GNRO-2012/00155, Response to Request for Additional Information (RAI) Set 43 Dated November 30, 2012

Response to Request for Additional Information (RAI) Set 43 Dated November 30, 2012

ML12354A465
Person / Time
Site:	Grand Gulf
Issue date:	12/18/2012
From:	Kevin Mulligan Entergy Operations
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
GNRO-2012/00155
Download: ML12354A465 (11)

Text

Entergy Operations, Inc.

P. O. Box 756 Port MS 39150 Kevin J. Mulligan Vice President, Operations Grand Gulf Nuclear Station Tel. (601) 437-6409 GNRO-2012/00155 December 18,2012 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Dear Sir or Madam:

Response to Requests for Additional Information (RAI) Set 43 dated November 30, 2012 Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29 NRC Letter, lLRequests for Additional Information for the Review of the Grand Gulf Nuclear Station, License Renewal Application," dated November 30,2012 (GNRI-2012/00254) (TAC No. ME7493)

Entergy OPerations, Inc. is providing, in Attachment 1, the response to the referenced Requests for Additional Information (RAI). Attachment 2 provides additional clarification requested by the staff as a result of responses provided to RAls in letter GNRO-20121001 05, dated September 13,2012.

This letter contains no new commitments. If you have any questions or require additional information, please contact Jeffery A. Seiter at 601-437-2344.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 18th day of December, 2012.

Sincerely,

~/

KJM/jas Attachments and cc: (see next page)

GNRO-2012/00155 Page 2 of2 Attachments:

1 Response to Requests for Additional Information (RAI) 2 Additional Clarification Information cc: with Attachments 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 Attachments 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 ATTN: 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 to GNRO-2012/00155 Response to Requests for Additional Information (RAI) to GNRO-2012/00155 Page 1 of 3 The format for the Requests 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.

RAI4.3-5c Background. In response to request for additional information (RAI) 4.3-5b, part (a), dated October 22, 2012, the applicant revised the time-limited aging analysis (TLAA) disposition of certain non-Class 1 non-piping components that have cumulative usage factor (CUF) values for their fatigue design. However, the applicant did not demonstrate that the design number of cycles used in the original analysis of the expansion joints in the standby liquid control (at tank outlet and pump inlets), high pressure core spray (diesel exhaust), and compressed air (air accumulators) will not be exceeded during the period of extended operation, as requested in RAI4.3-5b.

Issue. As identified in RAI 4.3-5b, the staff noted that it is not adequate to use the "qualified cycles" as a demonstration that the analyses for these expansion joints remains valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(i). In accordance with 10 CFR 54.21 (c)(1)(i), it must be demonstrated that the "design cycles" used in the analyses for these expansion joints 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 the plant actual operation would not exceed this 1200-cycle assumption is not an adequate justification under 10 CFR 54.21 (c)(1)(i>>.

The applicant also did not revise the updated final safety analysis report (UFSAR) supplement in LRA Section A.2.2.2 indicating that, for certain non-Class 1 expansion joints, the metal fatigue TLAAs have been projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii).

Request.

a.

Supplement RAI 4.3-5b's 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 demonstrating that the number of cycles used in the design analysis will not be exceeded during the period of extended operation (Le., cycles from plant actual operation will not exceed the number of design cycles).

b.

Revise UFSAR supplement in LRA Section A.2.2.2 indicating that, for certain non-Class 1 expansion joints, the metal fatigue TLAAs have been projected to the end of the period of extended operation in accordance with10 CFR 54.21 (c)(1 )(ii).

RAI4.3-5c RESPONSE:

a.

The following provides additional information on the disposition for the expansion joints at the three locations identified in the RAI: the standby liquid control (at tank outlet and pump inlets), high pressure core spray (diesel exhaust), and compressed air (air accumulators). This additional information also clarifies that the value of to GNRO-2012/00155 Page 2 of 3 "design cycles" in the RAI 4.3-5b response is the value for cycles in the original equipment specification and is not an input to the analyses.

The analysis for expansion joints for the standby liquid control system (at tank outlet and pump inlets) calculated how many cycles of movement of magnitude 0.125 inch that the expansion joints could withstand. The analysis does not use the design value of cycles (originally 1000), but uses the magnitude of movement (0.125 inch) as an input to determine how many cycles are allowed. The analysis concluded these expansion joints were qualified for over 400,000 cycles of movement (0.125 inch). Since the 400,000 cycles is much more than the value for which the expansion joints were originally specified <<1000 "design" cycles), the expansion joints were found acceptable. The standby liquid control system expansion joints do not experience thermal transients during normal plant operating and test conditions.

