ML091110584
| ML091110584 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/13/2009 |
| From: | Justin Poole Plant Licensing Branch III |
| To: | Meyer L Florida Power & Light Energy Point Beach |
| Poole Justin/DORL/LPL3-1/ 301-415-2048 | |
| References | |
| TAC ME0219, TAC ME0220 | |
| Download: ML091110584 (5) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 13, 2009 Mr. Larry Meyer Site Vice President FPL Energy Point Beach, LLC 6610 Nuclear Road Two Rivers, WI 54241-9516
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION FROM REACTOR SYSTEMS BRANCH RELATED TO LICENSE AMENDMENT REQUEST NO. 241 ALTERNATE SOURCE TERM (TAC NOS. ME0219 AND ME0220)
Dear Mr. Meyer:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated December 8, 2008, as supplemented by letters dated January 16 and 27, 2009, February 20,2009, and two letters on April 17, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML083450683, ML090160571, ML090280348, ML090540860, ML091100182,and ML091100215, respectively), FPL Energy Point Beach, LLC, submitted a request to revise the current licensing basis to implement the alternate source term through reanalysis of the radiological consequences of the Final Safety Analysis Report Chapter 14 accidents.
The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on April 22, 2009, it was agreed that you would provide the additional information within 30 days of the date of this letter.
The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2048.
Justin C. Poole, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301
Enclosure:
Request for Additional Information cc w/encl: Distribution via ListServ
REQUEST FOR ADDITIONAL INFORMATION POINT BEACH NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301
- 1. The licensee proposed to use RAVE methodology documented in WCAP-16259-P-A to determine rods in departure from nucleate boiling (DNB) for the locked rotor (LR) event. The Nuclear Regulatory Commission (NRC) safety evaluation report (SER) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML052340326),
approving WCAP-16259 indicated that the basis of NRC's acceptance of the WCAP is, in part, that "Westinghouse will maintain training guidelines that assure only qualified analysts perform and verify the analyses being performed." Discuss qualifications of the analysts and address how the analysts meet the Westinghouse training guidelines for use of the RAVE methodology documented in the WCAP.
- 2. The Westinghouse RAVE methodology contains three Westinghouse computer codes, SPNOVA, VIPRE, and RETRAN. Identify and provide the nodalization diagrams for use of SPNOVA, VIPRE, and RETRAN that deviated from those used for reference plants documented in applicable WCAP reports, and justify the deviations.
- 3. Section 6.2 of Enclosure 3 to the licensee's letter dated December 8, 2008, stated that in the analysis of the steam generator tube rupture (SGTR), "The equilibrium primary-to-secondary break flow is assumed to persist until 30 minutes after the initiation of the SGTR, at which time the operators have completed the actions necessary to terminate the steam release from the ruptured SG. Pressure between the ruptured SG and the primary system is such that the ruptured SG is not overfilled." The consequences of a SGTR depend largely on the ability of the operator to take necessary actions to terminate the primary-to-secondary break flow. The licensee did not indicate what is the operator action time from the start of the event assumed in the analysis to terminate break flow. If the break flow continues for an extended period of time, the secondary side of the SG may be filled and water may enter the steam line, which results in unanalyzed conditions. As a result of the January 1982 SGTR event at the Ginna plant, NRC questioned the assumptions used in the Ginna SGTR analysis, which assumed that the event is terminated in 30 minutes. In response to the NRC staff concerns, a subgroup of utilities in the Westinghouse Owner's Group was formed to address the licensing issues associated with a SGTR event on a generic basis. The subgroup submitted and NRC approved a topical report, WCAP-10698-P-A, "SGTR Analysis methodology to Determine the Margin to Steam Generator Overfill."
Discuss the SGTR mitigation strategy credited in the analysis and provide the basis for the mitigation strategy. Provide a sequence of the event listing the operator actions credited in the SGTR analysis and justify adequacy of the assumed operator actions and associated times. Also, discuss the computer code used to determine the margin to SG overfill and show it is an NRC-approved code. In addition, discuss whether the methodology documented in WCAP-1 0698 is applicable and needed to apply to the Point Beach plant for the SG overfill prevention or not. The Point Beach plant and Ginna plant are Westinghouse two-loop plants with similar rated thermal power levels. Among other Westinghouse plants, Enclosure
-2 Ginna performed and !\\IRC approved the SGTR reanalysis using the WCAP-1 0698 methodology. If the licensee determines that the WCAP methodology is not applicable and a reanalysis of the SGTR event is not needed, provide rationale for the determination. If the licensee determines that the reanalysis using the WCAP methodology is needed, it should provide the results of the SGTR reanalysis to the NRC staff for review.
- 4. In response (Enclosure 7 to letter of December 8,2008) to condition 1 on the use of RETRAN, the licensee indicated that RETRAN will be used in the analysis not only for the locked rotor event, but also for the following events: (1) excessive increase in steam flow; (2) steam line break; (3) loss of external electrical load; (4) loss of all alternating current power to the station auxiliaries; (5) loss of normal feedwater flow; (6) loss of reactor coolant flow; and (7) uncontrolled rod withdrawal at power.
