GNRO-2012/00128, Response to Requests for Additional Information (RAI) Set 38 Dated 10/03/2012

Response to Requests for Additional Information (RAI) Set 38 Dated 10/03/2012

ML12297A204
Person / Time
Site:	Grand Gulf
Issue date:	10/22/2012
From:	Mike Perito Entergy Operations
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
GNRO-2012/00128, TAC ME7493
Download: ML12297A204 (13)

Text

• Entergy

,.~. Entergy Operatio ns, Inc.

P. O. Box 756 Port Gibson, MS 39150 Michael Perito Vice President. Operations Grand Guif Nuclear Station Tel. (601) 437-6409 GNRO-2012/00128 October 22,20 12 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Response to Requests for Additional Information (RAI) Set 38 dated October 3, 2012 Grand Gulf Nuclear Station, Unit NO.1 Docket No. 50-416 License No. NPF-29

REFERENCE:

NRC Letter, "Requests for Additional Information for the Review of the Grand Gulf Nuclear Station License Renewal Application," dated October 3,201 2 (GNRI-2012/00219) (TAC No. ME7493)

Dear Sir or Madam:

Entergy Operations, Inc is providing, in the Attachment, the respon se to the referenced Requests for Additional Information (RAJ).

This letter contains no new commitments. If you have any questi ons or require additional information, please contact Christina L. Perino at 601-437-6299 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 22nd day of October, 2012.

Sincerely,

~MP/jas

Attachment:

Response to Requests for Additional Information (RAI) cc: (see next page)

GNRO-2012/00128 Page 2 of 2 cc: with Attachment Mr. John P. Boska, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8-C2 Washington, DC 20555 cc: without Attachment Mr. Elmo E. Collins, Jr.

Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 U.S. Nuclear Regulatory Commission ATTN: Mr. A. Wang, NRR/DORL Mail Stop OWFN/8 G14 11555 Rockville Pike Rockville, MD 20852-2378 U.S. Nuclear Regulatory Commission ATIN: Mr. Nathaniel Ferrer NRR/DLR Mail Stop OWFNI 11 F1 11555 Rockville Pike Rockville, MD 20852-2378 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150

Attachment to GNRO-2012100128 Response to Requests for Additional Information (RAI)

Attachment to GNRO-2012/00128 Page 1 of 10 The format for the Request for Additional Information (RAI) responses below is as follows. The RAI is listed in its entirety as received from the Nuclear Regulatory Commission (NRC) with background, issue and request subparts. This is followed by the Grand Gulf Nuclear Station (GGNS) RAI response to the individual question.

RAI4.3-5b Background. In its response to follow-up request for additional information (RAI) 4.3-5a dated September 4, 2012, Entergy Operations, Inc. (the applicant) provided a table that lists its non-Class 1 expansion joints. For several of these expansion joints under the evaluation column the applicant provided the usage factor associated with the design cycles for the non-Class 1 expansion joints. These expansion joints include the standby liquid control (at the pump discharge), standby diesel generators (turbocharger water outlets), high pressure core spray (standby service water supply to diesel, diesel air start and diesel fuel oil) and the leakage detection and controls system (exhaust blower inlet). As part of the evaluation, the applicant stated that based on the usage factor associated with the design cycles these expansion joints are acceptable for many more cycles than specified.

Furthermore, for several other expansion joints under the design cycles column the applicant provided the number of cycles for which the specific expansion joint was designed. In addition, under the evaluation column the applicant provided the number of cycles that a particular expansion joint was "qualified" for. These expansion joints include the standby liquid control (at tank outlet and pump inlets and at test tank), high pressure core spray (diesel exhaust) and compressed air (air accumulators). As part of the evaluation, the applicant stated that these expansion joints were qualified or determined to a number of cycles that is greater than the design cycles; therefore, the expansion joints are acceptable for many more cycles than specified.

Issue. Although the expansion joints may have a "small" usage factor when compared to the design limit of 1.0, the U.S. Nuclear Regulatory Commission (NRC or the staff) noted that a "small" usage factor does not support the disposition that the analysis remains valid for the period of extended operation (i.e., 10 CFR 54.21 (c)(1 )(i)). Similarly, the staff noted a component being "qualified" for many more cycles compared to the design cycles does not support the disposition that the analysis, which is based on the number of design cycles, remains valid for the period of extended operation (i.e., 10 CFR 54.21 (c)(1 )(i)).

