ML20206R853
| ML20206R853 | |
| Person / Time | |
|---|---|
| Issue date: | 05/10/1998 |
| From: | Delligatti M, Michael Waters, Withee C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML20206R845 | List: |
| References | |
| NUDOCS 9905210003 | |
| Download: ML20206R853 (9) | |
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6th International Conference on Nuclear Engineering ICONE-6 May 10-15,1998 Copyright @ 1998 ASME RECONSIDERATION OF DOSE ASSESSMENTS FOR FUTURE INDEPENDENT SPENT FUEL STORAGE INSTALLATION MULT!-ROW CASK ARRAYS l M. D. Waters, C. J. Withee, M. S. Delligatti Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards United States Nuclear Regulatory Commission Washington, DC 20555-0001 ABSTRACT The U. S. Nuclear Regulatory Commission (NRC) examined the methods used by independent spent fuel storage installation (ISFSI) licensees and applicants to calculate doses from multi-row cask arrays and found that some methods may not be adequate for demonstrating regulatory compliance for future ISFSI designs. Future ISFSIs are anticipated to be different from current ISFSis because the design-basis dose rates could approach regulatory dose limits as a result of larger storage arrays and/or shorter site boundary distances.
NRC requires licensees to perform both dose calculations and dose measurements to show compliance with the dose limits in Parts 20 and 72 of Title 10 of the Code of Federal l Reaulations (10 CFR Parts 20 and 72). Simplified modeling assumptions can affect the accuracy of calculated dose rates. NRC has accepted these approximations in calculations by current licensees partly because the maximum potential dose rates are generally more than an )
order of magnitude below the regulatory limits and because monitoring is performed during ISFSI operation. Calculations performed by some licensees incorporated the assumption that l the outer row of casks completely shields the radiation flux from the side of the inner cask rows.
NRC used MCBEND, a three-dimensional radiation transport Monte Carlo program, to analyze shielding interactions among casks in multi-row cask arrays. The analyses indicate that the side flux from the inner cask rows is only partially shielded by the outer row casks and can make a significant contribution to the calculated dose. The analyses also indicate that multi-row cask arrays exhibit dose peaking at certain angular positions relative to the array as a result of unshielded radiation from the inner cask rows. The partial shielding effects and the angular ,
dose peaks are unique to each ISFSI design and depend on cask design, cask array geometry, I source spectrum, and site boundary distances.
NRC expects more conservative and/or more accurate assumptions for dose calculations presented in future site-specific safety analysis reports and general licensee dose evaluations.
NRC may place guidance in its standard review plans and inspection manuals for use by NRC reviewers and inspectors in performing these safety reviews. This guidance will describe to NRC staff acceptable dose calculation methods that either assume no shielding credit from outer row casks, or that use a methodology that conservatively accounts for the angular variation in radiation dose rates.
9905210003 990329 C PDR ORG NECCN PDR The copyright law provides that no copyright exists in works prepared by an officer or employee of the U.S. Govemment as part of his or her cificial duties.
99057/0003
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- 1. BACKGROUND INFORMATION ON DRY CASK STORAGE An independent spent fuel storage installation (ISFSI) is a facility designed and constructed for the interim storage of spent nuclear fuel and associated radioactive materials. A typical ISFSI consists of dry storage casks placed in a 3.7- to 5.5-m (12- to 18-ft) pitched array on a reinforced concrete pad. The dry storage cask is engineered to provide radiological shielding, criticality control, source confinement, physical protection, and past.ive cooling of spent nuclear fuel during normal, off-normal, and accident conditions. Most U.S. dry storage cask designs consist of either multi-walled cylindrical casks or canister-module configurations. A typical multi-walled cylindrical cask contains 20 to 38 cm (8 to 15 inches) of steel or 8 to 13 cm (3 to 5 inches) of lead for gamma shielding and 10 to 15 cm (4 to 6 inches) of borated hydrogenous material for neutron shielding. The canister-module configuration typically consists of a sealed, transferable metal canister that is placed either horizontally or vertically into a concrete or metal module with a wall thickness of 60 to 90 cm (2 to 3 feet).
