ML093561396

Nonacceptance of Utilization of Areva Fuel and Associated Analysis Methodologies. (TAC ME2451) (TS-467)

ML093561396
Person / Time
Site:	Browns Ferry
Issue date:	12/23/2009
From:	Ellen Brown Plant Licensing Branch II
To:	Krich R Tennessee Valley Authority
Brown Eva, NRR/DORL, 415-2315
References
TAC ME2451, TS-467
Download: ML093561396 (7)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 23, 2009 Mr. R. M. Krich Vice President, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT UNIT 1 - NONACCEPTANCE OF UTILIZATION OF AREVA FUEL AND ASSOCIATED ANALYSIS METHODOLOGIES (TAC NO. ME2451) (TS-467)

Dear Mr. Krich:

By letter dated October 23, 2009, as supplemented by a letter dated November 17, 2009, the Tennessee Valley Authority submitted an amendment request for Browns Ferry Nuclear Plant, Unit 1. The proposed amendment would add the AREVA NP analysis methodologies to the list of approved methods to be used in determining the core operating limits. The purpose of this letter is to provide the results of the U.S. Nuclear Regulatory Commission (NRC) staff's acceptance review of this request. The acceptance review was performed to determine if there was sufficient technical information in scope and depth to allow the NRC staff to complete its detailed technical review. The acceptance review is also intended to identify whether the application has any readily apparent information insufficiencies in its characterization of the regulatory requirements or the licensing basis of the plant.

Consistent with Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), an amendment to the license (including the technical specifications) must fully describe the changes requested, and following as far as applicable, the form prescribed for original applications. Section 50.34 of 10 CFR addresses the content of technical information required.

This section stipulates that the submittal address the design and operating characteristics, unusual or novel design features, and principal safety considerations. The NRC staff has reviewed your application and concluded that the information delineated in the enclosure to this letter is necessary to enable the NRC staff to make an independent assessment regarding the acceptability of the proposed amendment request in terms of regulatory requirements and the protection of public health and safety and the environment.

In order to make the application complete, the NRC staff requests that TVA supplement the application to address the information requested in the enclosure by January 15, 2010. This will enable the NRC staff to complete its detailed technical review. If the information responsive to the NRC staff's request is not received by the above date, the application will not be accepted for review pursuant to 10 CFR 2.101, and the NRC staff will cease its review activities associated with the application. If the application is subseq'~ently accepted for review, you will be advised of any further information needed to support the NRC staff's detailed technical review by separate correspondence.

P. Swafford

- 2 The information requested and associated timeframe in this letter were discussed with Mr.

Daniel Green of your staff on December 21, 2009. If you have any questions, please contact the Browns Ferry Project Manager, Ms. Eva Brown, at (301) 415-2315.

Sincerely, Eva A. Brown, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-259

Enclosure:

Request for Supplemental Information cc: Distribution via Listserv

REQUEST FOR SUPPLEMENTAL INFORMATION AREVA FUEL CHANGE REQUEST TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT 1 TAC NO. ME2451

1.

Licensing Topical Report (LTR) ANP-2638P "Applicability of AREVA NP BWR

[boiling water reactor] Methods to Extended Power Uprate [EPU] Conditions," states that loss-of-coolant accident (LOCA) results are only weakly dependent on core average power. However, for the small break LOCA (SBLOCA) the analysis results are highly sensitive to the core average power level.

Since depressurization occurs through the automatic depressurization system (ADS) for SBLOCAs, the timing when low pressure injection systems reach rated flow is extended when the core steam generation rate is higher - as would be the case for EPU conditions. Based on the plant-specific power uprate and the ADS capacity, the limiting break for an EPU plant may be an SBLOCA. This was shown for Browns Ferry Unit 1 in the power uprate safety analysis report.