In order for the time-limited aging analyses of these expansion joints to remain valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(i), the applicant must demonstrate that the number of "design cycles" used in the current design analysis will not be exceeded after 60 years of operation (e.g., claiming a design cycle limit of 1200 thermal cycles for the high pressure core spray diesel exhaust joint without demonstrating that plant operation would not exceed this 1200 cycles is not an adequate justification under 10 CFR 54.21 (c)(1)(i)).

The staff noted that for the expansion joints in the standby liquid control system and high pressure core spray system, the license renewal application (LRA) does not provide AMR results for these expansion joints subject to "cracking-fatigue."

Request.

a. Supplement RAI 4.3-5a response to provide, for non-Class 1 expansion joints discussed above, adequate justification for the disposition of 10 CFR 54.21 (c)(1 )(i), by

Attachment to GNRO-2012/00128 Page 2 of 10 demonstrating that the design number of cycles used in the original analysis will not be exceeded during the period of extended operation (Le., cycles from plant operation will not exceed the number of design cycles). Alternatively, revise the TLAA disposition and include adequate justification.

b. Provide the AMR results for all non-Class 1 expansion joints in accordance with 10 CFR 54.21 (a)(1) or justify why it is not necessary to provide the AMR results.

RAJ 4.3-5b RESPONSE

a. The table below replaces the table provided in the response to RAI 4.3-5a with additional information added to address the request.

System Design Cycles Evaluation (location description)

Standby liquid control <1000 cycles of 0.125 The analysis qualifies the expansion joints for over (at tank outlet and pump inch movement 400,000 cycles of movement (0.125 inch).

inlets) Increasing the number of design cycles by a factor of 1.5 to account for the Period of Extended Operation (PEO) yields a value of <1500 cycles.

This value will remain below the analyzed number of cycles. Therefore, the TLAA remains valid for the PEO in accordance with 10 CFR 54.21(c)(1 )(i).

Standby liquid control (at 600 thermal cycles This nonsafety-related expansion joint for the test tank) 600 pressure cycles vented test tank is only filled with liquid at low temperature and it experiences no thermal or pressure transients during normal operation. The number of design cycles will not be exceeded during the PEO. Therefore, the TLAA remains valid for the PEO in accordance with 10 CFR 54.21 (c)(1 )(j).

Standby liquid control (at 1500 pressure cycles Usage factor calculated for these design cycles is pump discharge) 1500 thermal cycles 0.1. Increasing the number of cycles by a factor of 13,000 SRV actuations 1.5 to account for the PEO, the usage factor will Infinite pulsations become 0.15 which remains below the limit of 1.0.

The TLAA has been projected in accordance with 10 CFR 54.21 (c)(1 )(ii).

ECCS suction strainers (in 12,600 cycles combined As shown in LRA Table 4.3-1, the projected suppression pool) loadings from, seism ic, number of cycles for the PEO from SRV operation SRV operation and and earthquakes is significantly less than 12,600 LOCA as well as cycles. Therefore, the TLAA remains valid for the unlimited chugging PEO in accordance with 10 CFR 54.21(c)(1)(i).

cycles Standby diesel generators 1500 thermal cycles, Usage factor for these cycles was 0.04. Increasing (turbocharger water infinite number of cycles the number of these design cycles by a factor of outlets) for vibration loading 1.5 to account for the PEO, the usage factor will become 0.06 which remains below the limit of 1.0.

The TLAA has been projected in accordance with 10 CFR 54.21(c)(1 )(ii).

Standby service water 750 thermal cycles Usage factor for these cycles was 0.043.

supply to high pressure continuous vibration Increasing the number of design cycles by a factor core spray diesel of 1.5 to account for the PEG, the usage factor will become 0.065 which remains below the limit of 1.0.

Attachment to GNRO-2012/00128 Page 3 of 10 System Design Cycles Evaluation (location description)

The TLAA has been projected in accordance with 10 CFR 54.21 (c)(1 )(ii).

