ML20214C522

Topical Rept Evaluation of Rev 0 to TR 020, Methods for Analysis of BWR Lattice Physics. Collision Probability Module Code Acceptable for BWR Fuel Lattice Calculations

ML20214C522
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Site:	Oyster Creek
Issue date:	11/14/1986
From:	Office of Nuclear Reactor Regulation
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1. %,, UNITED STATES 8" n NUCLEAR REGULATORY COMMISSION

$ WASHINGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO GPU NUCLEAR CORPORATION TOPICAL REPORT TR 070. REVISION 0

" METHODS FOR THE ANALYSIS OF BOILING WATER REACTORS LATTICE PHYSICS" GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER AND LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION

. DOCKET NO. 50-219

1.0 INTRODUCTION

By letter dated November 25,1985 (Ref. 6), the GPU Nuclear Corporation (GPUN) submitted for review TR 020, Revision 0, " Methods for the Analysis of Boiling Water Reactors lattice Physics." The information in this report was supplemented by information submitted in response to questions from the NRC staff and its consultants (Ref. 4). The review by the staff of this report and the supplemental information was performed with the assistance of the staff's consultants from Brookhaven National laboratory (BNL).

The report describes the reactor physics related analyses performed at GPUN for BWR fuel lattices. These analyses are based on the Electric Power Research Institute (EPRI) version of the collision probability transport code CPM. The accuracy of the code, and its ability to perform eigenvalue, power distribution and isotopic depletion calculations is demonstrated via comparisons with results from experimental configurations) measurements and higher order (Monte (from operating) plants and Carlo calculations.

These comparative calculations were performed by GPUN, as well as by EPRI, as part of the original benchmarking of the code and for the Advanced Recycle Methodology Program.

2.0

SUMMARY

OF THE METHODOLOGY The CPM (Collision Probability Module) code was developed in Sweden by AB Atomenergi/Studsvik for the analysis of PWR and BWR fuel assemblies consisting of cylindrical rods in a square array, surrounded by water. In the BWR case, the external water region may contain a channel box, cruciform control rod, or boron curtain. The modelling combines fine group spectrum calculations for sub-regions of the assembly (e.g., fuel pin-cells), with a multigroup transport calculation for a partially homogenized heterogeneous assembly in two-dimensional (x-y) geometry. The version of the code used by GPUN, EPRI-CPM, is from the EPRI Advanced Recycle Methodology Program, i

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2 The calculational sequence for a typical BWR assembly involves three basic steps, with the spatial and energy detail becoming successively coarser as larger regions of the assembly are considered. These steps are termed the micro-group, macro-group, and two-dimensional assembly calculations.

Cruciform control rods are treated via a special subroutine, and the depletion of gadolinium bearing fuel pins requires an auxiliary calculation with the MICBURN code. A brief description of the calculations performed at each stage follows. /

The process begins with micro-group (69 groups) calculations for the various

- pin-cells present in the assembly (fuel, burnable absorber, water rod). A simplified annular geometry consisting of three or four regions is assumed.

For absorber pins or water rods the pin-cell is surrounded by a homogenized fuel buffer region, while for fuel pin-cells an outer zone accounting for the water gap and channel box can be included in addition to the requiar fuel-clad-moderator geometry if necessary.

The nuclear data for these calculations comes directly from the library file (CPMLIB3) which contains microscopic cross sections for 65 elements, including two pseudo isotopes representing non-saturating and slowly saturating fission products. The data are based primarily on ENDF/B-III. Resonance integral data for the four nuclides considered to be resonance absorbers (U-235, U-236, U-238, Pu-239) are tabulated as functions of temperature and potential 2

scattering, and are retrieved based on the application of equivalence theory i principles for the specific configuration.

The number of pin-cell micro-group calculations is determined by the user, but usually includes at a minimum, separate calculations for each different type

! of pin in the assembly, including fuel rods containing different initial enrichments, and an average pin-cell. The resultant fluxes are used to define macroscopic cross sections for spatially homogenized pin-cells in a reduced number of groups.

While in principle the data needed to perform the 2-D assembly calculation may be obtained from the above micro-group calculations, CPM includes an intermediate step, for geometries containing water gaps, whose ma.ior purpose is to reduce the number of energy groups required in the succeeding 2-D assembly calculation. The effect of the different water gap thickness on opposite

, sides of a BWR assembly is also accounted for in this step. This macro-group calculation is performed in cylindrical geometry with a user determined group structure (less than 26 groups 1 The fuel assembly is cylindricalized, i preserving the mean chords of th water gaps, and the volumes of each pin-cell layer of the fuel zone. Two calculations are performed, with the outermost annulus characteristic of the wide and narrow water gaps, respectively. The resultant fluxes are used to further condense the cross sections to the group structure used in the 2-0 (x-y) geometry calculation for the assembly.

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f The 2-D assembly calculation is performed with less than 13 e'nergy groups, and assumes diagonal symmetry. The geometry represents spatially smeared i

pin-cells, but otherwise explicitly includes the channel box, water gaps and any boron curtains. Cruciform control rods can be explicitly included; however, a prior homogenization of the highly heterogeneous wings is performed by the special Control Rod Collision Probability (CROCOP) subroutine in the 2-D group structure. The steel hub and end plates of the control rod can be explicitly represented. ,-

The local flux, rod-wise power distribution and bundle reactivity are determined in the 2-D assembly calculation. A fundamental mode buckling correction based on diffusion theory or the B 3method is used to modify the infinite lattice results to account for leakage.

i The depletion calculations assume 19 linear chains; five for heavy nuclides, 12 for explicit fission products and two for pseudo fission products. The 22 individually treated fission products account for 90% of the fission product absorption. A two-step predictor-corrector approach is employed which allows the use of larger time steps than is usually the case in cell burnup codes. The nuclide inventory in each fuel pin, burnable absorber and boron curtain is detennined separately based on the local fluxes. The

complicated spatial burnup of gadolinium bearing fuel rods requires an i auxiliary calculation outside of CFM. The MICBURN code which performs a detailed multi-zone calculation for a gadolinium pin surrounded by a fuel j

buffer zone serves this function. Both the gadolinium pin and the fuel buffer

zone are depleted to provide effective cross sections for the gadolinium as a function of burn-up which are then used in CPM.

