ML20195C427
| ML20195C427 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 11/06/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20195C414 | List: |
| References | |
| NUDOCS 9811170110 | |
| Download: ML20195C427 (7) | |
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UNITED STATES g
j NUCLEAR REGULATORY COMMISSION p
2 WASHINGTON, D.C. 20666-0001 4, Y SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING CORE SUPPORT PLATE WEDGE MODIFICATION j
FOR OYSTER CREEK NUCLEAR GENERATING STATION 17R OUTAGE DOCKET NO. 50-219
1.0 INTRODUCTION
By letters dated December 27,1996, and May 14,1998, respectively, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) submitted for NRC staff review the Electric Power Research Institute (EPRI) proprietary Reports TR-107284, "BWR Vessel and Internals Project, BWR Core Plate inspection and Flaw Evaluation Guidelines (BWRVIP-25)," December 1996, and TR-108722, "BWR Vessel and Internals Project, Top Guide / Core Plate Repair Design Criteria (BWRVIP-50)," May 1998. The BWRVIP-25 report was supplemented by a letter dated December 19,1997, in response to the staff's request for additional information (RAl), dated March 14,1997. The BWRVIP-25 report provides design information on the core plate geometries, weld locations, and potential failure locations for the several categories of boiling water reactors (BWR/2 through BWR/6). The BWRVIP-50 report provides the general design acceptance criteria for permanent or temporary repair of the BWR top guide or core plate.
By letter dated August 25,1998, as supplemented by letter dated September 14,1998, GPU Nuclear, Inc., (GPU or the licensee) submitted a proposed modification for the Oyster Creek Nuclear Generating Station (OCNGS) core support plate wedge during the OCNGS 17R outage for NRC staff review and approval. The licensee intends to use the BWRVIP-25 and BWRVIP-50 guidelines in the performance of core plate examinations, evaluations and modifications during the 17R outage.
The NRC staffs reviews of the BWRVIP-25 and -50 reports have not been completed at this time. The NRC staff is currently working on resolution of some outstanding issues with the BWRVIP on these documents. However, based on the staff's review of those documents to date, the staff has determined that these issues should not prevent licensees from following the guidance in the BWRVIP-25 and -50 reports with the understanding that the NRC staff has not issued final approval of the respective guidances. If concerns are found during the review of the BWRVIP-25 and -50 reports, and a licensee follows the BWRVIP-25 and -50 guidance, the NRC staff may request that the licensee also address these concerns from a plant-specific basis.
9811170110 981106 PDR ADOCK 05000219 P
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Enclosure
2.0 EVALUATION
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Background===
The rim holddown bolts of the core plate assembly are subjected to potential degradation due to intergranular stress corrosion cracking (IGSCC). Therefore, they require periodic inspection to assure their structuralintegrity. In accordance with the BWRVIP-25 report, GPU requested to implement a core plate wedge modification at OCNGS, as described in the BWRVIP-50 report.
The modification addresses the potentially degraded conditions of the core plate rim holddown bolts which affect the core plate !ateral alignment, as defined in the BWRVIP-25 report.
- Consistent with the option in BWRVIP-25, the licensee elected to perform the modification, in lieu of conducting an inspection of the holddown bolts. The modification is considered an altemative, pursuant to Section XI of the ASME Code, and as such, staff approval of the design change is required in accordance with 10 CFR 50.55a(a)(3)(i). The design and analysis of the wedge modification is in accordance with all pertinent requirements stated in the BWRVIP-50 report and in Appendix A thereto.
The modification involves the installation of eight wedge assemblies in the annulus between the core plate and the shroud, located adjacent to eight of the shield angles attached to the shroud.
