ML083220203

Relief Request for Modification to Core Shroud Stabilizer Assemblies

ML083220203
Person / Time
Site:	Hatch
Issue date:	12/08/2008
From:	Martin R Plant Licensing Branch II
To:	Ajluni M Southern Nuclear Operating Co
Martin R, NRR/DORL, 415-1493
References
TAC MD9579
Download: ML083220203 (7)

Text

December 8, 2008 Mr. M. J. Ajluni Manager, Nuclear Licensing 40 Inverness Center Parkway PO Box 1295 Birmingham, AL 35201

SUBJECT:

EDWIN I. HATCH NUCLEAR PLANT, UNIT NO 2, REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST FOR MODIFICATION TO CORE SHROUD STABILIZER ASSEMBLIES, (TAC NO. MD9579)

Dear Mr. Stinson:

By letter dated September 3, 2008 to the U.S. Nuclear Regulatory Commission (NRC), Southern Nuclear Operating Company (SNC, the licensee) submitted a request for authorization under the provisions of Title 10 of the Code of Federal Regulations 50.55a(a)(3)(i) for modification of the core shroud stabilizer assemblies (tie rods) for the Edwin I. Hatch Nuclear Plant, Unit 2. The NRC staff reviewed the information and identified that additional information is needed to complete the review. The NRC staff request for additional information (RAI) is enclosed. We request that SNC respond to this RAI within thirty (30) days of the date of this letter.

Sincerely,

/RA/

Robert E. Martin, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-366 cc: Distribution via ListServ

Mr. M. J. Ajluni Manager, Nuclear Licensing 40 Inverness Center Parkway PO Box 1295 Birmingham, AL 35201

SUBJECT:

EDWIN I. HATCH NUCLEAR PLANT, UNIT NO 2, REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST FOR MODIFICATION TO CORE SHROUD STABILIZER ASSEMBLIES, (TAC NO. MD9579)

Dear Mr. Stinson:

By letter dated September 3, 2008 to the U.S. Nuclear Regulatory Commission (NRC), Southern Nuclear Operating Company (SNC, the licensee) submitted a request for authorization under the provisions of Title 10 of the Code of Federal Regulations 50.55a(a)(3)(i) for modification of the core shroud stabilizer assemblies (tie rods) for the Edwin I. Hatch Nuclear Plant, Unit 2. The NRC staff reviewed the information and identified that additional information is needed to complete the review. The NRC staff request for additional information (RAI) is enclosed. We request that SNC respond to this RAI within thirty (30) days of the date of this letter.

Sincerely,

/RA/

Robert E. Martin, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-366 cc: Distribution via ListServ DISTRIBUTION:

PUBLIC RidsOgcRp Resource LPL2-1 R/F RidsAcrsAcnw_MailCTR Resource RidsNrrDorlLpl2-1 Resource RidsRgn2MailCenter Resource RidsNrrLAgLappert Resource RidsNrrDeEcvib Resource RidsNrrPMRMartin Resource RidsNrrDorlDPR Resource Accession Number: ML083220203 OFFICE NRR/LPL2-1/PM NRR/LPL2-1/LA DCI/CVIB/BC NRR/LPL2-1/BC NAME RMartin GLappert MMitchell, by memo dtd MWong (LOlshan for)

DATE 12/8/08 11/18/08 11/06/08 12/8/08 OFFICIAL RECORD COPY

REQUEST FOR ADDITIONAL INFORMATION RELATED TO REQUEST FOR AUTHORIZATION FOR MODIFICATION OF THE CORE SHROUD STABILIZER ASSEMBLIES EDWIN I. HATCH NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-366 By letter dated September 3, 2008 to the U.S. Nuclear Regulatory Commission (NRC)

(Agencywide Document Access and Management System (ADAMS) Accession No. ML082490720), Southern Nuclear Operating Company (SNC, the licensee), submitted a request for authorization under the provisions of Title 10 of the Code of Federal Regulations 50.55a(a)(3)(i) for modification of the core shroud stabilizer assemblies (tie rods) for the Edwin I.

Hatch Nuclear Plant, Unit 2 (Hatch-2). The licensee proposes to replace the tie rod upper supports with a modified upper support design capable of operation through the end of the renewed operating license term. The NRC staff has reviewed the information that the licensee provided to support the proposed request and requires additional information from the licensee.

1. Section 7.2.2 of Enclosure 1 in the September 3, 2008, letter indicates that SNC will inspect the upper support arm inner and outer corner radius locations during the 2011 refueling outage and on a 10-year interval thereafter. The technique used will be VT-1, visual examination, as described in the 2001 Edition of the American Society of Mechanical Engineers (ASME) Code, Section Xl, with 2003 Addenda. The upper support is fabricated using Alloy X-750 material.

Section 2, ABackground,@ in Enclosure 1 to the September 3, 2008, letter indicates that the cause of the cracking in the upper supports in the shroud stabilizer assembly was intergranular stress corrosion cracking (IGSCC) of the Alloy X-750 material. Alloy X-750 material is susceptible to IGSCC if subjected to sustained, large peak stress conditions.

The Boiling Water Reactor Vessel and Internals Project (BWRVIP) issued letters dated March 29, 2006, and April 3, 2006, requiring plants with core shroud tie rod repairs to inspect their repairs at their next scheduled refueling outage. These letters indicated that inspections should include all the same or similar locations where indications were observed at Hatch, Unit 1 (Hatch-1) during the units 2006 refueling outage and that consideration should also be given to other locations in the tie rod repair where X-750 material is used and which may experience high-sustained stresses.

a) The licensee is requested to identify all Alloy X-750 components, excluding the replacement tie rod upper support, in the primary vertical and horizontal load paths of the core shroud stabilizer assembly.

b) The licensee is requested to identify design changes that are proposed to reduce the total stress in the Alloy X-750 components, excluding the replacement tie rod upper support, to reduce their sustained, peak stresses to a value below the IGSCC susceptibility criteria.