. 3.0 EVALUATION

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i The CPM code is used by GPU Nuclear in its steady state BWR fuel cycle i analyses. The primary function of the code is to provide input data, l including fuel constants and detector response factors, for the GPUN i

three-dimensional core simulator, PSMS N0DE-B.

1 i The accuracy and adequacy of various aspects of EPRI-CPM and its models (e.g.,

i nuclear data, treatment of control rods and gadolinium, transport treatment of

pin-cells and fuel bundles, depletion) is demonstrated by comparisons to results from Monte Carlo calculations performed for Oyster Creek fuel assemblies by Exxon, and to measured results from power reactors and experimental configurations. Comparisons of eigenvalues (k-infinity or k f i

l pin power / fission rate distributions, and isotopic concentrations vs. burNu ~

l are presented in the report. Some of these results were generated by GPU, while others were taken from the EPRI/Studsvik benchmarking.

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! The review and evaluation considered the information presented in the topical

! report and in References 1, 2, and 3. The latter references provide detailed descriptions of the EPRI-CPM and MICBURN codes, and the results of the

' EPRI/Studsvik benchmark calculations, respectively. Additional information

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provided by GPUN in Reference 4 was also considered.

The EPRI-CPM code represents a state-of-the-art calculational model for the analysis of BWR fuel assemblies. In contrast to a numberAf currently 4

approved methodologies where separate transport and diffusion theory codes are employed for the pin-cell and assembly calculations, respectively, all calculations in CPM are based on transport theory. In addition, since the pin-cell spectrum calculations (performed in order to spatially homogenize the cell and to condense the group cross sections) are perfonned in conjunction i with the assembly calculation, they account for the specific geometry of the fuel bundle. This integrated approach therefore offers advantages over some

standard techniques. Extensive benchmarking of CPM by EPRI and AB Atomenergi/Studsvik demonstrated the accuracy of the code and led to its use as a " synthetic experiment" against which the performance of other codes could be judged. A slightly modified version, CPM-2, is included in the currently recommended ARMP procedures for steady-state BWR analyses. It is i also significant that the fuel assembly burn-up program CASMO (Ref. 5) has been approved by the staff. CASMO shares many common elements with CPM and has been shown to produce identical (or very similar) results to those of j CPM for most LWR situations.

The report presents results (and references to other results) of EPRI benchmarking calculations including power distributions in hot critical experiments, criticality values in cold critical experiments and isotopic i values from fuel experiments in the Yankee and Saxton reactors. These, and

! other benchmarking calculations by EPRI demonstrate that CPM provides an

! acceptable methodology for calculations in its prescribed areas. This review,

' as well as previous reviews of the code by BNL, has concluded that CPM is an l

acceptable methodology and code for lattice physics calculations.

The report f and Reference 4) also presents results of GPUN benchmarking calculations to demonstrate that GPUN has the ability to use the code properly as well as to further demonstrate the capability of the code to provide acceptable results. The comparisons presented include calculations for Oyster Creek and Hatch-1 power distribution measurements, and for Monte Carlo calculations (by Exxon) of Oyster Creek fuel assemblies providing k-infinity and power distribution values. The review of these benchmarking calculations (see Reference 4) has concluded that suitable comparisons were made and that there is a satisfactory agreement between the GPUN CPM calculation results and i the measurements or higher order calculations, and that GPUN has therefore demonstrated an acceptable ability to use the code.

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4.0 CONCLUSION

S The staff, with the assistance of consultants from BNL, has reviewed the GPUN topical report TR 020, Revision 0, submitted by GPUN to describe and justify the methodology to be used in licensing calculations involving lattice physics characteristics. The review evaluated the methodology and the ability of GPUN to use the methodology. Based on this review, the staff concludes that the CPM code as used by GPUN is acceptable for applicable BWR licensing calculations.

5.0 REFERENCES

1. "The Collision Probability Module EPRI-CPM," Advanced Recycle Methodology Program System Documentation, CCM-3, Part II.2, Chapter 6, Electric Power Research Institute, November 1975.

?. "MIC8 URN-Microscopic Burnup in Gadolinium Fuel Pins," Advanced Recycle Methodology Program System Documentation, CCM-3, Part II.7, Chapter 7, Electric Power Research Institute, November 1975.

3. "EPRI CPM Benchmarking," Advanced Recycle Methodology Program System Documentation, CCM-3, Part 1.1, Chapter 5 Electric Power Research Institute, November 1975.

4. Letter from R. F. Wilson (GPUN) to J. N. Donohew, Jr., (NRC) dated July 16,1986, " Reload Topical Report 020."

5. A. Ahlin, et al., "CASM0: A Fuel Assembly Burnu (Rev. Ed.), Studsvik Energiteknik AB (June 1978)p Program," AE-RF-76-4158

6. Letter from R. F. Wilson (GPUN) to J. Zwolinski, (NRC) dated November 25, 1985, " Reload Topical Report."

Principal Contributor: H. Richings Dated: November 14, 1986.

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