ORIGEN2: a revised and updated version of the Oak Ridge isotope generation and depletion code

ORIGEN2 is a versatile point depletion and decay computer code for use in simulating nuclear fuel cycles and calculating the nuclide compositions of materials contained therein. This code represents a revision and update of the original ORIGEN computer code which has been distributed world-wide beginning in the early 1970s. The purpose of this report is to give a summary description of a revised and updated version of the original ORIGEN computer code, which has been designated ORIGEN2. A detailed description of the computer code ORIGEN2 is presented. The methods used by ORIGEN2 to solve the nuclear depletion and decay equations are included. Input information necessary to use ORIGEN2 that has not been documented in supporting reports is documented

A concept for increasing the effective capacity of a unit area of a geologic repository

By processing spent fuel to remove the actinides, the thermal properties of the resulting high-level waste are substantially altered. In particular, the {open_quotes}thermal half-life{close_quotes} of the waste is reduced from centuries to about 30 years. This paper evaluates a High-Efficiency Waste Emplacement Concept (HEWEC) that takes advantage of the decrease in thermal half-life. The HEWEC is based on the observation that the waste loading per unit area of a repository is potentially limited by maximum allowable temperatures at several locations: the waste package (very near field), the rock surrounding the package and emplacement drifts (near field), and the large bulk of surrounding rock (far field). The first two are controlled by decay heat generated within years or decades of waste emplacement, primarily resulting from the fission products but with significant contributions from actinides. Far-field temperatures are controlled by decay heat generated over centuries, primarily from the actinides. While the critical temperature limit for spent nuclear fuel typically occurs within the package, it is close to limits in all other locations. However, if spent fuel without actinides (i.e., high-level waste) is emplaced in the repository, far-field temperatures no longer approach the limits, and waste loading is restricted by temperatures in near-field and very-near-field locations. If the repository is fully ventilated during operation, sufficient total decay heat can be removed to allow significantly more waste to be loaded in a unit area without exceeding temperature limits. Evaluation of HEWEC is based on analysis of several existing thermomechanical studies. It appears possible to increase the equivalent amount of waste loaded in a unit area of a repository by about a factor of 4.7 and application of the HEWEC precepts to unreprocessed spent fuel is not as effective, potentially increasing repository loading by only a factor of 1.2

User's manual for the ORIGEN2 computer code

This report describes how to use a revised version of the ORIGEN computer code, designated ORIGEN2. Included are a description of the input data, input deck organization, and sample input and output. ORIGEN2 can be obtained from the Radiation Shielding Information Center at ORNL

Comparison of experimentally determined spent-fuel compositions with ORIGEN 2 calculations

The specific experimental measurements of interest here involve the determination of parameters related to the actinide and fission product composition of samples from five elements taken from fuel assemblies discharged from the Turkey Point Unit 3 PWR. Two fuel assemblies were obtained for the purposes of nondestructive and destructive assay. These assemblies were initially fueled with 448 kg of UO/sub 2/ enriched to 2.556 wt % /sup 235/U and then irradiated for 851 full-power days. Five elements were then removed, and assay samples were taken from each element near the core midplane. The relevant parameters measured were /sup 148/Nd//sup 238/U, /sup 239/Pu//sup 238/U, and the isotopic compositions of U, Pu, Kr, and Xe. Fuel depletion calculations were performed using the updated ORIGEN2 PWR model. The burnup of the fuel was determined by adjusting the ORIGEN2 fuel burnup to match the experimentally determined /sup 148/Nd//sup 238/U ratio for each fuel element. The resulting burnup was then used to calculate the other experimentally determined parameters listed above. The agreement between ORIGEN2 and the experimental results is very good, with the average error for five samples being < 4% for most parameters. Based on this comparison, it appears that the ORIGEN2 computer code is capable of accurately calculating the composition of irradiated fuel from a modern PWR. However, well-characterized experimental measurements should continue to be obtained for validation purposes because the calculated values of many nuclides, particularly the minor actinides, still have significant uncertainties

Potential impact of ICRP-30 on the calculated risk from waste repositories

As a result of the large body of information that has been gathered since ICRP-2 was published (1959), the ICRP has undertaken the task of updating its radiation protection guidance. This update involves revision of the primary radiation guidance as well as the recalculation of intake limits (ICRP-30) based on update biological models, updated nuclide decay schemes, and a new method accounting for simultaneous dose to more than one organ. A detailed analysis of the impacts of ICRP-30 on waste repository safety and risk analyses would require an extensive and detailed study that has not yet been undertaken. Nevertheless, it is possible to identify, in an approximate manner, the impact of using ICRP-30 instead of 10 CFR 20/ICRP-2 in calculating the risk from radioactive repositories. Toward this end, the numerical guidance of ICRP-30 has been obtained and converted into RCG values for the general public using the same methods that were employed in deriving 10 CFR 20. The conversion was cross-checked by comparing 10 CFR 20 and ICRP-30-based values that were known to have remained the same. The most restrictive ICRP-30 RCGs were incorporated into the ORIGEN2 computer code, which was then used to calculate the toxicity of some radioactive materials of interest in waste repository considerations. As a basis for discussion, the toxicities of the spent fuel from a PWR and of the uranium ore required to make the fuel are given for both the 10 CFR 20 and ICRP-30-based RCGs. As is evident, the use of the revised RCGs reduces the toxicity of the spent fuel at times less than 100 years and increases the toxicity at times thereafter