Even if the original number of cycles specified (1000) is multiplied by the ratio of 60 years to 40 years (1.5), the resulting number (1500 cycles) does not exceed the acceptable qualified value (400,000 cycles) determined in the analysis. The number of cycles the analysis determined were acceptable (400,000) therefore will not be exceeded during the period of extended operation. Therefore, the analysis remains valid for the period of extended operation in accordance with 10 CFR 54.2 (c)(1)(i).

The analysis for the expansion joint for the high pressure core spray diesel exhaust determined the expansion joint was qualified for 4000 cycles of the specified movement. This 4000 cycles is determined independent of the value listed as "design cycles" in our response to RAI4.3-5b. The original specification specified that the expansion joint be able to withstand 1200 thermal cycles. Since the 4000 cycles determined by the calculation are more than originally specified 1200 "design" cycles, the expansion joints were deemed acceptable. Periodic monthly testing of the diesel through the period of extended operation will result in operating cycles less than the number of cycles the analysis determined acceptable (4000). Even if the original number of cycles specified is multiplied by the ratio of 60 years to 40 years (1.5), the resulting number (1800 cycles) does not exceed the value the analysis determined acceptable. Therefore, the analysis remains valid for the period of extended operation in accordance with 10 CFR 54.2 (c)(1)(i).

The analysis for the expansion joints for the compressed air accumulators calculated the expansion joints were qualified for over 10,000 cycles of the specified movement.

This 10,000 cycles was calculated independent of the "design cycles" listed in our response to RAI4.3-5b. The original specification identified 400 thermal and 300 dynamic cycles for the 40 years. Since the 10,000 cycles determined by the calculation are more than the expansion joints were originally specified to withstand (the 400 thermal and 300 dynamic "design" cycles), the expansion joints were determined acceptable. Even if the original number of cycles specified is multiplied by the ratio of 60 years to 40 years (1.5), the resulting number of cycles (600 and 450) does not exceed the value the analysis determined acceptable. The compressed air accumulator's expansion joints therefore will not exceed the 10,000 qualified cycles in the period of extended operation. Therefore, the analysis remains valid for the period of extended operation in accordance with 10 CFR 54.2 (c)(1)(i).

b.

The following changes are made to the LRA including section A.2.2.2 to reflect the RAI4.3-5 responses. Additions are shown with underline and deletions with strikethrough.

to GNRO-2012/00155 Page 3 of 3 4.3.2.2 Non-Piping Components Non-class 1 components other than piping system components require fatigue analyses if they were built to a section of the code such as ASME Section Ill, NC-3200 or ASME Section VIII, Division 2. A review of the non-Class 1 components other than piping identified non-Class 1 fatigue analysis applicable to expansion joints. Design specifications and calculations were identified for expansion joints with fatigue analyses for a bounding number of cycles, which were identified as time limited aging analyses. Evaluation of these certain analyses determined the number of analyzed cycles were adequate for 60 years of operation. Therefore, these non-Class 1 expansion joint TLAAs are valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(i). The analyses for other expansion joints were projected to 60 years of operation and the resulting CUF values remained less than 1.0. Therefore, these non-Class 1 expansion joint TLAAs have been projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii).

A.2.2.2 Non-Class 1 Metal Fatigue Non-class 1 components other than piping system components require fatigue analyses if they were built to a section of the code such as ASME Section III, NC-3200 or ASME Section VIII, Division 2. A review of the non-Class 1 components other than piping identified non-Class 1 fatigue analysis applicable to expansion joints. Design specifications and calculations were identified for expansion joints with fatigue analyses for a bounding number of cycles, which were identified as time limited aging analyses. Evaluation of these certain analyses determined the number of analyzed cycles were adequate for 60 years of operation. Therefore, these non-Class 1 expansion joint TLAAs are valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i). The analyses for other expansion joints were projected to 60 years of operation and the resulting CUF values remained less than 1.0. Therefore, these non-Class 1 expansion joint TLAAs have been projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii).

to GNRO-2012/00155 Additional Clarification Information to GNRO-2012/00155 Page 1 of 4 Issue: During a phone call on 11/20/2012, the Nuclear Regulatory Commission (NRC) asked if there were any components at Grand Gulf Nuclear Station (GGNS) that have been exempted/excluded from 10 CFR Part 50, Appendix J, Type 8 testing.

