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Title: Computed isotopic inventory and dose assessment for SRS fuel and target assemblies

Abstract

Past studies have identified and evaluated important radionuclide contributors to dose from reprocessed spent fuel sent to waste for Mark 16B and 22 fuel assemblies and for Mark 31 A and 31B target assemblies. Fission-product distributions after a 5- and 15-year decay time were calculated for a ``representative`` set of irradiation conditions (i.e., reactor power, irradiation time, and exposure) for each type of assembly. The numerical calculations were performed using the SHIELD/GLASS system of codes. The sludge and supernate source terms for dose were studied separately with the significant radionuclide contributors for each identified and evaluated. Dose analysis considered both inhalation and ingestion pathways: The inhalation pathway was analyzed for both evaporative and volatile releases. Analysis of evaporative releases utilized release fractions for the individual radionuclides as defined in the ICRP-30 by DOE guidance. A release fraction of unity was assumed for each radionuclide under volatile-type releases, which would encompass internally initiated events (e.g., fires, explosions), process-initiated events, and externally initiated events. Radionuclides which contributed at least 1% to the overall dose were designated as significant contributors. The present analysis extends and complements the past analyses through considering a broader spectrum of fuel types and a wider range of irradiationmore » conditions. The results provide for a more thorough understanding of the influences of fuel composition and irradiation parameters on fission product distributions (at 2 years or more). Additionally, the present work allows for a more comprehensive evaluation of radionuclide contributions to dose and an estimation of the variability in the radionuclide composition of the dose source term that results from the spent fuel sent to waste encompassing a broad spectrum of fuel compositions and irradiation conditions.« less

Authors:
 [1]; ;  [2]
  1. Westinghouse Savannah River Co., Aiken, SC (United States)
  2. SAIC (United States)
Publication Date:
Research Org.:
Westinghouse Savannah River Co., Aiken, SC (United States)
Sponsoring Org.:
USDOE, Washington, DC (United States)
OSTI Identifier:
149989
Report Number(s):
WSRC-TR-94-0456
ON: DE96003205; TRN: 96:001411
DOE Contract Number:
AC09-89SR18035
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: 19 Jun 1995
Country of Publication:
United States
Language:
English
Subject:
05 NUCLEAR FUELS; 40 CHEMISTRY; SAVANNAH RIVER PLANT; SPENT FUELS; RADIOACTIVE WASTES; REPROCESSING; SLUDGES; RADIOCHEMICAL ANALYSIS; TARGETS; FISSION PRODUCTS; RADIOISOTOPES; DISTRIBUTION FUNCTIONS; RADIOACTIVE WASTE STORAGE; ACTINIDES; S CODES; G CODES; RADIATION DOSES; INGESTION; INHALATION; RADIONUCLIDE KINETICS; ENVIRONMENTAL EXPOSURE PATHWAY; TANKS; INVENTORIES

Citation Formats

Chandler, M.C., Ketusky, E.T., and Thoman, D.C. Computed isotopic inventory and dose assessment for SRS fuel and target assemblies. United States: N. p., 1995. Web. doi:10.2172/149989.
Chandler, M.C., Ketusky, E.T., & Thoman, D.C. Computed isotopic inventory and dose assessment for SRS fuel and target assemblies. United States. doi:10.2172/149989.
Chandler, M.C., Ketusky, E.T., and Thoman, D.C. Mon . "Computed isotopic inventory and dose assessment for SRS fuel and target assemblies". United States. doi:10.2172/149989. https://www.osti.gov/servlets/purl/149989.
@article{osti_149989,
title = {Computed isotopic inventory and dose assessment for SRS fuel and target assemblies},
author = {Chandler, M.C. and Ketusky, E.T. and Thoman, D.C.},
abstractNote = {Past studies have identified and evaluated important radionuclide contributors to dose from reprocessed spent fuel sent to waste for Mark 16B and 22 fuel assemblies and for Mark 31 A and 31B target assemblies. Fission-product distributions after a 5- and 15-year decay time were calculated for a ``representative`` set of irradiation conditions (i.e., reactor power, irradiation time, and exposure) for each type of assembly. The numerical calculations were performed using the SHIELD/GLASS system of codes. The sludge and supernate source terms for dose were studied separately with the significant radionuclide contributors for each identified and evaluated. Dose analysis considered both inhalation and ingestion pathways: The inhalation pathway was analyzed for both evaporative and volatile releases. Analysis of evaporative releases utilized release fractions for the individual radionuclides as defined in the ICRP-30 by DOE guidance. A release fraction of unity was assumed for each radionuclide under volatile-type releases, which would encompass internally initiated events (e.g., fires, explosions), process-initiated events, and externally initiated events. Radionuclides which contributed at least 1% to the overall dose were designated as significant contributors. The present analysis extends and complements the past analyses through considering a broader spectrum of fuel types and a wider range of irradiation conditions. The results provide for a more thorough understanding of the influences of fuel composition and irradiation parameters on fission product distributions (at 2 years or more). Additionally, the present work allows for a more comprehensive evaluation of radionuclide contributions to dose and an estimation of the variability in the radionuclide composition of the dose source term that results from the spent fuel sent to waste encompassing a broad spectrum of fuel compositions and irradiation conditions.},
doi = {10.2172/149989},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jun 19 00:00:00 EDT 1995},
month = {Mon Jun 19 00:00:00 EDT 1995}
}

Technical Report:

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  • As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, aremore » not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.« less
  • The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2)more » Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO{sub 2} and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program equipment in the Savannah River Technology Center would need to be removed to accommodate pellet fabrication. This work would also be in a contaminated area.« less
  • The SAS2 control module and ORIGEN-S code of the SCALE code system have been used with the standard ORIGEN-S data libraries and the SCALE 27-group ENDF/B-V cross-section library to predict afterheat power and fuel inventories in uranium-fueled pressurized water reactor (PWR) fuel assemblies. Based on present comparisons with measured data and on previous experience, it is concluded that this combination of codes and data bases is properly qualified for the calculation of afterheat power in uranium-fueled PWR fuel assemblies. The prediction of fission product fuel inventory data for fuel samples from three PWR fuel assemblies appears to be adequate, butmore » the prediction of actinide inventory data is seen to be quite conservative with respect to measured data.« less
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