ML20149L183
| ML20149L183 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/21/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20149L181 | List: |
| References | |
| NUDOCS 9602230379 | |
| Download: ML20149L183 (2) | |
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[p* "'%qk UNITED STATES 7
S NUCLEAR REGULATORY COMMISSION E
'f WASHINGTON. D.C. 20S55-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO THE USE OF CASH 0-3/ SIMULATE-3 FOR RELOAD DESIGN METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORP 0 RATION THREE MILE ISLAND NUCLEAR STATION. UNIT NO. 1 DOCKET N0. 50-289
1.0 INTRODUCTION
By letters dated March 6, 1995 and October 5, 1995 (Refs. 1 & 2), the GPU Nuclear Corporation (GPUN), the licensee for Three Mile Island Nuclear Station, Unit 1 (THI-1), submitted Topical Report, TR-091, " Steady State Reactor Physics Methodology for THI-1," for NRC review and approval.
This report describes GPUN's methodology for using the NRC approved CASMO-3 and SIMULATE-3 computer codes for neutronics design analysis and implementation of the various description changes and revisions to physics methods that are consistent with the capabilities of the CASM0-3/ SIMULATE-3 design model.
The report also addresses the application of the CASM0-3/ SIMULATE-3 code package in modeling THI-1 cores and compares results to actual plant operating data.
2.0 BACKGROUND
TR-091 contains the documentation of the model and presents the benchmark results for TMI-1 measured physics data vs SIMULATE-3 predictions for cycles 1-10.
The report also compares CASMO-3 calculation of K-effective for a variety of critical experiments.
CASH 0-3 is a lattice physics code used by GPUN in determining neutronics input to SIMULATE-3 for pressurized water reactor (PWR) core performance analyses.
CASMO-3 uses a cross section library based on standard ENDF/8-IV cross sections to determine broad group input for SIMULATE-3.
The SIMULATE-3 code is based on a modified coarse mesh (nodal) diffusion theory calculation with i
coupled thermal hydraulic and doppler feedback. The code solves the two group diffusion equation with fuel assembly homogenization and baffle / reflector i
i modeling.
The code also performs core burnup depletion and performs 6 pin t
power distribution reconstruction, t
DR DOC 050 89 Enclosure P
PDR 9
, The two group model solves the neutron diffusion equation in three dimensions and the assembly homogenization employs the flux discontinuity correction factors from CASMO-3 to combine nodal flux shape and the heterogeneous flux distribution.
This concept is also applied to the baffle reflector region in both the axial and radial directions to eliminate the need for user supplied albedos or other adjustments at the core reflector interface.
SIMULATE-3 can be used to calculate the three dimensional pin-by-pin power distribution in a manner that accounts for the individual pin burnup and spectral effects.
SIMULATE-3 also calculates control rod worth, moderator, Doppler and xenon feedback effects.
TR-091 describes the calculations of the various core parameters used in safety analysis. The primary core parameters considered are the integrated radial and planar radial peaking factors, the moderator temperature coefficient of reactivity, the fuel temperature coefficient of reactivity, control element assembly (CEA) drop data, CEA ejection data, CEA scram reactivity worth, reactivity insertion for steamline break cooldown, and axial power distributions.
The methods which are used to develop uncertainties for these parameters are also described in TR-091. TR-091 also provides extensive benchmarking of the CASM0-3/ SIMULATE-3 neutronics models used in reload core analyses.
The measured data base consists of data from Cycles 1-10.
3.0 EVALUATION GPUN has performed extensive benchmarking using the CASMO-3/ SIMULATE-3 methodology.
This effort consisted of detailed comparisons of the calculated key physics parameters with measured data. These results were used to determine the 95/95 tolerance limits for application-to the calculation of the key PWR physics parameters.
This demonstrates the ability of GPUN to use the CASMO-3/ SIMULATE-3 computer program package for application to TMI-1.
In addition, the staff has already approved the use of this code package for a number of utilities to perform in house steady state physics analyses. These approvals were for essentially the same use of this package.
In addition, the uncertainties requested by GPUN for their methodology are conservative with respect to what is currently being used for nuclear analyses and as demonstrated in reference 2 the change over to the CASMO-3/ SIMULATE-3 analyses package will have a minimal effect on actual core operating limits.
Based on the results presented in TR-091 and based on the fact that l
CASM0-3/ SIMULATE-3 has been previously approved by the staff at other sites, the staff concludes that CASM0-3/ SIMULATE-3 methodology can be applied to PWR reactor physics calculations for TMI-l reload applications. The accuracy of this methodology has been demonstrated to be sufficient for use in licensing applications, including PWR reload physics analysis, generation of safety analysis inputs, startup predictions and reactor protection system setpoint updates.
Principal Contributor:
G. Schwenk Date: February 21, 1996 4
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