ML080250410

American Society of Mechanical Engineers (ASME) Code,Section XI, Inservice Inspection Program for the Third Ten-Year Inservice Inspection Interval and Associated 10 CFR 50.55a Requests - Replacement of Appendix H, Relief Requests / Alterna

ML080250410
Person / Time
Site:	Nine Mile Point
Issue date:	01/25/2008
From:	Laughlin G Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML080250410 (40)

Text

P.O. Box 63 Lycoming , NY 13093 January 25, 2008 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 American Society of Mechanical Engineers (ASME) Code,Section XI, Inservice Inspection Program for the Third Ten-Year Inservice Inspection Interval and Associated 10 CFR 50.55a Requests - Replacement of Appendix H, "Relief Requests / Alternatives"

REFERENCE:

(a) Letter from G. 1. Laughlin (NMPNS) to Document Control Desk (NRC) , dated December 14, 2007, American Society of Mechanical Engineers (ASME) Code,Section XI, Inservice Inspection Program for the Third Ten-Year Inservice Inspection Interval and Associated 10 CFR 50.55a Requests By letter dated December 14, 2007 (Reference a), Nine Mile Point Nuclear Station, LLC (NMPNS) submitted the Nine Mile Point Unit 2 (NMP2) Third Ten-Year Inservice Inspection (lSI) Plan and Schedule and requested NRC approval of the associated 10 CFR 50.55a requests pursuant to 10 CFR 50.55a(a)(3) or 10 CFR 50.55a(g)(6)(i), as applicable. The details of the nine (9) 10 CFR 50.55a requests associated with the third ten-year lSI program are contained in Appendix H of the lSI Plan and Schedule that was enclosed with Reference (a).

It has come to the attention ofNMPNS that there was a 10 CFR 50.55a request, identified as 2ISI-OIO, that was improperly included in Appendix H, and that another request, 21SI-00lA, was missing from Appendix H. Therefore, to rectify these errors, NMPNS hereby submits a replacement copy of Appendix H to the NMP2 Third Ten-Year lSI Plan and Schedule (see Enclosure 1) and requests that the NRC staff use this replacement copy for their review of the 10 CFR 50.55a requests associated with the third ten-year lSI program.

This letter does not contain any regulatory commitments.

Document Control Desk January 25, 2008 Page 2 Should you have any questions regarding the information in this submittal, please contact T. F. Syrell, Licensing Director, at (315) 349-5219.

G Jay Laughlin Manager Engineering Services GJL/DEV

Enclosure:

1. Replacement Copy of Appendix H to the Nine Mile Point Nuclear Station Unit 2 Third Ten-Year Inservice Inspection Plan and Schedule (CNG-NMP2-ISI-003, Revision 00) cc: S. J. Collins, NRC M. J. David, NRC Resident Inspector, NRC

ENCLOSUREl REPLACEMENT COPY OF APPENDIX H TO THE NINE MILE POINT NUCLEAR STATION UNIT 2 THIRD TEN-YEAR INSERVICE INSPECTION PLAN AND SCHEDULE (CNG-NMP2-ISI-003, Revision 00)

Nine Mile Point Nuclear Station, LLC January 25, 2008

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NINE MILE POINT NUCLEAR STATION, UNIT 2 CNG-NMP2-ISI*OO3 THIRD INSERVICE INSPECTION INTERVAL

. :SI " Rev. 00 INSERVICE INSPECTION Date: October 31, 2007 PLAN AND SCHEDULE Page H-1 of H-3 APPENDIX H - RELIEF REQUESTS I ALTERNATIVES H.D RELIEF REQUESTS I ALTERNATIVES During the First and Second Ten Year In-service Inspection Intervals, there were cases where component configuration and/or interference prevented the code required volume or surface area from being examined in it's entirety. In each case where such limitations were encountered, the details were documented on a Request for Relief or Alternate and submitted to the Nuclear Regulatory Commission as required by Title 10, Part 50, Section 55a of the Code of Federal Regulations for review and approval.

H.1 NMPNS has determined based on a detailed review that previous granted Requests for Relief or Alternates that certain limitations still exist and therefore , will require re-approval for the Third In-service Inspection Interval. Those requests for relief or Alternates on components which remain applicable for the Third In-service Inspection Interval have been provided within this Appendix.

Appendix H includes a listing that provides the identification and current status of each Request for Relief or Alternate submitted to the USNRC, and form an integral part of the current Inspection Plan and Schedule .

Note: Examination volume or surface area that cannot be examined due to interference by another component or part geometry, a reduction in examinatio.n coverage on any weld will be considered acceptable provided the reduction in coverage for that weld is less than 10%.

SUbjectof ASME Code Case N-460 Examination volume or surface area interferencethat does not meet the coverage requirements of Code Case N-460 , will be documented in the form of a Request for Relief per 10 CFR 50.55a (g)(4)(iv).

In cases where parts of the required examination areas cannot by effectively examined because of a combination of component design or current inspection technique limitations, NMPNS will continueto evaluate the development of new or improved examination techniques with the intent of applying these techniques where a practical improvement on the examination can be achieved .

File : APPENDIXHRO.doc

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NINE MILE POINT NUCLEAR STATION, UNIT 2 CNG-NMP2-ISI-003 THIRD INSERVICE INSPECTION INTERVAL Rev. 00

"-.J INSERVICE INSPECTION Date: October 31, 2007 PLAN AND SCHEDULE Page H-2 of H-3

• C LASS *t , 2 AND 3 RELlEf REQUEST SUM MARY STATUS

=

21SI-001A Reactor B-A , B1 .11 The permanent elimination of RPV Perform 100% of Granted TAC Formally Pressure circumferential welds longitudinal welds and 2 to MD3696 (2ISI-001) Vessel 3% of circumferential welds 11/05/07 that intersect longitudinal Note 5 welds 21SI-002 Formally Reactor B-A , B1.22 , Relief is requested from performing Perform exams to the (RR-IWB-7) Pressure B1.21 100% ofCRV maximum extent practical Vessel 2181-003 B-A, B1.30 Formally Reactor Perform exams to the Pressure Request Relief from performing 100% maximum extent practical (RR-IWB-3) ofCRV Vessel 21SI-004 B-O , B3 .90 Formally Reactor Relief is requested from performing Examine to the maximum (RR-IWB-2) Pressure 100% CRV extent practical Vessel 21SI-005 Formally Reactor B-O, B14.10 NMPNS proposes to Pressure Relief is requested from 100% perform additional (RR-IWB-1) examination CRO Housings Vessel examination as defined in the Relief Request 21SI-006 Reactor B-G-1, B6.40 .. .Reiief is requested from performing Perform exams to the Formally Pressure 100% CRV maximum extent practical (RR-IWB-13) Vessel 21SI-007 Various Formally B-F, B5.10, Relief is requested from ASME Section Implement Altemate Risk-(RR-RI -ISI-2) 85 .20 , 85.30, XI Class 1 and 2 piping examination Informed Inservice 85 .100, requirements .. Inspection Program

~~:l~g, 8-J ,89.11 ,

B9.21 , 89.31, 89 .32 , 89.40 C-F -1 , C5 .11, C5 .21, C5 .30, C5.40 C-F-2 , C5.51, C5 .61 , C5 .70, C5 .81 IGSCC "A" 21SI-008 Low Pressure CoG . Relief is requested from performing Perform exams to the Formally Core Spray 100% of the Code Required Surface maximum extent practical C6.20 (RR-IWC-5) and Residual Heat Removal 2151-009 High Pressure CoG Relief is requested from performing Perform exams to the Formally Core Spray, 100% of the Code Required Surface maximum extent practical C6.10 (RR-IWC-1) Low Pressure Core Spray, C-C Residual Heat C3 .10 Removal, Reactor Core Isolation Cooling File : APPENDIXHRO.doc

NINE MILE POINT NUCLEAR STATION, UNIT 2 CNG-NMP2-ISI-003

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THIRD INSERVICE INSPECTION INTERVAL lS I * , ' Rev. 00

~/ INSERVICE INSPECTION Date: October 31,2007 PLAN AND SCHEDULE Page H-3 of H-3

• • CLASS 1, 2 AND 3 <RELIEF REQUEST SUMI\IIARY STATUS Standard Legend Notes Note 1 Submitted to USNRC, awaiting USNRC Safety Evaluation Note 2 W ithdrawn from USNRC Consideration Note 3 Resubmitted following Request for Information, awaiting USNRC Evaluation Note 4 General- Unless the status identifies the Request for Relief as granted, implementation is prohib ited Note 5 Request 2ISI-001A has been updated to identify new ASME Section XI Edition/Addenda for inspection interval only.

