ML080570109

Safety Evaluation for Reactor Core Shroud Relief Request

ML080570109
Person / Time
Site:	Hatch
Issue date:	03/26/2008
From:	Melanie Wong NRC/NRR/ADRO/DORL/LPLII-1
To:	Madison D Southern Nuclear Operating Co
Martin R, NRR/DORL, 415-1493
References
GL-94-003, TAC MD6396
Download: ML080570109 (16)

Text

March 26, 2008 Mr. Dennis R. Madison Vice President - Hatch Edwin I. Hatch Nuclear Plant 11028 Hatch Parkway North Baxley, GA 31513

SUBJECT:

EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 (HATCH-1), SAFETY EVALUATION FOR REACTOR CORE SHROUD RELIEF REQUEST (TAC NO. MD6396)

Dear Mr. Madison:

By letter to the US Nuclear Regulatory Commission (NRC) dated August 14, 2007, as supplemented on December 18, 2007, January 30 and 31, and March 7, 2008, Southern Nuclear Operating Company, Inc., submitted a request for authorization to use modified core stabilizer assemblies and to inspect the upper support arm inner and outer corner radius locations using an American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, VT-1 examination technique at the first refueling outage following installation of the modified assemblies and at 10-year intervals, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(a)(3)(i).

Based on the NRC staffs review of the information provided in the submittals listed above, the staff finds that the proposed modification and the proposed inspection of the upper support arm will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative, is authorized for the remainder of the licensed operating period at Hatch-1, which ends on August 6, 2034.

All other requirements of ASME Code,Section XI, for which relief has not been specifically requested and approved, remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

Sincerely,

/RA/

Melanie C. Wong, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-321

Enclosure:

Safety Evaluation cc w/encl: See next page

- ML080580109

• SE provided by memo dated OFFICE NRR/LPL2-1/PM NRR/LPL2-1/LA NRR/CVIB/BC OGC NLO NRR/LPL2-1/BC NAME RMartin:nc MOBrien MMitchell MBaty MWong DATE 03/24/08 03/24/08 02/19/08*

03/12/08 03/26/08

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR REACTOR CORE SHROUD RELIEF REQUEST EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

DOCKET NO. 50-321

1.0 INTRODUCTION

In its letter dated August 14, 2007, Southern Nuclear Operating Company (SNC, the licensee),

requested authorization under the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(a)(3)(i) to modify the core shroud stabilizer assemblies for the Edwin I. Hatch Nuclear Plant, Unit 1 (Hatch-1). Additional information was submitted in letters dated December 18, 2007, January 30, and 31, and March 7, 2008. The August 14, 2007, letter proposed to replace all four Hatch-1 core shroud upper support assemblies and tie rod top nuts with a modified design during the units 2008 Refueling Outage (1RFO23) due to their potential for cracking. The August 14, 2007, letter indicates that the proposed modification is being submitted to the Nuclear Regulatory Commission (NRC) for review and approval as an alternative repair pursuant to 10 CFR 50.55a(a)(3)(i). The January 31, 2008, letter indicates that the augmented inspection requirements that were discussed in the January 30, 2008, letter are being included within the scope of the licensees request to the NRC for review and approval as an alternative repair pursuant to 10 CFR 50.55a(a)(3)(i). The March, 7, 2008, letter states that two of the four upper support assemblies were replaced during outage 1RFO23 and that, based on inspection results, that demonstrate acceptability of operation during the next operating cycle, the two remaining upper support assemblies will be replaced during outage 1RFO24.

2.0 REGULATORY REQUIREMENTS The inservice inspection (ISI) of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code), Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable editions and addenda as required by the 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Section 10 CFR 50.55a(a)(3) states in part that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The licensee submitted this relief request for the staff=s review and approval to use as an alternative repair under the provisions of 10 CFR 50.55a(a)(3)(i).

