 1 1/yr and 1.33 xl 0-10/yr, for 36 EFPY and 54 EFPY, respectively (See Table 2). Comparison of the NMP2 circumferential weld failure probability

(<2.5xl 0 12/yr for 54 EFPY) and the axial weld failure probability (1.33x1 0--°/yr for 54 EFPY) demonstrates that the circumferential weld failure probability is significantly less than that for the axial welds at 54 EFPY. Both the total probability of failure for the circumferential welds and axial welds is very low for 36 EFPY.

For 54 EFPY, the circumferential weld reliability is significantly higher than the axial welds. When compared with the results of those plants analyzed in BWRVIP-05, (total probability of failures for axial welds were typically from 2.5x10-10 to 2.5x10-' ) using conservative inputs, the probability of failure for NMP2 falls closer to the lower probability end of the BWRVIP-05 probability range.

Thus, the BWRVIP-05 NMP2 specific results as determined using the BWRVIP-05 methodology and subsequent BWRVIP responses to USNRC RAIs, are consistent with those in the NRC Independent Assessment. Both analyses conclude that the failure probability associated with circumferential welds is extremely small, and that it is orders of magnitude less than that for axial welds through the license renewal extended period of operation. It is concluded that the NMP2 circumferential weld satisfies, at the expiration of their original license and at the end of the license renewal extended period of operation, the limiting conditional failure probability for circumferential welds in the NRC staffs July 28, 1998 safety evaluation.

Table 2: Total Probability of Failure, BWRVIP Methodology 36 EFPY 54 EFPY Circumferential Welds <2.5xl 0-1 /yr <2.5xl 0-1Q/yr Axial Welds <2.5xl00-/yr 1.33x1 0-'u/yr Generic Letter 98-05, Criterion(2)

Demonstratelicensees 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 staff indicated that the potential for, and consequences of, non-design basis events not addressed in the BWRVIP-05 report should be considered. In particular, the NRC staff stated that non-design basis, cold, over-pressure transients should be considered. It is highly unlikely that a BWR would experience a cold, over-pressure transient. The NRC staff described several types of events that could be precursors to BWR RPV cold, over-pressure transients. These were identified as precursors because no cold, over-pressure event has occurred at a U.S. BWR. Also, the NRC staff identified one actual cold, over-pressure event that occurred during shutdown at a non-U. S. BWR. This event apparently included several operational errors that resulted in a maximum RPV pressure of 1150 psi with a temperature of 88'F. The BWRVIP responded with the conclusion that condensate and control rod drive (CRD) pumps could cause conditions that could lead to cold over-pressure events. This is summarized in the Final Safety Evaluation of BWRVIP-05 (TAC No. M93925), (Reference 2).

File: 21SI-001 ISI 001-5 of SI 001-10

Nine Mile Point Nuclear Station, Unit 2 10 CFR 50.55a Request Number 21SI-001 High pressure core spray injection has been used after the reactor has been shutdown during RPV cooldown with RPV temperature well above that required for leakage testing during the last two refueling outages to provide ALARA flushing of the injection piping. This procedure has multiple escalating contingencies built into the procedure to stop injection and includes procedures to prevent instrumentation problems from causing over-injection. Operator errors would need to occur before the vessel experiences high pressure. Thus, operator training would make this an unlikely source for over pressurization.

The Reactor Core Isolation Cooling (RCIC) system is steam turbine driven. During reactor cold shutdown conditions, no steam is available for operation of the system. Therefore, it is not plausible for the system to contribute to an over pressurization event while the unit is in cold shutdown.

During reactor cold shutdown conditions, the feedwater pumps are shutdown. It would require direct Operator action to start a feedwater pump and inject into the vessel. As discussed below, operating procedural restrictions, operator training and work control processes at NMP2 provide appropriate controls to minimize the potential for RPV cold over-pressurization events.

