This work reviewed the historical literature associated with the Dragon experiment and Water Boiler reactors operated at Los Alamos during the Manhattan Project. Frisch's invited talk given at the Fast Burst Reactor Conference held the University of New Mexico in Albuquerque, NM in 1969 is quoted. From the literature review, basic models for the Dragon experiment and for a Water Boiler type assembly (aqueous homogeneous reactor) were created that can be used for conducting multi-physics simulations for criticality excursion studies. This methodology utilizes the coupled neutronic-hydrodynamic method to perform a time-dependent dynamic simulation of a criticality excursion. MCNP was utilized to calculate important nuclear kinetic parameters that were incorporated into the models. Simulation results compared reasonably well with historic data.

In collaboration with Argonne National Laboratory (ANL), Los Alamos National Laboratory (LANL) is assisting in the design and development of portions of the second ANL Bubble Experiment to be performed in late 2020 at ANL. The ANL Bubble Experiment, as called in this report, is a series of direct electron irradiations of a uranyl sulfate solution to produce radiolysis-induced gas bubbles of hydrogen and oxygen. The gas bubbles formed in the solution enhance mixing and heat transfer. The study of the dynamics, shape, and size of radiolysis-induced gas bubbles is of great importance to understand and characterize the thermal and fluid behavior of the solution, especially for the solution based, neutron-induced fission production technique for Mo-99. During the first experiment performed in 2014, ANL’s 35 MeV electron linear accelerator provided average powers of 6, 12, and 15 kW using a rastered beam to homogeneously heat the 15 x 15 x 80-cm uranyl-sulfate solution. Gas bubble size, shape, velocity, solution temperature, and hydrogen and oxygen concentrations were recorded during the irradiations. The details of the experimental setup and results for the first experiment are described in the ANL reports, Design and Construction of Experiment for Direct Electron Irradiation of a Uranyl Sulfate Solution: Bubble Formation and Thermal Hydraulics Studies and Experimental Results for Direct Electron Irradiation of a Uranyl Sulfate Solution: Bubble Formation and Thermal Hydraulics Studies. The purpose of this report is to describe the issues and lessons learned associated with the first experiment and elaborate on designs to improve and obtain more accurate experimental results for the upcoming experiment.

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.

of feedback coefficients that serve as coupling coefficients relating the neutron population and the physical mechanisms that drive reactivity effects, such as fissile material temperature and density changes. The operational characteristics of such nuclear systems usually manifest themselves when perturbations between system power (neutron population) and system operating conditions arise. Successful operation of such systems requires the establishment of steady equilibrium conditions. However, prior to obtaining the desired equilibrium (steady-state) conditions, an approach from zero-power (startup) must occur. This operational regime may possess certain limiting system conditions that must be maintained to achieve effective startup. Once steady-state is achieved, a key characteristic of this operational regime is the level of stability that the system possesses. Finally, a third operational regime, shutdown, may also possess limiting conditions of operation that must be maintained. This report documents the operational characteristics of a “generic” Accelerator Driven Fissile Solution (ADFS) system during the various operational regimes of startup, steady-state operation, and shutdown. Typical time-dependent behavior for each operational regime will be illustrated, and key system parameters, such as response times, will be quantified. A generalized linear systems analysis of steady-state operations will be performed to evaluate the level of stability of ADFS systems. This information should provide a basic understanding of typical ADFS system operational behavior, and facilitate the development of monitoring procedures and operator aids.

This report describes some of the characteristics of fissile solutions systems, namely: the ability of fuel to flow rapidly during operation due to thermal forces and radiolytic gas dynamics drives the physics of these systems; exhibition of high negative reactivity feedback due to fuel temperature increase and radiolytic gas generated void (both decrease fuel density); evidence of well-damped systems and bounded reactivity excursions result in bounded (new) steady-state operating condition; they are very docile and slow to respond due to long neutron lifetime and high thermal inertia of the fuel (specific heat of fuel is an order of magnitude greater than solid fuels); operation requires large excess reactivity to be available ($5.00 or more); and finally they are sensitive to auxiliary systems such as cooling water, gas handling, and water makeup; small changes can have large effects due to the large feedbacks

Dynamic System Simulation (DSS) utilizes state-variables described by differential or difference equations to model system evolution. DSS techniques have been applied to develop a family of models to examine the time-dependent operational behavior of fissile solution systems. The base, or generic model, predicts the behavior of an aqueous homogeneous reactor (AHR), configured with a single cooling loop, operating at atmospheric pressure. A model specific for SUPO (Super Power), a uranium solution fueled AHR that operated at Los Alamos National Laboratory from 1951 to 1974, tailored from the generic model demonstrated close correlation with experimental data for steady-state operation1. SUPO is considered the benchmark for steady-state AHR operation. Subsequently specific models were tailored from the generic model for a variety of historical reactors including KEWB (Kinetics Experiments Water Boiler) “A-2” and “B-5” cores and Silene, which is considered the benchmark for AHR pulse operations. These models demonstrated DSS techniques could reliably be extended to different core geometries (SUPO and KEWB “A-2” were spherical; KEWB “B-5” and Silene cylindrical)2. This family of models also shows close correlation with experimental data in all modes of fissile solution system operation including pulse, free evolution, and steady-state. System response due to rate and amplitude of reactivity insertion closely matches experimental data. This report describes additional extensions of this family of models to include a wider design space for fissile solution systems. Pressurized cores, cooling schemes involving multiple cooling loops of various geometries, and accelerator-driven sub-critical system concepts are modeled.

from a variety of design decisions. This provides a method to assist in optimization of the system design. Once design has been generated in some detail the C++ version of the system model may then be implemented in a LabVIEW user interface to evaluate operator controls and instrumentation and operator recognition and response to off-normal events. Taken as a set of system models the DSS, Visual Studio, and LabVIEW progression provides a comprehensive set of design support tools.

