ML090080390
| ML090080390 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 02/05/2009 |
| From: | Mark Kowal Plant Licensing Branch 1 |
| To: | Entergy Nuclear Operations |
| Boska J, NRR, 301-415-2901 | |
| References | |
| RR-3-44, TAC ME0011 | |
| Download: ML090080390 (8) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 5, 2(Ul Vice President, Operations Entergy Nuclear Operations, Inc.
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NO.3 - RELIEF REQUEST (RR) NO. RR-3-44 FOR REACTOR COOLANT SYSTEM PRESSURE TEST (TAC NO. ME0011)
Dear Sir or Madam:
By letter dated October 20, 2008, Entergy Nuclear Operations, Inc. (the licensee) submitted Relief Request No. RR 3-44 to the Nuclear Regulatory Commission (NRC) on the reactor coolant system pressure test applicable to Indian Point Nuclear Generating Unit No.3 (IP3) for the third 10-year inservice inspection (lSI) interval. The relief pertains to the boundary subject to test pressurization during performance of a system leakage test conducted at or near the end of the inspection interval. The American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) requires the test to extend to all ASME Code Class 1 pressure retaining components within the system boundary. In lieu of this, the licensee has proposed an alternative to pressurize up to the normal operating system boundary which would exclude a small segment of the Class 1 pressure boundary from attaining test pressure. However, the visual examination during pressurization would include all components within the system boundary. The NRC staff finds that the licensee's proposed alternative provides reasonable assurance of structural integrity and is, therefore, acceptable.
Based on the information provided in the licensee's submittal, the NRC staff concludes that the licensee's compliance with this section of the ASME Code would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, pursuant to Title 10 of the Code of Federal Regulations, Section 50.55a(a)(3)(ii), the NRC staff authorizes the lSI program alternative proposed in RR 3-44 for the third 1O-year lSI interval at IP3, which ends in July 2009, as compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC safety evaluation is provided in the enclosure.
- 2 If you have any questions regarding this approval, please contact the Indian Point Project Manager, John Boska, at (301) 415-2901.
Sincerely, Mark G. Kowal, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286
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 REQUEST FOR RELIEF RR-3-44 ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO.3 DOCKET NO. 50-286
1.0 INTRODUCTION
By letter dated October 20, 2008, Agencywide Documents Access and Management System (ADAMS) accession number ML083030053, Entergy Nuclear Operations, Inc. (the licensee) submitted Relief Request (RR) 3-44 to the Nuclear Regulatory Commission (NRC) on the reactor coolant system (RCS) pressure test applicable to Indian Point Nuclear Generating Unit No.3 (IP3) for the third 10-Year inservice inspection (lSI) interval. The relief pertains to the boundary subject to test pressurization during performance of a system leakage test conducted at or near the end of the inspection interval. The American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, requires the test to extend to all ASME Code Class 1 pressure retaining components within the system boundary. In lieu of this, the licensee has proposed an alternative to pressurize up to the inboard isolation valve which would exclude a small segment of the Class 1 pressure boundary from attaining test pressure.
These excluded segments include ASME Code Class 1 components associated with piping connections to the RCS from the safety injection (SI) system, the residual heat removal (RHR) system, the pressurizer spray piping, and drain lines in the RCS. However, the visual examination during pressurization would include all components within the system boundary.
2.0 REGULATORY REQUIREMENTS As specified in Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g), inservice inspection of nuclear power plant components shall be performed in accordance with the requirements of ASME Code,Section XI, except where specific written relief has been granted by the 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 (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. As stated in 10 CFR 50.55a(g)(5)(iii),
if the licensee has determined that conformance with certain Code requirements is impractical for its facility, the licensee shall notify the Commission and submit information to support the determinations.
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 preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the Enclosure
- 2 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 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) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed in paragraph (b) of that section. The lSI Code of Record for the third 10-year lSI interval for IP3 is the 1989 Edition of the ASME Code,Section XI. The third 10-year lSI interval for IP3 started on July 21, 2000, and is scheduled to end on July 21,2009.
3.0 TECHNICAL EVALUATION
Components for Which Relief is Requested All ASME Code Class 1 components in the system pressure boundary between isolation valves identified in Attachment 1 to Relief Request RR 3-44.
ASME Code Requirements Table IWB-2500-1, Examination Category B-P, Item Numbers B15.51 (Piping) and B15.71 (Valves) requires that a system hydrostatic pressure test be performed once each 10-year inspection interval in accordance with the requirements of IWB-5222. Furthermore, Note 2 of Table IWB-2500-1 states that the pressure retaining boundary during the system hydrostatic test shall include all Class 1 components within the system boundary. This would require the normally closed process and drain lines and connections to be opened/bypassed and pressurized.
Licensee's Request for Relief A system leakage test shall be conducted at or near the end of each inspection interval, prior to reactor startup. The segment of Class 1 piping between an inboard and an outboard isolation valve including the valves in the system boundary for the RHR system, SI system, and RCS will be visually examined for evidence of past leakage and/or leakage during the system leakage test conducted with the isolation valves in the position required for normal reactor startup.
Licensee's Basis for Requesting Relief The Class 1 piping and connections are equipped with isolation valves (including check valves) which provide double isolation of the reactor coolant pressure boundary (RCPB). These valves are generally maintained in the closed position during normal plant operation. The piping outboard of the first isolation valve is not normally pressurized. Under normal operating conditions, the piping and connections are subject to reactor coolant system pressure and temperature only if leakage through the inboard valves occurs. To perform the Code required pressure test, it would be necessary to manually open the inboard valves or install temporary jumper hoses around check valves to pressurize the piping and connections between the two isolation valves.
- 3 The components and piping connected to the RCS, such as loop drain lines, the SI system, and the RHR system for which relief is requested are the portion of piping between an inboard and an outboard isolation valve including the valves.