3.0 TECHNICAL EVALUATION

3.1 ASME Code Requirement The core shroud tie rod assemblies are not included as a repair option under Section XI of the ASME Code. However, the licensee=s proposed alternative was designed to comply with the ASME Code,Section III, Division 1, Subsection NG, 2001 edition, and 2003 addenda and Code Case N-60-5, AMaterials for Core Support Structures,Section III, Division 1.@

3.2 Licensee=s Basis for Requesting Relief

=

Background===

Industry experience has shown that boiling-water reactor (BWR) core shroud welds are subject to intergranular stress corrosion cracking and/or irradiation-assisted stress corrosion cracking (IGSCC/IASCC). As a result, the staff issued Generic Letter (GL) 94-03, AIntergranular Stress Corrosion Cracking of Core Shrouds at Boiling Water Reactors,@ which led BWR facilities to perform detailed inspections and analyses of the core shrouds. Hatch-1 installed core shroud tie rod assemblies in 1994 on a pre-emptive basis in lieu of ultrasonic (UT) inspection of the core shroud H1 through H8 horizontal welds. The tie rods functionally replace the core shroud horizontal H1 through H10 welds. The General Electric Company (GE) designed and installed the Hatch-1 tie rod assemblies. GE provided core shroud repairs using tie rods to many other domestic BWR plants. By letters dated September 2, 1994, and September 23, 1994, SNC submitted a repair of the Hatch-1 core shroud by installation of four tie rods to the staff for review and approval. The staff, in its letter dated September 30, 1994, found the proposed repair acceptable as allowed by the requirements of 10 CFR 50.55a(a)(3)(i).

In spring 2006, cracking was discovered in two of the four tie rod upper supports at Hatch-1 during an in-vessel visual inspection (IVVI). The apparent root cause was IGSCC in the Alloy X-750 tie rod upper support material. Alloy X-750 material is susceptible to IGSCC if subjected to sustained, large peak stress conditions. GE conducted an evaluation to determine if the potential IGSCC in the Alloy X-750 tie rod structural components of other BWR core shroud repairs designed by GE could be a reportable condition under 10 CFR Part 21. GE used the criterion provided in the BWR Vessel and Internals Project (BWRVIP) report BWRVIP-84, ABWR Vessel and Internals Project, Guidelines for Selection and Use of Materials for Repairs to BWR Internals, BWRVIP-84,@ for the IGSCC susceptibility assessment of the Alloy X-750 components in the tie rod vertical load path.

Based on the Hatch-1 finding, GE revised the assessment of the GE tie rod upper support design life and determined that the potential for a high peak surface stress existed for the Hatch-1 tie rod design. The potential for high peak stress in the original tie rod upper bracket design at Hatch-1 was attributed to the lack of a specified radius at the corner junction between horizontal and vertical legs of the bracket which creates a high stress concentration. This high peak stress reduced the design life of the original tie rod upper support. Tie rod inspections performed during each refueling outage could justify continued operation on a cycle-by-cycle basis. However, the licensee has determined that the most prudent course of action is pre-emptive replacement of the tie rod upper supports with a modified upper support design capable of operation through the end of the units renewed operating license term (2034).

GE conducted an extent of condition review to determine if other Alloy X-750 tie rod components had similar potential for high peak stress. GE identified that the root radii of the threads in the tie rod threaded components of the original Hatch-1 design may be smaller than the nominal values used in previous design evaluations. Hence, the licensee proposed to include a modified tie rod nut that incorporates an improved locking mechanism. To improve IGSCC resistance, the new tie rod nuts will include a specified root radius sufficient to minimize the peak principle stress to within the same criterion as used for the upper support.

3.3 Technical Evaluation of Modified Tie Rod Assembly 3.3.1 Design Objectives The objectives of the proposed tie rod repair modifications were to design and install replacement upper support assemblies and tie rod top nuts that will remain resistant to IGSCC through the end of the units current operating license (i.e., until 2034) and to ensure that the replacement components interface correctly with the existing core shroud repair hardware.

