ML093650086
| ML093650086 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/07/2010 |
| From: | Chernoff H Plant Licensing Branch 1 |
| To: | Joyce T Public Service Enterprise Group |
| Ennis R, NRR/DORL, 415-1420 | |
| References | |
| TAC ME1270, TAC ME1271, TAC ME1272 | |
| Download: ML093650086 (7) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 7,2010 Mr. Thomas Joyce President and Chief Nuclear Officer PSEG Nuclear P.O. Box 236, N09 Hancocks Bridge, NJ 08038
SUBJECT:
SAFETY EVALUATION OF RELIEF REQUESTS FOR THE THIRD 10-YEAR INTERVAL OF THE INSERVICE INSPECTION PROGRAM FOR SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 (TAC NOS. ME1270, ME1271 AND ME1272)
Dear Mr. Joyce:
By letter dated May 12, 2009, PSEG Nuclear LLC (the licensee) submitted relief requests SC-13R-91 and S1-13R-92 which proposed alternatives to certain requirements specified in Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) for the inservice inspection (lSI) of components at Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. The subject relief requests are for the respective third 1O-year interval of the lSI program at Salem Unit Nos. 1 and 2. For Salem Unit No.1, the third interval began on May 19, 2001, and will end on May 20, 2011. For Salem Unit No.2, the third interval began on November 27,2003, and will end on November 27,2013.
The U.S. Nuclear Regulatory Commission staff has completed its review of the subject relief requests as documented in the enclosed Safety Evaluation (SE). Our SE concludes the following:
- 1)
With respect to relief request SC-13R-91, the proposed alternative provides reasonable assurance of the structural integrity of the subject components. Furthermore, the NRC staff also concludes that the licensee's compliance with the ASME Code requirements would result in hardship without a compensating increase in the level of quality and safety. Therefore, pursuant to Section 50.55a(a)(3)(ii) of Title 10 of the Code of Federal Regulations (10 CFR), the proposed alternative is authorized for the remainder of the respective third 1O-year lSI interval for Salem Unit Nos. 1 and 2.
- 2)
With respect to relief request S1-13R-92, the proposed alternative provides 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 third 1O-year lSI interval for Salem Unit NO.1.
All other requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including a third party review by the Authorized Nuclear Inservice Inspector.
- 1. Joyce
- 2 If you have any questions concerning this matter, please contact the Salem Project Manager, Mr. Richard Ennis, at (301) 415-1420.
Sincerely,
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4arolclK. Chernoff, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUESTS FOR THE THIRD 10-YEAR INTERVAL OF THE INSERVICE INSPECTION PROGRAM PSEG NUCLEAR LLC SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311
1.0 INTRODUCTION
By letter dated May 12, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML091410451), PSEG Nuclear LLC (PSEG or the licensee) submitted relief requests SC-13R-91 and S1-13R-92 which proposed alternatives to certain requirements specified in Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) for the inservice inspection (lSI) of components at Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. Specifically, relief request SC-13R-91 (which is applicable to Salem Unit Nos. 1 and 2) and relief request S1-13R-92 (which is applicable to Salem Unit No.1) propose alternatives to ASME Code requirements for system leakage testing conducted at or near the end of each inspection interval.
The subject relief requests are for the respective third 1O-year interval of the lSI program at Salem Unit Nos. 1 and 2. For Salem Unit No.1, the third interval began on May 19, 2001, and will end on May 20,2011. For Salem Unit No.2, the third interval began on November 27,2003, and will end on November 27,2013.
2.0 REGULATORY EVALUATION
The lSI of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(g), except where specific relief has been granted by the Nuclear Regulatory Commission (NRC or Commission) pursuant to 10 CFR 50.55a(g)(6)(i). Pursuant to 10 CFR 50.55a(a)(3), 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.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the Enclosure
- 2 preservice examination requirements, set forth in the ASIVIE 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 regulation requires that inservice examination of.components and system pressure tests conducted during the first 1O-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) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The Code of Record for the third 10-year lSI interval for Salem Unit Nos. 1 and 2 is the ASME Code,Section XI, 1998 Edition through 2000 Addenda.
3.0 TECHNICAL EVALUATION
3.1 Relief Request SC-13R-91 Licensee's Request Relief request SC-13R-91 pertains to ASME Code Class 1 reactor coolant system (RCS) double isolation vent and drain valves and the piping between the valves. The valves are normally closed during plant operation. The outboard valve is pressurized only if the inboard valve is open (or leaks by the valve seat).