The EXEM BWR-2000 LOCA analysis methodology is described by LTR EMF-2361 (P)(A),

"EXEM BWR-2000 ECCS [emergency core cooling system] Evaluation Model." This LTR states:

SBLOCA PCTs [peak cladding temperatures] are bound when the conservatism included in the EM methodology is applied. This result is acceptable because small break events are not limiting in BWRs and the test evaluated simulated an extremely small break in which core uncovery and the resulting heat-up is minor such that the conservatism (Appendix K coefficients) are not allowed to raise fuel temperature to values of concern.

When the LTR language is considered in the context of the Unit 1 LOCA analyses at EPU conditions, EXEM BWR-2000 does not appear to be applicable. First, for Unit 1 at EPU conditions, the limiting break is a small break. This appears contrary to the basis for the staff approval of EXEM BWR-2000.

Second, at EPU conditions the core heat-up is not rapidly terminated because blowdown times are prolonged for SBLOCA. Therefore, the core uncovery persists for a longer duration and the Appendix K assumptions (e.g., the 20-percent increase in decay heat) will contribute to significant heatup and high-fuel temperatures. This appears to conflict with the disposition of the SBLOCA qualification results in the LTR.

Enclosure

- 2 Therefore, it does not appear that EXEM BWR-2000 is applicable to analyze the limiting LOCA event for Unit 1. Provide the SBLOCA analyses for Unit 1 using acceptable methods.

2.

Provide the LOCA results for hydrogen generation/core wide oxidation.

3.

The statements regarding the transition core effects on the LOCA analyses (EMF-2950(P), "Browns Ferry Units 1, 2, and 3 Extended Power Update LOCA Break Spectrum Analysis," Section 2), would inherently impose similar performance conclusions on the legacy fuel. By this logic, the current licensing basis analysis should demonstrate similar performance to the fuel transition analysis. Provide the report describing the previous licensing basis (in this case, the previous licensing basis is at EPU conditions) LOCA analysis and supplement the analysis by accounting for model differences that cause the results in break spectrum, location, geometry, and results to differ.

4.

Single failure analyses do not account for the ADS unavailability. Please provide the failure modes and effects analysis for the ADS with respect to the limiting postulated SBLOCA, and for the postulated high pressure coolant injection (HPCI) line break.

Justify not analyzing the failure of the ADS system in toto, or even a single ADS valve.

5.

Table 2.3, Cycle Specific Reload Evaluation Methodologies, of the Reload Safety Analysis Report (RSAR, ANP-2864(P), "Browns Ferry Unit 1 Cycle 9 Reload Safety Analysis") lists more transients than Table 2.1 indicates are analyzed on a cycle-specific basis. Provide an explanation for the discrepancy.

6.

The approved AREVA nuclear design method topical reports do not appear to describe a method for calculating the standby liquid control system (SLCS) cold shutdown margin (CSDM).

Provide a description of the SLCS CSDM calculation method. The methodology description should include:

a. A discussion of the codes used;

b. A discussion of the nuclear data that must be generated (e.g., lattice parameters for given boron concentrations);

c. A list of pertinent analysis assumptions (e.g., active core averaged boron weight),

d. A justification of the method accuracy;

e. A discussion on the treatment of short-lived highly absorbing nuclides (e.g., xenon and samarium); and,

f. A discussion of the assumed core thermal-hydraulic conditions (e.g.,

temperature of 68 degrees F).

- 3 This description should address the application of the method to the co-resident GE14 fuel.

7.

Provide an analysis of SLCS shutdown margin at 68 degrees F, consistent with definition of rodded shutdown margin appearing in Unit 1 Technical Specifications (TSs).

Demonstrate the capability to maintain subcritical configuration in cold conditions.

8.

Clarify the following language in the TS: "... latest approved versions applicable to BFN." Several of the references listed have supplements and addenda. Address why the supplements and addenda were not included and provide a listing of the latest approved version as well as the latest approved supplements and/or addenda to the listed topical reports.

Also, clarify the process that is followed when a new version, supplement, or addendum is approved by the NRC in the midst of the generation of the cycle operating limits report for the next cycle.

9.

