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Final Safety Evaluation for Westinghouse Electric Company

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Final Safety Evaluation for Westinghouse Electric Company June 27, 2007 Mr. James A. Gresham, Manager Regulatory Compliance and Plant Licensing Westinghouse Electric Company P.O. Box 355 Pittsburgh, PA 15230-0355 SUBJECT: FINAL SAFETY EVALUATION FOR WESTINGHOUSE ELECTRIC COMPANY (WESTINGHOUSE) TOPICAL REPORT (TR) CENPD-132 SUPPLEMENT 4-P-A, ADDENDUM 1-P, “CALCULATIVE METHODS FOR THE CE [COMBUSTION ENGINEERING] NUCLEAR POWER LARGE BREAK LOCA EVALUATION MODEL - IMPROVEMENT TO 1999 LARGE BREAK LOCA EM STEAM COOLING MODEL FOR LESS THAN 1 IN/SEC CORE REFLOOD” (TAC NO. MD2161) Dear Mr. Gresham: By letter dated May 11, 2006, Westinghouse submitted TR CENPD-132, Supplement 4-P-A, Addendum 1-P, “Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model - Improvement to 1999 Large Break LOCA EM Steam Cooling Model for Less Than 1 in/sec Core Reflood,” to the U.S. Nuclear Regulatory Commission (NRC) staff. By letter dated June 7, 2007, an NRC draft safety evaluation (SE) regarding our approval of TR CENPD- 132, Supplement 4-P-A, Addendum 1-P, was provided for your review and comment. By email dated June 11, 2007, from William Slagle, Senior Fuel Licensing Engineer, Westinghouse indicated a few minor editorial comments on the draft SE. The resolution of these minor editorial comments were incorporated into the final SE enclosed with this letter. The NRC staff has found that CENPD-132, Supplement 4-P-A, Addendum 1-P is acceptable for referencing in licensing applications for Westinghouse designed pressurized water reactors to the extent specified and under the limitations delineated in the TR and in the enclosed final SE. The final SE defines the basis for our acceptance of the TR. Our acceptance applies only to material provided in the subject TR. We do not intend to repeat our review of the acceptable material described in the TR. When the TR appears as a reference in license applications, our review will ensure that the material presented applies to the specific plant involved. License amendment requests that deviate from this TR will be subject to a plant-specific review in accordance with applicable review standards. In accordance with the guidance provided on the NRC website, we request that Westinghouse publish accepted proprietary and non-proprietary versions of this TR within three months of receipt of this letter. The accepted versions shall incorporate this letter and the enclosed final SE after the title page. Also, they must contain historical review information, including NRC requests for additional information and your responses. The accepted versions shall include an "-A" (designating accepted) following the TR identification symbol. J. Gresham -2- If future changes to the NRC's regulatory requirements affect the acceptability of this TR, Westinghouse and/or licensees referencing it will be expected to revise the TR appropriately, or justify its continued applicability for subsequent referencing. Sincerely, /RA/ Ho K. Nieh, Deputy Director Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project No. 700 Enclosure: Final SE cc w/encl: See next page Westinghouse Electric Project No. 700 cc: Mr. Gordon Bischoff, Manager Owners Group Program Management Office Westinghouse Electric Company P.O. Box 355 Pittsburgh, PA 15230-0355 [email protected] 12/21/05 J. Gresham -2- June 27, 2007 If future changes to the NRC's regulatory requirements affect the acceptability of this TR, Westinghouse and/or licensees referencing it will be expected to revise the TR appropriately, or justify its continued applicability for subsequent referencing. Sincerely, /RA/ Ho K. Nieh, Deputy Director Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project No. 700 Enclosure: Final SE cc w/encl: See next page DISTRIBUTION: PUBLIC PSPB Reading File RidsNrrDpr RidsNrrDprPspb RidsNrrPMJThompson RidsNrrLADBaxley RidsOgcMailCenter RidsAcrsAcnwMailCenter RidsNrrDssSnpb SRosenberg (Hardcopy) ADAMS ACCESSION NO.: ML071730336 *No major changes to SE input. NRR-043 OFFICE PSPB/PM PSPB/PM PSPB/LA SNPB/BC* PSPB/BC DPR/DD NAME JThompson TMensah DBaxley AMendiola TMensah for HNieh for SRosenberg Jhopkins DATE 6/ 26 /07 6/ 26 /07 6/25/07 4/17/07 6/ 26 /07 6/ 27 /07 OFFICIAL RECORD COPY FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT (TR) CENPD-132, SUPPLEMENT 4-P-A, ADDENDUM 1-P, “CALCULATIVE METHODS FOR THE CE [COMBUSTION ENGINEERING] NUCLEAR POWER LARGE BREAK LOCA [LOSS-OF-COOLANT ACCIDENT] EVALUATION MODEL [EM] - IMPROVEMENT TO 1999 LARGE BREAK LOCA [LBLOCA] EM [EVALUATION MODEL] STEAM COOLING MODEL FOR LESS THAN 1 IN/SEC CORE REFLOOD” WESTINGHOUSE ELECTRIC COMPANY PROJECT NO. 700 1.0 INTRODUCTION AND BACKGROUND The Westinghouse Electric Company (Westinghouse) submitted a TR revision, by letter LTR-NRC-06-25, dated May 11, 2006 (Reference 1), titled "Submittal of CENPD-132 