File: APPENDIXHRO.doc

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

A. COMPONENT IDENTIFICATION:

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 8ottom Head Radial Plate to Shell 1 8-A 81 .11 2RPV-A8 Shell 1 to Shell 2 8-A 81.11 2RPV-AC Shell 2 to Shell 3 8-A 81.11 2RPV-AD Shell 3 to Shell 4 8-A 81 .11 B. APPLICABLE CODE REQUIREMENTS:

The applicable ASME Code ,Section XI, for the Nine Mile Point Nuclear Station (NMPNS), Unit 2, Third 10-Year Interval , In-Service Inspection Program is the 2001 Edition through the 2003 Addenda.

In accordance with the provisions of 10 CFR 50.55a , "Codes and Standards," paragraph 10 CFR 50.55a(a)(3), Nine Mile Point Nuclear Station , LLC (NMPNS), Unit 2 requests permanent relief for the remaining license period and the license renewal period of extended operation, from the requirement of ASME Code Section XI, Subarticle IW8-2500, Table IW8-2500-1 , Volumetric Examination of Examination Category B-A, "Pressure Retaining Welds in Reactor Vessel",

Examination Item Number 81 .11, "Circumferential Shell Welds ." See Figure 1 for weld locations.

Subarticle IW8-2500 requires components specified in Table IW8-2500-1 to be examined. Table IW8-2500-1 requires volumetric examination of all RPV shell circumferential welds each inspection interval (i.e., Examination Category 8-A, Item No. 81 .11).

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 8WRVIP-05, "8WR 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 8WR RPV circumferential welds are orders of magnitude lower than that of the longitudinal welds . The NRC staff conducted an independent risk-informed , probabilistic fracture mechan ics assessment (PFMA) of the analysis contained in BWRVIP-05 (Reference 1), and the results are documented in the final safety evaluation of the 8WRVIP-05 report (TAC No. M93925) , (Reference 2). This assessment concluded that the probability of failure of the 8WR RPV circumferential welds is orders of magnitude lower than that of the axial shell welds and the added risk caused by not lSI 001A-1 of IS1001A-10

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

D. BASIS FOR RELIEF AND ALTERNATIVE EXAMINATIONS:

Proposed Alternative In accordance with 10 CFR 50.55a(a)(3)(i), and cons istent with information contained in NRC Generic Letter 98-05, (Reference 4) and in the NRC safety evaluation for BWRVIP-74-A (Reference 10) Nine Mile Point Nuclear Station will implement the following alternate provis ions 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 in-service inspection (lSI) requirement of 10 CFR 50.55a(g) based on the NRC Safety Evaluation . (Reference 2)

The lSI examination requirements of the ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.12, "Reactor Pressure Vessel Shell Longitud inal 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 longitud inal 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 00 surface or where access is practical from the internal 10 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 2001 Edition through 2003 Addenda of ASME Section XI, Appendix VIII as modified by the Performance Demonstration Initiative (POI) 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. Add itionally, the NRC assessment demonstrated that inspection of BWR RPV circumferential welds does not measurably affect the probability of failure.

The USNRC 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 lSI 001A-2 of IS1001A-10

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI*001A volumetric examination of RPV shell circumferential welds, ASME Section XI, Table IWB-2500-1 ,

Examination Category B-A, Item B1.11. The USNRC stated in that Boiling Water Reactor (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 probabil ity for circumferential welds in the NRC staffs 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 spec ified in the NRC staffs July 28, 1998, safety evaluation, (Criterion 2).

This request also demonstrates that the safety criteria spec ified in BWRVIP-74-A, (Reference 9) and the October 18, 2001 (Reference 10) will continue to be met for the extended period of 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 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 lSI requirements of the ASME Code Section XI 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) - Conditional Failure Probability 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 condition failure probabilities, a comparison of the chem istry values and the predicted f1uence 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 f1uence 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 f1uence and crack growth) .

lSI 001A-3 of IS1001A-10

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-001A For plants with Reactor Pressure Vessels fabricated by CSI Nuclear Company (CSIN), the peak end-of-license neutron fluence for circumferential welds used in the NRC PFM analysis was 18 5.1x10 n/crrr', At NMP2 , the highest f1uence anticipated at the end of the current license period 17 2 (36 EFPY) is 5.8x1 0 n/cm and at the end of the license renewal extended period of operation is 17 2 8.6x1 0 n/cm (Reference 7). Thus , embrittlement due to f1uence 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 SWRVIP-05 Limiting Plant Specific Analysis and Independent Assessment Fracture Analysis limiting case for the current license term . The chemistry factor, LiRTNDT, mean ART, and ART are calculated consistent with the guidelines of Regulatory Gu ide 1.99, Rev. 2. The data used for the evaluation based on the SWRVIP-05 methodology are also shown in Table 1.

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, nfern< 5.8x10 8.6x10" 5.1x10'* 6.06x10 9.03x1017 Initial RTNDT' *50 -50 -65 -40 -40 OF Chemistry Factor 54 54 134.9 95 95 Cu% 0.04* 0.04* 0.1 0.07* 0.07*

Ni% 0.82" 0.82" 0.99 0.89" 0.89*

t::. RTNDT (OF) Monte Carlo Predicted 109.5 Monte Carlo Predicted Mean ART (OF) 4.89 9.7 44.5 37.16 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 f1uence 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 (4.89°F) as compared to the NRC Final Safety Evaluation (SE) for CSIN plants (44.5 OF) at the end of the orig inal license period . Comparison of the NM P2 specific data and the data used in the NRC Final Safety Evaluation indicates that the difference is the f1uence at the end of 36 EFPY and the chem istry 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).

lSI 001A-4 of IS1001A-10

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI*001A 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 2.0x1 0.7 for 32 EFPY for CBIN fabricated vessels. The conditional probability of failure was 1.0x1 0-6 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 1x10*3/y r. This results in a total probability of failure of 5x10*12/ 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/y r to 10-4/y r. 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 of in-service 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 f1uence 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 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 1x10.7 for 36 EFPY and 54 3/yr.

EFPY . The BWRVIP frequency of over-pressurization was determined to be 1x10* This gives a total probability of failure for NMP2 of less than 2.5x1 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*11/yr and 1.33 x10*10/yr, for 36 EFPY and 54 EFPY, respectively (See Table 2). Comparison of the NMP2 circumferential weld failure probability

<<2.5x10*12/yr for 54 EFPY) and the axial weld failure probability (1.33x10*10/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 10 7 axial welds were typically from 2.5x10. 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 RAls, 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 <2.5x1 O"~/yr <2.5xlO-'~/yr

<2.5x10* 1l/Yr l U/yr Axial Welds 1.33x10' lSI 001A-5 of IS1001A-10

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-001A Criterion (2) - Limiting the Frequency of Cold Over-pressure Events Demonstrate licensees have implemented operator training and established procedures 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).

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 condit ions, 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.

lSI 001A-6 of IS1001A-10

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-001A Pressure testing of the RPV is classified as an "Infrequently Performed Test or Evolution" wh ich 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 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 and 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 lSI 001A-7 of IS1001A-10

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-001A 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, techn ical 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 staff's JUly 28, 1998, safety evaluation.

