ML063340294
| ML063340294 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 11/29/2006 |
| From: | Marinos E NRC/NRR/ADRO/DORL/LPLII-1 |
| To: | Christian D Virginia Electric & Power Co (VEPCO) |
| Lingam, Siva NRR/DORL 415-1564 | |
| References | |
| TAC MD2673 | |
| Download: ML063340294 (10) | |
Text
November 29, 2006 Mr. David A. Christian Senior Vice President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
SURRY POWER STATION, UNIT NO. 2 - FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL RELIEF REQUEST CMP-007 (TAC NO. MD2673)
Dear Mr. Christian:
By letter dated July 27, 2006, Virginia Electric and Power Company (VEPCO) submitted Relief Request CMP-007 for the fourth 10-year inservice inspection (ISI) interval at Surry Power Station, Unit No. 2 (Surry 2). In Relief Request CMP-007, the licensee requested relief for the inspection of several welds of Residual Heat (RHR) Exchangers 2-RH-E-1A and 2-RH-E-1B and Regenerative Heat Exchanger (RHX) 2-CH-E-3 due to excessive personnel radiation exposure and geometric examination difficulties at Surry 2. The Nuclear Regulatory Commission (NRC) staff has completed its review of this relief request, and the NRC staffs evaluation and conclusions are contained in the enclosed Safety Evaluation.
The NRC staff has determined that imposing certain American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code) requirements are a hardship without a compensating increase in quality and safety and that to impose the ASME Code requirements would be a hardship on the licensee. The NRC staff also concludes that the licensee s proposed alternative provides reasonable assurance of leak tightness of the subject RHX and RHR heat exchangers. Therefore the licensees proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the Surry 2 fourth 10-year ISI interval. All other requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Sincerely,
/RA/
Evangelos C. Marinos, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-281
Enclosure:
Safety Evaluation
cc w/encl: See next page November 29, 2006 Mr. David A. Christian Senior Vice President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
SURRY POWER STATION, UNIT NO. 2 - FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL RELIEF REQUEST CMP-007 (TAC NO. MD2673)
Dear Mr. Christian:
By letter dated July 27, 2006, Virginia Electric and Power Company (VEPCO) submitted Relief Request CMP-007 for the fourth 10-year inservice inspection (ISI) interval at Surry Power Station, Unit No. 2 (Surry 2). In Relief Request CMP-007, the licensee requested relief for the inspection of several welds of Residual Heat (RHR) Exchangers 2-RH-E-1A and 2-RH-E-1B and Regenerative Heat Exchanger (RHX) 2-CH-E-3 due to excessive personnel radiation exposure and geometric examination difficulties at Surry 2. The Nuclear Regulatory Commission (NRC) staff has completed its review of this relief request, and the NRC staffs evaluation and conclusions are contained in the enclosed Safety Evaluation.
The NRC staff has determined that imposing certain American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code) requirements are a hardship without a compensating increase in quality and safety and that to impose the ASME Code requirements would be a hardship on the licensee. The NRC staff also concludes that the licensee s proposed alternative provides reasonable assurance of leak tightness of the subject RHX and RHR heat exchangers. Therefore the licensees proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the Surry 2 fourth 10-year ISI interval. All other requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Sincerely,
/RA/
Evangelos C. Marinos, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-281
Enclosure:
Safety Evaluation cc w/encl: See next page Distribution: RidsAcrsAcnwMailCenter RidsRgn2MailCenter(EGuthrie)
Public RidsNrrPMSLingam RidsNrrCvib(MMitchell)
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RidsOgcRp RidsNrrLpl2-1(EMarinos) SLee, EDO Rgn II ADAMS ACCESSION NO: ML063340294 *memo transmitting SE dated. NRR-028 OFFICE NRR/LPL2-1/PM NRR/LPL2-1/LA NRR/CVIB/BC OGC NRR/LPL2-1/BC NAME SLingam:spl MOBrien MMitchell J. Martin EMarinos DATE 11/22 /06 11 /22 /06 9/25/06* 11/20/06 11 / 29 /06 OFFICIAL RECORD COPY
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST CMP-007 SURRY POWER STATION, UNIT NO. 2 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281
1.0 INTRODUCTION
By letter dated July 27, 2006 (Agencywide Documents Access and Management System Accession No. ML062090375), the Virginia Electric and Power Company (the licensee) submitted a relief request (Relief Request CMP-007) associated with the reactor vessel inservice inspection (ISI) of the welds for the fourth 10-year ISI interval at Surry Power Station, Unit No. 2 (Surry 2). Relief Request CMP-007 pertains to a relief for the inspection of several welds of Residual Heat (RHR) Exchangers 2-RH-E-1A and 2-RH-E-1B and Regenerative Heat Exchanger (RHX) 2-CH-E-3 due to excessive personnel radiation exposure and geometric examination difficulties at Surry 2.