The following specific lines are included in the request for relief.
(i)
Small Size Class 1 System Drain Lines (shown on lSI sketches 1-4103, 1-4209, 1-4308, and 1-4408).
(ii)
Piping segment consisting of 14 inch diameter piping between RHR inlet valves 730 and 731.
(iii)
Piping segments in the SI Loops Low Head Check Valves 897A through 8970 and Upstream piping (1-4101,1-4202,1-4301, and 1-4401).
(iv)
Piping segments in the SI Loops High Head Check Valves 8578 through 857F and 857N and Upstream piping (1-4302,1-4406,1-4207,1-4107,1-4306, and 1-4102).
(v)
Auxiliary Spray and Alternate Charging Line Check Valves 211, and 21OC and Upstream piping (1-4508, 1-4601).
In pressurizing the piping segments and the valves, to the Code-required test pressure, the licensee would be subject to hardship or unusual difficulty without a compensating increase in the level of quality and safety as stated below.
The affected components are located inside containment. Tests performed inside the radiologically restricted area increase the total exposure to plant personnel while they are modifying and restoring system lineups, as well as contaminate test equipment.
Use of single valve isolation from systems with lower design pressures could result in over-pressurization of these systems and damage to permanent plant equipment.
Use of single valve isolation is a significant personnel safety hazard.
There are no test connections for testing the piping between check valves in the RHR system and the SI system, and thus, pressurization using an external source would require significant modification to the piping segments.
The licensee proposes an alternative method for the pressurization boundary for specified Class 1 piping. The proposed method will leave the barriers intact for the visual examination rather than opening or bypassing the first isolation barrier prior to the examination. This modified approach will result in significant personnel exposure savings as well as minimizing the risk of personnel injury or contamination associated with opening or bypassing these normally closed isolation devices. Since these pressure tests are performed at the end of a refueling outage, elimination of the requirement to open or bypass these isolation devices will also minimize the impact on the outage duration.
-4 Licensee's Proposed Alternative The licensee proposes to perform a system leakage test of the Class 1 systems and components during the current 10 year inspection interval in accordance with the Code of Record with the isolation valves in the normally closed position. The RCPB piping and connections up to the first isolation valve will be visually examined with the isolation valves in the normally closed position. This examination will be performed at the nominal operating pressure associated with 100% reactor power after satisfying the Code-required hold time.
Additionally, the portion of piping between the two isolation valves will be VT-2 examined during the required inservice pressure test of the corresponding system, which will also be performed during the 2009 refueling outage.
NRC Staff Evaluation
The Code of Record, the 1989 Edition of the ASME Code,Section XI, Table IWB-2500-1, Category B-P, Item numbers B15.51 and B15.71 requires system hydrostatic testing of Class 1 pressure retaining piping and valves once per 1O-year interval. The licensee has adopted the NRC approved ASME Code Case N-498-4, "Alternative Requirement for 1O-year System Hydrostatic Testing for Class 1, 2, and 3 SystemsSection XI, Division 1," which allows a system leakage test in lieu of the system hydrostatic 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. However, in Relief Request RR 3-44, the licensee proposed an alternative to the boundary subject to test pressurization required under the Code of Record for the RCS drain lines, and the piping segments in SI and RHR systems between an inboard and an outboard isolation valve in the system boundary. The line configuration, as outlined, 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. As requested in RR 3-44, with the inboard isolation valve closed during the system leakage test, 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 may cause safety concerns for the personnel performing the examination. Alternatively, the line segments between the isolation valves could be separately pressurized to the required test pressure by a hydrostatic pump or a jumper hose, but there are no test connections between the isolation valves to attach either of these.
One factor that supports the acceptability of the licensee's proposal is that the segments of Class 1 pressure boundary between the inboard and outboard isolation valves in RHR and Sl systems that are not tested to the Code-required test pressure, would be pressure tested at the associated system's operating pressure during the RHR system inservice test and the Sl system functional test during the refueling outage. 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 (IGSCC), primary water stress corrosion cracking (PWSCC), or thermal fatigue, that is likely to affect the welds in the subject segments.
- 5 The subject isolation valves are located inside the containment, and any manual actuation (opening and closing) of these valves would expose plant personnel to undue radiation exposure during modification and restoration of system lineups. The NRC staff concurs with the licensee's finding that compliance with the Code requirement would result in hardship without a compensating increase in the level of quality and safety. The licensee, however, has proposed an alternative to visually examine (VT-2) for leaks in the isolated portion of the subject segments of piping with the inboard and outboard isolation valves in the normally closed position which 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. The NRC staff believes that the licensee's proposed alternative will provide reasonable assurance of structural integrity for the RCS drain lines and the piping segments in SI and RHR systems between an inboard and an outboard isolation valve including the valves while maintaining personnel radiation exposure to as low as reasonably achievable.
4.0 CONCLUSION
Based on the above review, the NRC staff concludes that test pressurization during the system leakage test of the Class 1 pressure retaining components within the system boundary of RCS drain lines and piping segments in SI and RHR systems between an inboard and an outboard isolation valve, including the valves, as required by the ASME Code would result in hardship to the licensee without a compensating increase in the level of quality and safety. The licensee's proposed alternative in RR 3-44 provides a reasonable assurance of structural integrity for the subject components. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the proposed alternative in RR 3-44 is authorized for the third 1O-year lSI interval of IP3, as compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. 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 Contributor: Prakash Patnaik Date: February 5, 2f['f)
' ML090080390 OFFICE LPL1-1/PM LPL1-1/LA CPNB/BC OGC LPL1-1/BC NAME
..IBoska SLittie AHiser STurk MKowal DATE 1/08/09 1/12/09 1/26/09 2/03/09 2/05/09