Supplement 4-P-A, Addendum 1-P, 'Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model Improvement to 1999 Large Break LOCA EM Steam Cooling Model for Less Than 1 in/sec Core Reflood'" and "CENPD-132 Supplement 4-NP-A, Addendum 1-NP, 'Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model Improvement to 1999 Large Break LOCA EM Steam Cooling Model for Less Than 1 in/sec Core Reflood'" (Proprietary/Non-Proprietary). The revision documented both a proposed and an optional change to the Westinghouse 1999 LBLOCA EM for CE-designed plants. The change would remove some of the conservatism in the approved steam cooling heat transfer component model found in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix K, by including spacer grid heat transfer effects for reflood rates less than 1 in/sec. This change was proposed because of the calculated adverse consequences to the emergency core cooling system (ECCS) performance using the 1999 EM for the CE 16x16 Next Generation Fuel (NGF) design. The ECCS calculations were adversely impacted by the increase in the core hydraulic pressure loss, the increase in the core cross-sectional flow area, and the decrease in the fuel rod cladding outside diameter. In particular, the core reflood calculations during a LBLOCA were adversely impacted by the CE 16x16 NGF design changes and the core reflood rates that were used to calculate the reflood heat transfer coefficients for the hot rod decreased. The CE 16x16 NGF design changes were estimated by Westinghouse to have an insignificant impact on the ECCS performance peak cladding temperature (PCT). However, the impact on the ECCS -2- performance maximum cladding local oxidation percentage for the hot rod rupture node was estimated by Westinghouse to be large enough to warrant specific consideration. Westinghouse provided additional information, in response to the NRC staff’s request for additional information (RAI), to supplement and clarify portions of the submittal by letter LTR-NRC-07-2 dated January 10, 2007 (Reference 2), letter LTR-NRC-07-7 dated February 1, 2007 (Reference 3), and by letter LTR-NRC-07-18 dated March 28, 2007 (Reference 4). 2.0 REGULATORY EVALUATION The ECCS is designed to provide protection against postulated LOCAs caused by ruptures in the primary system piping. The functional requirements for the ECCS performance, under all LOCA conditions postulated in the design, must satisfy 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.” The ECCS calculated cooling performance is based on an acceptable EM for which there is sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor coolant system during a LOCA or, in this case, an ECCS EM developed in conformance with Appendix K to 10 CFR Part 50. The three specific ECCS performance criteria important to this safety evaluation are: 1. The calculated maximum fuel element cladding temperature does not exceed 2200 EF. 2. The calculated total local oxidation of the cladding does not exceed 17 percent of the total cladding thickness before oxidation. Total local oxidation includes pre-accident oxidation as well as oxidation that occurs during the course of the accident. 3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam does not exceed 1 percent of the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react. The impact of the CE 16x16 NGF design changes on the ECCS performance maximum cladding local oxidation percentage for the hot rod rupture node was estimated to be large enough to warrant specific consideration for item 2, above. The following additional requirements were considered for an Appendix K based model: 1. 10 CFR Part 50, Appendix K.D.5.b states “during refill and during reflood when reflood rates are less than one inch per second, heat transfer calculations shall be based on the assumption that cooling is only by steam, and shall take into account any flow blockage calculated to occur as a result of cladding swelling or rupture as such blockage might affect both local steam flow and heat transfer.” 2. 10 CFR Part 50, Appendix K.C.4.e states “after CHF [Critical Heat Flux] is first predicted at an axial fuel rod location during blowdown, the calculation shall not use nucleate boiling heat transfer correlations at that location subsequently during the blowdown even if the calculated local fluid and surface conditions would apparently justify the reestablishment of nucleate boiling. Heat transfer assumptions characteristic of return -3- to nucleate boiling (rewetting) shall be permitted when justified by the calculated local fluid and surface conditions during the reflood portion of a LOCA.” 3.0 TECHNICAL EVALUATION 3.1 1999 EM Steam Cooling Model for Core Reflood Rate Less Than 1 in/sec. • The 1999 EM steam cooling model is a component of the Appendix K ECCS calculational method, and is applied to the hot rod rupture node elevation and above when the core reflood rate is less than 1 in/sec.
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