Summary In summary, the NMP2 spec ific chemistry, f1uence and ART were compared against the NRC staffs July 28, 1998 , safety evaluation 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 ident ify 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 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).

E. IMPLEMENTATION SCHEDULE:

Pursuant to 10 CFR 50.55a(a)(3)(i), Nine Mile Point Nuclear Station, LLC, Unit 2 requests permanent relief for the current license period and the license renewal extended period of operation. NMPNS has demonstrated that the criteria specified in GL 98-05 (Reference 4) are met for the initial operating license period, and that the criteria of BWRVIP-74-A (Reference 9) are met for the entire additional extended period of operation. Therefore, the requested duration of the proposed alternative is justified .

F. PRECEDENTS:

The NRC has previously approved a number of similar requests, includinq 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) lSI 001A-8 of IS1001A-10

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-001A G. ATTACHMENTS:

Figure 1 - Location of Reactor Pressure Vessel Welds 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, Gus C. Lainas (NRC) to Carl Terry, BWRVIP Chairman, USNRC 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, Jack R. Strosnider (NRC) to Carl Terry, BWRVIP Chairman , USNRC 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 Regulations, Nuclear Regulatory Commission, 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.

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.

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 Carl Terry, BWRVIP Chairman, "Acceptance for Referencing of EPRI Proprietary Report TR-113596, "BW R Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines (BWRVIP-74)"

and Appendix A, "Demonstration of Compliance with the Techn ical Information Requirements of the License Renewal Rule (10 CFR 54.21)," dated October 18,2001 .

11. NRC Letter to NMPNS , "Nine Mile Point Nuclear Station , Unit NO.2 - Authorization Under 10 CFR 50.55a(a)(3)(i) For Proposed Alternative Reactor Pressure Vessel Circumferential Shell Weld Volumetric Examinations (TAC No. MD3696), dated November 5,2007.

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 2151-002 Proposed Request for Relief In Accordance with 10 CFR 50.55a(g)(5)(iii)

A. COMPONENT IDENTIFICATION:

System: Reactor Recirculation System Code Class: ASME Code Class 1 Component

Description:

Two (2) Circumferent ial and Two (2) Meridional Reactor Vessel Bottom Head Welds:

Components Affected:

2RPV-DB Meridional bottom head radial plate-to- B-A B1.22 bottom head radial plate 2RPV-DC Meridional bottom head radial plate-to- B-A B1.22 bottom head radial plate 2RPV-DG Circumferential bottom head dollar B-A B1.21 plate-to-bottom head dollar plate 2RPV-DR Circumferential bottom head dollar B-A B1 .21 plate-to-bottom head dollar plate B. APPLICABLE CODE REQUIREMENTS:

The ASME Code,Section XI, applicable to the Nine Mile Point Nuclear Station (NMPNS), Unit 2, Third Ten-Year In-Service Inspection Program is the 2001 Edition through the 2003 Addenda .

Sub-article IWB-2500, Table IWB-2500-1 requires essentially 100% volumetric examination of all reactor vessel bottom head welds each inspection interval (i.e., Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, Item No's. B1 .21, Circumferential Head Welds , and B1.22, Meridional Head Welds) .

C. REASON FOR REQUEST FOR RELIEF:

In accordance with the provisions of 10 CFR 50.55a, "Codes and Standards ," paragraph 10 CFR 50.55a(g)(5)(iii) , Nine Mile Point Nuclear Station, LLC (NMPNS), Unit 2 requests relief from the requirement of ASME Code Section XI, 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 .21,"Circumferential Head Welds" and B1.22 , "Meridional Head Welds."

D. BASIS FOR RELIEF AND ALTERNATIVE EXAMINATIONS:

The original design of the reactor pressure vessel bottom head assembly does not provide accessibility for the manual volumetric examinations on the bottom head circumferential and File: 21SI-002 lSI 002-1 of lSI 002-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-002 meridional welds due to interference with the Control Rod Drive (CRD) penetrations and the Reactor Pressure Vessel (RPV) support skirt.

Only approximately 12" to 24" on' each end of circumferential bottom head dollar plate welds 2RPV-DG and 2RPV-DR can be examined due to interference with the CRD penetration hous-ings.

Approximately one foot cannot be examined on each of the other bottom head meridional welds 2EPV-DB and 2RPV-DC due to interference with the RPV support skirt.

Table 1 below provides the percent of Code Required Volume (CRV) achieved during the inspection interval and based upon the results has essentially remained unchanged from the previous inspection interval.

Table 1 Percent of Code Required Volume Achieved COMPONENT IDENTIFICATION PERCENT OF CODE REQUIRED VOLUME ACHIEVED 2RPV-DB 89% coverage achieved 2RPV-DC 82% coverage achieved 2RPV-DG 19% coverage achieved 2RPV-DR 21% coverage achieved The subject examinations have been completed to the maximum extent possible, additional coverage is not possible without redesign of the reactor vessel bottom head Proposed Alternative:

Nine Mile Point Nuclear Station will continue to implement to the maximum extent possible the inservice examination requirements of ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.21 and B1.22, In summary, NMPNS has concluded that based on the permanent limitations and the percent of examination coverage achieved that continued approval of the request for relief is justified in accordance with 10 CFR 50.55a(g)(5)(iii).

E. IMPLEMENTATION SCHEDULE Relief is requested for the Third Ten-Year Inservice Inspection Interval. (April 5, 2008 to April 4, 2018)

F. PRECEDENTS This request for relief was originally submitted as RR-IWB-7, under NMPNS letter dated July 30, 1999 and relief was granted. (Reference 1).

G. ATIACHMENTS :

Figure 1 - Typical weld locations H. REFERENCES

1. USNRC Safety Evaluation, dated March 3, 2000, TAC No. MA6273 File: 21SI-002 lSI 002-2 of lSI 002-3

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-003 Proposed Request for Relief In Accordance with 10 CFR 50.55a(g)(5)(iii)

A. COMPONENT IDENTIFICATION:

System: Reactor Pressure Vessel Code Class: ASME Code Class 1 Component

Description:

Reactor Pressure Vessel Flange to Shell Weld Components Affected :

2RPV-AE Reactor Pressure Vessel Flange to B-A B1.30 Shell Weld B. APPLICABLE CODE REQUIREMENTS:

The ASME Code,Section XI, applicable to the Nine Mile Point Nuclear Station (NMPNS), Unit 2, Third In-Service Inspection Program is the 2001 Edition through the 2003 Addenda.

Sub-article IWB-2500, Table IWB-2500-1 requires essentially 100% volumetric examination of the reactor vessel flange to shell weld each inspection interval (i.e., Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, Item No. B1.30, Shell to Flange Weld) .

C. REASON FOR REQUEST FOR RELIEF:

In accordance with the provisions of 10 CFR 50.55a, "Codes and Standards," paragraph 10 CFR 50.55a(g)(5)(iii), Nine Mile Point Nuclear Station, LLC, Unit 2 requests relief from the requirement of ASME Code Section XI, Sub article IWB-2500 , Table IWB-2500-1, Volumetric Examination of Examination Category B-A, "Pressure Retaining Welds in Reactor Vessel", Examination Item Number B1.30, "Shell to Flange Weld ."

D. BASIS FOR RELIEF AND ALTERNATIVE EXAMINATIONS:

The original design configuration of the shell to flange weld does not allow access from both sides of the weld due to the inside diameter taper from the flange forging to the thinner upper shell course.

Examination of the shell to flange weld was performed to the maximum extent poss ible from both the Reactor Pressure Vessel shell course and from the flange seal surface. Because of unparallel surfaces above the shell to flange weld it is impossible to achieve further coverage without redesign of the reactor pressure vessel flange.

The percent of Code Required Volume achieved during the inspection interval has essentially remained unchanged. Table 1 below provides the results of the percent of CRV achieved .

File: 21SI-003 lSI 003-1 of lSI 003-2

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-003

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IOENTIFI*CATION,* . ': ,A,CHIEVEDDURINGINSPECTION INTERVAL .