High pressure core spray 750 thermal cycles Usage factor for these cycles was 0.0265.

(diesel air start) continuous vibration Increasing the number of design cycles by a factor of 1.5 to account for the PEa, the usage factor will become 0.04 which remains below the limit of 1.0.

The TLAA has been projected in accordance with 10 CFR 54.21 (c)(1 )(ii).

High pressure core spray 750 thermal cycles Usage factor for these cycles was 0.011.

(diesel fuel oil) Increasing the number of design cycles by a factor of 1.5 to account for the PEa, the usage factor will become 0.017 which remains below the lim it of 1.0.

The TLAA has been projected in accordance with 10 CFR 54.21 (c)(1 )(ii).

High pressure core spray 750 thermal cycles This portion of the air intake is not heated during (non-safety related diesel vibration diesel operation. These expansion joints at the air air intake) intake are not expected to experience thermal cycles such that the number of design cycles will not be exceeded during the PEa. Therefore, the TLAA remains valid for the PEa in accordance with 10 CFR 54.2HcH1 Hi).

High pressure core spray 750 thermal cycles This isolated portion of the drain line is not heated (non-safety related diesel vibration during diesel operation. These expansion joints at cooling water drain) the drain are not expected to experience thermal cycles such that the number of design cycles will not be exceeded during the PEa. Therefore, the TLAA remains valid for the PEa in accordance with 10 CFR 54.21 (c)(1 )(i).

High pressure core spray 1200 thermal cycles The expansion joints were analyzed and (diesel exhaust) determined to be qualified for over 4000 cycles. If the design cycles are assumed to increase by a factor of 1.5 to 1800 during the PEa, they still remain below the analyzed number. Therefore, the TLAA remains valid for the PEa in accordance with 10 CFR 54.21 (c)(1)(i).

Leakage detection and 500 thermal cycles Usage factor for these cycles was 0.0051.

control system (exhaust 500 dynamic cycles Increasing the number of design cycles by a factor blower inlet) of 1.5 to account for the PEa, the usage factor will become less than 0.008 which remains below the limit of 1.0. The TLAA has been projected in accordance with 10 CFR 54.21 (c)(1 )(ii).

Compressed air (air 400 thermal Over 10,000 cycles were determined to be allowed accumulators) 300 dynamic by analysis. If the design cycles are assumed to increase by a factor of 1.5 during the PEa, the resulting cycles (600 and 450) remain below the analyzed number. Therefore, the TLAA remains valid for the PEa in accordance with 10 CFR 54.2HcH1 Hi).

Heater, Vents, and Drains 600 thermal cycles This system will be heated up as part of plant System INSR affecting SR startup. As shown in LRA Table 4.3-1, plant (a2) (at heater drain startups are limited to less than 120. Therefore, pumps) the TLAA remains valid for the PEa in accordance with 10 CFR 54.21 (c)( 1)(i).

Attachment to GNRO-2012/00128 Page 4 of 10 System Design Cycles Evaluation (location description)

Condensate and 600 thermal cycles This system will be heated up as part of plant Feedwater System INSR startup. As shown in LRA Table 4.3-1, plant affecting SR (a2) (at startups are limited to less than 120. Therefore, condensate pumps) the TLAA remains valid for the PEO in accordance with 10 CFR 54.21(c)(1)(i).

Extraction Steam System 2000 cycles This system will be heated up as part of plant I NSR affecting SR (a2) startup. As shown in LRA Table 4.3-1, plant (turbine extraction steam) startups are lim ited to less than 120. Therefore, the TLAA remains valid for the PEO in accordance with 10 CFR 54.21 (c)(1)(i).

Condensate and 3000 cycles This system will be heated up as part of plant Feedwater System INSR startup. As shown in LRA Table 4.3-1, plant affecting SR (a2) startups are limited to less than 120. Therefore, (feedpump turbine the TLAA remains valid for the PEO in accordance exhaust) with 10 CFR 54.2Hc)(1)(i).

b. As requested, the following rows are being added to the LRA to identify further information in the Metal Fatigue-TLAA section for these expansion joints or flexible connections. In addition to adding the rows for cracking-fatigue, two rows are being added for the High Pressure Core Spray (HPCS) diesel generator system to correct an omission of rows for loss of material for these expansion joints in Table 3.3.2-16.

