ML20128D225

Safety Evaluation Accepting Methodology Described in Topical Rept EA-CA-91-0001-M, Steady State Core Physics Methods for BWR Designs & Analysis

ML20128D225
Person / Time
Site:	River Bend
Issue date:	12/02/1992
From:	Attard A Office of Nuclear Reactor Regulation
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ML20127E324	List: ML20127E318 ML20128D202 ML20128D225
References
NUDOCS 9212070186
Download: ML20128D225 (2)

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} UNITED STATES NUCLEAR REGULATORY COMMISSION 1

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. O WASHINGTON. D.C. 20846

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8 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO STEADY-STATE CORE PHYSICS METHODS FOR BWR DESIGN AND ANALYSIS -

TOPICAL REPORT EA-CA-91-0001-M GULF STATES UTILITIES COMPANY RIVER BEND STATION. UNIT 1 DOCKET NO. 50-458

1.0 INTRODUCTION

By letter dated January 30, 1991, Gulf States Utilities (GSU) submitted the topical report EA-CA-91-0001-M, entitled " Steady State Core Physics Methods for BWR Design and Analysis." The topical report describes the methods and '

models proposed for GSU application to the reload analyses of the River Bend Station (RBS). The report includes a description of-the computer codes and modes, and the qualification of the methods via comparison to steady-state operating reactor benchmarks and River Bend plant data. The GSU methods for performing the licensing analyses of the loss of feedwater heating, rod withdrawal and fuel misloading events are also provided together with the a basis for the RBS shutdown margin and standby liquid control system technical '

specifications.

2.0 STAFF EVALUATION The topical report EA-CA-91-0001-M included calculations of detailed lattice physics for RBS. -Core analysis _ methodology employing the CASMO computer code ,

led to the determination of homogenized nodal cross-sections, pin-wise power distributions and traversing ir. core probe instrumentation response factors.

CASMO was also used to perform fuel assembly calculations over the expected range of fuel burnup at 0.0, 0.40 and 0.70-void fraction.

The computer code MICBURN was used to perform one-dimensional multi-region-dealetion= calculations of the gadolina absorber rods. The output cross-: -

sections data from MICBURN forms the input to CASMO.- The core performance -

analyses are performed with the-SIMULATE-E three-dimensional coupled

- neutronics/ thermal-hydraulics computer code. The code FIBWR was used to-determine assembly flow rates and bypass flow. -SIMTRAN-E is used to calculate two-dimensional-spatia 1' integration data'that forms the-kinetics input to the-one-dimensional RETRAN-02 model.

The staff reviewed and evaluated the methodology _ described in topical report EA-CA-91-0001-M _for analyzing and qualifying core physics methods for BWR designs, and its intended use at RBS. The staff was aided by the technical L assistance of- Brookhaven Nationa1' Laboratory (BNL). The evaluation and' .

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e findings are described in detail in the BNL technical evaluation report (TER) which is enclosed as an attachment to this report.

3.0 CONCLUSION

The staff finds the methodology described in topical report EA-CA-91-0001-H,

" Steady-State Core Physics Methods for BWR Designs and Analysis," acceptable for reload licensing analyses for River Bend Station, subject to the applicable limitations and restrictions presented below:

1. The TER conclusion restricts the use of the methodology described in the topical report EA-CA-0001-M to RBS-cores containing fuel types similar.to those included in the benchmarking data base. If different fuel designs are introduc.d or operating conditions vary from those included in the benchmark data base, it is the licensee's responsibility to confirm the-adecuacy of the methodology, and to demonstrate that the accuracy of 3recictions and calculational uncertainties are within the domain of the 3enchmarking data base.

2. In order to provide a conservative analysis of the control' rod withdrawal error (CRWE) event, the SIMULATE-E modified coarse mesh diffusion theory should be used to determine the operating limit minimum critical power r

ratio (0LMCPR) for the CRWE event.

3. A reevaluation of the fuel loading error event wiil be required for River Bend core designs that are outside the range of the approved generic analysis. .

Attachment:

Technical Evaluation Report Principal Contributor: Anthony C. Attard, Reactor Systems Branch, NRR Date: December 2,1992 L

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