[en] One of the concerns facing accelerator-driven transmutation systems (ADSs) is whether the radiotoxicity of materials produced during the transmutation process poses more of a concern than does the radiotoxicity of the spent nuclear fuel (SNF) itself. Most of the common fission products (or FPs) are emitters of beta radiation, but additionally, some of the radionuclides generated during spallation are alpha emitters. Thus, both ingestion and inhalation radiotoxicity of the materials produced during spallation could be significant. Typically, ingestion is considered to be more significant than inhalation radiotoxicity for long-term storage/disposal (such as in a repository) because the greatest potential biological hazard to humans occurs when the isotope is absorbed in nearby ground water or brine and transported from the repository to drinking water. Nonetheless, inhalation radiotoxicity is also important to analyze in case of a breach of containment inside the accelerator facility and/or for short-term (i.e., above-ground) storage concerns. Thus, this study calculated the radiotoxicity of spallation products (or SPs) from three different targets: lead-bismuth eutectic (LBE), LBE-cooled tungsten, and LBE-cooled lead

[en] The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

[en] 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

[en] This article presents several reasonable cases in which four mechanisms - dissolution, physical mixing, adsorption, and precipitation (either chemical change or evaporation) - might concentrate fissile material in and around a disposal container for radioactive waste at the proposed repository at Yucca Mountain, Nevada. The possible masses, concentrations, and volume are then compared to criticality limits. The cases examined evaluate the geologic barrier role in preventing criticality since engineered options for preventing criticality (e.g., boron or gadolinium neutron absorber in the disposal container) are not considered. The solid concentrations able to form in the natural environment are insufficient for criticality to occur because (a) solutions of ²³⁵U and ²³⁹Pu are clearly not critical; (b) physical mixing of fissile material with the entire potential iron oxide (as goethite - FeOOH) in a waste package is not critical; (c) the adsorption of ²³⁹Pu on consolidated iron oxide in a waste package is not critical; (d) the adsorption of ²³⁵U on consolidated iron oxide in a waste package is not critical when accounting for reduced adsorption because of carbonates at high pH; (e) the filtration of iron oxide colloids, containing fissile material, by the thin invert material is not critical; (f) insufficient retention through precipitation of ²³⁵U or ²³⁹Pu occurs in the invert; (g) adsorption of ²³⁵U and ²³⁹Pu on devitrified or clinoptolite-rich tuff below the repository is not critical; (h) the average precipitation/adsorption of ²³⁵U as uranyl silicates in the tuff is not critical by analogy with calcite deposition in lithophysae at Yucca Mountain; and (i) precipitation/adsorption (caused by cyclic drying) as uranyl silicates on fracture surfaces of the tuff is not critical by analogy with the oxidation of UO₂, migration of U^VI, and precipitation in fractures at the Nopal I ore deposit in Mexico

[en] This paper presents the mass, concentration, and volume required for a critical event to occur in homogeneous mixtures of fissile material and various other geologic materials. The fissile material considered is primarily highly enriched uranium spent fuel; however, ²³⁹Pu is considered in some cases. The non-fissile materials examined are those found in the proposed repository area at Yucca Mountain, Nevada: volcanic tuff, iron rust, concrete, and naturally occurring water. For ²³⁵U, the minimum critical solid concentration for tuff was 5 kg/m³ (similar to sandstone), and in goethite, 45 kg/m³. The critical mass of uranium was sensitive to a number of factors, such as moisture content and fissile enrichment, but had a minimum, assuming almost 100% saturation and >20% enrichment, of 18 kg in tuff as Soddyite (or 9.5 kg as UO₂) and 7 kg in goethite. For ²³⁹Pu, the minimum critical solid concentration for tuff was 3 kg/m³ (similar to sandstone); in goethite, 20 kg/m³. The critical mass of plutonium was also sensitive to a number of factors, but had a minimum, assuming 100% saturation and 80-90% enrichment, of 5 kg in tuff and 6 kg in goethite