Additions are shown with underline.

Ta bl e 3 .2. 2 - 1 R eSI'd uaI H eat RemovaISiystem NUREG-Aging Effect Aging Component Intended Requiring Management 1801 Table 1 Type Function Material Environment Management Program Item Item Notes Expansion Pressure Stainless Treated Cracking- Metal Fatigue VII.E3.A- 3.3.1-2 .Q joint boundary steel water (int) fatigue -TLAA 62 Table 3.2.2-2 Low Pressure Core Spray System NUREG-Aging Effect Aging Component Intended Requiring Management 1801 Table 1 Type Function Material Environment Management Program Item Item Notes Expansion Pressure Stainless Treated Cracking- Metal Fatigue VII.E3.A- 3.3.1-2 .Q joint boundary steel water (int) fatigue -TLAA 62 Table 3.2.2-3 High Pressure Core Spray System NUREG-Aging Effect Aging Component Intended Requiring Management 1801 Table 1 Type Function Material Environment Management Program Item Item Notes Expansion Pressure Stainless Treated Cracking- Metal Fatigue VII.E3.A- 3.3.1-2 .Q joint boundary steel water (int) fatigue -TLAA 62

Attachment to GNRO-2012/00128 Page 5 of 10 Table 3.2.2-4 Reactor Core Isolation Cooling System Aging Effect Aging NUREG-Component Intended Requiring Management 1801 Table1 Type Function Material Environment Management Program Item Item Notes Expansion Pressure Stainless Treated Cracking- Metal Fatigue VII.E3.A- 3.3.1-2 Q joint boundary steel water (int) fatigue -TLAA 62 Table 3.3.2-2 Standby Liquid Control System Aging Effect Aging NUREG-Component Intended Requiring Management 1801 Table Type Function Material Environment Management Program Item 1 Item Notes Sodium Expansion Pressure Nickel oentaborate Cracking- Metal Fatigue --- --- G joint boundary alloy fatigue -TLAA solution (int)

Expansion Pressure Stainless Sodium Cracking- Metal Fatigue --- --- .ti joint boundary steel oentaborate fatigue -TLAA solution (int)

Table 3.3.2-4 Leakage Detection and Control System Aging Effect Aging NUREG- Table Component Intended Requiring Management 1801 1 Type Function Material Environment Management Program Item Item Notes Expansion Pressure Stainless Condensation Cracking- Metal Fatigue --- --- .ti joint boundary steel tio.ll fatigue -TLAA Table 3.3.2-7 Standby Service Water System Aging Effect Aging NUREG- Table Component Intended Requiring Management 1801 1 Type Function Material Environment Management Program Item Item Notes Flexible Pressure Stainless Raw water Cracking- Metal Fatigue --- --- .ti connection boundary steel tio.ll fatigue -TLAA Table 3.3.2-11 Compressed Air System Aging Effect Aging NUREG- Table Component Intended Requiring Management 1801 1 Type Function Material Environment Management Program Item Item Notes Flexible Pressure Stainless Condensation Cracking- Metal --- --- .ti connection boundary steel tio.ll fatigue Fatigue -

TLAA Table 3.3.2-15 Standby Diesel Generator System Aging Effect Aging NUREG- Table Component Intended Requiring Management 1801 1 Type Function Material Environment Management Program Item Item Notes Expansion Pressure Nickel Treated Cracking- Metal Fatigue --- --- G joint boundary alloy water (int) fatigue -TLAA

Attachment to GNRO-2012/00128 Page 6 of 10 Table 3.3.2-16 HPCS Diesel Generator System Aging Effect Aging NUREG- Table Componen Intended Requiring Management 1801 1 Type Function Material Environment Management Program Item Item Notes Flexible Pressure Stainless ronnection boundary steel Air indoor Cracking- Metal Fatigue --- ---  !::!