3.3.2 Design Criteria The modified upper supports and tie rod top nuts comply with the criteria delineated in the BWRVIP-02-A, ABWR Vessel and Internals Project, BWR Core Shroud Repair Design Criteria, Revision 2,@ and BWRVIP-84 reports with no exceptions taken. The original codes and design standards used for construction of the original tie rod assemblies were delineated in GE Specification 25A5572, Revision 2, which was included in the August 14, 2007, core shroud repair submittal. The original codes and design standards remain applicable to the proposed modifications, as well as other more recent standards (e.g., BWRVIP-84 report).

3.3.3 Description of Repair Components and Design Features The geometry of the upper support replacement hardware (upper support and tie rod nut) were shown in Figures 1 and 2 of the GE stress analysis report GE-NE-0000-0061-6346-R2-P (Enclosure 2 to August 14, 2007, letter). These newly-designed components incorporate features that improve their ability to resist IGSCC. These features included: (1) a large fillet radius at the corner of the upper support arm; (2) increased width and thickness of the upper support; (3) elimination of sharp edges; and (4) a larger root radius of the tie rod nut threads.

The original tie rod installation required that cutouts be made in the shroud head flange to accommodate the upper support arms, which hang over the shroud flange. The width of the cutouts will be increased to accommodate the increased width of the modified upper supports.

The original design analysis was submitted by the licensee in letters dated September 2, 1994, and September 23, 1994. The staff, in a letter dated September 30, 1994, approved this design analysis. Subsequently, a technical concern was identified and evaluated regarding the possibility of gaps at the lower core shroud welds during normal operation. By letters dated February 20, 1996, and March 4, 1996, SNC notified the NRC that the concern would be resolved by increasing the shroud stabilizer mechanical preload by applying additional torque to the shroud stabilizer nut. In its letter dated May 30, 1996, the NRC concurred with this resolution and the adjustment was performed during the units 1996 Refueling Outage (1RFO16).

The licensee claimed in their August 14, 2007, letter that the proposed current modification has an insignificant affect on the following attributes that were evaluated in the original design analysis:

(1) Seismic model (2) Dynamic analysis (3) Bypass flow (4) Load case and load combinations (5) Shroud deflections (6) Effect of the tie rod modification on reactor pressure vessel stress analysis (7) Effect of the tie rod modification on core shroud shell, shroud head and shroud support plate (8) Loose parts considerations (9) Flow induced vibration (10) Radiation Effects 3.3.4 Structural Evaluation Finite element analysis (FEA) and/or manual calculations were used to structurally analyze the modified upper support assembly components. The original FEA of the upper support brackets used the COSMOS finite element code. The mesh size in the original model was coarse and not suitable for capturing peak stresses. A revised FEA of the replacement upper support bracket with refined mesh sizes has been performed using the ANSYS computer program.

Details of the analysis, such as input criteria, applied loading, material properties, boundary conditions, and analysis methods were described in the GE stress analysis report (Enclosure 2 to August 14, 2007, letter). The licensee also contracted Structural Integrity Associates, Inc.

(SIA) to perform an independent third-party review of the GE upper support finite element analysis. SIA developed a separate ANSYS model and their results compared favorably to the GE results for the maximum principle tensile stress. The ANSYS program is qualified for use on safety-related components.

The replacement hardware components (upper support and tie rod nut) were evaluated for their susceptibility to IGSCC. The BWRVIP-84 report limits the allowable peak stress for Alloy X-750 material to 0.8 times the minimum yield stress (Sy) at operating temperature. The design goal established by the licensee was to maintain sustained peak stress below 0.6 times Sy for all new Alloy X-750 upper support components and Alloy X-750 tie rod nuts, thereby providing margin to the criteria in the BWRVIP-84 report.

The replacement hardware components were also evaluated against ASME Code-allowable stresses. The design stress intensity (Sm) and Sy values for Alloy X-750 material were specified in accordance with Code Case N-60-5. This is consistent with the BWRVIP-84 report.

The membrane and bending stresses were calculated for these components and shown to meet the ASME Code allowable stress limits.

