Benchmarked analyses of neutron and secondary gamma rays for a spent-fuel shipping cask using MORSE-CGA-PC and the DABL69 cross sections
- United Engineers and Constructors Inc., Philadelphia, PA (United States)
The potential neutron radiation dose rates around a spent-fuel shipping cask are a fundamental consideration for the design of such a cask. Neutron radiation from spent fuel is predominantly from decay of {sup 242}Cm and{sup 244}Cm. Secondary gamma can be produced during neutron transport. The combination of the MORSE-SGC computer program and a 32-neutron-group, 18-gamma-group, P5 library derived from ENDF/B-IV could reasonably predict the neutron transport and secondary gamma production and transport, based on actual experimental data. The experimental cask was designed to contain three pressurized water reactor (PWR) or seven boiling water reactor fuel assemblies. The neutron source employed in the experiment was {sup 252}Cf, whose spectrum is similar to the fission spectrum expected from PWR spent fuel. This paper discusses analyses of the same experimental data using the MORSE-CGA program, the DABL69 cross-section set, and a surface crossing estimator. The DLC-130/DABL69 cross-section was employed as the most suitable, readily available, broad-group library. The DABL69 contains 46-neutron and 23-gamma-ray energy groups and the Legendre expansion coefficient for angular distribution is 5(P5). Furthermore, this analysis was done using a personal computer (PC) version of MORSE-CGA, with runs taken to the point of reducing fractional standard deviations to 10% or less. The purpose of this paper is to compare calculated radiation dose rate with available measured data.
- OSTI ID:
- 7036525
- Report Number(s):
- CONF-920606-; CODEN: TANSA
- Journal Information:
- Transactions of the American Nuclear Society; (United States), Vol. 65; Conference: American Nuclear Society annual meeting, Boston, MA (United States), 7-12 Jun 1992; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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SPENT FUEL CASKS
SHIELDING
ACCURACY
BENCHMARKS
CALIFORNIUM 252
COMPUTER CODES
COMPUTERIZED SIMULATION
CONCRETES
CROSS SECTIONS
DESIGN
EFFICIENCY
M CODES
MONTE CARLO METHOD
RADIATION DOSES
SPENT FUELS
STEELS
THERMAL NEUTRONS
THREE-DIMENSIONAL CALCULATIONS
ACTINIDE ISOTOPES
ACTINIDE NUCLEI
ALLOYS
ALPHA DECAY RADIOISOTOPES
BARYONS
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CALIFORNIUM ISOTOPES
CASKS
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ELEMENTARY PARTICLES
ENERGY SOURCES
EVEN-EVEN NUCLEI
FERMIONS
FUELS
HADRONS
HEAVY NUCLEI
IRON ALLOYS
IRON BASE ALLOYS
ISOTOPES
MATERIALS
NEUTRONS
NUCLEAR FUELS
NUCLEI
NUCLEONS
RADIOISOTOPES
REACTOR MATERIALS
SIMULATION
SPONTANEOUS FISSION RADIOISOTOPES
YEARS LIVING RADIOISOTOPES
420204* - Engineering- Shipping Containers