2.0 REGULATORY REQUIREMENTS Inservice inspection of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code), Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as required by Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(g), except where specific relief has been granted by the Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if: (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.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The ASME Code of record for the Surry 2 Fourth 10-year ISI Program, which began on May 10, 2005, is the 1998 edition of Section XI of the ASME Code, through the 2000 addenda.
3.0 TECHNICAL EVALUATION
Request for Relief No. CMP-007 Component Identification:
RHR Exchangers (2-RH-E-1A and 2-RH-E-1B) welds and RHX (2-CH-E-3) welds:
RHR 2-RH-E-1A Weld Nos. Description ASME Code ASME Code Class Category/Item 1-A01 Head Circumferential C-A/C1.20 2 Weld 1-A02 Shell Circumferential C-A/C1.10 2 Weld 1-A05, 1-A06, 1-A07, & Reinforcing Plate C-B/C3.31 2 1-A08 Welds to Nozzle and Vessel RHR 2-RH-E-1B Weld Nos. Description ASME Code ASME Code Class Category/Item 1-B01 Head Circumferential C-A/C1.20 2 Weld 1-B02 Shell Circumferential C-A/C1.10 2 Weld 1-B05, 1-B06, 1-B07, & Reinforcing Plate C-B/C3.31 2 1-B08 Welds to Nozzle and Vessel RHX 2-CH-E-3 Weld Nos. Description ASME Code ASME Code Class Category/Item 1-04, 1-17, & 1-19 Circumferential Head B-B/B2.51 1 Welds 1-03, 1-18, & 1-22 Tubesheet-to-Shell B-B/B2.80 1 Welds 1-06, 1-08, 1-09, Nozzle-to-Vessel B-D/B3.160 1 1-11, 1-13, & 1-15 Welds 1-01, 1-21, & 1-24 Head Circumferential C-A/C1.20 2 Welds 1-02, 1-20, & 1-23 Tubesheet-to-Shell C-A/C1.30 2 Welds
ASME Code Requirements ASME Code,Section XI, Examination Categories B-B and B-D from Table IWB-2500-1 and C-A and C-B from Table IWC-2500-1 require that 100 percent volumetric or surface examinations be performed on the welds and nozzle inside radius areas.
Licensees Basis for Relief Request (As Stated)
The subject welds are shown in Figures 11, 21, and 31 for stainless steel components 2-RH-E-1A, 2-RH-E-1B and 2-CH-E-3, respectively.
The Regenerative Heat Exchanger (2-CH-E-3) provides preheat for the normal charging water flowing into the Reactor Coolant System (RCS). The Residual Heat Exchangers are designed to cool the RCS during plant shut down operations.
A feasibility study has been performed within the ASME and prepared by Westinghouse Owner's Group (WOG) project MUHP 5093, Working Group Inservice Inspection Optimization Action 97-01, ISI-03-06, BC03-338, "Technical Basis for Revision of Inspection Requirements for Regenerative and Residual Heat Exchangers," August 2004. Technical justification for eliminating the surface and volumetric inspections of the Residual and Regenerative Heat Exchangers is supported in this report. The components at Surry Power Station (i.e., 2-RH-E-1A and 1B; 2-CH-E-3) are typical of the heat exchangers described by fabrication, geometric design, inspection requirements, and geometric restrictions.