3.3.5 Materials and Fabrication For the proposed modification, the licensee used the following materials:

(1) Alloy X-750 -Tie rod upper support main load path bearing parts and miscellaneous smaller parts that are not in main load path (2) Alloy X-750 -Tie rod nuts (3) Alloy X-750 -Tie rod upper support dowel pins (4) Type 316 stainless steel -Tie rod upper hex nuts The above-listed materials have been used for many other reactor internal components and have demonstrated good resistance to stress corrosion cracking in laboratory testing and long-term service experience in the non-welded and low-sustained operating stress conditions.

Nickel-based Alloy X-750 and Type 316 austenitic stainless steel are acceptable per the BWRVIP-84 report and Section III of the ASME Code. The proposed materials for the replacement parts are consistent with those used in the original Hatch-1 tie rod design, which was found acceptable by the staff as documented in the SE dated September 30, 1994.

Consistent with the fabrication requirements specified in the BWRVIP-84 report, the licensee proposed to not utilize any avoidable crevices in the upper bracket design. If crevices are inherently present, the licensee proposed to implement the following requirements:

(1) the design of crevice will not have any sensitized areas and only IGSCC resistant materials will be used, and (2) to the extent practical, stagnant conditions will be minimized by venting.

The licensee stated that there are no welds in the replacement upper support assemblies, and they will be procured and processed to prevent sensitized material by meeting the requirements of the BWRVIP-84 report.

3.4 Pre-Modification Inspection Since the Hatch-1 core shroud stabilizers were recently inspected in 2006 and the core shroud horizontal welds were inspected and were found redundant to the core shroud stabilizers, only examinations necessary to support successful installation of the proposed modification will be performed at 1RFO23.

3.5 Post-Modification Inspection A post-modification inspection prior to reactor vessel (RV) reassembly will include a general post-maintenance visual inspection and recording of the fit of the core shroud support hardware onto the core shroud to confirm that there are no interferences at the support locations and that the installation is in accordance with the requirements of the modification drawings and GE Installation Specification 26A7163. The staff understands that the licensees inspection will verify, as a minimum, the following attributes:

a. all retainer clips and latches are in place for the upper spring, the mid-support, the lower spring, and the tie rod nut,

b. the upper spring, the mid-support, and the lower spring are all in contact with the RPV wall,

c. the support plate gussets have not been damaged during the operation to modify the core shroud stabilizers,

d. contact exists between the lower support clevis pin and hook and on both sides of the hook,

e. contact exists between the mid-support and shroud, and between upper support and shroud and,

f. the "as-left" inspection cleanliness is equal to or better than the "as-found" inspection.

3.6 Inspections During Subsequent Refueling Outages In the first refueling outage following installation of the modified tie rod upper supports, the licensee will inspect the tie rod assemblies in accordance with the requirements defined in report BWRVIP-76, BWR Vessel and Internals Project BWR Core Shroud Inspection and Flaw Evaluation Guidelines, Section 3.5, Option 1 or 2. The licensee will also repeat the post-modification inspections (items a through f).

In addition, the shroud support plate gussets associated with the anchorages for the shroud tie rods will be re-examined using an enhanced visual inspection (EVT-1) technique at 1RFO23, 1RFO24, 1RFO28 and 1RFO32.

3.6 Conclusion Based on the technical justification, the licensee concludes that the proposed modification provides an acceptable level of quality and safety, pursuant to the requirements specified in 10 CFR 50.55a(a)(3)(i).

4.0 STAFF EVALUATION 4.1 Safety Evaluation of the Modified Tie Rod Assembly 4.1.1 Structural and Design Evaluation The design analysis indicated that the licensee used the following aspects of the design criteria for the modification of the upper core shroud support and tie rod nuts:

(A) Maximum tensile principal stress due to sustained normal condition loads for IGSCC evaluation of Alloy X-750 materials (B) Stress intensities for the upper core shroud support for ASME Code compliance In the current modification, the licensee used a criterion for IGSCC susceptibility of Alloy X-750 material which specifies a maximum threshold stress limit that is more conservative than the existing value specified in the BWRVIP-84 report. By creating an upper core shroud support fillet radius, the maximum principal stress is reduced significantly in the upper support which ultimately reduces the susceptibility to IGSCC. The staff reviewed the maximum principal stress values for all the Alloy X-750 materials that are used in the current modification and concludes that these principal stress values comply with IGSCC design criterion.

