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Title: MICROSHIELD ANALYSIS TO CALCULATE EXTERNAL RADIATION DOSE RATES FOR SEVERAL SPENT FUEL CASKS.

Abstract

Abstract not provided.

Authors:
; ;
Publication Date:
Research Org.:
Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
1267224
Report Number(s):
SAND2007-0589C
524063
DOE Contract Number:
AC04-94AL85000
Resource Type:
Conference
Resource Relation:
Conference: Proposed for presentation at the Waste Management Conference held February 25 - March 1, 2007 in Tucson, AZ.
Country of Publication:
United States
Language:
English

Citation Formats

Weiner, Ruth F., Osborn, Douglas., and Marincel, Michelle K.. MICROSHIELD ANALYSIS TO CALCULATE EXTERNAL RADIATION DOSE RATES FOR SEVERAL SPENT FUEL CASKS.. United States: N. p., 2007. Web.
Weiner, Ruth F., Osborn, Douglas., & Marincel, Michelle K.. MICROSHIELD ANALYSIS TO CALCULATE EXTERNAL RADIATION DOSE RATES FOR SEVERAL SPENT FUEL CASKS.. United States.
Weiner, Ruth F., Osborn, Douglas., and Marincel, Michelle K.. Mon . "MICROSHIELD ANALYSIS TO CALCULATE EXTERNAL RADIATION DOSE RATES FOR SEVERAL SPENT FUEL CASKS.". United States. doi:. https://www.osti.gov/servlets/purl/1267224.
@article{osti_1267224,
title = {MICROSHIELD ANALYSIS TO CALCULATE EXTERNAL RADIATION DOSE RATES FOR SEVERAL SPENT FUEL CASKS.},
author = {Weiner, Ruth F. and Osborn, Douglas. and Marincel, Michelle K.},
abstractNote = {Abstract not provided.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 2007},
month = {Mon Jan 01 00:00:00 EST 2007}
}

Conference:
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  • The purpose of this MicroShield analysis is to calculate the external radiation, primarily gamma, dose rate for spent fuel casks. The reason for making this calculation is that currently all analyses of transportation risk assume that this external dose rate is the maximum allowed by regulation, 10 mrem/hr at 2 m from the casks, and the risks of incident-free transportation are thus always overestimated to an unknown extent. In order to do this, the program by Grove Software, MicroShield 7.01, was used to model three Nuclear Regulatory Commission (NRC) approved casks: HI-STAR 100, GA-4, and NAC-STC, loaded with specific sourcemore » material. Dimensions were obtained from NUREG/CR-6672 and the Certificates of Compliance for each respective cask. Detectors were placed at the axial point at 1 m and 2 m from the outer gamma shielding of the casks. In the April 8, 2004 publication of the Federal Register, a notice of intent to prepare a Supplemental Yucca Mountain Environmental Impact Statement (DOE/EIS-0250F-S1) was published by the Office of Civilian Radioactive Waste Management (OCRWM) in order to consider design, construction, operation, and transportation of spent nuclear fuel to the Yucca Mountain repository [1]. These more accurate estimates of the external dose rates could be used in order to provide a more risk-informed analysis. (authors)« less
  • Version 00 QBF calculates and plots in a short running time, three dimensional radiation dose rate distributions in the form of contour maps on specified planes resulting from cylindrical sources loaded into vehicles or ships. Shielding effects by steel walls and shielding material layers are taken into account in addition to the shadow effect among casks. This code system identifies the critical points on which to focus when designing the radiation shielding structure and where each of the spent fuel shipping casks should be stored. The code GRAPH reads the output data file of QBF and plots it using themore » HGX graphics library. QBF unifies the functions of the SMART and MANYCASK codes included in CCC-482.« less
  • From 1984 through 1989, the Pacific Northwest Laboratory was involved in performance testing (heat transfer and shielding) spent-fuel metal storage casks and horizontal concrete storage modules. Gamma and neutron dose rates were measured at the surfaces and near the casks/modules during the performance tests. The purpose of this paper is to combine the dose rate data and summarize the results and conclusions of the total data base generated during the cask/module performance tests. The paper identifies metal spent-fuel storage casks that were performance tested under US Department of Energy (DOE), Electric Power Research Institute (EPRI), and Virginia Power sponsorship; themore » CASTOR-1C cask tested in the Federal Republic of Germany; and horizontal concrete modules tested by Carolina Power and Light, DOE, and EPRI.« less
  • Neutron and gamma dose rates from typical rail and truck spent fuel transport casks are reported for a variety of spent PWR fuel sources and cask conditions. The IF 300 rail cask and NLI 1/2 truck cask were selected for use as appropriate cask models. All calculations (cross section preparation, generation of spent fuel source terms, radiation transport calculations, and dose evaluation) were performed using various modules of the SCALE computational system. Conditions or parameters for which there were variations between cases include: detector distance from cask, spent fuel cooling time, the setting of fuel or neutron shielding cavities tomore » either wet or dry, the cobalt content of assembly materials, normal fuel assemblies and consolidated cannisters, the geometry mesh interval size, and the order of the angular quadrature set. 13 refs., 6 figs., 9 tabs.« less