Two ISFSI license options are available under 10 CFR Part 72 (1998). Utilities and other entities may apply for a site-specific license that employs a dry storage cask designed for a specific site and specific fuel assemblies. A utility that has a 10 CFR Part 50 license also has the option to use a general license to store fuel in a dry storage cask, the design of which has been approved and has received a certificate of compliance (COC) from the NRC. The NRC issues COCs for cask designs that meet applicable 10 CFR Part 72 requirements. At this time, there are 10 operating ISFSis in the United States that use one or more of 11 NRC-approved dry storage cask designs. On the basis of spent fuel pool storage capacity data, the NRC staff expects that approximately 15 to 25 additional power reactor licensees could require storage capabilities extemal to their pools in the next 10 years (Department of Energy,1995). Several nuclear utilities plan to construct ISFSis and use onsite dry cask storage systems to meet this need. In addition, the U.S. Department of Energy (DOE) has submitted a topical report on a plan to store spent fuel from multiple utilities at a temporary centralized interim storage facility if pending legislation is enacted (1997). Private Fuel Storage, L.L.C., has submitted an application to store spent fuel from eight utilities at an away-from-reactor ISFSI site (1997).
- 11. CURRENT OFFSITE DOSE REGULATIONS The regulations that limit offsite doses for ISFSis are 10 CFR 72.104 and 10 CFR 20.1301 (1998). The requirements in 10 CFR 72.104 specify that the annual dose equivalent to any individual who is located beyond the controlled area must not exceed 0.25 mSv (25 mrem) to the whole body,0.75 mSv (75 mrem) to the thyroid, and 0.25 mSv (25 mrem) to any other organ. The requirements in 10 CFR 20.1301 apply to all NRC licensees and specify that the total effective dose equivalent to individual members of the public does not exceed 1.0 mSv (100 mrem) in a year. In addition, it requires that the dose in any unrestricted area from extemal sources does not exceed 0.02 mSv (2 mrem) in any hour. 1 i
The calculated dose rates for each operating ISFSI site range between one to five orders of l magnitude below the 0.25-mSv/yr (25-mrem /yr) limit. These ISFSis have been designed to i store between 8 to 120 dry storage casks, with the nearest person at a distance between 425 m (1400 ft) to approximately 2400 m (8000 ft). These relatively large distances from the public ;
and small-array designs facilitate compliance with the limit. The NRC has examined the ;
potential for the overall trend of estimated dose rates to increase as new ISFSI sites are 1 constructed to meet utility needs. A relative increase in offsite doses for new ISFSI sites could
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' be caused by three primary factors: (1) dry storage cask designs that have less shielding and/or higher spent fuel source terms, (2) ISFSI sites that have a relatively higher number of casks, I and (3) the need for shorter distances between the ISFSI and the site boundary. I For example, the centralized interim storage facility planned by the DOE is designed to store between 5,300 to 7,800 dry storage casks and the private fuel storage facility is designed to store approximately 4,000 dry storage casks. In addition, future ISFSI site boundaries could be as close as 100 m (328 ft) and meet boundary requirements in 10 CFR 72.106. l lit. OFFSITE DOSE CALCULATIONS The general and site-specific licenseos must demonstrate during licensing proceedings that offsite dose rates will not exceed regulatory limits. Vendors who apply for a COC must submit calculations for a generic lSFSI array to demonstrate that regulatory compliance is achievable.
A utility using an approved cask under the general license must perform site-specific calculations that assess the dose to members of the public and demonstrate regulatory compliance before fuel loading. The applicant for a site-specific license must submit bounding offsite dose calculations that assess the dose to members of the public and demonstrate regulatory compliance before license approval. Both the peneral and the site-specific license holders must verify compliance by radiation monitoring during operation.