The licensee used the ASME Code design criteria that are specified in the ASME Code,Section III, Subsection NG, Core Support Structure, 2001 edition through and including the 2003 addenda for the current core shroud tie rod modifications. All the stress intensity values for the components that are used in the current core shroud tie rod modifications complied with the ASME Code,Section III, Subsection NG design criteria. Therefore, the staff concluded that the licensee=s current core shroud tie rod modifications are structurally qualified to meet the ASME Code,Section III design criteria, and this design is consistent with the original design basis of the tie rod modification.

To determine the principle stress for the design analysis, GE used the computer code ANSYS for performing a finite element analysis (FEA). The stress analysis modeled the bearing interface of the horizontal arm of the upper support with that of the flange using contact elements with a coefficient of friction specified in Section 5.3.1 of Enclosure 2 to the August 14, 2007, letter. In the December 18, 2007, letter the licensee evaluated the impact of the coefficient of friction on the analysis. The licensee indicates:

A parametric assessment was performed using coefficient of friction values of 0.2, 0.3, and 0.4. The results of these analyses demonstrated that the change in the Pm+Pb+Q+F stress in the upper support due to the above friction coefficients is negligible (< 1%). In the case of the tie rod nut threads, the stress ratio to yield varied nominally from 0.57 to 0.64 for the above friction values. This range of stress ratios is still well within the BWRVIP-84 IGSCC threshold of 0.8. The friction factor used is a typical value for such applications and is consistent with the GE standard design specification for core support.

Based on the licensee=s response and sensitivity analysis, the staff concluded that licensee=s stress analysis has adequately modeled the interaction of the horizontal arm of the upper support and the shroud flange.

The licensees vendor, GE, used ANSYS Finite Element Computer Code, Version 10.0 to determine the total stress on safety-related components. The staff discussed the GE method of validating this computer code in a September 27, 2007, safety evaluation to M. A. Balduzzi, Entergy Nuclear Operations (ADAMS Accession Number: ML072420161). GE indicated that ANSYS has been used for several RPV internals evaluations (e.g., Clinton Power Station, Unit 1

- Core Shroud Repair, Docket No. 50-461). In addition, ANSYSs applicability and validity was demonstrated by running a series of verification cases (over 200) that exercise the elements and options used in the finite element code. The verification cases were extracted from textbooks in which classical or theoretical solutions are published.

Since GE has validated the ANSYS computer code, the staff concluded that its use for determining principle stresses in the safety analysis is acceptable.

The IGSCC analysis that was performed in Enclosure 2 to the licensees August 14, 2007, letter did not include the tie rod upper support dowel pin. Since this component was fabricated from Alloy X-750 material, it could be susceptible to IGSCC. In a letter dated January 30, 2008, the licensee evaluated the component for its susceptibility to IGSCC. The licensee concluded that the stresses on the dowel pin are significantly less than the IGSCC threshold criteria.

Since stresses on the tie rod upper support dowel pin are below the IGSCC threshold criteria, the staff concluded that they are acceptable for use in the modified tie rod assembly.

The original core shroud stabilizer design analysis in the August 2, 1994, letter discussed the impact of the core shroud stabilizer on downcomer flow. Since this was not discussed in the August 14, 2007, letter, the staff requested that the licensee evaluate the impact of the proposed modification to the core shroud stabilizer assembly design on downcomer flow.

In the January 30, 2008, letter, the licensee indicated that the proposed modification is bounded by the analysis contained in the original core shroud stabilizer design submitted in their letter dated August 2, 1994, as it relates to downcomer flow and water inventory in the downcomer.

Since the licensee has addressed the impact of the proposed modification to the core shroud stabilizer assembly design on downcomer flow, the staff concluded that this issue has been resolved.