Although RETRAN was approved by NRC on a generic basis, the licensee should provide a discussion to address the adequacy of the specific plant application of RETRAN for analysis of the events identified above as event (1) through (7) by showing that: the analysts using RETRAN to perform the analysis are adequately qualified; the values of input parameters appropriately represent the plant conditions or reflect limiting core operating conditions when applicable; the results of thermal-hydraulic and system responses for the analysis are within the approved applicable ranges of RETRAN; and there is no mathematical unstable conditions. (The same requests are applied to the use of VIPRE for licensing applications).
Generic letter (Gl) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," outlines a process that a licensee can use to remove cycle-specific parameters from the plant-specific Technical Specifications (TS) to a licensee-controlled document entitled, "Core Operating Limit Report" (CORl). A necessary element of that process is that a licensee includes specific methodologies in TSs. In accordance with the Gl guidance, justify that the topical reports that documented the approved RETRAN and VIPRE codes are not referenced in TS 5.6.4, "Core Operating Limits Report (COlR)".
In addressing compliance of condition 2 on the use of VIPRE, the licensee indicated that "continued applicability of the input assumptions is verified on a cycle-by-cycle basis using the Westinghouse reload methodology described in WCAP-9272P-A. Following the Gl gUidance, justify that the topical report, WCAP-9272-A, is not referenced in TS 5.6.4.
- 5. Section 6.3 of Enclosure 3 to the licensee's letter dated December 8, 2008, stated that "30%
of the fuel rods in the core are assumed to suffer damage due to DNB" in the lR radiological analysis.
Provide a basis for the assumption of 30 percent fuel failure due to DNB used in the lR radiological analysis. Discuss the methodology for determination of the number of failed fuel rods due to DNB during a lR event. Justify that the core with assumed 30 percent fuel failure maintains in a coolable geometry during a lR event.
- 6. In response (Enclosure 7 to letter of December 8, 2008) to condition 1 on the use of VIPRE, the licensee indicated that the WRB-1 correlation with the associated safety departure from nucleate boiling ratio (DNBR) limit of 1.17 was used in the DNB analysis for the Point Beach
- 3 14X14 422V+ fuel. Discuss the design of 14X14 422V+ fuel, compare it with the fuel designs that are acceptable for use of the WRB-1 correlation, and justify that the use of WRB-1 for the 14x14 422V+ fuel is within the applicable range of the WRB correlation with the associated safety DNBR limit of 1.17. If the application of the WRB-1 correlation was previously approved by NRC, provide the author, date and title of the NRC SER approving the WRB-1 correlation for use in the DNBR analysis for the Point Beach 14X14 422V+ fuel.
- 7. In addressing compliance (Enclosure 7 to letter of December 8,2008) with condition 5 on the use of RAVE, the licensee indicated that the impact of exceeding 30 percent void fraction limit was investigated and it was determined to be conservative with respect to over pressurization during a LR event. Discuss the impact study of the void fraction on over pressurization and provide the results to support the conclusion that the void fraction greater than 30 percent will result in a higher reactor coolant system pressure during a LR over pressurization event.
- 8. The NRC SER (from NRC to W. J. Johnson (Westinghouse) in letter dated November 26, 1990) approving the use of SPI\\lOVA imposed the following conditions:
(a) the comparison must include both the advanced nodal code (ANC) calculational uncertainty and the ANC/SPNOVA calculated difference in isothermal temperature coefficient (ITC), if the SPNOVA ITC uncertainty is determined by comparison to ANC; (b) additional benchmarking is required, if SPNOVA is applied to fuel and core designs that differ significantly from those included in the benchmark data discussed in Section 3.2.1 of the SER; and (c) the uncertainties of transient application of SPNOVA are required to assure an acceptable margin to the fuel safety limits and must be provided in event-specific submittal.
Address the compliance with the above conditions.
May 13, 2009 Mr. Larry Meyer Site Vice President FPL Energy Point Beach, LLC 6610 Nuclear Road Two Rivers, WI 54241-9516
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION FROM REACTOR SYSTEMS BRANCH RELATED TO LICENSE AMENDMENT REQUEST NO. 241 ALTERNATE SOURCE TERM (TAC NOS. ME0219 AND ME0220)
Dear Mr. Meyer:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated December 8, 2008, as supplemented by letters dated January 16 and 27, 2009, February 20, 2009, and two letters on April 17, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML083450683,ML090160571,ML090280348, ML090540860,ML091100182,and ML091100215, respectively), FPL Energy Point Beach, LLC, submitted a request to revise the current licensing basis to implement the alternate source term through reanalysis of the radiological consequences of the Final Safety Analysis Report Chapter 14 accidents.
The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on April 22, 2009, it was agreed that you would provide the additional information within 30 days of the date of this letter.
The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2048.
Sincerely, IRAI Justin C. Poole, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301
Enclosure:
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