License holders and COC applicants have treated offsite dose calculations as simple source problems and used radiation transport computer programs that simulate the transmission of both direct-beam and air-scattered radiation. The SKYSHINE series of programs have been the predominant programs used by dry storage cask designers and users. The SKYSHINE series includes SKYSHINE II, SKYSHINE Ill, and MicroSkyshine. Other codes that have been used include ANISIN and RA" KERN. The SKYSHINE programs model the transmission of gamma radiation from multip.e ooint sources through an infinite medium of air. SKYSHINE 11 and lil also have the ability to n.Nel neutron transmission. The SKYSHINE air-scattered calculations employ a pre-calculated database of energy and angle-dependent coefficients of air-scattered dose rates from mono-directional point sources. The database of coefficients is typically calculated from curve fitting procedures applied to several Monte Carlo air-scattering computations (Lampley,1981).
To calculate offsite dose rates with SKYSHINE, the surface fluxes on the side and top of the dry storage cask are converted into equivalent point sources. A separate radiation shielding transport program is typically used to calculate these radial and axial source components.
SKYSHINE uses these source components to calculate offsite doses at specified distances.
The total dose contribution from a single dry storage cask is the sum of the SKYSHINE values calculated for the radial and axial flux components.
IV. MULTI-ROW CASK ARRAY PROBLEM Problem The cumulative dose from a multi-cask array ISFSI is typically determined by multiplying the ;
calculated radial and axial dose values by the number of casks that contribute to the offsite l dose rate for each respective component. Examination of several NRC-approved cask designs l and ISFSI applications reveals that some past offsite dose calculations have assumed that the side flux components from inner-row casks are completely shielded by the outside row of casks
in the array. As a result, these radial flux components were not included in the overall dose values calculated in some ISFSI and vendor applications.
Analysis To assess the impact of this assumption, ground-level gamma dose rates generated by the side flux component of an inner-row cask were investigated at angular intervals along a 250-m (820-ft) circumference around a representative cask array. The analysis considered air-scattered radiation around, and direct-beam Gamma radiation streaming between, an outer row of casks.
A previous scoping analysis showed that radiation reflection off other casks in the same representative array is not significant with respect to the dose rate at 250 m (820 ft). Therefore, radiation scattering off adjacent casks was ignored in the following analysis in order to conserve computer run time.
The cask array model consisted of nine casks in a three-by-three array with a 5.5-m (18-ft) center-to-center pitch, as shown in Figure 1. A standard air volume 5,000 m (16,400 ft) in height encompassed the array. The cask design consisted of a vertical metal cask that is 5.2 m (17 ft) in length and 2.4 m (8.0 ft) in outside diameter, as shown in Figure 2. The radial wall of the cask consisted of approximately 20 cm (8 in) of carbon steel and 10 cm (4 in) of a borated polymer. The center cask in the array represented the inner-row cask for the analysis and contained a volumetric gamma source equivalent to 24 pressurized-water reactor spent fuel assemblies with 40,000 MWD /MTU bumup and a 5-year cooling time. This gamma source results in a midplane surface dose rate of approximately 1.0 mSv/hr (100 mrem /hr). The surrounding eight casks in the array were modeled with the same outer cask dimensions, but as perfect absorbers.
The gamma dose rates were calculated with MCBEND, a three-dimensional Monte Carlo radiation transport program. MCBEND uses point-energy photon cross section libraries and contains splitting and Russian-roulette reduction methods that increase Monte Carlo efficiency.
The use of MCBEND for dry cask shielding and skyshine problems has been supported by several comparison and benchmark studies (Wouters,1996).
Dose rates were calculated in an annulus volume at a 250-m (820-ft) radius from the center cask in the array. The annulus was 1.8 m (5.9 ft) in height and had an inner and outer radius of 249 and 251 ra (817 and 823 ft), respectively. The annulus was divided into 10-degree
' increments from 0 to 90 degrees, and the radiation flux was tallied by MCBEND in each incremental volume. A second calculation was performed for a single, stand-alone cask at 250 m (820 ft) by removing the surrounding casks from the model.
