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Title: Radiation dose rate distributions of spent fuel dry casks estimated with MAVRIC based on detailed geometry and continuous-energy models

Journal Article · · Annals of Nuclear Energy (Oxford)
 [1];  [1];  [1];  [1];  [1];  [1]
  1. Univ. of Florida, Gainesville, FL (United States)
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This study presents a detailed comparison of the dose rate distributions of a dry TN–32 fuel cask with two geometry models and two cross-sectional datasets. The accuracies of radiation dose rate estimation and computational efficiencies of each geometry model with two cross-sectional data-sets are compared. The use of automated variance reduction techniques can significantly improve the computational efficiency of such a realistic, deep penetration problem that involves radiation transport from different volumetric sources, thereby eliciting only a small statistical error. Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) is a computational sequence within the SCALE 6.2 code package based on consistent adjoint driven importance sampling (CADIS), a type of automated variance reduction technique. Homogenous and full fuel assembly models are built herein, and two nuclear cross-section libraries (V7–200N47G and continuous energy) are applied in this work. Based on the detailed comparisons, we found that neutron dose rate estimation is more dependent on geometry modeling than on cross-section data. For neutron-induced gamma rays, the dose rate distribution depends on both the spatial self-shielding effect and the cross-section library. The primary gamma rays respectively contribute to the total dose rate by ~91% and ~ 99% on side and top surfaces, and the dose rate accuracy is more dependent on the cross-section library than on geometry modeling. In terms of the computation efficiency and efforts spent on geometry modeling, the homogenous fuel assembly model with the MG library can produce an acceptable dose rate distribution, but the detailed fuel assembly model with the continuous-energy library is required for more precise dose rate estimation.

Research Organization:
Univ. of Michigan, Ann Arbor, MI (United States)
Sponsoring Organization:
USDOE National Nuclear Security Administration (NNSA), Office of Nonproliferation and Verification Research and Development (NA-22)
Grant/Contract Number:
NA0002534
OSTI ID:
1525310
Journal Information:
Annals of Nuclear Energy (Oxford), Vol. 117, Issue C; ISSN 0306-4549
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English
Citation Metrics:
Cited by: 9 works
Citation information provided by
Web of Science

References (7)

Automated Variance Reduction of Monte Carlo Shielding Calculations Using the Discrete Ordinates Adjoint Function journal February 1998
Monte Carlo variance reduction with deterministic importance functions journal January 2003
Monte Carlo Shielding Analysis Capabilities with MAVRIC journal May 2011
Shielding Benchmark Calculations with SCALE/MAVRIC and Comparison with Measurements for the German Cask CASTOR® HAW 20/28 CG journal December 2009
Surface Dose Rate Calculations of a Spent-Fuel Storage Cask by Using MAVRIC and Its Comparison with SAS4 and MCNP journal July 2011
Implementation of Surface Detector Option in Scale SAS4 Shielding Module journal March 2000
A comparison of dose rate calculations for a spent fuel storage cask by using MCNP and SAS4 journal December 2008

Cited By (1)

A Methodology for Optimizing the Management of Spent Fuel of Nuclear Power Plants Using Dry Storage Casks journal February 2019

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