ML091050488
| ML091050488 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 04/15/2009 |
| From: | Justin Poole Plant Licensing Branch III |
| To: | Hale S Florida Power & Light Energy Point Beach |
| Poole Justin/DORL/LPL3-1/ 301-415-2048 | |
| References | |
| Download: ML091050488 (3) | |
Text
From:
Justin Poole Sent:
Wednesday, April 15, 2009 1:59 PM To:
'Steve_Hale@fpl.com'; 'Flentje, Fritzie'
Subject:
Draft RAIs on LAR 241 Alternate Source Term from Reactor Systems Branch
- Steve, By letters dated December 8, 2008, January 16 and 27, 2009, and February 20, 2009, FPL Energy submitted a license amendment application for Point Beach Nuclear Plant Units 1 and 2 to revise the current licensing basis to implement the alternate source term through reanalysis of the radiological consequences of the FSAR Chapter 14 accidents.
The Reactor Systems Branch has been reviewing the submittal and has determined that additional information is needed to complete its review. We would like to discuss the questions, in draft form below, with you in a conference call.
This e-mail aims solely to prepare you and others for the proposed conference call. It does not convey a formal NRC staff position, and it does not formally request for additional information.
Justin C. Poole Project Manager NRR/DORL/LPL3-1 U.S. Nuclear Regulatory Commission (301)415-2048 email: Justin.Poole@nrc.gov
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
DRAFT
- 1. The licensee proposed to use RAVE methodology documented in WCAP-16259-P-A to determine rods in DNB for the locked rotor event. The NRC safety evaluation report (SER, ADAMS ML052340326) approving WCAP-16259 indicated that the basis of NRCs acceptance of the WCAP is, in part, that Westinghouse will maintain training guidelines that assure only qualified analyst perform and verify the analyses being performs. 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 licensees 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 to terminate the steam release from the ruptured SG.
Pressure between the rupture SG and the primary system is such that the rupture SG is not overfilled (underline is added for emphasis). 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 result 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 Owners 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-10698 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, Ginna performed and NRC approved the SGTR reanalysis using the WCAP-10698 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, 2009) 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) excess Increase in steam flow; (2) steam line break; (3) loss of external electrical load; (4) loss of AC power to the station auxiliary; (5) loss of normal feedwater; (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 used 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 Specification, 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 (TRs) 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 6.2 of Enclosure 3 to the licensees 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 locked rotor (LR) radiological analysis.
Provide a basis for the assumption of 30% 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% fuel failure maintains in a coolable geometry during a LR event.
- 6. In response (Enclosure 7 to letter of December 8, 2009) to condition 1 on the use of VIPRE, the licensee indicated that the WRB-1 correlation with the associated safety DNBR limit of 1.17 was used in the DNB analysis for the Point Beach 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, 2009) with condition 5 on the use of RAVE, the licensee indicated that the impact of exceeding 30% 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% will result in a higher RCS 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 SPNOVA imposed the following conditions:
(a) the comparison must include both the ANC calculational uncertainty and the ANC/SPNOVA calculated difference in ITC, if the SPNOVA isothermal temperature coefficient (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.
DRAFT