J. M. East and K. R. O’Kula
Westinghouse Safety Management Solutions LLC
Aiken, SC 29804-5388

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Accident analysis for U.S. Department of Energy (DOE) nuclear facilities is an integral part of the overall safety basis developed by the contractor to demonstrate facility operation can be conducted safely. An appropriate documented safety analysis for a facility discusses accident phenomenology, quantifies source terms arising from postulated process upset conditions, and applies a standardized, internationally-recognized database of dose conversion factors (DCFs) to evaluate radiological conditions to offsite receptors.

Until recently, most DOE accident analyses have applied DCFs based on Federal Guidance Report 11 (FGR 11). In general, the latter contains dose per unit intake values that are numerically identical to recommendations and limits published by the International Commission on Radiological Protection (ICRP) in ICRP Publications 26, 30 and 48 (plutonium and related elements). However, revised metabolic models for uptake of radioactivity, and associated DCFs have been recently been published in ICRP Publications 68 and 72.

This paper is a preliminary assessment of the impact of the new ICRP guidance in the context of accident analysis. Prototypic high-level waste source terms are used as the basis for assessing the impacts of the new DCF recommendations.

In this study, total effective dose equivalents (TEDEs) using two sets of inhalation committed effective dose equivalent (CEDE) DCFs are used with prototypic source terms from high-level waste accident analyses. The first set applies the inhalation CEDE DCFs from FGR 11, and the second set is based on ICRP Publications 68 and 72 with the added capability of incorporating dose coefficients for ten particle sizes and ten commitment periods.

The study examined two generic release scenarios. The first scenario is a release to the environment after passing through a filtration system (filtered release). In this scenario, only the change in particle size is considered. The reduction in source material due to filtration is not credited. The filtered particulates are assumed to have a deposition velocity of 0.001 m/s. This dry deposition velocity corresponds to a particle with an approximate aerodynamic equivalent diameter (AED) of 0.2 mm to 0.4 mm based on Savannah River Site (SRS) parameters. The second release scenario is a release to the environment without filtration (unfiltered release). The unfiltered particulates are assumed to have a deposition velocity of 0.01 m/s. This dry deposition velocity corresponds to a particle with an approximate AED of 2 mm to 4 mm again based on SRS parameters.

The FGR 11 methodology uses the same DCF file for all ages in the general public population and for all release conditions. All particles are assumed to have a size of 1.0 mm activity median aerodynamic diameter (AMAD). Since the ICRP Dose Coefficient Database includes inhalation CEDE DCFs for a range of particle sizes, DCF files in the ICRP 68/72 methodology were tailored according to filtration release conditions. Thus, for the unfiltered release a 1 mm AMAD particle is assumed, and for the filtered release, a 0.1 mm AMAD particle is used.

Three HLW source terms are processed under both dose evaluation methodologies. The first is dominated by plutonium and cesium radionuclides. The second source term contains one-third to one-fourth the concentration of plutonium and cesium radionuclides compared to the first, and the third is dominated by americium and curium constituents. The resulting dose due to these source terms is largely due to plutonium.

In the FGR 11 analysis, plutonium is taken to have a lung clearance class of years (Y), which is based on the assumption of the material being in the oxide form. In ICRP Publication 68, a distinction is made between the insoluble and soluble plutonium oxide. Insoluble plutonium is assigned to lung absorption type medium (roughly corresponding to lung clearance class weeks or W) while the soluble plutonium is assigned to lung absorption type slow (roughly corresponding to lung clearance class Y). Soluble plutonium (the lower DCF) is applicable only with firing the plutonium media under very high temperatures. Therefore, for the ICRP 68/72 analysis, the plutonium form is assumed to have a lung absorption type medium (M). For all other radionuclides with more than one lung clearance class or lung absorption type, the highest inhalation CEDE DCF from the oxide and nitrate compounds is selected.

The MACCS computer code, Version 1.5.11.1, is the software applied to process the source term inputs. MACCS is used in a Latin Hypercube Sampling (LHS) mode to sample hourly, SRS-specific meteorological records from a five-year quality-assured database. The meteorological data file and a spatial grid, configured for sufficient resolution and containing evaluation points at specific distances of interest, are evaluated as inputs. The calculated dose is the 95^th percentile, direction-independent dose at the closest boundary distance from the closest point in a group of facilities, to the site boundary.

For cases 1 and 2 that are dominated by plutonium, the TEDEs for filtered release cases calculated with ICRP 68/72 DCFs are higher (5% to 35%) than those calculated with FGR 11 DCFs. However, the ICRP 68/72 doses for the same case are lower (~50%) for unfiltered releases. For the americium-curium dominated source term, the ICRP 68/72 TEDEs are 15% to 60% less than FGR 11 doses, for both filtered and unfiltered cases, respectively.

This analysis, although preliminary, indicates that changes in dose conversion factors alone will not necessarily result in appreciable changes to the consequences calculated in the context of accident analyses. Other important factors include accident phenomenology, filtration assumptions, and environmental modeling.