From the shell side 52% CRY Achieved 2RPV-AE-SS From the flange side 100% CRY Achieved 2RPV-AE-FS Proposed Alternative Nine Mile Point Nuclear Station will continue to implement to the maximum extent possible the inservice examination requirements of the ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.30 as follows :

Perform essentially 100% volumetric examination of flange to shell weld 2RPV-AE from the flange side during the first inspection period, and to the extent possible conduct volumetric examination of the shell to flange weld from the shell side during the third inspection period.

In summary, NMPNS has concluded that based on the permanent limitations, the percent of examination coverage achieved that continued approval of the request for relief is justified in accordance with 10 CFR 50.55a(g)(5)(iii).

E. IMPLEMENTATION SCHEDULE Relief is requested for the Third Ten-Year Inservice Inspection Interval (April 5, 2008 to April 4, 2018)

F. PRECEDENTS This request for relief was originally submitted as RR-IWB-3, under NMPNS letter dated July 30, 1999 and granted under USNRC Safety Evaluation, dated March 3, 2000 (Reference 1).

G. ATTACHMENTS :

None H. REFERENCES

1. USNRC Safety Evaluation, dated 03/03/00, TAC No. MA6273.

File: 21SI-003 lSI 003-2 of lSI 003-2

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-004 Proposed Request for Relief In Accordance with 10 CFR 50.55a(g)(5)(iii)

A. COMPONENT IDENTIFICATION:

System: Reactor Pressure Vessel Code Class: ASME Code Class 1 Component Description : Thirty-three (33) reactor pressure vessel nozzle-to-shell welds Components Affected:

RCS N1 Nozzle 2 B-'D B3.90 RCS N2 Nozzles 10 B-O B3.90 MSS N3 Nozzles 4 B-O B3.90 FWS N4 Nozzles 6 B-D B3.90 CSL N5 Nozzles 1 B-D B3.90 RHS N6 Nozzles 3 B-O B3.90 ICS N7 Nozzles 1 B-O B3.90 ISC N9 Nozzles 2 B-O B3.90 RPV N10 Nozzles 1 B-O B3.90 CSH N16 Nozzles 1 B-O B3.90 RPVCH N8 Nozzles 1 B-O B3.90 RPVCH N18 Nozzles 1 B-O B3.90 B. APPLICABLE CODE REQUIREMENTS:

The ASME Code,Section XI, applicable to the Nine Mile Point Nuclear Station (NMPNS), Unit 2, Third In-Service Inspection Program is the 2001 Edition through the 2003 Addenda .

Sub-article IWB-2500, Table IWB-2500-1 requires essentially 100% volumetric examination of the nozzle to shell welds each inspection interval (i.e., Examination Category B-O, Full Penetration Welds of Nozzles in Vessels (Inspection Program B), Item No. B3.90, Nozzle-to-Vessel Welds).

C. REASON FOR REQUEST FOR RELIEF:

In accordance with the provisions of 10 CFR 50.55a, "Codes and Standards ," paragraph 10 CFR 50.55a(g)(5)(iii), Nine Mile Point Nuclear Station , LLC (NMPNS), Unit 2 requests relief from the requirement of ASME Code Section XI, Sub article IWB-2500, Table IWB-2500-1 , Volumetric Examination of Examination Category B-O, "Full Penetration Welds of Nozzles in Vessels (Inspection Program B)", Examination Item Number B3.90, "Nozzle-to-Vessel Welds ."

D. BASIS FOR RELIEF AND ALTERNATIVE EXAMINATIONS:

The volumetric examination of these Reactor Pressure Vessel nozzle-to-shell welds is limited to varying extents due to nozzle-to-shell blend areas, vessel scanner tracks, close proximity of other nozzles, limited access from the nozzle side of welds and mechanical limitations .

Table 1 identifies each nozzle to shell weld that was examined to the maximum extent possible with the principal deterrent to achieving Code Compliance being the design configuration of the weld joints .

File: 21SI-004 lSI 004-1 of lSI 004-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-004 Proposed Alternative Nine Mile Point Nuclear Station will continue to implement to the maximum extent possible the inservice examination requirements of the ASME Code Section XI, Table IWB-2500-1 ,

Examination Category B-D, Item No. B3.90, utilizing the latest UT techniques and equipment.

NMPNS has demonstrated that the permanent limitations and the percent of examination coverage achieved has essentially remained unchanged and will continue to be met. Therefore, the requested duration of the proposed alternative is justified 2RPV-KA05 64.5% CRY achieved 2RPV-KA06 64.5% CRY achieved 2RPV-KA07 64.5% CRY achieved 2RPV-KA13 65.3% CRY achieved 2RPV-KA14 64% CRY achieved 2RPV-KA16 65% CRY achieved 2RPV-KA17 64.7% CRVachieved 2RPV-KA20 66.3% CRY achieved 2RPV-KA21 58% CRY achieved 2RPV-KA22 64.7% CRY achieved 2RPV-KA23 65.3% CRY achieved 2RPV-KA24 64.5% CRY achieved 2RPV-KA27 63% CRY achieved 2RPV-KA31 64% CRY achieved 2RPV-KA32 70% CRVachieved 2RPV-KA33 65.8% CRY achieved In summary, NMPNS has concluded that based on the permanent limitations, the percent of examination coverage achieved that continued approval of the request for relief is justified in accordance with 10 CFR 50.55a(g)(5)(iii).

File: 21SI-004 lSI 004-2 of lSI 004-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-004 E. IMPLEMENTATION SCHEDULE:

Relief is requested for the Third Ten-Year Inservice Inspection Interval (April 5, 2008 to April 4, 2018)

F. PRECEDENTS:

This request for relief was originally submitted as RR-IWB-2, under NMPNS letter dated July 30, 1999 and granted under NRC Safety Evaluation , dated March 3, 2000 (Reference 1).

G. AITACHMENTS:

None H. REFERENCES

1. USNRC Safety Evaluation , dated 03/03/00, TAC No. MA6273 File: 21SI-004 151 004-3 of 151 004-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-005 Proposed Request for Relief In Accordance with 10 CFR 50.55a(g)(5)(iii)

A. COMPONENT IDENTIFICATION:

System: Reactor Pressure Vessel Code Class: ASME Code Class 1 Component

Description:

Reactor Vessel pressure retaining welds in control rod housings Components Affected:

2RPV-CRDH001A Pressure retaining welds in Control 8-0 814 .10 through Rod Housings 2RPV-CRDH040A 2RPV-CRDH0018 Pressure retaining welds in Control 8-0 814.10 through Rod Housings 2RPV-CRDH0408

8. APPLICABLE CODE REQUIREMENTS:

The applicable ASME Code,Section XI, for Nine Mile Point Nuclear Station (NMPNS), Unit 2, Third 10-Year Interval , In-Service Inspection Program is the 2001 Edition through the 2003 Addenda.

Sub-article IW8-2500, Table IWB-2500-1 requires essentially 100% volumetric or surface examination of 10 percent of peripheral CRD housings each inspection interval (Le., Examinat ion Category 8-0, Pressure Retaining Welds in Control Rod Housings, Item No. 814.10, Welds in CRD Housing) .

C. REASON FOR REQUEST FOR RELIEF:

In accordance with the provisions of 10 CFR 50.55a , "Codes and Standards," paragraph 10 CFR 50.55a(g)(5)(iii), Nine Mile Point Nuclear Station, LLC (NMPNS), Unit 2 requests relief from the requirement of ASME Code Section XI, Sub-article IW8-2500, Table IWB-2500-1 , Volumetric or Surface Examination of Examination Category 8-0, "Pressure Retaining Welds in Control Rod Housings" , Examination Item Number 814 .10, "Welds in CRD Housing."

D. BASIS FOR RELIEF AND ALTERNATIVE EXAMINATIONS:

Limited accessibility for all 40 peripheral CRD housing welds are due to inherent obstructions caused by surrounding cables, tubing , and foundations which are not practical to remove or replace.