PLACE FIG.1 HERE PLACE FIG.2 HERE Results and Findinas
- A history of over 150 million particles was tracked to generate the angular variation of radiation dose rates shown in Figure 3. The plotted angles represent the center of each incremental volume. The dose rates ranged from 0.014 mSv/yr,(1.4 mrem /yr) to 0.085 mSv/yr (8.5 mrem /yr) and have an average statistical uncertainty of 12 percent. The dose rate peaks at 25 and 65 degrees indicate significant dose contribution from unshielded, direct line-of-sight radiation streaming between the outer row of casks. The lower dose rates at 5,45, and 85
degrees indicate the shielding effect of the outer casks. The stand-alone single cask had an average dose of 0.14 mSv/yr (14 mrem /yr).
PLACE FIG.3 HERE V. CONCLUSION l The results show that the side flux component of a typical inner-row cask is not completely shielded. Therefore, the side flux component from casks in the second row of a multi-row array can contribute to the offsite dose at particular angles. The degree of impact from the side )
components of inner-row casks is site-specific and depends on factors such as cask !
dimensions, ISFSI array dimensions, and offsite distances. However, examination of offsite dose calculations performed by current site-specific ISFSI licensees indicates that a calculated i increase in offsite dose resulting from inclusion of the second row of casks in the calculations is not significant with respect to regulatory dose limits at current operating sites. This is due to the relatively large site-boundary distances and small array sizes. Therefore, licensing conclusions j for current ISFSI sites remain unchanged. In addition, calculated doses are verified by radiation l monitoring at the ISFSI site during operation.
The NRC expects future calculations to either assume no shielding credit from outer row casks i or use a methodology that conservatively accounts for the angular variation in radiation dose rates. These assumptions will be important if new ISFSI designs cause calculated dose values to approach regulatory limits. The NRC staff plans to incorporate guidance for its staff in the
" Standard Review Plan for Dry Cask Storage Systems" (U.S. NRC,1997) and the " Standard Review Plan for Spent Fuel Dry Storage Facilities" (U.S. NRC,1996). These documents are guidance used by the staff to review dry cask storage design applications and site-specific ISFSI applications. The NRC staff also plans to incorporate information into future inspection procedures that provide guidance for inspection of the offsite dose evaluations by general l licensees.
VI. REFERENCES l
- 1. Title 10, Code of Federal Reaulations. Part 72, " Licensing Requirements for the i Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste," Office of the Federal Register, Washington, D.C.
- 2. Department of Energy, " Spent Fuel Storage Requirements 1994-2042," DOE /RW-0431-Rev.1, June 1995.
- 3. Department of Energy, " Centralized Interim Storage Facility Topical Safety Analysis Report," Revision 0, submitted to the U.S. Nuclear Regulatory Commission on May 1, 1997, Docket 72-21.
- 4. Private Fuel Storage, L.L.C., " Safety Analysis Report - Private Fuel Storage Facility" Revision 0, submitted to the U.S. Nuclear Regulatory Commission on June 20,1997, Docket 72-22.
- 5. Title 10, Code of Federal Reaulations. Part 20, " Standards for Protection Against Radiation," Office of tb Federal Registrar, Washington, D.C.
- 6. Lampley, C.M., "The SKYSHINE-Il Procedure: Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray, and Secondary Gamma-Ray Dose Rates in Air," NUREG/CR-0781, January 1981.
- 7. Wouters R. and Quoidbach A.,"Skyshine Analysis of a Spent Fuel Dry Storage Facility with MCBEND." Proceedings of the 1996 Topical Meeting for Radiation Protection and Shielding - Advances and Applications in Radiation Protection and Shielding. April 21-25,1996.
- 8. U.S. Nuclear Regulatory Commission, " Standard Review Plan for Dry Cask Storage Systems," NUREG-1536, January 1997.
- 9. U.S. Nuclear Regulatory Commission, " Draft Standard Review Plan for Spent Fuel Dry Storage Facilities," NUREG-1567, October 1996.
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