The wedge assemblies are fabricated from solution annealed 300 series stainless steel and X-750 alloy steel. The wedges essentially replace the function of the rim holddown bolts and limit the lateral displacement of the core plate, by transmitting the lateral fuel load from the core plate directly to the shroud. The licensee has also stated that the modification will accommodate all reactor design conditions and fuel configurations currently specified in the plant licensing design basis, and that the installation will not require any modifications or alterations to existing reactor intemals The design of the wedge assemblies and their location in a region of low-flow regime is such as to preclude loose parts in case of wedge assembly failure.
The wedges replace structurally the lateral load resistance provided by the bolts and render bolt inspection unnecessary. The BWRVIP-25 report corcludes that the addition of the wedges provides the necessary lateral support to the core to prevent misalignment that could hinder insertion of the control rods, even if all rim holddown bolts fail. Verticallift of the core plate in this eventuality is limited by the guide tubes. The staff has completed a preliminary review of the conclusions of the BWRVIP-25 report and finds them acceptable for operation of OCNGS through fuel cycle 17.
The licensee has performed an evaluation pursuant to 10 CFR 50.59 and has determined that the proposed modification does not involve an unreviewed safety question as follows:
- 1. The installation of the core plate wedges will not increase the probability of occurrence or the consequences of an accident previously analyzed. The nore plate wedges are being installed as a proactive measure to address future potential that some core plate structural components might degrade, and to eliminate the need for inspections that would be difficult to perform. The core plate wedges are installed between the core plate and shroud and are positively locked into position. They have no moving parts and provide a redundant load path for the lateral
3-loads. As such, the wedges provide additional assurance that lateral core plate displacements will be limited to acceptable values. Therefore, the wedge installation will not increase the probability of an accident 6 occur, nor the consequences of an accident, if one does occur.
- 2. The installation of the core plate wedges does not create a possibility for an accident or malfunction of a different type than previously identified in the SAR. The core plate wedges were designed such that they meet all applicable UFSAR criteria. The core plate wedges provide an additionalload path for lateral constraint of the core plate. The wedges are fabricated from stress corrosion resistant meterial and have low applied stresses during normal operation. There is no welding in the construction or installation of the wedges. All parts are locked in place by means of mechanical devices. Installation and inspection procedures will ensure proper installation of the wedges. As such, the possibility of a different type of accident or malfunction is not created. Functions of other safety related systems are not affected.
- 3. The installation of the core plate wedges will not decrease the margin of safety as defined in the bases of any Technical Specification. The Technical Specifications and their bases do not address or discuss the core plate or wedges and are not affected by the installation of the wedges. No safety analysis referenced in the bases will change. No design allowable or licensed acceptance limit for the plant will be exceeded as a result of this modification.
Accordingly, GPU has concluded that a license amendment is not required to effectuate this modification. The staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied.
2.1 Materials and Inspection Evaluation 2.1.1 Materials and Fabrication The licensee stated that the materials specified for use in the wedge modification assemblies are fabricated from solution annealed 300 series stainless steel and Alloy X-750, are resistant to stress corrosion cracking, and have been used successfully in the BWR reactor coolant system environment. Further, no welding is permitted in the fabrication or installation of the modification and special controls and process qualifications are imposed in the fabrication of the modification to assure acceptable material surface conditions after machining.
The stat a ruewed the OCNGS core plate wedge modification materials specified and the fabrication processes described in the submittal, and concludes that they are reasonable and in accordance with current industry practice.
2.1.2 Pre-Modification and Post-Modification Inspections The licensee describes three (Pre-Modification, Post-Modification, and Subsequent Outages) inspections that are to be performed in support of the proposed modification. These inspections are summarized below:
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1.
Pre-Modification inspection: Visual examinations will be performed prior to installation to confirm that the area around each installation site is free of obstructions and debris 1
and that the core plate and shroud are structurally sound. In addition, the as-built gaps between the core plate and shroud will be measured to permit final wedge machining to obtain an installation gap of 0.02 to 0.03 inches.
2.
Post-Modification Inspection: Prior to vessel reassembly, visual inspections will be performed to confirm proper installation of each wedge.
3.