During normal cold shutdown conditions, RPV level and pressure are controlled with the Control Rod Drive (CRD) 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 cooled to less than 212°F. If either of these systems were to fail, the Operator would adjust the other system to control level. Under these conditions, the CRD system typically injects water into the reactor at rate of approximately 63 gpm. This slow injection rate allows the operator sufficient time to react to unanticipated level changes and, thus, significantly reduces the possibility of an event that would result in a violation of the pressure/temperature limits.

The Standby Liquid Control (SLC) system is another high-pressure water source to the RPV.

However, there are no automatic starts associated with the system that can occur with the reactor shutdown. SLC injection requires an Operator to manually start the system from the Control Room or from the local test station. Additionally, the injection rate of the SLC pump is approximately 42 gpm, which would give the Operator ample time to control reactor pressure in the case of an inadvertent injection.

Pressure testing of the RPV is classified as an "Infrequently Performed Test or Evolution" which ensures that these tests receive special management oversight and procedural controls to maintain the plant's level of safety within acceptable limits. The pressure test is conducted so that the required temperature bands for the pressure increases are achieved and maintained prior to increasing pressure. During performance of an RPV pressure test, level and pressure are controlled using the CRD and RWCU systems using a "feed and bleed" process. Increase in pressure is limited to less than 50 psig per minute. Reactor coolant pump starts are also prohibited with the reactor vessel in a solid-water condition. These practices minimize the likelihood of exceeding the pressure-temperature limits during performance of the test.

NMP2 has taken steps to reduce the potential for LTOP events through procedural controls and personnel training.

Operating procedural restrictions, operator training and work control processes at NMP2 provide appropriate controls to minimize the potential for RPV cold over pressurization events. During normal cold shutdown conditions, reactor water level, pressure, and temperature are maintained within established bands in accordance with operating procedures. The Operations procedure governing Control Room activities requires that Control Room Operators frequently monitor for indications and alarms to detect abnormalities as early as possible. This procedure also requires that the Shift Manager be notified immediately of any changes or abnormalities in indications.

Furthermore, changes that could affect reactor level, pressure, or temperature can only be File: 21SI-001 ISI 001-6 of SI 001-10

Nine Mile Point Nuclear Station, Unit 2 10 CFR 50.55a Request Number 21SI-001 performed under the knowledge and direction of the Shift Manager or Control Room Supervisor.

Therefore, any deviations in reactor water level or temperature from a specified band will be promptly identified and corrected. Finally, plant conditions and on-going activities that could affect critical plant parameters are discussed at each shift turnover. This ensures that on-coming Operators are cognizant of activities that could adversely affect reactor level, pressure, or temperature.

Procedural controls for reactor temperature, level, and pressure are an integral part of Operator training. Specifically, Operators are trained in methods of controlling water level within specified limits, as well as responding to abnormal water level conditions outside the established limits.

Additionally, Control Room Operators receive training on brittle fracture limits and compliance with the Technical Specification pressure-temperature limits curves. Plant-specific procedures have been developed to provide guidance to the Operators regarding compliance with the Technical Specification requirements on pressure-temperature limits.

During plant outages, the 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 per procedures. At NMP2, the outage scheduler schedules outage work items.

Senior Reactor Operators (SRO) assigned to the Work Control Center provide oversight of outage schedule development to avoid conditions that could adversely impact reactor water level, pressure, or temperature. From the outage schedule, a daily schedule is developed listing the work activities to be performed. These daily schedules are reviewed and approved by SROs and a copy is maintained in the Control Room. Changes to the schedule require SRO review and approval.

During outages, work is coordinated through the Work Control 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 possibly affect reactor level decay heat removal during refueling outages. The Control Room Operator is required to provide positive control of reactor water level and pressure within the specified band, including restoration actions being taken. Pre-job briefings are conducted for complex work activities, such as RPV pressure tests that have the potential of affecting critical RPV parameters. Pre-job briefings are attended by the cognizant individuals involved in the work activity. Expected plant responses and contingency actions to address unexpected conditions, or responses that may be encountered, are included in the briefing discussion.