[en] Modeling of nuclear criticality was omitted from performance assessment calculations for the Waste Isolation Pilot Plant (WIPP), a repository for waste contaminated with transuranic radioisotopes, located in southeastern New Mexico, based on arguments of low probability and low consequence. Low-probability arguments are presented here. Guidance provided by the Environmental Protection Agency (EPA) - the regulator of WIPP - allowed either qualitative 'credibility' arguments or quantitative probability estimates when screening features, events, and processes such as criticality. Although information to quantitatively evaluate the probability of a criticality event was mostly lacking, qualitatively reasoned discussion of the inability to assemble a critical configuration of fissile material was accepted by the EPA. Specifically, after disposal and prior to an inadvertent human intrusion into the repository, there is no credible mechanism to move radioisotopes (and particularly, fissile material) since only small amounts of brine enter the repository, as adequately demonstrated in calculations over the years. An inadvertent human intrusion (an event that must be considered because of safety regulations) might allow a large pressure gradient to move more brine through the repository, but there is still no credible mechanism to counteract the natural tendency of the material to disperse during transport. Unfavorable physical conditions on concentrating fissile material include low initial solid concentration of fissile material, small mass of fissile material transported over 10 000 yr, and insufficient physical compaction; unfavorable hydrologic conditions include the limited amount of brine available to transport fissile material. Unfavorable geochemical conditions on concentrating the fissile radioisotopes include lack of sufficient adsorption and water chemistry conducive to precipitation

[en] Highlights: • Tested here are 4 methods for estimating critical rod position, in Monteburns, of a reactor fuel array. • Inverse multiplication methods better predict critical rod position at the cost of more iterations. • A polynomial fit technique can predict most plutonium isotopics to within 5%. - Abstract: This burnup study examined the effect of a predicted critical control rod position on the nuclide predictability of several axial and radial locations within a 4 × 4 graphite moderated gas cooled reactor fuel cluster geometry. To achieve this, a control rod position estimator (CRPE) tool was developed within the framework of the linkage code Monteburns between the transport code MCNP and depletion code CINDER90, and four methodologies were proposed within the tool for maintaining criticality. Two of the proposed methods used an inverse multiplication approach – where the amount of fissile material in a set configuration is slowly altered until criticality is attained – in estimating the critical control rod position. Another method carried out several MCNP criticality calculations at different control rod positions, then used a linear fit to estimate the critical rod position. The final method used a second-order polynomial fit of several MCNP criticality calculations at different control rod positions to guess the critical rod position. The results showed that consistency in prediction of power densities as well as uranium and plutonium isotopics was mutual among methods within the CRPE tool that predicted critical position consistently well. While the CRPE tool is currently limited to manipulating a single control rod, future work could be geared toward implementing additional criticality search methodologies along with additional features.

[en] The Next Generation Safeguards Initiate (NGSI) of the United States Department of Energy has funded a multi-laboratory/university collaboration to quantify plutonium content in spent fuel (SF) with non-destructive assay (NDA) techniques and quantify the capability of these NDA techniques to detect pin diversions from SF assemblies. The first Monte Carlo based spent fuel library (SFL) developed for the NGSI program contained information for 64 different types of SF assemblies (four initial enrichments, burnups, and cooling times). The maximum amount of fission products allowed to still model a 17x17 Westinghouse pressurized water reactor (PWR) fuel assembly with four regions per fuel pin was modelled. The number of fission products tracked was limited by the available memory. Studies have since indicated that additional fission product inclusion and asymmetric burning of the assembly is desired. Thus, an updated SFL has been developed using an enhanced version of MCNPX, more powerful computing resources, and the Monte Carlo-based burnup code Monteburns, which links MCNPX to a depletion code and models a representative 1∕8 core geometry containing one region per fuel pin in the assemblies of interest, including a majority of the fission products with available cross sections. Often in safeguards, the limiting factor in the accuracy of NDA instruments is the quality of the working standard used in calibration. In the case of SF this is anticipated to also be true, particularly for several of the neutron techniques. The fissile isotopes of interest are co-mingled with neutron absorbers that alter the measured count rate. This paper will quantify how well working standards can be generated for PWR spent fuel assemblies and also describe the spatial plutonium distribution across an assembly. More specifically we will demonstrate how Monte Carlo gamma measurement simulations and a Monte Carlo burnup code can be used to characterize the emitted gamma spectrum and the asymmetries experienced in the second SFL.