Explain why the following topical reports are not included in TS 5.6.5:

EMF-2209(P), Rev. 2, Addendum 1, "SPCB Additive Constants for ATRIUM-10 Fuel," May 1, 2008; EMF-CC-074(P)(A), "BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2," Vol. 4, Siemens Power Corporation, August 2000; EMF-85-74(P), Supplement 1(P)(A) and Supplement 2 (P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation, February 1998; BAW-10255(P)(A), Revision 2, "Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code," Framatome ANP, May 2008; and BAW-10247(P)(A), Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA, May 2008.

10.

Section 4.8.1 of ANP-2860P, "Browns Ferry Unit 1 Summary of Responses to Requests for Additional Information" states that for fast pressurization transients surface heat flux calculations are provided to the licensee and that appropriate linear heat generation rate factor (LHGRFAC) values are developed based on Global Nuclear Fuel (GNF) thermal-mechanical analyses. Provide a description of the analysis procedures that are used to demonstrate that the GNF fuel meets applicable thermal-mechanical licensing limits. Describe the number of cases that were benchmarked in the safety analysis.

This description should address both fast and slow transients.

11.

Provide the results of the Thermal Mechanical Analyses for GE14 fuel, which are not contained in the Reload Safety Analysis Report.

- 4

12.

Provide the reference, Mneimneh, GNF, letter to McNelley, TVA, "Revised LHGR Limits for BF1 Transition," MJM-TVA-ER1-09-39.

13.

Provide additional details regarding the control rod drop accident (CRDA) analysis.

Results of the CRDA analysis appear inconsistent with the methodology. Provide an updated description of the analytic method that explains how the maximum number of rods exceeding 170 cal/g is determined, and what assumptions are used to make this determination.

14.

The approved methodology relies on parameterization of generic analysis results.

Provide the parameterized function. Given the application to GE14, justify the applicability of the generic analyses to modern fuel bundle designs; in particular describe how the differences in bundle fuel mass are accounted for in the parameterized function.

15.

Describe the core conditions that are evaluated to determine the limiting values for the following parameters: (1) control blade worth, (2) four bundle local peaking factor, (3) Doppler coefficient, and (4) delayed neutron fraction.

16.

Discuss the relationship between the maximum dropped rod reactivity worth and the cycle analyses provided in the fuel cycle design report. This documentation should specify the limiting rod, the method used to identify the limiting rod, the limiting point in exposure, and any consideration given for operational flexibilities (e.g., suppressing power in a leaking fuel bundle).

17.

The American Society of Mechanical Engineers overpressure analysis using ATRIUM-10 fuel and AREVA methods credits the failure of a direct scram as the limiting single failure. NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for the Nuclear Power Plants", Chapter 5.2.2 assumes a reactor scram on the second safety-grade scram signal. This is a condition of the analysis, not an assumed failure.

Identify the limiting single failure assumed in this analysis.

18.

The RSAR dispositions criticality for new and spent fuel storage based on previous analyses. Address why the use of previous analyses are acceptable.

19.

The RSAR dispositions Final Safety Analysis Report (FSAR) Section 14.5.2.8 - Pressure Regulator Failure Downscale - by stating that this event is eliminated as an Anticipated Operational Transient by installation of a digital fault-tolerant main turbine electro-hydraulic control system. Address why this event was eliminated and whether the modification was made to support the transition to AREVA fuel.

20.

The RSAR dispositions FSAR Chapter 14.5.6.4 - RCP [Reactor Coolant Pump] Rotor Seizure Accident - based on the fact that its consequences are bounded by the LOCA accident; however, the acceptance criteria are listed in terms of minimum critical power ratio and peak pressure. Provide an explanation for the discrepancy.

ML093561396 NRR-106 OFFICE LPL2-2/PM LPL-WB/LA SRXB/BC SNPB/BC LPL2-2/BC NAME TOrf for EBrown RSoia for BClaylon GCranslon AMendiola SLingam for TBoyce DATE 12/23/09 12/23/09 12/23/09 12/23/09 12/23/09