Evaluation of the Acceptability of Potential Depleted Uranium Hexafluoride Conversion Products at the Envirocare Disposal Site

The purpose of this report is to review and document the capability of potential products of depleted UF{sub 6} conversion to meet the current waste acceptance criteria and other regulatory requirements for disposal at the facility in Clive, Utah, owned by Envirocare of Utah, Inc. The investigation was conducted by identifying issues potentially related to disposal of depleted uranium (DU) products at Envirocare and conducting an initial analysis of them. Discussions were then held with representatives of Envirocare, the state of Utah (which is a NRC Agreement State and, thus, is the cognizant regulatory authority for Envirocare), and DOE Oak Ridge Operations. Provisional issue resolution was then established based on the analysis and discussions and documented in a draft report. The draft report was then reviewed by those providing information and revisions were made, which resulted in this document. Issues that were examined for resolution were (1) license receipt limits for U isotopes; (2) DU product classification as Class A waste; (3) use of non-DOE disposal sites for disposal of DOE material; (4) historical NRC views; (5) definition of chemical reactivity; (6) presence of mobile radionuclides; and (7) National Environmental Policy Act coverage of disposal. The conclusion of this analysis is that an amendment to the Envirocare license issued on October 5, 2000, has reduced the uncertainties regarding disposal of the DU product at Envirocare to the point that they are now comparable with uncertainties associated with the disposal of the DU product at the Nevada Test Site that were discussed in an earlier report

Once-through CANDU reactor models for the ORIGEN2 computer code

Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

Proposed classification scheme for high-level and other radioactive wastes

The Nuclear Waste Policy Act (NWPA) of 1982 defines high-level radioactive waste (HLW) as: (A) the highly radioactive material resulting from the reprocessing of spent nuclear fuel....that contains fission products in sufficient concentrations; and (B) other highly radioactive material that the Commission....determines....requires permanent isolation. This paper presents a generally applicable quantitative definition of HLW that addresses the description in paragraph (B). The approach also results in definitions of other waste classes, i.e., transuranic (TRU) and low-level waste (LLW). A basic waste classification scheme results from the quantitative definitions

Decay characteristics of once-through LWR and LMFBR spent fuels, high-level wastes, and fuel-assembly structural material wastes

The decay characteristics of spent fuel, high-level waste, and fuel-assembly structural material (cladding) waste are presented in the form of ORIGEN2 output tables for (1) a pressurized water reactor operating on a once-through cycle with low-enrichment uranium feed, (2) a boiling-water reactor operating on a once-through cycle with low-enrichment uranium feed, and (3) a liquid-metal fast breeder reactor being fueled with depleted uranium enriched with discharged light water reactor plutonium on a once-through basis. The decay characteristics given include the mass (g), radioactivity (Ci), thermal power (W), photon activity (photons/s and MeV/W-s in 18 energy groups), and neutron activity (neutrons/s) from (..cap alpha..,n) and spontaneous fission events. The first three characteristics are given for each element and for the principal nuclide contributors to the activation products, actinides, and fission products. Also included are a summary description of the ORIGEN2 reactor models that form the basis for the calculated results and a physical description of the fuel assemblies for the three reactors

Overall Assessment of Actinide Partitioning and Transmutation for Waste Management Purposes

A program to establish the technical feasibility and incentives for partitioning (i.e., recovering) actinides from fuel cycle wastes and then transmuting them in power reactors to shorter-lived or stable nuclides has recently been concluded at the Oak Ridge National Laboratory. The feasibility was established by experimentally investigating the reduction that can be practicably achieved in the actinide content of the wastes sent to a geologic repository, and the incentives for implementing this concept were defined by determining the incremental costs, risks, and benefits. Eight US Department of Energy laboratories and three private companies participated in the program over its 3-year duration. A reference fuel cycle was chosen based on a self-generated plutonium recycle PWR, and chemical flowsheets based on solvent extraction and ion-exchange techniques were generated that have the potential to reduce actinides in fuel fabrication and reprocessing plant wastes to less than 0.25% of those in the spent fuel. Waste treatment facilities utilizing these flowsheets were designed conceptually, and their costs were estimated. Finally, the short-term (contemporary) risks from fuel cycle operations and long-term (future) risks from deep geologic disposal of the wastes were estimated for cases with and without partitioning and transmutation. It was concluded that, while both actinide partitioning from wastes and transmutation in power reactors appear to be feasible using currently identified and studied technology, implementation of this concept cannot be justified because of the small long-term benefits and substantially increased costs of the concept