4.1.2 Materials and Fabrication Section 6.3, "Materials Fabrication," in Enclosure 1 to the licensees August 14, 2007, letter indicates that the replacement hardware conforms specifically to the conditions described in Sections 3.5.2, 3.6.2 and 3.6.3 of the NRC's safety evaluation for the BWRVIP-84 report, dated September 6, 2005. These sections of the staffs safety evaluation describe requirements for surface preparation techniques to reduce susceptibility to IGSCC and fatigue in cold-worked austenitic stainless steel and Alloy X-750 material and to reduce the susceptibility to cracking of electrical discharged machined Alloy X-750 material.

The staff requested that the licensee provide a description of the surface preparation techniques used to reduce the susceptibility to IGSCC and fatigue in cold-worked austenitic stainless steel and Alloy X-750 material in the replacement hardware. The staff also requested that the licensee provide a description of the surface preparation techniques and qualification tests performed to reduce the susceptibility to cracking of electrical discharged machined Alloy X-750 material in the replacement hardware.

In the December 18, 2007, letter, the licensee indicated:

For those austenitic stainless steel components in the assembly, surface cold work was addressed by solution annealing of the components subsequent to machining.

For the Alloy X-750 components, surface preparation was addressed as follows:

The material was supplied in the precipitation-hardened condition, and therefore no surface preparation techniques (e.g., machining, grinding) were applied prior to age hardening.

Final machining passes were limited to 0.010.

Grinding was controlled in accordance with a written procedure, such that the process used successively finer grit with the final grit size being #120 (or finer).

For the case of electro-discharge machining of the Alloy X-750 components, the process was qualified. Qualification consisted of performing a cut using the maximum heat input parameters, removing a metallographic cross-section, and examining the cut surface. Since no fissures or cracks were found, the process was considered acceptable.

The staff found this response acceptable because the licensee has implemented heat treatment and surface preparation techniques to reduce the susceptibility to IGSCC and fatigue in cold-worked austenitic stainless steel and Alloy X-750 material and qualified the electrical discharge machining operation for Alloy X-750 material.

In the December 18, 2007, letter, licensee identified the following Alloy X-750 components, excluding the replacement tie rod upper support assembly and nuts, in the primary vertical and horizontal load paths of the core shroud stabilizer assembly:

Lower Spring Clevis Pin Upper Spring Jack Bolt (Upper Stabilizer Spring to Upper Support)

Upper Contact (Upper Stabilizer Spring to RPV Wall)

The licensee indicated that the total sustained tensile stresses in these Alloy X-750 components are less than the IGSCC initiation threshold criteria. Therefore, no design changes to these components were proposed.

Since the total sustained tensile stresses in these Alloy X-750 components are less than the IGSCC initiation threshold criteria, the staff agreed that no design changes to these components are necessary.

4.1.3 Pre-Modification and Post-Modification Inspection Since the licensee will be installing new stabilizer assembly supports and tie rod top nuts that are designed to tensile stresses below the IGSCC initiation threshold criteria, the pre-modification and the post-modification inspections proposed by the licensee are acceptable.

4.1.4 Inspections During Subsequent Refueling Outages The licensee has committed to perform inspections at the first refueling outage following the installation of the modified tie rod upper supports in accordance with the requirements defined in the staff-approved BWRVIP-76 report. Implementation of the inspection guidelines of the BWRVIP-76 report provide adequate information regarding the structural integrity of the core shroud modifications.

In the December 18, 2007, letter, the licensee indicated that the next re-inspection of the replaced hardware, as well as inspection of the other vertical and horizontal load Alloy X-750 components, will be performed following 10 additional years of operation, approximately 12 years after the upcoming 2008 outage. The peak stresses in all of the components in the vertical and horizontal load path will now be below the threshold for IGSCC. Therefore, these re-inspections will also be performed using ASME Code,Section XI, visual VT-3 techniques.