.U fatigue -TLAA Flexible Pressure Stainless ronnection boundary steel Condensation Cracking- Metal Fatigue --- ---  !::!

fatigue -TLAA Flexible Pressure Stainless Condensation Loss of Compressed rvll.D.AP- 3.3.1- .Q

"'onnection boundary steel material Air Monitoring

~ 56 Flexible Pressure Stainless ronnection boundary steel Fuel oil Cracking- Metal Fatigue --.  !::!

fatigue -TLAA Flexible Pressure Stainless Fuel oil Loss of Diesel Fuel WII.H1.AP- 3.3.1- &

ronnection boundary steel material Monitoring 303

~ Ii Table 3.3.2-19-2 Standby Liquid Control System, Nonsafety-Related Components Affecting Safety-Related Systems Aging Effect Aging NUREG-Component Intended Requiring Management 1801 Table Type Function Material Environment Management Program Item litem Notes Expansion Pressure Stainless Treated Cracking- Metal Fatigue --- ---  !::!

joint boundary steel water (int) fatigue -TLAA Table 3.3.2-19-28 HPCS Diesel Generator System, Nonsafety-Related Components Affecting Safety-Related Systems Aging Effect Aging NUREG- Table Component Intended Requiring Management 1801 1 Type Function Material Environment Management Program Item Item Notes Flexible Pressure Stainless Raw water Cracking- Metal Fatigue --- ---  !::!

connection boundary steel .U fatigue -TLAA Table 3.4.2-2-10 Extraction Steam System, Nonsafety-Related Components Affecting Safety-Related Systems Aging Effect Aging NUREG- Table Component Intended Requiring Management 1801 1 Type Function Material Environment Management Program Item Item Notes Expansion Pressure Nickel Steam (int) Cracking- Metal Fatigue --- --. G joint boundary alloy fatigue -TLAA

Attachment to GNRO-2012/00128 Page 7 of 10 RAJ B.1.9-4a Background. By letter dated August 15, 2012, the applicant responded to RAI B.1.9-4 that addresses the applicant's recent operating experience related to an indication in one of the residual heat removal system to reactor pressure vessel nozzles (N06B-KB weld).

In its response, the applicant stated that due to industry concerns with cracking in dissimilar metal welds, the industry committed to an accelerated inspection program and details of this program are provided in Boiling Water Reactor Vessel and Internals Project (BWRVIP)-222, "Accelerated Inspection Program for BWRVIP-75-A Category C Dissimilar Metal Welds Containing Alloy 182", July 2009. The applicant also stated that during 2012, the nozzle discussed in RAI B.1.9-4 received extensive weld crown reduction and surface preparation which enabled the examination to detect the subject flaw.

Issue. It is not clear to the staff how the inspection schedule for Category C dissimilar metal welds was accelerated and what guidance document is used for the inspection method. In addition, the LRA does not identify the implementation of the accelerated inspections, which are specified in BWRVIP-222, as an enhancement to the existing BWR Stress Corrosion Cracking Program.

The staff is also not clear if the applicant's program includes any other weld that has limited inspection coverage as was the case with Weld N06B-KB prior to the weld crown reduction.

Reguest.

a. Describe how the inspection schedule for Category C dissimilar metal welds was accelerated. As part of the response, clarify whether the inspection schedule in the accelerated inspections meets or is more conservative than the inspection schedule that would be used in accordance with BWRVIP-75-A.

In addition, describe the inspection method and the reference of the guidance document for the inspection method.

b. Justify why the LRA does not identify the implementation of the accelerated inspection program, which is described in BWRVIP-222, as an enhancement to the existing BWR Stress Corrosion Cracking Program.

c. Clarify whether any other welds in the scope of the program have limited inspection coverage as was the case with Weld N06B-KB prior to the weld crown reduction. If so, justify why the limited inspection coverage is acceptable to manage cracking of such welds.

RAJ B.1.9-4a RESPONSE

a. The inspection schedule for Category C dissimilar metal (OM) welds was based on Section 9 of BWRVIP-222 "Accelerated Inspection Program for BWRVIP-75-A Category C Dissimilar Metal Welds Containing Alloy 182" which was issued in 2009. This guidance was based upon results of prior inspections and consideration of the effectiveness of applied stress mitigation measures against intergranular stress corrosion cracking (IGSCC) initiation and growth. This accelerated schedule was more conservative than the inspection schedule published in BWRVIP-75-A. The BWRVIP-

Attachment to GNRO-2012/00128 Page 8 of 10 222 schedule specified the Category C OM welds must be examined before June 2015.