ASME Code,Section XI, Table IWB-2500-1, Examination Category B-P, Item Numbers B15.51 (Piping) and B15.71 (Valves), requires that a system leakage test be conducted once each 1O-year inspection interval in accordance with the requirements of IWB-5222. IWB-5222(b) requires that the pressure-retaining boundary during the system leakage test (conducted at or near the end of each inspection interval) extend to all Class 1 components within the system boundary. To meet this requirement, the inboard isolation valve would have to be opened during the system leakage test.
As an alternative to the requirements in IWB-5222(b), the licensee has proposed to perform visual leakage examination of the affected components, with the inboard isolation valve maintained in the normally closed position. The visual examinations would be performed at or near the end of the inspection interval.
The licensee request stated, in part, that:
The requirement to extend the system leakage test boundary to the outboard valve on these vent and drain connections results in a hardship without a compensating increase in the level of quality and safety. Repositioning the inboard manual valves before and after the test will take considerable time and will result in an increase in radiological dose to plant personnel. These off-normal configurations may also contribute to the risk of delaying normal plant start-up because of the critical path time and effort required to ensure system configuration is restored.
PSEG Nuclear, LLC estimates that complying with the current IWB-5222(b) requirement would result in an accumulated dose of approximately 0.75 person
- 3 rem for Salem Unit 1 and an additional 0.75 person-rem at Salem Unit 2 during each respective outage when the end of interval pressure test would be completed.
NRC Staff Evaluation
The 1998 Edition through 2000 Addenda of ASME Code,Section XI, requires that all Class 1 components within the RCS boundary undergo a system leakage test at or near the end of each inspection interval. However, in relief request SC-13R-91, the licensee proposed an alternative to the boundary subject to test pressurization required under the Code for the RCS vent and drain lines. The valve configuration provides double-isolation of the RCS. Under normal plant operating conditions, the subject pipe segments would see RCS temperature and pressure only if leakage through an inboard isolation valve occurs. With the inboard isolation valve closed during the system leakage test (as proposed in the subject relief request), the segment of piping between an inboard and an outboard isolation valve would not get pressurized to the required test pressure during a system leakage test. In order to perform the ASME Code-required test, it would be necessary to manually open each inboard isolation valve to pressurize the corresponding pipe segment. Pressurization by this method would preclude double valve isolation of the RCS and could potentially cause safety concerns for the personnel performing the examination. In addition, since the isolation valves are located inside containment, manual actuation (opening and closing) of these valves would expose plant personnel to undue radiation exposure during conduct of such activities. Another means of pressurizing the line segment between the isolation valves to the required test pressure would be to use a hydrostatic pump; however, there are no test connections between the valves to attach a pump.
As an alternative, the licensee has proposed to visually examine (i.e., VT-2) the isolated portion of the subject segments of piping for leaks with the inboard and outboard isolation valves in the normally closed position. These examinations would indicate any evidence of past leakage during the operating cycle as well as any active leakage during the system leakage test if the inboard isolation valve leaks. In addition, the Salem Technical Specifications (TSs) for RCS operational leakage monitoring (TS 3/4.4.6.2 for Salem Unit 1 and TS 3/4.4.7.2 for Salem Unit 2) provide reasonable assurance that appropriate actions (including plant shutdown) would be taken if leakage exceeded specified limits. Based on these considerations, the NRC staff concludes that the proposed alternative provides reasonable assurance of the structural integrity of the subject components. Furthermore, since the proposed alternative is consistent with maintaining personnel radiation exposure as low as reasonably achievable, and opening of the inboard isolation valves could potentially cause safety concerns for the personnel performing the examination, the NRC staff also concludes that the licensee's compliance with the ASME Code requirements would result in hardship without a compensating increase in the level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the proposed alternative is authorized for the remainder of the respective third 1O-year lSI interval for Salem Unit Nos. 1 and 2.
- 4 3.2 Relief Request S1-13R-92 Licensee's Request Relief request S1-13R-92 pertains to portions of the ASME Code Class 1 RCS safety injection (Sl) lines as shown in Figure 1 in Attachment 2 to the licensee's application dated May 12, 2009.
ASME Code,Section XI, Table IWB-2500-1, Examination Category B-P, Item Numbers B15.51 (Piping) and B15.71 (Valves), requires that a system leakage test be conducted once each 1O-yearinspection interval in accordance with the requirements of IWB-5222. IWB-5222(a) requires that the system leakage test (conducted at or near the end of each inspection interval) be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power.