Each of the 40 peripheral CRD housing has two welds. Therefore, eight welds are required to be examined . Assuming Code Case N-460 minimum coverage allowable of 90%, eight (8) full weld examinations equals a minimum requirement of 720 total percentage points.

File: 21SI-005 lSI 005-1 of lSI 005-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 2151-005 Portions of six (6) additional welds were examined to the extent possible, such that, fourteen (14) welds were actually exam ined. (See Table 1 below) . Examination coverage ranged from 27% to 100% . The total of examined percentage points summed to 953, thus exceeding the 720 percentage points required. Although the use of an inspection mirror achieved 100% coverage on three of the welds (thus reducing the original population for which relief is sought from 8 to 5) this request is still required.

This request for relief has been updated to reference the applicable Code edition and addenda and the percentage of coverage that was achieved during the previous inspection interval.

Table 1 Percent of Code Required Volume Achieved Original Sample Additional Sample 2RPV-CRDH007A 54% Coveraoe achieved 2RPV-CRDH001A 43% Coverage achieved 2RPV-CRDH0078 54% Coveraqe achieved 2RPV-CRDH0018 80% Coveraqe achieved 2RPV-CRDH0036A 54% Coverage achieved 2RPV-CRDH004A 43% Coverage achieved 2RPV-CRDH00368 54% Coveraqe achieved 2RPV-CRDH0048 75% Coveraqe achieved 2RPV-CRDH0037A 54% Coveraqe achieved 2RPV-CRDH005A 100% Coverage achieved 2RPV-CRDH00378 54% Coveraqe achieved 2RPV-CRDH0058 100% Coveraqe achieved 2RPV-CRDH0038A 100%Coveraqe achieved 2RPV-CRDH00388 100% Coverage achieved Proposed Alternative:

Nine Mile Point Nuclear Station will continue to implement the following alternate provisions for the subject CRD Housing weld examinations.

The lSI examination requirements of the ASME Code Section XI , Table IW8-2500-1 , Exam inat ion Category 8-0, Item No. 814.10, shall be performed, to the maximum extent possible, and shall include:

Partial examinations of 10% of the welds plus six (6) additional welds, such that the aggregate total is greater than or equal to eight full examinations (720 total percentage points .)

E. IMPLEMENTATION SCHEDULE:

NMPNS has demonstrated that the permanent limitations and the percent of examination coverage achieved has essentially remained unchanged and will continue to be met. Therefore, the requested duration of the proposed request for relief is justified.

Relief is requested for the Third Ten-Year In-service Inspection Interval (April 5, 2008 to April 4, 2018)

F. PRECEDENTS:

This request for relief was originally submitted as RR-IW8-1 , under NMPNS letter dated July 30, 1999 and relief was granted (Reference 1).

G. ATTACHMENTS:

Figure 1 - Unit 2 Plan View H.

REFERENCES:

1. USNRC Safety Evaluation , dated March 3, 2000 , TAC No. MA6273 File: 21SI-005 lSI 005-2 of lSI 005-3

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- Typical of 185 Locations File: 21SI-005 lSI 005-3 of lSI 005-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-006 Proposed Request for Relief In Accordance with 10 CFR 50.55a(g)(5)(iii)

A. COMPONENT IDENTIFICATION:

System: Reactor Pressure Vessel Code Class: ASME Code Class 1 Component

Description:

2MSS*REV1 (2RPV-TF001 thru 2RPV-TF076) , Threads in Reactor Vessel Flange Components Affected:

2RPV-TF001 thru Threads in Reactor Vessel Flange 8-G-1 B6.40 2RPV-TF076 B. APPLICABLE CODE REQUIREMENTS :

The applicable ASME Code,Section XI, for the Nine Mile Point Nuclear Station (NMPNS), Unit 2, Third In-Service Inspection Program is the 2001 Edition through the 2003 Addenda .

Sub-article IW8-2500 requires components specified in Table IWB-2500-1 to be examined. Table IWB-2500-1 requires volumetric examination of essentially 100% of the volume described by Figure IWB-2500-12 of Examination Category B-G-1, "Pressure Retaining Bolting Greater Than 2 in. Diameter", Examination Item Number 86.40 , "Threads in Reactor Vessel Flange.

C. REASON FOR REQUEST FOR RELIEF:

In accordance with the provisions of 10 CFR 50.55a, "Codes and Standards ," paragraph 10 CFR 50.55a(g)(5)(iii) , Nine Mile Point Nuclear Station, LLC (NMPNS) , Unit 2 requests relief from the requirement of ASME Code Section Xl , Sub-article IW8-2500 , Table IWB-2500-1 , Volumetric Examination of Examination Category 8-G-1 , "Pressure Retaining Bolting Greater Than 2 in.

Diameter", Examination Item Number B6.40, "Threads in Reactor Vessel Flange ."

D. BASIS FOR RELIEF AND ALTERNATIVE EXAMINATIONS:

The groove that the o-ring seal is placed in limits the accessibility of the transducers used to ultrasonically interrogate this base material. As a result, 100% volumetric interrogation is deemed impractical.

Nine Mile Point Nuclear Station has considered the consequences of a failure of this system and finds that, due to the conservatism of design inherent to the reactor pressure vessel, catastrophic failure of this component is considered highly unlikely (as reflected in the FSAR choice of the design basis accident.) Therefore , further analysis of the consequences of failure of the reactor pressure vessel flange threads is not required.

File: 21SI-006 lSI 006-1 of lSI 0006-2

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-006 Proposed Alternative:

Nine Mile Point Nuclear Station will continue to implement to the maximum extent possible the in-service examination requirements of the ASME Code Section XI, Table IW8-2500-1 , Examination Category 8-G-1 , Item No. 86.40 for each of the 76 ligament areas, ie., CRV =90.2%.

In summary , NMPNS has concluded that based on the percent of examination coverage achieved that continued approval of the request for relief is justified in accordance with 10 CFR 50.55a(g)(5)(iii) .

Lastly, These examinations document interrogated volumes greater than 90%, but less than 100%, in all cases. There are no additional techniques that could be utilized to increase the volume examined for each of the ligament areas.

E. IMPLEMENTATION SCHEDULE:

Relief is requested for the Third Ten-Year In-service Inspection Interval (April 5, 2008 to April 4, 2018)

F. PRECEDENTS:

This request for relief was originally submitted under RR-IWB-13 , under NMPNS letter dated July 30,1999 and granted under NRC Safety Evaluation, dated March 3, 2000 (Reference 1).

G. ATIACHMENTS:

None H.

REFERENCES:

1. USNRC Safety Evaluation. Dated March 3, 2000, TAC No. MA6273 File: 2151-006 151 006-2 of 151 0006-2

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 2151-007 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

A. COMPONENT IDENTIFICATION System: Various ASME Code Class 1 and 2 Systems Code Class: ASME Code Class 1 and 2 Component Description : ASME Code Class 1 and 2 Piping Welds Components Affected:

Various ASME Code Class 1 Piping Welds B-F B5.10, B5.20, B5.30, B5.100, B5.110, B5.120 Various ASME Code Class 1 Piping Welds B-J B9.11, B9.21, B9.31, B9.32, B9.40 Various ASME Code Class 2 Piping Welds C-F-1 C5.11, C5.21, C5.30, C5.40 Various ASME Code Class 2 Piping Welds C-F-2 C5.51, C5.61, C5.70, C5.81 Various IGSCC Category A Piping Welds A N/A B. APPLICABLE CODE REQUIREMENTS The applicable ASME Code,Section XI, for the Nine Mile Point Nuclear Station (NMPNS) , Unit 2, Third In-Service Inspection Program is the 2001 Edition through the 2003 Addenda .

C. REASON FOR REQUEST FOR RELIEF In accordance with the provisions of 10 CFR 50.55a, "Codes and Standards," paragraph 10 CFR 50.55a(a)(3),

Nine Mile Point Nuclear Station, LLC (NMPNS), Unit 2 requests relief from the requirement of ASME Code Section XI, Sub-article IWB-2500 and IWC-2500, Tables IWB-2500-1 and IWC-2500-1, Examination Categories B-F, B-J, C-F-1 and C-F-2, "Pressure Retaining Welds in Piping", and IGSCC Category "A" welds.

ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 currently contain the requirements for examination of piping components via nondestructive examination (NDE). The previously approved RI-ISI program (Reference 1) will be substituted for Class 1 and Class 2 piping (Examination Categories B-F, B-J, C-F-1, C-F-2 and IGSCC Category A welds) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section XI Code will be unaffected.

D. BASIS FOR RELIEF AND ALTERNATIVE EXAMINATIONS Pursuant to 10 CFR 50.55a(a)(3), NRC approval of the Nine Mile Point Nuclear Station, Unit 2 Alternate Risk-Informed In-service Inspection program (RI-ISI) as an alternative to the current 2001 Edition through the 2003 Addenda, ASME Section XI inspection requirements for Class 1, Examination Category B-F and B-J, Class 2, Examination Category C-F-1 and C-F-2, and IGSCC Examination Category "A" piping welds is requested. The Nine Mile Point Unit 2 RI-ISI Program has been developed in accordance with the EPRI methodology contained in EPRI TR-112657, "Risk-Informed In-service Inspection Evaluation Procedure" (Reference 2). It was approved for use at Nine Mile Point Nuclear Station during the second inspection period of the second ten-year inspection File: 21SI-007 lSI 007-1 of lSI 007-5

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-007 interval and is still applicable for the third in-service inspection interval. The Nine Mile Point Unit 2 specific RI-ISI program is summarized in Table 1. The RI-ISI program has been updated consistent with the intent of NEI-04-05 (Reference 3) and continues to meet EPRI TR-112657 and Regulatory Guide 1.174 risk acceptance criteria.

The ASME Code Section XI Code required minimum percentage (50%) was completed in the second period of the second lSI Interval and the remaining fifty percent (50%) of the RI-ISI program welds were completed by the end of the third Inspection Interval, as required by Reference 1. This Relief Request is to align the RI-ISI Interval and the Code Year with the Third Interval lSI program. 100% of the RI-ISI Program weld examinations will be completed in the Third Inspection Interval.

NMPNS will continue to implement the Risk-Informed Inservice Inspection Program in accordance with ASME Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B,Section XI, Division 1."

In addition, NMPNS intends to perform additional examinations required due to the identification of flaws, which are determined to exceed the acceptance standards, during the current refueling outage prior to the units return to service.

The ultrasonic examination volume to be used based on degradation mechanism and component configuration will be the examinat ion figures specified in Section 4 of EPRI TR-112657 .

NMPNS intends that for ultrason ic examination procedures , equipment, and personnel used to detect and size flaws in piping welds shall be qualified by performance demonstration in accordance with ASME Section XI Appendix VIII , "Performance Demonstration for Ultrasonic Examination Systems." The volumetric scanning will be in both axial and circumferential directions to detect flaws in these orientations.

As part of the RI-ISI living program update, the delta risk assessment was re-evaluated and was determined to continue to meet the delta risk acceptance criteria of EPRI TR-112657.

E. IMPLEMENTATION SCHEDULE Relief is requested for the Third Ten-Year In-service Inspection Interval of the Nine Mile Point Unit 2 In-service Inspection Program, beginning April 5, 2008 and scheduled to end April 4, 2018.

F. PRECEDENTS USNRC previously approved the Nine Mile Point Nuclear Station, Unit 2 Alternate Risk-Informed In-service Inspection Program via Reference 1.

Per our commitment in section 4 of our original relief request (Reference 5), NMPNS during the re-evaluation process following each inspection period, considers both the plant and industry operat ing experience and updates the RI-ISI program as required .

G. AITACHMENTS Table 1 -Inspection Location Selection Comparison ASME Section XI Code and EPRI TR-112657 by Risk Category."

H. REFERENCES

1. USNRC Letter dated May 31, 2001 "Nine Mile Point Nuclear Station, Unit NO.2 - Approval to Use a II Risk-Informed Inservice Inspection Program for the Second 10-Year Interval" (TAC No. MB0297).

2. EPRI TR-112657. Electric Power Research Institute Report for Alternative Requirements of Risk-Informed In-service Inspection Evaluation Procedure, EPRI, Polo Alto, CA: 1999, Rev B-A.

3. NEI-04-05 , "Living Program Guidance to Maintain Risk-Informed In-service Inspection Programs for Nuclear Plant Piping Systems", dated April 2004 .

File: 21SI-007 lSI 007-2 of lSI 007-5

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 2151-007

4. Request for Additional Information Related to Byron Station, Units 1 and 2, Request for relief 13R-02, TAC Ns MD3855 and MD3856, dated August 8,2007

5. NMPC Letter dated October 16,2000, Request for Authorization to Use Risk-Informed Inservice Inspection Alternative."

File: 21SI-007 lSI 007-3 of lSI 007-5

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 2151-007 Table 1 Inspection Location Selection Comparison Between ASME Section XI Code and EPRI TR-112657 by Risk Cateaorv System Risk Consequence Failure Potential Code Category 1~'. Approved RI-ISllnterval New RI-ISllnterval Category Rank Rank OMs Rank Weld RI-ISI Other Weld I RI-ISI Other Count Count 001-ASS 68 Low Medium None Low C-F-2 4 0 4 0 002-CSH 2 High Hloh TASCS Medium 8-J 11 3 11 3 002-CSH 4 Medium Hioh None Low 8-J , C-F-1 6 1 6 1 002-CSH 68 Low Medium None Low C-F-2 164 0 165 0 002-CSH 7 Low Low None Low C-F-1 4 0 4 0 003-CSL 2 Hiqh Hiah TASCS Medium 8-J 8 2 8 2 003-CSL 4 Medium Hiah None Low 8-J, C-F-1 10 1 10 1 003-CSL 68 Low Medium None Low 8-J . C-F-2 114 0 114 0 003-CSL 7 Low Low None Low C-F-1 4 0 4 0 004-DER 2 Hiah Hiqh TASCS Medium 8-J 1 1(e) 1 1(e) 004-DER 7 Low Low None Low 8-J 2 0 2 0 005-FWS 2 Hiah Hioh TASCS Medium 8-J 31 8 31 8 005-FWS 4(1) Medium (Hiah) Hiah FAC Low (Hioh) 8-J 4 0 4 0 005-FWS 4 Medium Hioh None Low 8-J 43 5 43 5 005-FWS 58 Medium Medium CC Medium 8-J 2 1 om 0 005-FWS 58 Medium Med ium TASCS Medium 8-J 11 1 11 2 005-FWS 58 Medium Medium TASCS, CC Medium 8-J 4 0 4il 0 005-FWS 68 Low Medium None Medium 8-J 6 0 8(fl 0 007-ICS 2 HiQh Hlch TT,TASCS Medium 8-J 9 3 10 (0) 3 007-ICS 4 Medium Hiah None Low 8-J, C-F-2 41 5 43(h) 5 007-ICS 58 Medium Medium TT, TASCS Medium 8-J 3 1 4'i) 1 007-ICS 68 Low Medium None Low 8-J , C-F-2 223 0 223 0 007-ICS 7 Low Low None Low N/A 1 0 1 0 008-ISC 2 Hiah Hiah TASCS Medium 8-J 3 1 3 1 008-ISC 4 Medium Hiah None Low 8-J 2 1 2 1 008-ISC 58 Medium Medium TASCS Medium 8-J 3 1(e) 3 1(e) 008-ISC 68 Low Medium None Low B-F ,8-J 11 0 11 0 013-MSS 4 Medium Hioh None Low B-J 84 9 84 9 013-MSS 58 Medium Medium TASCS Medium B-J 10 1 10 1 013-MSS 68 Low Medium None Low 8-F , 8-J, C-F-2 238 0 243(a) 0 013-MSS 7 Low Low None Low B-J 8 0 10(a) 0 011-RCS 4(2) Medium (Hiah) Hiah IGSCC Low (Medium) 8-J 1 1 1 1 011-RCS 4 Medium Hiah None Low 8-J 105 10 105 10 010-RDS 2 High Hich CC Medium B-J 1 1 on>> 0 010-RDS 4 Medium Hiah None Low 8-J 1 1 2(b) 1 010-RDS 68 Low Medium None Low C-F-2 76 0 76 0 014-RHS 2 Hlch Hiah EC Medium 8-J 4 1 4 1 014-RHS 2 Hiqh Hioh TASCS Medium 8-J 22 6 22 6 014-RHS 4 Medium Hioh None Low 8-J , C-F-1, C-F-2 77 8 77 8 014-RHS 58(3) Medium(H iah) Medium TASCS ,FAC Medium C-F-2 17 0 17 0 014-RHS 58 Medium Medium TASCS Medium C-F-2 208 23 208 23 014-RHS 68(3) Low(High) Medium FAC Low(Hiahl 8-J, C-F-2 3 0 3 0 014-RHS 6A Low Low TASCS Medium N/A 16 0 16 0 File: 21SI-007 lSI 007-4 of lSI 007-5