[en] The use of both reactors and accelerator-driven systems (ADSs) for the transmutation of spent nuclear fuel has been examined for decades. The advantage of burning the material in a reactor is that it can provide more reliable electricity generation than an ADS (which is subject to beam interruptions). The advantage of an ADS is that the material does not have to maintain constant reactivity (i.e., it can be subcritical, and the k_eff can decrease over a cycle) and thus, operations are inherently more 'safe' (i.e., turning off the accelerator stops the production of additional fission neutrons). Using a combination of (1) existing or new light-water reactors (LWRs) to burn at least the plutonium and (2) an ADS for the remaining material allows both advantages to be utilized and is referred to as a two-tiered approach. In addition to power production, burning a fraction of the plutonium in the LWR helps the ADS by decreasing the fission-to-capture ratio of the actinides being transmuted. At ∼0.8, this ratio causes the reactivity in the accelerator to remain fairly constant for cycles of up to 2 to 3 years. The use of three different fuels (mixed oxide, thorium/plutonium oxide, and non-fertile fuel) in a simple LWR is addressed in this paper, as is the use of several different actinide streams for the non-fertile fuel (plutonium by itself and plutonium plus minor actinides). In addition, the effect of each type of fuel/material stream on the ADS is addressed, as is the transmutation of long-lived fission products in this two-tiered environment. (authors)

[en] For the purpose of NDA simulations of the ROK fuel assemblies, we have developed an axial burnup distribution to represent the pins themselves based on gamma scans of rods in the G23 assembly. For the purpose of modeling the G23 assembly (both at ORNL and LANL), the pin-by-pin burnup map as simulated by ROK is being assumed to represent the radial burnup distribution. However, both DA and NDA results indicate that this simulated estimate is not 100% correct. In particular, the burnup obtained from the axial gamma scan of 7 pins does not represent exactly the same 'average' pin burnup as the ROK simulation. Correction for this discrepancy is a goal of the well-characterized assembly task but will take time. For now, I have come up with a correlation for 26 axial points of the burnup as obtained by gamma scans of 7 different rods (C13, G01, G02, J11, K10, L02, and M04, neglecting K02 at this time) to the average burnup given by the simulation for each of the rods individually. The resulting fraction in each axial zone is then averaged for the 7 different rods so that it can represent every fuel pin in the assembly. The burnup in each of the 26 axial zones of rods in all ROK assemblies will then be directly adjusted using this fraction, which is given in Table 1. Note that the gamma scan data given by ROK for assembly G23 included a length of ∼3686 mm, so the first 12 mm and the last 14 mm were ignored to give an actual rod length of ∼366 cm. To represent assembly F02 in which no pin-by-pin burnup distribution is given by ROK, we must model it using infinitely-reflected geometry but can look at the effects of measuring in different axial zones by using intermediate burnup files (i.e. smaller burnups than 28 GWd/MTU) and determining which axial zone(s) each burnup represents. Details for assembly F02 are then given in Tables 2 and 3, which is given in Table 1 and has 44 total axial zones to represent the top meter in explicit detail in addition to the other 26 zones. Note that the MCNP files for F02 were created using the Monte Carlo burnup linkage code Monteburns, which saves MCNP input files with detailed compositions as a function of burnup. The 'intermediate burnup files' produced for F02 include a cooling time of 27 years. The axial location of 5 spacers was also included in the ROK F02 assembly in which each spacer contained a length of 3.81 cm. Note that due to the nature of Monteburns, which was run in a special fashion for this problem, the step number increments after the 27 year decay, so the second column of Table 2 refers to the step number that should be used in the Monteburns files.