As an alternative to the requirements of IWB-5222(a), the licensee has requested that for portions of the SI systems that are continually pressurized during an operating cycle, the pressure associated with the statically-pressurized Sl systems may be used in lieu of the pressure corresponding to 100% rated reactor power. The test pressure will be 650 pounds per square inch gauge (psig), corresponding to the minimum operating pressure for the affected components.
Note, the licensee request was based on Code Case N-731. On June 2,2009, the NRC staff issued a notice in the Federal Register (74 FR 36303) which, in part, proposed a change to 10 CFR 50.55a to incorporate by reference a proposed revision (Revision 16) to Regulatory Guide (RG) 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1."
As described in the notice, the NRC concluded, in accordance with the process for review of ASME Code Cases, that each of the Code Cases listed in Table 1 of the notice are technically adequate and consistent with current NRC regulations. Table 1 includes Code Case N-731 as one of the Code Cases approved unconditionally for use. However, it is uncertain if the rule will be finalized prior to the proposed implementation of the subject relief request (i.e., Salem Unit NO.1 spring outage). As such, the NRC staff has reviewed the proposed alternatives based on the merits of the information provided by the licensee.
The licensee's request stated, in part, that:
In order to obtain the pressure corresponding to 100% rated reactor pressure a jumper (temporary connections) would have to be installed between the reactor coolant system (RCS) and the volume between the first-off check valves and accumulator isolation valves. This lineup is not allowed by Technical Specifications (all vents and drains are required to remain closed) in Mode 3, the mode the RCS would have to be in to be at the required pressure. The volume of pipe must be otherwise pressurized using hydrostatic testing pumps; this would result in excessive dose, unnecessary special test procedures and unnecessary expenditure of plant resources during the ascension to power phase following a refueling outage. Testing in accordance with IWB-5221 (a) is not required for an adequate level of quality and safety because the associated components are designed to the full pressure rating of the [reactor coolant pressure boundary (RCPB)]. Additionally, these segments are isolated from the full RCS pressure
- 5 during normal operations and are subject to ASME Code required VT-2 (Visual) inspections which are performed each refueling outage. These inspections would be sufficient to identify structural defects.
NRC Staff Evaluation
The 1998 Edition through 2000 Addenda of ASME Code,Section XI, requires that all Class 1 components within the RCS boundary undergo a system leakage test at or near the end of each inspection interval. The system leakage test is required to be performed at a test pressure not less than the nominal operating pressure of the RCS corresponding to 100% rated reactor power and shall include all Class 1 components within the RCS boundary.
As an alternative, the licensee has proposed to pressure test at a pressure associated with the statically-pressurized system of the accumulator tank in lieu of the Code-required pressure corresponding to 100% rated reactor power. The test pressure will be approximately 650 psig corresponding to the minimum operating pressure for the affected components. The NRC staff believes that the VT-2 visual examination conducted during system leakage test for the subject systems would indicate any evidence of past leakage during the operating cycle as well as any active leakage during the system leakage test at a lower leak rate to initiate further corrective action. Another mitigating factor in accepting the test pressure at system operating pressure in lieu of the Code-required test pressure is based on the fact that there is no known degradation mechanism, such as intergranular stress corrosion cracking, primary water stress corrosion cracking, or thermal fatigue that is likely to affect the welds in the subject segments. In addition, as discussed in the licensee's request, the level and pressure of the SI accumulators are continuously monitored and any leakage would be investigated and identified in accordance with station operating procedures. Based on these considerations, the NRC staff concludes that the proposed alternative provides 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 third 1O-year lSI interval for Salem Unit No.1.
4.0 CONCLUSION
The following summarizes the NRC staff conclusions based on the technical evaluation discussed above.
With respect to relief request SC-13R-91, the proposed alternative provides reasonable assurance of the structural integrity of the subject components. Furthermore, the NRC staff also concludes that the licensee's compliance with the ASME Code requirements would result in hardship without a compensating increase in the level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the proposed alternative is authorized for the remainder of the respective third 10-year lSI interval for Salem Unit Nos. 1 and 2.
With respect to relief request S1-13R-92, the proposed alternative provides 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 third 1O-year lSI interval for Salem Unit NO.1.
- 6 All other requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including a third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: P. Patnaik R. Ennis Date: January 7, 2010
ML093650086 OFFICE LPL1-2/PM LPL1-2/LA CSGB/BC LPL1-2/BC NAME REnnis ABaxter RTavlor (w/edits)
HChernoff DATE 12/31/09 115/10 1/6/10 1/7/10