As stated in the Westinghouse report, these components were designed and installed before the imposition of the inservice inspection requirements by Section XI and are not designed for performance of ultrasonic and surface examination. The small diameter of the vessel and nozzles of the Regenerative Heat Exchanger makes a meaningful ultrasonic examination very time consuming and dose intensive. The physical limitations would substantially diminish the ability to discriminate flaw indications from geometry existing around the joint. Referring to the Residual Heat Exchangers, interference with the lower support and interference with inlet and outlet pipes leads to only partial coverage for examination of the head and shell circumferential welds.
Furthermore, these components are located in high radiation fields. The estimated personnel dose to perform interval [ASME] Code inspections on the Regenerative Heat Exchanger is 12.0 man-rem, and it is estimated that 4.5 man-rem would be required to meet the inspection requirements per interval for the Residual Heat Exchanger. In view of the significant dose required to be expended for limited examination providing questionable results, the value of performing the [ASME] Code required examinations is minimal.
- 1. Figures 1, 2, and 3 are not included in this safety evaluation (SE) and can be found in the licensees letter dated July 27, 2006.
Two other factors presented in the Westinghouse report for these components were considered by the ASME committee - flaw tolerance and risk assessment.
Fracture evaluations were performed for the components using finite element models and fracture calculations. It was concluded that the heat exchangers have a large flaw tolerance and that significant leakage would be expected long before any failure occurred. Fatigue crack growth was determined to be extremely slow even in the most highly stressed region. Thus, detailed inspections are not required to ensure heat exchanger integrity.
A risk evaluation was performed using the accepted methodology applied for Risk Informed ISI piping inspection programs. The following conclusions were made:
- Safety equipment required to respond to the potential event is unaffected.
- Potential for loss of pressure boundary integrity is negligible.
- No safety analysis margins are changed.
- Leakage before full break is expected (no core damage consequences associated with leakage).
Thus, elimination of the subject inspections would not be expected to result in a significant increase in risk.
There have been no through-wall leaks on these components or components of similar design as reported in industry and as discussed in the Westinghouse report. The only related leak in the United States occurred in January 2004 at San Onofre Unit 3 on the letdown line exiting the Regenerative Heat Exchanger.
This failure was caused by excessive vibration on the piping line and is not an indication of failure on the actual heat exchanger.
All of these welds and the nozzle inner radius section have received some type of nondestructive examination during inservice or preservice inspection. The pressure retaining welds on the Regenerative Heat Exchanger received preservice volumetric examinations as outlined in the attached table2. Since the preservice [examinations], visual VT-2 examinations have been performed in accordance with NRC approved relief requests. Some examinations were limited in coverage but these limitations would again create reduced coverage today.
Licensees Proposed Alternative Examination (As Stated)
In accordance with the provisions of 10 CFR 50.55a(a)(3)(ii), approval is requested to use an alternative to the requirements in Table IWB 2500-1 for Categories B-B and B-D pertaining to the Regenerative Heat Exchanger and Table IWC 2500-1 for Categories C-A and C-B pertaining to the Residual and
- 2. The table reference in the licensees basis for relief is not included in this SE and can be found in the licensees letter dated July 27, 2006.
Regenerative Heat Exchangers. Complying with the [ASME Code] required examination would result in hardship without a compensating increase in quality and safety due to excessive personnel radiation exposure and geometric examination difficulties. Specifically, a VT-2 examination will be performed as an acceptable alternative to the [ASME] Code required examination.