During Subsequent Refueling Outages: During the next refueling outage, a baseline visual inspection (VT-3) will be performed on accessible areas of the wedges to confirm j
the post-modification inspections. During subsequent outages, inspections are I
expected to be limited in scope.
The staff has reviewed the proposed inspection scope described in the submittal, and finds that, for the pre-and post-modification inspections, the licensee should perform an " enhanced VT-1" visualinspection per the latest BWRVIP-03 guidelines. The subsequent refueling outages inspections should also be performed in accordance with the latest (i.e., as of the date of the inspection) BWRVIP-50 guidelines. The staff concludes that, with the above caveats, that the licensee's proposed inspections are reasonable and meet or exceed current industry practice.
2.2 Core Performance Evaluation The licensee provided MPR-1957 Rev. O, which documents the design of the Oyster Creek core plate wedge modification. Section 6 of the MPR-1957 report, System Evaluations, discussed the potential core plate horizontal displacement and core bypass flow with the core plate wedges installed.
The licensee evaluated the maximum transient and permanent core plate displacement during a seismic event with wedges installed in Section 5 of the MPR-1957 report. The maximum i
calculated transient displacement of the fuellcore plate is limited to 0.529 inch and the permanent core plate displacement is 0.375 inch. The staff notes that the displacement results of the seismic analyses compare well with the GE test data which are discussed in the GE proprietary report GENE-771-44-0894 Rev. 2. GE performed full scale tests to determine allowable horizontal displacements of the top guide and core plate using actual control rod and fuel channel in as-designed configurations. The testing was performed to determine the consequences to control rod scram times during a seismic event with fuel channel displacements. Displacements greater than the maximum allowable amount established by GE would inhibit control rod insertion. The licensee's calculated transient and permanent i
displacements of the core plate with the wedges installed are within the allowable displacements calculated by GE.
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The licensee evaluated the maximum leakage path and maximum amount of leakage through a postulated fully cracked vertical weld in the H5/H6A shroud segment during normal operating conditions. The maximum leakage path was estimated to be 0.495 in. This flow area was then 2
used to evaluate the maximum coolant flow that could bypass the core. The licensee estimated that the total leakage from this flow area was approximately 30 gpm or 0.02% of the total core flow. The staff agrees with the licensee that the total postulated leakage through a vertical weld in the H5/H6A shroud segment is insignificant and will not affect the performance of the Emergency Cors Cooling System.
2.3 Load and Stress Analysis The shroud was evaluated with the installed wedge modification under differential pressure loads, radial restraint loads, seismic fuel loads, and lateral loads due to a recirculation line break. These loads were combined in accordance with the plant licensing basis specification.
The Service Level D load combinations included a combination of Safe Shutdown Earthquake (SSE) and Main Steam line Loss of Coolant Accident (LOCA) loads, and another combination of SSE and Recirculation Line (RL) LOCA loads.
The loads on the shroud were evaluated for the intact condition and flawed conditions of the shroud. Two flawed conditions were evaluated (1) the case where all circumferential welds and vertical welds in the H5/H6A shroud section were assumed completely failed, and (2) the case where only the vertical welds in H5/H6A were assumed to fail and the circumferential welds remained intact.
The licensee developed finite element models of the shroud with various boundary conditions to reflect the state of the welds. The models included the effects of previously installed tie-rod restraints. The licensee evaluated the loads in the wedges and radial restraints (bumpers), and i
the stresses in the shroud under various loading conditions and weld integrity assumptions.
These were evaluate d on an elastic basis, both statically and dynamically, using the public domain program ANSYS.
The core plate assembly was also modeled, including the top plate, the outer ring, the stiffener beams and the stabilizer bar components. The modelincluded the wedge restraints between the core plate assembly and the shroud. The core plate assembly was modeled as a full 360' model, since the assembly is not symmetric, and the wedge assemblies were not located symmetrically around the circumference. These components were also evaluated on an elastic basis both statically and dynamically using the program ANSYS.