According to sub-section IWA-2213 in Section XI of the ASME Code, VT-3 techniques are adequate for determining the general mechanical and structural condition of the component. In their January 30, 2008, letter, the licensee indicated that existing flaws in the upper support arm corner radius were detected with a VT-3 examination and confirmed by an enhanced visual examination (EVT). The staff is concerned that a VT-3 examination would not detect IGSCC and requested that future examinations of the upper support arm corner radius be conducted using an examination technique that is more capable of detecting IGSCC. The licensee proposed to inspect the upper support arm inner and outer corner radius locations using a VT-1 examination technique as described in the 2001 edition of the ASME Code,Section XI with 2003 addenda at the first refueling outage following installation of the modified tire rod upper support assemblies and at 10-year intervals following the 1RFO24. The application of a VT-1 examination technique is an enhancement to the inspections required by ASME Code,Section XI for RV core support structures, such as core shrouds.Section XI of the ASME Code indicates that the VT-1 examination technique is conducted to detect cracks and provides greater resolution ability than the VT-3 examination. Since the VT-1 examination provides greater resolution than the VT-3 examination and that the VT-3 examination was able to detect the existing flaws in the upper support arm corner radius, the staff concluded that the licensees proposal to inspect the upper support arm inner and outer corner radius locations using a VT-1 examination technique is acceptable.

5.0 CONCLUSION

S The staff found that the modified core shroud stabilizer assemblies for Hatch-1 are acceptable for the following reasons:

(A) The newly-designed upper supports have a large fillet radius at the corner of the support arm and are wider and thicker with no sharp edges. The newly designed tie rod nuts have threads with a larger root radius. These upper support and tie rod nut design changes reduce the peak stress and thereby increase resistance to IGSCC.

(B) The licensee=s modified core shroud stabilizer assemblies are structurally qualified to meet the ASME Code,Section III design criteria, and this design is consistent with the original design basis for the tie rod modification.

(C) The inspection guidelines of the BWRVIP-76 report provide adequate information regarding the structural integrity of the core shroud modifications.

(D) The inclusion in the ISI program of the inspection of the upper support arm inner and outer corner radius locations using an ASME Code,Section XI, VT-1 examination technique at 10-year intervals will provide for detection of IGSCC, if it were to occur.

Based on the above discussion, the staff concludes that the modified core stabilizer assemblies and the proposed inspection of the upper support arm inner and outer corner radius locations using an ASME Code,Section XI, VT-1 examination technique at the first refueling outage following installation of the modified assemblies and at 10-year intervals will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative, is authorized for the remainder of the licensed operating period at Hatch-1, which ends on August 6, 2034.

Principal Contributor: B. Elliot Date: March 26, 2008

Edwin I. Hatch Nuclear Plant, Units 1 & 2 cc:

Mr. Dennis R. Madison Vice President - Hatch Edwin I. Hatch Nuclear 11028 Hatch Parkway North Baxley, GA 31513 Laurence Bergen Oglethorpe Power Corporation 2100 E. Exchange Place P.O. Box 1349 Tucker, GA 30085-1349 Mr. R. D. Baker Manager - Licensing Southern Nuclear Operating Company, Inc.

P.O. Box 1295 Birmingham, AL 35201-1295 Resident Inspector Plant Hatch 11030 Hatch Parkway N.

Baxley, GA 31531 Harold Reheis, Director Department of Natural Resources 205 Butler Street, SE., Suite 1252 Atlanta, GA 30334 Steven M. Jackson Senior Engineer - Power Supply Municipal Electric Authority of Georgia 1470 Riveredge Parkway, NW Atlanta, GA 30328-4684 Mr. Reece McAlister Executive Secretary Georgia Public Service Commission 244 Washington St., SW Atlanta, GA 30334 Arthur H. Domby, Esq.

Troutman Sanders Nations Bank Plaza 600 Peachtree St, NE, Suite 5200 Atlanta, GA 30308-2216 Chairman Appling County Commissioners County Courthouse Baxley, GA 31513 Mr. Jeffrey T. Gasser Executive Vice President Southern Nuclear Operating Company, Inc.

P.O. Box 1295 Birmingham, AL 35201-1295 Mr. K. Rosanski Resident Manager Oglethorpe Power Corporation Edwin I. Hatch Nuclear Plant P.O. Box 2010 Baxley, GA 31515