Validation of Spent-Fuel Isotopics Predicted by the SCALE-4 Depletion Sequence

Conference · Sat Nov 09 23:00:00 EST 1991 · Transactions of the American Nuclear Society

OSTI ID:7117192

The Standardized Computer Analyses for Licensing Evaluation (SCALE) code system is used extensively to perform away-from-reactor safety analyses (particularly criticality safety, shielding, and heat transfer analyses) for spent light water reactor fuel. Spent-fuel characteristics, such as radiation sources, heat generation sources, and isotopic concentrations, can be computed within SCALE using the SAS2 control module. At user-defined time steps, the SAS2 sequence performs a radiation transport analysis (via XSDRNPM-S) to obtain appropriate cross sections and spectral parameters for an ORIGEN-S point-depletion analysis. Each ORIGEN-S case produces the burnup-dependent fuel composition to be used in the next spectral calculation. Thus, the burnup-dependent cross sections and spectral parameters generated for ORIGEN-S are dependent on the initial enrichment, specified power history, and assembly model input to SAS2. A final ORIGEN-S case is used to perform the complete depletion/decay analysis using the burnup-dependent cross sections developed by the iterative scheme. The latest version of the SAS2 control module as released with SCALE-4 is denoted SAS2H. The purpose of this paper is to report recent SAS2H/ORIGEN-S calculations that were performed to compare pressurized water reactor (PWR) spent-fuel isotopic concentrations with measured data determined by radiochemical analyses.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• OSTI ID: 7117192
• Report Number(s): CONF-911107--
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 64; ISSN TANSA; ISSN 0003-018X
• Publisher: American Nuclear Society
• Country of Publication: United States
• Language: English

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050900 -- Nuclear Fuels-- Transport
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11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100* -- Power Reactors
Nonbreeding
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Boiling Water Cooled
210200 -- Power Reactors
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42 ENGINEERING
420203 -- Engineering-- Handling Equipment & Procedures
97 MATHEMATICS AND COMPUTING
ACTINIDE ISO
ACTINIDE NUCLEI
ACTINIDES
ALPHA DECAY RADIOISOTOPES
AWAY-FROM-REACTOR STORAGE
BURNUP
BWR TYPE REACTORS
CALVERT CLIFFS-1 REACTOR
CALVERT CLIFFS-2 REACTOR
CHEMICAL ANALYSIS
COMPUTER CODES
COMPUTERIZED SIMULATION
CRITICALITY
CROSS SECTIONS
ELEMENTS
ENERGY SOURCES
ENERGY TRANSFER
ENRICHED URANIUM
ENRICHED URANIUM REACTORS
EVEN-ODD NUCLEI
FISSION PRODUCTS
FUELS
HEAT TRANSFER
HEAVY NUCLEI
INTERNAL CONVERSION RADIOISOTOPES
ISOMERIC TRANSITION ISOTOPES
ISOTOPE ENRICHED MATERIALS
ISOTOPES
ITERATIVE METHODS
MATERIALS
METALS
MINUTES LIVING RADIOISOTOPES
NUCLEAR FUELS
NUCLEI
Nuclear Criticality Safety Program (NCSP)
O CODES
OBRIGHEIM REACTOR
POST-IRRADIATION EXAMINATION
POWER REACTORS
PWR TYPE REACTORS
QUANTITATIVE CHEMICAL ANALYSIS
RADIATION SOURCES
RADIOACTIVE MATERIALS
RADIOASSAY
RADIOCHEMICAL ANALYSIS
RADIOISOTOPES
REACTOR MATERIALS
REACTORS
S CODES
SHIELDING
SIMULATION
SPENT FUEL STORAGE
SPENT FUELS
SPONTANEOUS FISSION RADIOISOTOPES
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URANIUM 235
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