The BWRVIP-075-A schedule required periodic inspection of a smaller percentage of the Category C OM weld population. All Category C OM welds at GGNS (34 total) were inspected by ultrasonic examination as detailed in BWRVIP-222 Section 9.

b. The accelerated inspection program based on BWRVIP-222 has been completed at GGNS. The last of the BWRVIP-222 accelerated inspections was completed during the refueling outage in 2012. There are no other welds remaining that would qualify for inspection under an accelerated inspection program. The use of BWRVIP-222 is not an enhancement to the Boiling Water Reactor (BWR) Stress Corrosion Cracking Program since no GGNS Category C OM welds will require an accelerated inspection in the period of extended operation. The inspection schedule of BWRVIP-75-A applies going forward.

c. Two Category C OM welds in the scope of the program have limited inspection coverage that do not meet the requirements of 90% or greater inspection coverage as detailed in ASME Code Case N-460. In both of these Category C OM welds, inspection coverage is at least 75% of the weld. GGNS received approval from the NRC (dated May 25, 2010) for this inspection coverage submitted in GGNS Relief Request GG-ISI-007. In the relief request response, the NRC concluded that based on the volumetric coverage obtained, it is reasonable to conclude that if significant service-induced degradation had occurred, evidence of it would have been detected by the examinations that were performed. Furthermore, the staff determined that the examinations performed provide reasonable assurance of structural integrity of the subject welds.

RAI8.1.11-6 Background. By letter dated August 15, 2012, the applicant responded to RAI B.1.9-2a that, in part, addresses the applicant's aging management for the stainless steel and nickel alloy thermal sleeves and thermal sleeve extensions of reactor vessel nozzles [recirculation inlet, core spray inlet, and residual heat removal (RHR)/Iow-pressure coolant injection (LPCI) nozzles].

In its response, the applicant amended the LRA and indicated that the BWRVIP, along with the Water Chemistry Control - BWR Program, is credited to manage cracking due to stress corrosion cracking and intergranular stress corrosion cracking in the thermal sleeves and thermal sleeve extensions of the reactor vessel nozzles. The applicant also stated the recirculation inlet, core spray inlet, and RHR/LPCI nozzles are in part, formed by the internal leg of the Y-shaped safe ends for those nozzles.

As described in Appendix C of the LRA, the applicant's response to Action Item No.5 of BWRVIP-42-A states that the BWRVIP has developed strategies to ensure the integrity of inaccessible welds [associated with the LPCI nozzle and thermal sleeve]. The applicant also stated that these strategies are included in Section 3 of BWRVIP-42, Revision 1 and it has committed to programs described as necessary in the BWRVIP reports to manage the effects of aging during the period of extended operation.

Issue. In its review, the staff noted that the applicant's AMR line items for the thermal sleeve components in LRA Table 3.1.2-1 do not include the thermal sleeve and thermal sleeve

Attachment to GNRO-2012/00128 Page 9 of 10 extension of the reactor vessel feedwater nozzle. In addition, neither the LRA or applicant's RAI response provide a clear description of the "strategies" in BWRVIP-42, Revision 1 to manage the inaccessible thermal sleeve welds of the LPCI nozzle and other reactor vessel nozzles as applicable.

The staff also needs additional information to clarify how the applicant will inspect the thermal sleeves and thermal sleeve extensions. In case inspections are not performed on the thermal sleeve components, leakage analyses are necessary to ensure that the intended functions of the thermal sleeve components are adequately maintained in a consistent manner with the guidance in Section 3.2.4 of BWRVIP-18-A.

Request.

a. Clarify why the applicant's AMR line items for the thermal sleeve components in LRA Table 3.1.2-1 do not include the thermal sleeve and thermal sleeve extension of the reactor vessel feedwater nozzle.

b. Describe the "strategies" in BWRVIP-42, Revision 1 to clarify how the implementation of the strategies will manage cracking of the inaccessible thermal sleeve components of the LPCI nozzle and other reactor vessel nozzles as applicable. Include the following in the description, as applicable.