NRC Staffs Evaluation The ASME Code,Section XI, Examination Categories B-B and B-D from Table IWB-2500-1, and C-A and C-B from Table IWC-2500-1 require a 100 percent volumetric or surface examination of the weld length of the RHR 2-RH-E-1A and 2-RH-E-1B, and RHX 1-CH-E-3 welds and nozzle inside radius areas. The licensee proposed an alternative to the ASME Code requirements for ASME Code,Section XI, Table IWB-2500-1, Categories B-B and B-D pertaining to the RHX and Table IWC 2500-1 for Categories C-A and C-B pertaining to the RHX and RHR heat exchangers. The licensee noted that complying with the ASME Code-required examination would result in a hardship without a compensating increase in quality and safety due to excessive personnel radiation exposure and geometric examination difficulties. A VT-2 visual examination was proposed by the licensee in lieu of the ASME Code requirements. The RHR heat exchanger and RHX shells are made of American Society for Testing and Materials (ASTM) A213 TP 304 stainless steel3 and the nozzles are 3-inch schedule 160 piping of similar material.
The physical limitations would substantially diminish the ability to discriminate flaw indications from geometry existing around the joint. Regarding the RHR heat exchangers, interference with the lower support and interference with inlet and outlet pipes would permit only partial coverage for examination of the head and shell circumferential welds. For example4, the joint design of the nozzle weld joined to a 9.25-inch outside diameter (OD) x 0.675-inch thick vessel configuration of the weldolet precludes axial ultrasonic (UT) examination in either direction.
This limits the UT examination to a single axial scan from the vessel side of the nozzle.
The Westinghouse Owners Group (WOG) report noted that the RHR heat exchangers and RHXs were designed and installed before ISI requirements by ASME Code,Section XI were required to be implemented by industry. As a result, the design of the RHR heat exchangers and RHXs have limited ultrasonic and surface examinations. The small diameter of the vessels and nozzles of the heat exchangers made it difficult to perform UT and surface examinations and limited meaningful results. The examinations are very time consuming and result in high dose rates to the personnel and technicians preparing and performing the examinations since these components are located in high radiation fields. The estimated personnel dose to perform 10-year ISI interval ASME Code inspections on the RHR heat exchangers is 12.0 man-rem. The licencee estimated that 4.5 man-rem would be required to meet the inspection requirements per 10-year ISI interval for the RHX. Therefore, considering the limited examinations providing questionable results and dose expended, the value of performing the ASME Code-required examinations is minimal.
- 3. Information on material was obtained from a licensee letter dated August 25, 2003 (ML032471647).
- 4. Example of the joint design was obtained from a licensee letter dated August 25, 2003.
Two other factors presented in the WOG report for these components were considered by the ASME committee: flaw tolerance and risk assessment. Fracture evaluations were performed for the components using finite element models and fracture calculations. It was concluded that the heat exchangers have a large flaw tolerance and that significant leakage would be expected long before any failure occurred. Fatigue crack growth was determined to be extremely slow even in the most highly stressed region. The WOG report concluded that detailed inspections are not required to ensure heat exchanger integrity.
In the WOG report, a risk evaluation was performed using the accepted methodology applied for risk informed ISI piping inspection programs. The following conclusions were made in the WOG report:
- Safety equipment required to respond to the potential event is unaffected.
- Potential for loss of pressure boundary integrity is negligible.
- No safety analysis margins are changed.
- Leakage before full break is expected (no core damage consequences associated with leakage).
The WOG report further concluded that elimination of the subject inspections would not be expected to result in a significant increase in risk.
The licensee noted that there have been no industry reports of through-wall leaks in these components or components of similar design as discussed in the WOG report.
Because of these inspection difficulties, there is a history of licensees requesting relief for these ASME Code requirements for the RHX and RHR heat exchangers. These relief requests have similarly described the above difficulties associated in performing the ASME Code-required examinations. Although these relief requests are plant-specific, the basis of these reliefs all include common concerns with metallurgical properties, geometry, accessibility, and significant radiation exposure. For the RHR heat exchangers, over 60 percent of licensee reliefs cite significant radiation exposure as one of the principle reasons for requesting relief from the ASME Code-required examinations. For the RHX, 90 percent of licensees cite geometry and accessability as the basis for relief.