Enclosure

c) The licensee is requested to identify when the components identified in the response to RAI-1a were previously inspected for IGSCC and to identify the results from the inspection.

d) The licensee is requested to identify the proposed frequency of inspection for the Alloy X-750 components identified in RAI-1a and what type of inspection will be performed to ensure that potential IGSCC in these Alloy X-750 components will be promptly identified.e) The licensee is requested to explain why VT-1 examination was chosen as the inspection method for detecting IGSCC at the upper support arm inner and outer corner radius locations.

2. Please identify the water chemistry (i.e., hydrogen addition, noble metal addition, etc.)

controls that have been instituted, or will be instituted, at Hatch-2 to reduce the susceptibility of Alloy X-750 and austenitic stainless steel to IGSCC. What impact does this water chemistry control have on the susceptibility to IGSCC of the replacement tie rod upper support and tie rod top nuts?

3. In the licensees August 14, 2007, letter requesting approval of the Hatch-1 core shroud stabilizer modification, the licensee indicated that its Post-Modification Inspection Plan, Prior to RPV Assembly, would include an inspection of the support plate gusset and attachment welds. Since the integrity of the support plate gusset and attachment welds are necessary for maintaining tie rod preload, the staff believes inspection of these plates and welds is necessary. Section 7.2.1 of Enclosure 1 to the September 3, 2008, letter does not include inspection of the gusset plate welds and attachment welds. The NRC staff requests that the licensee either include the gusset plate and attachment welds in the reinspection plan or explain why reinspection is not necessary. The licensee is requested to identify its plan for reinspection of the gusset plate and attachment welds at future refueling outages.

4. Section 4.0 in Enclosure 3 to the September 3, 2008 letter indicates that several components (i.e., tie rod nut, support, etc.) in the Hatch-2 core shroud stabilizer modification will be fabricated using Alloy XM-19 austenitic stainless steel material (these components in the Hatch-1 core shroud stabilizer modification were fabricated using Alloy X-750 and Type 316 austenitic stainless steel material). Section 6.3 of Enclosure 1 in the September 3, 2008, letter indicates that surface cold work in austenitic stainless steel material was addressed by controlling machining in accordance with demonstrated procedures or solution annealing of the component subsequent to machining. For all components fabricated using austenitic stainless steel material, the licensee is requested to identify the heat treatment and surface cold work limitations for machining that were instituted to prevent IGSCC. Describe the tests performed to demonstrate that the heat treatment and surface cold work limitations will prevent IGSCC.

5. Section 5.2.1 in Enclosure 3 to the September 3, 2008 letter, indicates that no specific criterion for prevention of IGSCC in Alloy XM-19 material is specified in BWRVIP-84, BWR Vessel and Internals Project Guidelines for Selection and Use of Materials for Repair to BWR Internal Components, however, General Electric-Hitachi Nuclear Energy (GEH) has specified a maximum plastic strain limit to assure the materials are not susceptible to IGSCC. Please provide data and analyses that demonstrate that the

maximum plastic strain limit for Alloy XM-19 material will assure that the Alloy XM-19 material is not susceptible to IGSCC.

6. Section 3 of Enclosure 3 and Section 3.3 of Enclosure 1, show the replacement upper support components and contain a brief description of the components that will be repaired/replaced. Please provide a drawing (or pictorial) that illustrates the differences between the original and the proposed replacement upper components for comparison.

Also identify, dimensionally, the proposed changes to the shroud head flange.

7.a. Section 4.1.1.2 of Enclosure 1 states (in two places) that details of the GE structural analysis and results are provided in Attachment 2. Please confirm that the GE structural analysis is contained in Enclosure 3 of the application and not Attachment 2, or provide attachment 2.

7.b. Section 4.1.1.2 of Enclosure 1 also states that Structural Integrity Associates, Inc.

performed an independent third party review of the GE analysis and developed a separate ANSYS finite element analysis, the results of which compared favorably to the GE results.

Please provide a comparison of the results of the two analyses for the replacement upper support assembly parts for NRC staff review.

8.a Please describe in detail the proposed changes to the core shroud.

8.b Please clarify whether the proposed modification impacts the existing shroud stress analysis and how it has been documented.

8.c Please povide a summary of the stresses in the shroud flange due to analyzed loading conditions and due to the proposed geometry changes to the shroud head flange, along with a comparison to ASME code allowable values.

8.d Please confirm that, with the exception of the proposed modification to the shroud head flange, there are no other contact areas between the upper core shroud support assembly and core shroud that have been changed.

9. Section 5.3.1 of the stress analysis report (Enclosure 3) states that: The gap between the upper support and the shroud head flange is conservatively not included in the model. The contact between the top surface of the upper support and the shroud head flange would reduce the stresses in the upper support.

a) Please confirm that this is the vertical gap between the upper support and the shroud head flange.

b) Please verify that the statement, that the contact between the top surface of the upper support and the shroud head flange would reduce the stresses in the upper support, refers to the rotation of the upper support arms.

c) Please explain, under what conditions this gap is expected to close and why a horizontal force due to upper head shroud area thermal expansion against the upper support arms has not been considered.