RADIATION ANALYSIS FOR A GENERIC CENTRALIZED INTERIM STORAGE FACILITY

Scott G. Gillespie TRW Environmental Safety Systems, Inc. 2650 Park Tower Drive Vienna, VA 22180 (703)205-3 846

Robert G. Eble Duke Engineering & Services, Inc. 400 S. Tryon Street Charlotte, NC 28202 (703)204-8 657

Patricia Lopez TRW Environmental Safety Systems, Inc. 2650 Park Tower Drive, Suite 800 Vienna, VA 22180 (703)204-8562

SUMMARY

INTRODUCTION

This paper documents the radiation analysis performed for the storage area of a generic Centralized Interim Storage Facility (CISF) for commercial spent nuclear fuel (SNF) to establish the CISF Protected Area, Unrestricted Area, and Restricted Area boundaries. In order to model a large (6000 cask) storage array with a reasonable amount of analysis time and effort, a simplified calculational model was developed for the CISF.

DESCRIPTION OF ANALYSLS

The CISF is designed to accommodate several different types of SNF storage systems. In order to simplrfy the calculation of dose rates from the storage area, the Westinghouse Large PWR Multi- Purpose Canister (MPC) is selected as a “representative” storage system, since sufficient information is contained in its Safety Analysis Report’ to allow accurate modeling, and the surface dose rates on the MPC either envelop or are consistent with other storage systems 25.

The source term, geometry, and shield compositions are taken from Reference 1. The storage area, as shown in Figure 1, consists of 6000 vertical storage casks placed on 40 concrete pads. The casks are 3.81 m (12.5 ft) in diameter and 5.5 m (18 ft) high. The QAD6 point kernel transport code is used to calculate gamma dose rates and fluxes from the storage mode, and the MCNP’ Monte Carlo transport code is used to calculate neutron dose rates and fluxes. Axial (top vent) fluxes are inputted into the &-scattered dose calucations. A simple geometric model is used to estimate the shielding effect of intervening cask rows in order to reduce the number of calculations required for the “back rows” of casks.

SYE

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government Neither the United States Government nor any agency thereof, nor any of their employees, make any warranty, express or implied, or assumes any legal liabili- ty or responsibility for the accuracy, completeness, or usefulness of any information, appa- ratus, product, o r process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessar- ily state or reflect those of the United States Government or any agency thereof.

The total calculated gamma dose rates from the storage array are shown in Table 1. Direct gamma dose rates are calculated using QAD, and air-scattered dose rates are calculated using the MicroSkyshine' code. The &-scattered neutron dose rates were determined by SKYSHINE-El? calculations to be negligible compared to gamma dose rates.

To establish the regulatory boundary distances for the CISF, the gamma dose rates from the storage array are compared to dose limits in 1OCFR72l1 for the Protected Area (25 mdyr), and 10CFR2012 for the Unrestricted Area (500 m/yr, or 0.25 mr/hr), and the Restricted Area (2 mr/hr). The results are:

0

0

0

Restricted Area Boundary: Unrestricted Area Boundary: Protected Area Boundary:

50 meters 200 meters 700 meters

To verify the conservatism of using the QAD-generated skyshine source term, an additional air- scattered gamma dose calculation is performed using an axial gamma source term calculated using MCNP. The total dose rates using the QAD model are slightly nonconservative (2-4 percent) close to the storage array, but are conservative by about 20 percent at larger distances.

CONCLUSIONS

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2.

3.

4.

For SNF storage arrays using current storage cask designs, neutron doses are negligible compared to gamma doses at large distances.

For large close-packed storage arrays, only the first few rows of casks contribute significantly to the gamma dose.

The air-scattered gamma dose component is only significant compared to the direct component at large distances from the storage array.

The QAD point-kernel technique, coupled with the MicroSkyshine line-beam response function approximation provides an adequate methodology for calculating gamma radiation doses from a large SNF storage may with a reasonable amount of effort.

REFERENCES

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2.

Westinghouse Government and Environmental Services Company, MPC Project, Large OST and OSS Safe0 Analysis Report, MPC-CD-02-016, Revision 1, June 1996

Portland General Electric Company, Trojan Independent Spent Fuel Storage Installation License Application, Revision 0, March 1996

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7.

8.

9.

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Pacific Nuclear Fuel Services, Inc., Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUH-003, Revision 2, November 1993

Nuclear Assurance Corporation, Safety Analysis Report for the NAC Storable Transport Cask, Revision 2, October 1993

Holtec International, lOCFR71 Safety Analysis Report for Packaging for the Holtec International Storage, Transport and Repository Cask System (HI-STAR 100 Cask System), Holtec Report HI-951251, Revision 0, January 1995

Oak Ridge National Laboratory, RSIC Computer Code Collection, QAD-CGGP, A Combinatorial Version of QAD-PSA, A Point Kernel Code System for Neutron and G a m - Ray Shielding Calculations Using the GP Buildup Factor, CCC-493

Oak Ridge National Laboratory, RSIC Computer Code Collection, MCNP 4A, Monte Carlo N-Particle Transport Code System, CCC-200

Grove Engineering, Inc., MicroSkyshine Manual, 1987 (Updated March 1994)

Oak Ridge National Laboratory, RSIC Computer Code Collection, SKYSHINE-III, Calculation of the Effects of Structure Design on Neutron, Primury G a m - R a y and Secondary G a m - R a y Dose Rates in Air, CCC-289, February 1982

Civilian Radioactive Waste Management System, Management and Operating Contractor, Interim Storage Facility Design Requirements Document, Revision 1 , January 1997

Code of Federal Regulations, Title 10, Part 20.105(a) and (b) - Permissible levels of radiation in unrestricted areas

Code of Federal Regulations, Title 10, Part 72.104(a) - Criteria for radioactive materials in effluents and direct radiation from an ISFSI or M R S

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Table 1. Gamma Dose Rates from CISF Storage Area

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Figure 1. CISF Storage Area Layout Model

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