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 2151*007 Table 1 Inspection Location Selection Comparison Between ASME Section XI Code and EPRI TR-112657 by Risk Category System Risk Consequence Failure Potential Code Category 1~*. Approved RI-ISllnterval New RI-ISI Interval Category Rank Rank OMs Rank Weld RI-ISI Other Weld RI-ISI Other Count Count 014-RHS 6B Low Medium None Low B-J, C-F-2 537 0 536(e) 0 014-RHS 7 Low Low None Low B-J, C-F-1, C-F-2 104 0 104 0 006-RPV 2 HiQh Hlqh IGSCC , CC Medium B-J 21 6 Old) 0 006-RPV 4 Medium HiQh None Low B-J 3 0 3 0 006-RPV 4(2) Medium(HiQh) Hiqh IGSCC Low(Medium) B-F 9 2 30(d) 4 006-RPV 6B Low Medium None Low B-J 1 0 1 0 009-SLS 2 High High TASCS Medium B-J 3 1 3 1 009-SLS 4 Medium High None Low B-J 14 2 14 2 009-SLS 5B Medium Medium TASCS Medium B-J 7 1 7 1 009-SLS 6B Low Medium None Low B-J 26 0 26 0 012-WCS 2 High High TASCS Medium B-J 10 0 10 0 012-WCS 2 High High TASCS,IGSCC Medium B-J 8 5 8 5 012-WCS 4(2) Medium(High) High IGSCC Low(Medium) B-J 10 5 10 5 012-WCS 4(1) Medium(HiQh) Hiqh FAC Low(HiQh) B-J 4 0 4 0 012-WCS 4 Medium HiQh None Low B-J 75 4 75 4 012-WCS 5B(3) Medium(High) Medium TASCS,FAC Medium(High) B-J 4 0 4 0 012-WCS 5B Medium Medium TASCS Medium B-J 25 3 25 3 012-WCS 6B(3) Medium(HiQh) Medium FAC Low(HiQh) B-J 2 0 2 0 012-WCS 6B Low Medium None Low B-J 15 0 15 0 012-WCS 7 Low Low None Low B-J 8 0 8 0 2482 125 2492 120 Notes to Table 1:

(a) MSS - main steam drain modification added five (5) welds to RC 6 and 7.

(b) RDS - one (1) weld moved from RC 2 to RC 4 as a result of CC updated evaluation; (c) Socket welds require only a VT-2 examination each outage, in conjunction with system leakage test (d) RPV - twenty-one (21) welds moved from RC 2 to RC 4 as a result of CC updated evaluation; (e) RHS - one (1) weld deleted from RC 6 as a result of piping modification; (t) FWS - two (2) welds were moved from RC 5 to RC 6 as a result of CC updated evaluation; (g) ICS - one (1) weld added to RC 2 as a result of piping and valve modification ;

(h) ICS - two (2) welds added to RC 4 as a result of piping and valve modification; (i) ICS - one (1) weld added to RC 5 as a result of piping and valve modification; (j) FWS- CC mechanism removed as a result of CC update evaluation ;

File: 21SI-007 lSI 007-5 of lSI 007-5

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 2151-008 Proposed Request for Relief In Accordance with 10 CFR 50.55a(g)(5)(iii)

A. COMPONENT IDENTIFICATION:

System: Low Pressure Core Spray (LPCS) and Residual Heat Removal (RHS)

Code Class: ASME Code Class 2 Component

Description:

LPCS and RHS valve body welds Components Affected :

2CSL *HCV118, VWHCV118-C Valve Body Welds C-G C6.20 2CSL *HCV118 , VWHCV118-D Valve Body Welds C-G C6.20 2CSL *HCV118, VWHCV118-LW Valve Body Welds C-G C6.20 2CSL*HCV119, WVHCV119-C Valve Body Welds C-G C6.20 2CSL *HCV 119, WVHCV-119-D Valve Body Welds C-G C6.20 2CSL *HCV119, WVHCV119-LW Valve Body Welds C-G C6.20 2CSL *MOV112, VWMOV112-C Valve Body Welds C-G C6.20 2CSL *MOV112, VWMOV112-D Valve Body Welds C-G C6.20 2CSL *MOV112, VWMOV112-LW Valve Body Welds C-G C6.20 2CSL*V121, VWV121-C Valve Body Welds C-G C6 .20 2CSL *V121, VBW121-LW Valve Body Welds C-G C6.20 2RHS *V376, VWV376-LW Valve Body Welds C-G C6.20 2RHS*V378, VWV378-LW Valve Body Welds C-G C6.20 2RHS*MOV8A, VWMOV8A-D Valve Body Welds C-G C6.20 2RHS *MOV1 C, VWMOV1 C-C Valve Body Welds C-G C6.20 2RHS*MOV1 C, VWMOV1 C-D Valve Body Welds C-G C6 .20 2RHS*MOV1C, VWMOV1C-LW Valve Body Welds C-G C6.20 2RHS*MOV2A, VWMOV2A-C Valve Body Welds C-G C6.20 2RHS*MOV2A, VWMOV2A-D Valve Body Welds C-G C6.20 2RHS *MOV8A, VWMOV8A-C Valve Body Welds C-G C6.20 File: 21SI-008 lSI 008-1 of lSI 008-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-008 B. APPLICABLE CODE REQUIREMENTS:

The ASME Code,Section XI, applicable to the Nine Mile Point Nuclear Station (NMPNS), Unit 2, Third In-Service Inspection Program is the 2001 Edition through the 2003 Addenda.

Sub-article IWC-2500 requires components specified in Table IWC-2500-1 to be examined. Table IWC-2500-1 requires essentially 100% surface examination of all components in each piping run examined under Examination Category C-F each inspection interval (i.e., Examination Category C-G, Pressure Retaining Welds in Pumps and Valves, Item No. C6.20 Valve Body Welds).

C. REASON FOR REQUEST FOR RELIEF:

In accordance with the provisions of 10 CFR 50.55a, "Codes and Standards," paragraph 10 CFR 50.55a(g)(5)(iii), Nine Mile Point Nuclear Station, LLC (NMPNS) , Unit 2 requests relief from the requirement of ASME Code Section XI, Sub-article IWC-2500, Table IWC-2500-1, 100% Surface Examination of Examination Category C-G, "Pressure Retaining Welds in Pumps and Valves, "

Examination Item Number C6.20, "Valve Body Welds.

Accessibility to the valve body welds is limited due to permanent interferences .

D. BASIS FOR RELIEF AND ALTERNATIVE EXAMINATIONS:

A significant portion of the required code coverage has been achieved, as noted below in Table 1, for the twenty welds for which relief is requested . This coverage assures an acceptable level of inservice structural integrity. To increase the percent of coverage, major redesign and modification would be required without a compensating increase in the level of quality or safety.