In anticipation of proposed ASME Code Case N-706, Alternative Examination Requirements of Table IWC-2500-1 for PWR Stainless Steel Residual and Regenerative Heat Exchanger, Division 1, the NRC staff contracted Pacific Northwest National Laboratory (PNNL) to perform a study5 regarding the issues of inspection and the value of continued volumetric and/or surface examinations of pressure-retaining shell welds from the exterior surface of the RHX and RHR heat exchangers. ASME Code Case N-706 allows the licensee to perform VT-2 visual examinations in lieu of volumetric or surface examinations and is based on the Westinghouse topical report cited in the licensees submittal.
- 5. The study titled Assessment of ASME Code Examinations on RHX and RHR by PNNL is not included in this report.
In the study PNNL concluded that the Westinghouse evaluations agree with the PNNL risk evaluation to the extent that both predict relatively small contributions to risk (Core Damage Frequency of 2.38E-10 or less). The PNNL study further concluded that with respect to the RHX, the failure probability for the RHX shell is low; however, some nozzle locations may be sensitive to thermal fatigue and a higher failure potential would be expected. The consequence of failure for the RHX is moderate (conditional core damage probability between 1E-6 and 1E-4). Ordinarily the resulting risk categorizations would suggest that some selected RHX nozzles be considered for inspection. However, the radiation burden associated with the nozzles is very high. The thermal fatigue loading was accounted for in the design of the RHX nozzles and utilities monitor the number of occurrences of letdown and charging design thermal transients. In addition, the number and magnitude of the actual thermal transient events seen by these nozzles is less than the conservative values assumed in the design analysis. In this light, the challenges to component integrity and inspection benefits (reductions in failure probability and risk) do not appear to offset the high radiation burden associated with performing volumetric examinations of these components.
The NRC staff determined, based on the PNNL study, that the licensees basis for relief was acceptable. To impose the ASME Code requirements would be a hardship without a compensating increase in quality and safety and the licensees proposed alternative provides reasonable assurance of leak tightness of the subject RHX and RHR heat exchangers.
4.0 CONCLUSION
S For Relief Request CMP-007, the NRC staff concludes that ASME Code requirements are a hardship without a compensating increase in quality and safety and that to impose the ASME Code requirements would be a hardship on the licensee. The NRC staff also concludes that the licensee s proposed alternative provides reasonable assurance of leak tightness of the subject RHX and RHR heat exchangers. Therefore the licensees proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the Surry 2 fourth 10-year ISI interval. All other requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: T. McLellan Date:
Surry Power Station, Units 1 & 2 cc:
Ms. Lillian M. Cuoco, Esq. Office of the Attorney General Senior Counsel Commonwealth of Virginia Dominion Resources Services, Inc. 900 East Main Street Building 475, 5th Floor Richmond, Virginia 23219 Rope Ferry Road Waterford, Connecticut 06385 Mr. Chris L. Funderburk, Director Nuclear Licensing & Operations Support Mr. Donald E. Jernigan Dominion Resources Services, Inc.
Site Vice President Innsbrook Technical Center Surry Power Station 5000 Dominion Blvd.
Virginia Electric and Power Company Glen Allen, Virginia 23060-6711 5570 Hog Island Road Surry, Virginia 23883-0315 Senior Resident Inspector Surry Power Station U. S. Nuclear Regulatory Commission 5850 Hog Island Road Surry, Virginia 23883 Chairman Board of Supervisors of Surry County Surry County Courthouse Surry, Virginia 23683 Dr. W. T. Lough Virginia State Corporation Commission Division of Energy Regulation Post Office Box 1197 Richmond, Virginia 23218 Dr. Robert B. Stroube, MD, MPH State Health Commissioner Office of the Commissioner Virginia Department of Health Post Office Box 2448 Richmond, Virginia 23218