Robert Kimpland

Criticality of a 237Np Sphere

Abstract A series of critical-mass experiments using a 6-kg neptunium sphere was performed on the Planet vertical-assembly machine at Los Alamos National Laboratory (LANL). The purpose of the experiments was to obtain a better estimate of the critical mass of 237Np. The configurations that were studied included surrounding the neptunium sphere with highly enriched uranium (HEU) shells as well as reflecting it with iron and polyethylene. An additional experiment using a 4.5-kg α-phase plutonium sphere surrounded with HEU was performed to demonstrate how well the computer transport code and the existing cross-section data for uranium and plutonium could reproduce the experiment. For some of the configurations, the prompt-neutron decay constants at delayed critical were measured. These experiments provided an integral measurement of the cross sections for 237Np in the fast-energy and possibly in the intermediate-energy regions. The measured keff from these experiments was compared with the calculated keff from the Monte Carlo N-Particle (MCNP) transport code using ENDF/B-V and ENDF/B-VI and cross-section data evaluated by the Nuclear Theory and Applications group (T-16) at LANL. In all the neptunium experiments, the calculated keff values based on ENDF/B-VI data were ~1% lower than the experimental keff. After adjusting the cross sections for neptunium and 235U to match the bare neptunium/HEU experiment as well as Godiva keff criticality and spectra indexes, the MCNP code yielded a value of 57 ± 4 kg for the bare critical mass of 237Np.

Dynamic Analysis of Nuclear Excursions in Underground Repositories Containing Plutonium

A recent study performed at the Los Alamos National Laboratory postulates that plutonium-239 stored in underground repositories could lead to a nuclear explosion of up to a few hundred gigajoules. The study suggests that plutonium originally contained in glass logs could escape its containment and disperse into the surrounding native rock of the repository. This dispersion would then lead to an autocatalytic process that ultimately would lead to a catastrophic nuclear explosion. A computer model that simulates this autocatalytic process has been developed at the Los Alamos Critical Experiments Facility. The model has been used to determine the fission yield of such an event and the effects of that yield on the repository. The goal of this work is to quantify the consequences of the autocatalytic process, not to determine the probability of such an event occurring.

Analysis of the Effects of Radiolytic-Gas Bubbles on the Operation of Solution Reactors for the Production of Medical Isotopes

Abstract One of the primary methods to produce medical isotopes, such as 99Mo, is by irradiation of uranium targets in heterogeneous reactors. Solution reactors present a potential alternative to produce medical isotopes. The Medical Isotope Production Reactor (MIPR) concept has been proposed to produce medical isotopes with lower uranium consumption and waste than those in heterogeneous reactors. Commercial production of medical isotopes in solution reactors requires steady-state operation at ~200 kW. At this power regime, fuel-solution temperature increase and radiolytic-gas bubble formation introduce a negative reactivity feedback that has to be mitigated. A model based on the point reactor kinetic equations has been developed to investigate these reactivity effects. This model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA) and shows the feasibility of solution reactors for the commercial production of medical isotopes.

Reactivity analysis of solution reactors for medical-radioisotope production

One of the primary methods to produce medical isotopes, such as 99 Mo, is by irradiation of uranium targets in heterogeneous reactors. Solution reactors present a potential alternative to produce medical isotopes. The medical isotope production reactor concept has been proposed to produce medical isotopes with lower uranium consumption and waste than the corresponding fuel consumption and waste in heterogeneous reactors. Commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, fuelsolution temperature increase and radiolytic-gas bubble formation introduce a negative reactivity feedback that has to be mitigated. This work analyzes the reactivity effects on the operation of solution reactors for the production of medical isotopes and provides some reactor characteristics that may mitigate the negative reactivity feedback introduced by the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. � 2003 Elsevier B.V. All rights reserved.

Dynamic System Simulation of Fissile Solution Systems

A time-dependent dynamic system simulation model of fissile solution systems is described. The model is composed of four coupled sub-models: neutron kinetics, radiolytic gas generation, core thermal, and plenum models. The performance of the model is compared against experimental data of SUPO, KEWB, Silene, and HRE aqueous homogeneous reactors. Model extensions to address accelerator-driven subcritical systems are also discussed. AHR conceptual designs incorporating all “lessons learned” from experimental history and modelling is presented as the ideal design for Mo-99 production.

Criticality characteristics of mixtures of plutonium, silicon dioxide, Nevada tuff, and water

The parameters that determine when critical mixtures of {sup 239}Pu, SiO{sub 2}, and water and mixtures of {sup 239}Pu, Nevada tuff, and water are capable of sustaining an increasing neutron chain reaction as may be caused by a positive void coefficient at constant temperature are established. A single canister is considered that is loaded with up to 75 kg of {sup 239}Pu. A survey of critical spherical mixtures of plutonium, SiO{sub 2}, tuff, and water at constant temperature is created and these results are examined to determine the mixtures that might be autocatalytic. Regions of criticality instability are identified that have the possibility of autocatalytic power behavior. A positive void coefficient is possible for a very limited range of wet systems.

Concept for LEU Burst Reactor

Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

Robust Medical Isotope Production System

The success of this theoretical undertaking provided confidence that the behavior of new and evolving designs of fissile solution systems may be accurately estimated. Scaled up versions of SUPO, subcritical acceleratordriven systems, and other evolutionary designs have been examined.

Critical experiments with highly enriched uranium and matrix elements (Si, Mg, Al, Gd, and Fe)

Abstract Scientists at the Los Alamos National Laboratory measured the critical masses of square prisms of highly enriched uranium diluted in various X/235U with matrix material and polyethylene. The configuration cores were 22.86 and 45.72 cm square and were reflected with 8.13-cm-thick and 10.16-cm-thick side polyethylene reflectors, respectively. The configurations had 10.16-cm-thick top and bottom polyethylene reflectors. For some configurations, the Rossi-α, which is an eigenvalue characteristic for a particular configuration, was measured to establish a reactivity scale based on the degree of subcriticality. These experiments provided critical mass data in the thermal energy range for systems containing Si, Mg, Al, Gd, and Fe. The measured keff from these experiments was compared with the calculated keff from MCNP using ENDF/B-V and ENDF/B-VI cross-section data. The observed biases were +0.005 Δk and +0.008 Δk for Si, +0.0006 Δk and +0.008 Δk for Al, +0.0023 Δk for Mg, +0.004 Δk and +0.01304 Δk for Gd, and +0.0123 Δk and -0.00106 Δk for Fe.

Critical mass experiment using U-235 foils-SiO{sub 2}-polyethylene plates

Fissile material in waste is frequently encountered in decontamination and decommissioning activities. Thousands of drums containing radioactive waste are stored in storage facilities throughout the DOE complex. The amount of fissile material in each drum is generally small because of the criticality safety limits that have been calculated using neutron transport computer codes such as MCNP, KENO, or ONEDANT. No experimental critical data are available to assure the correctiveness of the calculations for those systems containing fissile material (U-235, U-233, and Pu-239) in contact with matrix material (Al{sub 2}O{sub 3}, CaO, MgO, and SiO{sub 2}) in the drums. The purpose of the U-235 foil-SiO{sub 2}-polyethylene experiment is to provide experimental data to validate the computer transport codes and the cross section data.