TABLE 1 PERCENT CODE REQUIRED SURFACE ACHIEVED COMPONENT IDENTIFICATION CODE REQUIRED COMPONENT IDENTIFICATION CODE REQUIRED SURFACE SURFACE ACHIEVED ACHIEVED 2CSL *HCV118, VWHCV118*C 2RHS*V376, VWV376-LW 86% MT coverage 82% coverage 2CSL*HCV118, VWHCV118-D 2RHS*V378, VWV378-LW 86% MT coverage 81% coverage 2CSL*HCV118 , VWHCV118*LW 2RHS*MOV8A, VWMOVSA-D 76% PT coverage 80% PT coverage 2CSL*HCV119, WVHCV119-C 2RHS*MOV1 C , VWMOV1 e-C 60% PT coverage 70% PT coverage 2CSL *HCV119, WVHCV119*D 2RHS*MOV1 C , VWMOV1 CoD 80% PT coverage 84% PT coverage 2CSL *HCV119, WVHCV119*LW 2RHS*MOV1 C , VWMOV1 C-LW 82% MT coverage 81% PT coverage 2CSL *MOV112 , VWMOV112*C 2RHS*MOV2A , VWMOV2A-C 80% MT coverage 60% PT coverage 2CSL *MOV112, VWMOV112*D 2RHS*MOV2A , VWMOV2A-D 60% PT coverage 80% PT coverage 2CSL*MOV112 , VWMOV112*LW 2RHS*MOV8A , VWMOV8A-C 87% MT coverage 60% PT coverage 2CSL*V121 , VWV121-C 80% MT coverage 2CSL*V121, VBW121 -LW 87% MT coverage NMPNS has demonstrated that the permanent limitations and the percent of examination coverage achieved has essently remained unchanged and will continue to be met. Therefore, the requested duration of the proposed alternative is justified.

File: 21SI-008 lSI 008-2 of lSI 008-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-008 Proposed Alternative Nine Mile Point Nuclear Station will implement the lSI examination requirements of the ASME Code Section XI, Table IWC-2500-1, Examination Category C-G, Item No. C6.20, to the maximum extent possible.

E. IMPLEMENTATION SCHEDULE:

Relief is requested for the Third Ten-Year Inservice Inspection Interval (April 5, 2008 to April 4, 2018)

F. PRECEDENTS:

This request for relief was originally submitted as RR-IWC-5, Part 3, under NMPNS letter dated July 30, 1999 and granted under NRC Safety Evaluation, dated March 3, 2000 (Reference 1).

G. ATTACHMENTS:

None H.

REFERENCES:

1. USNRC Safety Evaluation, dated 03/03/00, TAC No. MA6273 File: 2151-008 151 008-3 of 151 008-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 2151-009 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(ii)

A. COMPONENTIDENTIFICATION:

System: High Pressure Core Spray (HPCS), Low Pressure Core Spray (LPCS), Residual Heat Removal (RHS) and Reactor Core Isolation Cooling (RCIC).

Code Class: ASME Code Class 2 Component

Description:

Thirty-four (34) weldments, on six (6) separate pumps ; seven (7) integral attachment (Cat. C-C) welds, and 27 pressure retaining (Cat. C-G) welds:

Components Affected:

2CSH*P1 PW207 Pump Casing Welds C-G C6.10 PW208 PW209 PW212 PW217 PW218 PW219 2CSL*P1 PW311 Pump Casing Welds C-G C6.10 PW312 PW315 PW316 PW319 2RHS*P1A PW111A, Pump Casing Welds C-G C6.10 PW112A PW113A PW116A PW118A PW121A Welded Attachment C-C C3.30 2RHS*P1 B PW111B Pump Casing Welds C-G C6.10 PW112B PW113B PW116B PW118B PW121B Welded Attachment C-C C3.30 2RHS*P1C PW111C Pump Casing Welds C-G C6.10 PW112C PW113C PW116C PW118C PW121C Welded Attachment C-C C3.30 File: 21SI-009 lSI 009-1 of lSI 009-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 2151-009 2ICS *P1 PW400 Welded Attachment C-C C3.30 PW401 PW402 PW403 B. APPLICABLE CODE REQUIREMENTS:

The applicable ASME Code ,Section XI, for the Nine Mile Point Nuclear Station (NMPNS), Unit 2, Third In-Service Inspection Program is the 2001 Edition through the 2003 Addenda.

Sub-article IWC-2500 , Table IWC-2500-1 , requires essentially 100% Surface Examination of Pump Casing Welds each inspection interval (l.e., Examination Category C-G, "Pressure Retaining Welds in Pumps and Valves," Examination Item Number C6.10, "Pump Casing Welds ."

Sub-article IWC-2500, Table IWC-2500-1 , requires essentially 100% Surface Examination of Pump Welded Attachments each inspection interval (Le., Examination Category C-C, "W elded Attachments for Vessels, Piping, Pumps and Valves," Examination Item Number C3.30, "Pump Welded Attachments)".

C. REASON FOR REQUEST FOR RELIEF:

In accordance with the provisions of 10 CFR 50.55a, "Codes and Standards," paragraph 10 CFR 50.55a(a)(3), Nine Mile Point Nuclear Station , LLC (NMPNS), Unit 2 requests relief, from the requirement of ASME Code Section XI, Sub article IWC-2500, Table IWC-2500-1, Surface Examination of Examination Category C-G, "Pressure Retaining Welds in Pumps and Valves, "

Examination Item Number C6.10, "Pump Casing Welds" and Surface Examination of Examination Category C-C, "Welded Attachments for Vessels, Piping , Pumps and Va lves," Examination Item Number C3.10, "Pump Welded Attachments."

D. BASIS FOR RELIEF AND ALlERNAliVE EXAMINATIONS:

The pumps are installed in a concrete pit, thereby making the exterior of the casing welds and entire integral attachment welds inaccessible for surface examination. Examination of the casing welds would require either disassembly or removal from the pit. Examination of the integral attachment welds would require lifting the pump from the pit. The hardships associated with pump disassembly of lifting from the pit would far exceed any beneficial safety improvements that might be achieved by such an examination. For the integral attachments on pump ICS*P1, approximately 17% of each of the four welds is inaccessible. The pump design utilizes U shaped attachments that limit access to the entire weld surface.

Since these pumps are subject to testing per ASME O&M Code , foss of integrity of the pump casing welds would be detected during quarterly pressure, differential, and flow rate testing .

Failure of integral attachments welds would be detected by quarterly vibration measurements.

Furthermore, pump casing integrity is verified during system leakage testing .

Proposed Alternative:

Nine Mile Point Nuclear Station will continue to implement to the maximum extent poss ible the in-service surface examination requirements of the ASME Code Section XI, Table IWC-2500-1, Examination Category CoG, Item No. C6.1 0 and Examination Category C-C, "Welded Attachments for Vessels, Piping, Pumps and Valves," Examination Item Number C3.30, "Pump File: 21SI-009 lSI 009-2 of lSI 009-3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval Request Number 21SI-009 Welded Attachments " on welds of pumps that become access ible when disassembled for routine maintenance .

In summary , NMPNS has concluded that based on the permanent limitations and the percent of examination coverage achieved that continued approval of the request for relief is justified and presents a hardship without a compensating increase in safety, in accordance with 10 CFR 50.55a (a)(3)(ii).

E. IMPLEMENTATION SCHEDULE Relief is requested for the Third Ten-Year Inservice Inspection Interval. (April 5, 2008 to April 4, 2018)

F. PRECEDENTS:

This request for relief was originally submitted as RR-IWC-!, under NMPNS letter dated July 30, 1999 and granted under NRC Safety Evaluation , dated March 3, 2000 (Reference 1).

G. ATTACHMENTS:

None H.

REFERENCES:

1. USNRC Safety Evaluation, dated March 3, 2000, TAC No. MA6273.

File: 21SI-009 lSI 009-3 of lSI 009-3