GGNS Response The components that have been exempted/excluded from 10 CFR Part 50, Appendix J Type 8 testing are identified and discussed in items a, b, and c below.

a)

The components listed below are inservice inspection (lSI) ports integral to the guard pipes in containment penetrations 5,6,7,8,9, 10, 14, 17, 18, 19 and 87.

821 PPN420A, lSI Inspection Port, Main Steam Line "A" - containment penetration 5 821 PPN421A, lSI Inspection Port, Main Steam Line "A" - containment penetration 5 821PPN4208, lSI Inspection Port, Main Steam Line "8" - containment penetration 6 821 PPN421 8, lSI Inspection Port, Main Steam Line "8" - containment penetration 6 821 PPN420C, lSI Inspection Port, Main Steam Line "C" - containment penetration 7 821 PPN421 C, lSI Inspection Port, Main Steam Line "C" - containment penetration 7 821 PPN420D, lSI Inspection Port, Main Steam Line "0" containment penetration 8 821 PPN421 0, lSI Inspection Port, Main Steam Line "0" - containment penetration 8 821 PPN422A, lSI Inspection Port, Feed Water "A" - containment penetration 9 821 PPN423A, lSI Inspection Port, Feed Water "A" - containment penetration 9 821 PPN4228, lSI Inspection Port, Feed Water "8" - containment penetration 10 821 PPN4238, lSI Inspection Port, Feed Water "8" - containment penetration 10 E12PPN406, lSI Inspection Port, Residual Heat Removal (RHR) Suction-containment penetration 14 E12PPN407, lSI Inspection Port, RHR Suction - containment penetration 14 to GNRO-2012/00155 Page 2 of4 E51PPN411, lSI Inspection Port, Main Steam Supply to Reactor Core Isolation Cooling (RCIC) Turbine - containment penetration 17 E51PPN412, lSI Inspection Port, Main Steam Supply to RCIC Turbine-containment penetration 17 E12PPN413, lSI Inspection Port, Abandoned RHR to Head Spray Penetration -

containment penetration 18 E12PPN414, lSI Inspection Port, Abandoned RHR to Head Spray Penetration -

containment penetration 18 E12PPN426, lSI Inspection Port, Main Steam Line Drain (Main Steam Isolation Valves [MSIV]) - containment penetration 19 E12PPN427, lSI Inspection Port, Main Steam Line Drain (MSIV) - containment penetration 19 G33PPN406A, lSI Inspection Port, Reactor Water Cleanup (RWCU) Pump Suction from Recirc Loops - containment penetration 87 G33PPN406B, lSI Inspection Port, RWCU Pump Suction from Recirc Loops -

containment penetration 87 The original configuration of these lSI ports consisted of gasketed inner and outer covers plates pulled together with two bolts. The original configuration was modified and the current configuration consists of the inner cover plate, with the gasket removed, welded to the guard pipe and the gasketed outer cover plate connected to the inner cover plate with bolts. The American Society of Mechanical Engineers (ASME) pressure boundary function is performed by the welded inner cover plate bearing against the inside surface of the guard pipe when the guard pipe is pressurized. The configuration of the lSI ports with the welded inner cover plate and guard pipe is considered part of the containment pressure boundary, but no longer meets the criteria that require 10 CFR Part 50 Appendix J, Type B testing.

These lSI ports are integral to the guard pipes and are included as part of the component "Guard piping" in license renewal application (LRA) Table 3.5.2-1. As shown in Table 3.5.2-1, the Containment Leak Rate and Containment ISI-IWE programs will manage the relevant effects of aging during the period of extended operation (PEO).

b)

Two blind flanges listed below are associated with the containment upper and lower personnel air lock's outer bulkhead, containment penetration numbers 2 and 3 respectively.

M23Y002FLG - "Blind Flange" Upper Personnel Air Lock - containment penetration 2 M23Y001 FLG - "Blind Flange" Lower Personnel Air Lock - containment penetration 3 to GNRO-2012/00155 Page 30f4 These blind flanges are tested during the overall airlock (barrel) leak rate test. These blind flanges are integral to the personnel airlocks and are included as "Containment personnel lock" in LRA Table 3.5.2-1. As shown in Table 3.5.2-1, the Containment Leak Rate and Containment ISI-IWE programs will manage the relevant effects of aging during the PEO.

c)

The components listed below are restricting orifice plates with double O-ring seals in the process piping lines associated with containment penetrations 23,24,27,32, and 67.

