Publications

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This paper describes the fissile mass and concentration necessary for a critical event to occur outside containers disposed in a bedded salt repository. The criticality limits are based on modeling mixtures of water, salt, dolomite, concrete, rust, and fissile material using a neutron/photon transport computational code. Several idealized depositional configurations of fissile material in the host rock are analyzed: homogeneous spheres and heterogeneous arrangements of plate fractures in regular arrays. Deposition of large masses and concentrations are required for criticality to occur for low enriched 235U enrichment. Homogeneous mixtures with deposition in all the porosity are more reactive at high enrichments of 235U and 239Pu. However, unlike typical engineered systems, heterogeneous configurations can be more reactive than homogeneous systems at high enrichment when deposition occurs in only a portion of the porosity and the total porosity is small, because the relationship between the porosity of the fractures and matrix also strongly influences the results.

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The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.

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A simplified and bounding methodology for analyzing the pressure buildup and hydrogen concentration within an unvented 2R container was developed (the 2R is a sealed container within a 6M package). The specific case studied was the gas buildup due to alpha radiolysis of water moisture sorbed on small quantities (less than 20 Ci per package) of plutonium oxide. Analytical solutions for gas pressure buildup and hydrogen concentration within the unvented 2R container were developed. Key results indicated that internal pressure buildup would not be significant for a wide range of conditions. Hydrogen concentrations should also be minimal but are difficult to quantify due to a large variation/uncertainty in model parameters. Additional assurance of non-flammability can be obtained by the use of an inert backfill gas in the 2R container.

Preliminary shielding calculations were performed for a prototype National Spent Nuclear Fuel Program (NSNFP) transport cask. This analysis is intended for use in the selection of cask shield material type and preliminary estimate of shielding thickness. The radiation source term was modeled as cobalt-60 with radiation exposure strength of 100,000 R/hr. Cobalt-60 was chosen as a surrogate source because it simultaneous emits two high-energy gammas, 1.17 MeV and 1.33 MeV. This gamma spectrum is considered to be large enough that it will upper bound the spectra of all the various spent nuclear fuels types currently expected to be shipped within the prototype cask. Point-kernel shielding calculations were performed for a wide range of shielding thickness of lead and depleted uranium material. The computational results were compared to three shielding limits: 200 mrem/hr dose rate limit at the cask surface, 50 mR/hr exposure rate limit at one meter from the cask surface, and 10 mrem/hr limit dose rate at two meters from the cask surface. The results obtained in this study indicated that a shielding thickness of 13 cm is required for depleted uranium and 21 cm for lead in order to satisfy all three shielding requirements without taking credit for stainless steel liners. The system analysis also indicated that required shielding thicknesses are strongly dependent upon the gamma energy spectrum from the radiation source term. This later finding means that shielding material thickness, and hence cask weight, can be significantly reduced if the radiation source term can be shown to have a softer, lower energy, gamma energy spectrum than that due to cobalt-60.

Based on general arguments presented in this report, nuclear criticality was eliminated from performance assessment calculations for the Waste Isolation Pilot Plant (WIPP), a repository for waste contaminated with transuranic (TRU) radioisotopes, located in southeastern New Mexico. At the WIPP, the probability of criticality within the repository is low because mechanisms to concentrate the fissile radioisotopes dispersed throughout the waste are absent. In addition, following an inadvertent human intrusion into the repository (an event that must be considered because of safety regulations), the probability of nuclear criticality away from the repository is low because (1) the amount of fissile mass transported over 10,000 yr is predicted to be small, (2) often there are insufficient spaces in the advective pore space (e.g., macroscopic fractures) to provide sufficient thickness for precipitation of fissile material, and (3) there is no credible mechanism to counteract the natural tendency of the material to disperse during transport and instead concentrate fissile material in a small enough volume for it to form a critical concentration. Furthermore, before a criticality would have the potential to affect human health after closure of the repository--assuming that a criticality could occur--it would have to either (1) degrade the ability of the disposal system to contain nuclear waste or (2) produce significantly more radioisotopes than originally present. Neither of these situations can occur at the WIPP; thus, the consequences of a criticality are also low.

Transuranic (TRU) waste generated by the handling of plutonium during research on or production of U.S. nuclear weapons will be disposed of in the Waste Isolation Pilot Plant (WIPP). This paper describes the physical and radiological properties of the TRU waste that will be deposited in the WIPP. This geologic repository will accommodate up to 175,564 m3 of TRU waste, corresponding to 168,485 m3 of contact-handled (CH-) TRU waste and 7079 m3 of remote-handled (RH-) TRU waste. Approximately 35% of the TRU waste is currently packaged and stored (i.e. legacy) waste, with the remainder of the waste to be packaged or generated and packaged in activities before the year 2033, which is the closure time for the repository. These wastes were produced at 27 U.S. Department of Energy (DOE) siles in the course of generating defense nuclear materials. The radionuclide and nonradionuclide inventories for the TRU wastes described in this paper were used in the 1996 WIPP Compliance Certification Application (CCA) performance assessment calculations by the Sandia National Laboratories/New Mexico (SNL/NM).