Operator training curriculum (lesson plan/simulator scenario) cover basic theory and application of brittle fracture, vessel thermal stress, operational transient procedures including high water level, technical specifications and heat up and cool down (Technical Specification pressure/temperature curve adherence). During simulator scenarios, the crews demonstrate skills, knowledge and abilities with regard to responding to potential low temperature/high pressure events. Additionally, the training is used to enforce management's expectation for strict procedural compliance and conservative decision making.

Based on the above discussion, the frequency of cold over-pressure events is limited to the amount specified in the NRC staffs July 28, 1998, safety evaluation.

Summary In summary, the NMP2 specific chemistry, fluence and ART were compared against the NRC SE values and found to be bounded demonstrating that the NRC SE conclusions regarding failure probability have been satisfied. In addition, an NMP2 specific probabilistic fracture mechanics evaluation was performed to determine the probability of failure when subjected to an LTOP event during the original licensed term and the license renewal extended period of operation and has confirmed that NMP2 has taken steps to reduce the potential for LTOP events through procedural controls and personnel training. In addition, an evaluation to identify the sources for File: 21SI-001 ISI 001-7 of ISI 001-10

Nine Mile Point Nuclear Station, Unit 2 10 CFR 50.55a Request Number 21SI-001 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 life of the current operating license through the license renewal extended period of 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).

6. Duration of Proposed Alternative Pursuant to 10 CFR 50.55a(a)(3)(i), Constellation Energy requests permanent relief for the current license period and the license renewal extended period of operation for NMPNS, Unit 2.

Although Reference 4 permits BWR licensees to request permanent relief for the remaining term of the existing initial operating license, NMPNS has demonstrated that the criteria specified in Reference 4 will continue to be met for the entire additional extended period of operation.

Therefore, the requested duration of the proposed alternative is justified.

Precedents The NRC has previously approved a number of similar requests, including the following:

• LaSalle Station Units 1 and 2, NRC letter dated January 28, 2004 (TAC Nos. MB9755 and MB9756)

• Duane Arnold Energy Center, NRC letter dated January 6, 2005 (TAC No. MC2181)

• Columbia Generating Station, NRC letter dated June 1, 2005 (TAC No. MC3916)

" 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)

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, Gus C. Lainas (NRC) to Carl Terry, BWRVIP Chairman, "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, Jack R. Strosnider (NRC) to Carl Terry, BWRVIP Chairman, "Supplement to Final Safety Evaluation of BWR Vessel and Internals Project BWRVIP-05 Report (TAC No.

MA3395), dated March 7, 2000.

4. United States Nuclear Regulatory Commission, Office of Nuclear Reactor Regulations, 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. NRC Letter to NMPNS, "Nine Mile Point Nuclear Station, Unit No. 1, Issuance of Amendment Re: Pressure-Temperature Limit Curves", (TAC Nos. MB6687), dated October 27, 2003.

File: 21SI-001 ISI 001-8 of ISI 001-10

Nine Mile Point Nuclear Station, Unit 2 10 CFR 50.55a Request Number 21SI-001

6. NRC Letter to 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 Report No.: SIR-06-394, "Technical Justification for Elimination of Nine Mile Point Unit 2 Reactor Pressure Vessel Circumferential Weld Inspections," dated October 2006.

8. Viper Computer Code, Version 1.2, Structural Integrity Associates, January 1998.

File: 21SI-001 ISI 001-9 of ISI 001-10

Reactor Pressure Vessel 0 Weld Locations 270 180 I I.

SHELL RINGS AE 32'~I I CD Z 32Z..

BM "1M ELV Shell Ring#4 350 DEG N14D DU F 288. I 252 N14C J*

i N14B

@ BN DEG DEG D @170 (nZ11 Ring#3 330 DEG mShell

• " ADV _ 210_______ DE D5012 309 190 39 BJ DEG l Z '30T' ELV JB J ELV 270 cS 210 150 120 CI DEG 240 DEG DEG DEG AC31* 50 DEG 3"

35 Vertical1 288' N12D " i Weld BE . 12Cc N1B ' D, ELV ýD Shell Ring#2 SKIRT DA,DB,DC,DDDE,DF)