The staff has reviewed the load combinations, modeling, boundary conditions, and the method of analysis of the core plate assembly and shroud with the installed wedge assemblies, and j
concludes that they are reasonable and in accordance with current industry practice.
I The licensee also demonstrated that the highest stresses in the shroud, core plate assembly, and the wedge assemblies determined under the relevant ASME Service Levels A, B and D i-load combinations meet the requirements of Subsection NB of the ASME Boiler and Pressure Vessel Code,1989 Edition. The maximum load in the highest loaded wedge assembly, and the
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largest shroud stresses at the wedge assembly locations, were shown to be substantially lower than the Service Level D allowable stress under the specified weld boundary conditions and load combinations. The highest stressed component in the core plate assembly under the SSE+RL LOCA load combination was shown to have an adequate margin with respect to the Service Level D allowable (safety factor = 1.18).
The staff finds that the licensee has demonstrated an adequate margin in the shroud and the core plate assembly with the installed wedge assemblies under all Service load combinations.
The staff finds the replacement of the rim holddown bolts with the wedge assemblies i
acceptable.
3.0 CONCLUSION
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The NRC staff concludes that the proposed OCNGS modification, to install core support plate wedges to structurally replace the lateral resistance provided by the rim hold down bolts, is acceptable for one operating cycle. The staff conclusion is based upon their preliminary review of the BWRVIP-25 and BWRVIP-50 reports to date. Resolution of any remaining issues with respect to the BWRVIP-25 and -50 reports will be discussed with the licensee when they are l
identified. Based on the sbove, the NRC staff has concluded that there is reasonable l
assurance that plant operation in this matter, for one operating cycle, poses no undue risk to the health and safety of the public.
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Based on the above evaluation and pursuant to 10 CFR 50.55a(a)(3)(i), the staff finds that the requested action is an acceptable attemative to Section XI of the ASME Boiler and Pressure l
Vessel Code and provides an acceptable level of quality and safety.
l The NRC staff is continuing its review of the BWRVIP-25 and -50 reports, and, while the NRC l
staff has not identified any major deficiencies in the BWRVIP's technical assessments, it has l
not yet made a determination as to their acceptability. Therefore, if concems are found during the review of the BWRVIP-25 and -50 reports, and the licensee follows the BWRVIP-25 and -50 guidance, the NRC staff may request that the licensee address these concems from a plant-specific basis.
Principal Contributor: C. E. Carpenter K.A.Kavanagh M. Hartzman Date: November 6, 1998 t
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4.0 REFERENCES
- 1. Letter dated August 25,1998, GPU Nuctear to NRC, regarding proposed modification for the Oyster Creek Nuclear Generating Station (OCNGS) core support plate wedge during the OCNGS 17R outage for NRC staff review and approval.
- 2. Letter dated December 27,1E 36, as supplemented by letter dated December 19,1997, BWRVIP to NRC, regarding staff review of the EPRI proprietary Report TR-107284, "BWR Vessel and Internals Project, BWR Core Plate Inspection and Flaw Evaluation Guideline (BWRVIP-25)," December 1996, which provides design information on the core plate geometries, weld locations, and potential failure locations for the several categories of boiling water reactors (BWR/2 through BWR/6).
- 3. Letter dated May 14,1998, BWRVIP to NRC, regarding staff review of the EPRI proprietary Report TR-108722, "BWR Vessel and Internals Project, Top Guide / Core Plate Repair Design Criteria (BWRVIP-50)," May 1998, which provides the general design acceptance criteria for permanent or temporary repair of the BWR top guide or core plate.
- 4. MPR Associates Report MPR-1957, " Design Submittal for Oyster Creek Core Plate Wedge Modification, Revision 0," August 1998.
- 5. Charnley, J.E., General Electric, " Justification of Allowable Displacements of the Core Plate and Top Guide Shroud Repair," Proprietary, November 16,1994.
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