1. If the program includes leakage analyses to manage cracking of the thermal sleeve components (e.g., for inaccessible locations), describe the results of the leakage analyses to demonstrate that cracking of the thermal sleeve components does not affect the intended functions of these components.

2. If the program includes inspections to manage cracking of the thermal sleeve components, describe the method and frequency of the inspections to demonstrate the adequacy of the inspections. As part of the response, clarify whether any of the thermal sleeve welds of the recirculation inlet, core spray, RHR/LPCI and feedwater nozzles can be examined using ultrasonic testing that is applied on the outer surface of the associated piping and safe ends.

c. Ensure that the LRA is consistent with the response.

RAI 8.1.11-6 RESPONSE

a. Thermal sleeves, which include the portions of the sleeves considered to be extensions, for the feedwater inlet nozzles use a two-stage piston ring mounted in the thermal sleeve in conjunction with an interference fit between the sleeve and safe end. The sleeves are not welded to the nozzle so they are not part of the pressure boundary. The thermal sleeves protect the feedwater nozzles from thermal oscillations over the life of the plant, but this function is not required in response to any design basis event and is not an intended function for license renewal. BWRVIP-06-A, Safety Assessment of BWR Reactor Internals, demonstrates that the feedwater sparger has no safety functions and the failure of the sparger would not significantly impact other safety-related components within the vessel. The thermal sleeve is not explicitly included in this justification.

However, the sparger is much larger than the thermal sleeve and restrains the end of the thermal sleeve so the consequences of a failure of the thermal sleeve would be less significant than the failure of the sparger. Since the thermal sleeve has no safety functions and its failure would not significantly impact safety-related components within the vessel, the thermal sleeves have no intended functions for license renewal and are not subject to aging management review. As a result no Aging Management Review (AMR) line items are included in the LRA.

Attachment to GNRO-2012/00128 Page 10 of 10

b. The strategies for managing cracking of the inaccessible thermal sleeve components of the LPGI nozzle presented in BWRVIP-42, Rev. 1, are based on BWRVIP-168, BWR Vessel and Internals Project, Guidelines for Disposition of Inaccessible Core Spray Piping Welds in BWR Internals. BWRVI P-168 was the basis for inspection requirements for inaccessible core spray welds included in BWRVIP-18, Rev. 1, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines, which is under NRG review.

BWRVIP-42, Rev. 1, includes strategies for both leakage evaluation and inspection.

Both strategies are based on the conditions of similar plant-specific accessible welds, that is, they are applied when cracking of the accessible welds is identified. Guidance on the identification of similar plant-specific accessible welds and scope expansion criteria for similar plant-specific accessible welds is included in BWRVIP-42, Rev. 1.

1. Leakage from known flaws as well as from assumed cracks in partially accessible and inaccessible welds must be evaluated to ensure that the leakage is bounded by plant-specific leakage margins. Observed cracking in similar plant-specific accessible welds is used as an indicator of the amount of leakage expected at the inaccessible welds. This leakage plus leakage from other sources must be less than the limit required to ensure the peak clad temperature remains within acceptable limits. BWRVIP-42, Rev. 1 includes guidance on calculating leakage from known flaws, and guidance on estimating leakage from assumed flaws in inaccessible welds.

2. The population of similar plant-specific accessible welds is inspected with the methods and frequency established for these welds in BWRVIP-42. The degree of cracking observed in similar plant-specific accessible welds in the LPGI coupling assembly is used as an indicator of the expected amount of cracking in the inaccessible welds. When the expected amount of cracking in the inaccessible welds reaches a specified limit, actions must be taken to inspect or repair the inaccessible weld and ensure adequate structural margins are maintained.

BWRVIP-42, Rev. 1 provides guidance for the performance of structural evaluations of known flaws and the determination of when action is required for the inspection or repair of inaccessible welds.

None of the thermal sleeve welds of the recirculation inlet, core spray or RHR/LPGI nozzles can be examined using ultrasonic testing that is applied on the outer surface of the associated piping and safe ends. As indicated in part a above, the feedwater thermal sleeves are not welded to the nozzles nor are they subject to aging management review.

c. The LRA requires no changes to reflect these responses.