These process piping lines are fluid filled and terminate below minimum drawdown level of the suppression pool, thereby maintaining a water seal following a loss of coolant accident. Therefore, as identified in UFSAR Table 6.2-49, 10 CFR Part 50, Appendix J Type 8 testing is not required.

E120003A, RHR "A" Pump Test Return line to Suppression Pool-containment penetration 23 E1200038, RHR "8" Pump Test Return Line to Suppression Pool-containment penetration 67 E120003C, RHR "C" Pump Test Return Line to Suppression Pool-containment penetration 24 E210004, LPCS Test Return Line to Suppression Pool-containment penetration 32 E220005, HPCS Test Return Line to Suppression Pool - containment penetration 27 These restricting orifices are evaluated in Section 3.2 of the LRA as part of their respective piping system (i.e., Residual Heat Removal "RHR", High Pressure Core Spray "HPCS", and Low Pressure Core Spray "LPCS") aging management review.

These orifices are included in the component grouping "Orifice" in LRA Tables 3.2.2-1, 3.2.2-2 and 3.2.2-3. As shown in those tables, Water Chemistry Control-8WR program will manage the effects of aging on these orifices during the PEO.

As described in LRA Section 2.1.2.4.1, "Packing, gaskets, component mechanical seals, and O-rings are typically used to provide a leak-proof seal when components are mechanically joined together. These items are commonly found in components such as valves, pumps, heat exchangers, ventilation units or ducts, and piping segments. In accordance with American National Standards Institute (ANSI) 831.1 and the ASME 80iler and Pressure Vessel Code Section III, the subcomponents of pressure retaining components are not considered pressure-retaining parts... ". Therefore, these O-ring subcomponents are not relied on to perform a license renewal intended function and are not subject to aging management review.

to GNRO-2012/00155 Page 4 of4 Issue: During a phone call on 11/20/2012, the NRC asked if we could add additional information in LRA Section A.1.15 (Containment Leak Rate), "program description" regarding industry criteria similar to what was provided in letter GNRO-2012/00105.

GGNS Response LRA section A.1.15 and corresponding section B.1.15, "Containment Leak Rate Program" are revised as shown below. Additions are shown with underline.

A.1.15 Containment Leak Rate Program The Containment Leak Rate Program provides for detection of loss of material, cracking, and loss of function in various systems penetrating containment. The program also provides for detection of age-related degradation in material properties of gaskets, a-rings, and packing materials for the primary containment pressure boundary access points.

Containment leakage rate tests (LRT) are performed to assure that leakage through the containment and systems and components penetrating primary containment does not exceed allowable leakage limits sPecified in the plant technical specifications. An integrated leak rate test (ILRT) is performed during a period of reactor shutdown at the frequency specified in 10 CFR Part 50, Appendix J, Option B, based upon the criteria in Regulatory Guide 1.163. NEI 94-

01. and ANSI 56.8-1994. Performance of the integrated leak rate test per 10 CFR Part 50, Appendix J demonstrates the leak-tightness and structural integrity of the containment. Local leak rate tests (LLRT) are performed on isolation valves and containment access penetrations at frequencies that comply with the requirements of 10 CFR Part 50, Appendix J, Option B.

8.1.15 CONTAINMENT LEAK RATE Program Description The Containment Leak Rate Program is an existing program that provides for detection of loss of material, cracking, and loss of function in various systems penetrating containment. The program also provides for detection of age-related degradation in material properties of gaskets, a-rings, and packing materials for the primary containment pressure boundary access points.

Containment leakage rate tests (LRT) are performed to assure that leakage through the containment and systems and components penetrating primary containment does not exceed allowable leakage limits specified in the plant technical specifications. An integrated leak rate test (ILRT) is performed during a period of reactor shutdown at the frequency specified in 10 CFR Part 50, Appendix J, Option B, based upon the criteria in Regulatory Guide 1.163. NEI 94-01, and ANSI 56.8-1994. Performance of the integrated leak rate test per 10 CFR Part 50, Appendix J demonstrates the leak-tightness and structural integrity of the containment. Local leak rate tests (LLRT) are performed on isolation valves and containment access penetrations at frequencies that comply with the requirements of 10 CFR Part 50, Appendix J, Option B.