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.

An experiment has been performed combining highly enriched uranium, a hydrogenous moderator (polyethylene), and concrete. The purpose of the experiment was to provide additional criticality data that can be used to verify and validate criticality safety evaluations in support of the decommissioning of nuclear facilities throughout the Department of Energy complex. In this experiment, criticality was observed as a function time due to the curing and drying processes that occurred in the concrete.

One of the primary methods to produce medical isotopes, such as Mo-99, 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 similar to 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.

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 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, fuel-solution 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.

Critical masses of square-prisms of highly enriched uranium diluted in various X/235U ratios with matrix material and polyethylene were measured. The Configuration cores were 22.86-cm and 45.72-cm square and were reflected with 8.1 3-cm and 10.1 6-cm thick side polyethylene reflectors, respectively. The configurations had 10.1 6-cm thick top and bottom polyethylene reflectors. For some configurations, the Rossi-a, which is an eigenvalue value characteristic for a particular configuration, was measured to establish a reactivity scale based on the degree of subcriticality . Finally, the critical mass experiments are compared with values calculated with MCNP and ENDF/B-VI cross-sections.

Radioactive waste containing fissile material is frequently encountered in decontamination and decommissioning activities. For the most part, this waste is placed in containers or drums and stored in storage facilities. The amount of fissile material in each drum is generally small because of criticality safety limits that have been calculated with computer transport codes such as MCNP,1 KENO,2 or ONEDANT.3 To the best of our knowledge, no experimental critical mass data are available to verify the accuracy of these calculations or any calculations for systems containing fissile material (U-235, Pu-239, U-233) in contact with matrix material such as Al2O3, CaO, SiO2, Al, MgO, etc. The experiments presented in this paper establish the critical masses of highly enriched uranium foils diluted to various X/235U ratios with polyethylene and SiO2, polyethylene and aluminum, polyethylene and MgO, polyethylene and Gd, polyethylene and Fe, and moderated and reflected with polyethylene. In addition, these critical mass experimental data will be used to validate cross section data.

Energy under contract W-7405-ENG-36. By acceptance of this article, the publisher recognizes that the U.S. Government retains a nonexclusive, royaltyfree license to publish or reproduce the published form of this contribution, or to allow others to do so, for U.S. Government purposes. Los Alamos National Laboratory requests that the publisher identify this article as work performed under the auspices of the U.S. Department of Energy. Los Alamos National Laboratory strongly supports academic freedom and a researcher's right to publish; as an institution, however, the Laboratory does not endorse the viewpoint of a publication or guarantee its technical correctness. FORM 836 (10/96) LA-UR-99-4060 Approved for public release; distribution is unlimited.

This experiment demonstrated how the neutron multiplication of a system increases as moderated material is placed between highly enriched uranium foils. In addition, this experiment served to demonstrate the hand-stacking technique and approach to criticality be remote operation. This experiment was designed by McLaughlin in the mid-seventies as part of the criticality safety course that is taught at the Los Alamos Critical Experiments Facility. The H/{sup 235}U ratio for this experiment was 215, which is the ratio at which the minimum critical mass for this configuration occurs.

A series of reactivity worth measurements, involving small samples of {sup 233}U metal, were performed using the Los Alamos National Laboratory Critical Experiments Facility`s Solution High-Energy Burst Assembly (SHEBA) research reactor. The first objective of the experiments was to investigate the behavior of {sup 233}U in the intermediate neutron energy spectrum. The second objective was to demonstrate that SHEBA could be used to perform worth experiments in support of the spent nuclear fuel (SNF) burnup credit effort. The burnup credit effort, which is pursuing the approval to transport commercial SNF in significantly higher capacity casks, has expressed interest in using SHEBA to perform worth experiments. The proposed worth experiments would be used to validate data on the behavior of fission products commonly found in SNF, such as {sup 149}Sm, {sup 153}Eu, and {sup 143}Nd. Samples of these materials and other fission products would be placed in SHEBA, and their reactivity worths would be measured. These experiments must be able to cover the range of neutron energy spectra expected in SNF casks. SHEBA, which has an energy spectrum similar to a pressurized water reactor, can be modified to alter the local energy spectrum around a sample. The SHEBA/{sup 233}U experiments weremore » also performed to demonstrate this capability.« less

An accident analysis has been performed for the nuclear criticality safety class (NCSC) foil experiment. The Los Alamos Critical Experiments Facility (LACEF) performs this experiment regularly during its 2-, 3-, and 5-day nuclear criticality safety classes. This accident analysis is part of an effort to modify the NCSC foil experiment plan so that the experiment may be operated at delayed critical. Currently, the NCSC foil experiment may only be operated up to a neutron multiplication of 100. The purpose of the accident analysis is to ensure that any accidental nuclear excursion does not exceed the boundary of the safety envelope described in the LACEF safety analysis report (SAR). The experiment consists of very thin, highly enriched (93% {sup 235}U) uranium metal foils (23 X 23 X 0.008 cm) interleaved between Lucite plates (36 X 36 X 1.27 cm). The fuel foils and Lucite plates are stacked vertically to form a critical assembly. Extra Lucite plates placed at the top and bottom of the assembly act as vertical reflectors. The assembly is operated remotely with the use of a general-purpose vertical-lift platform machine. The accident scenario consists of one additional fuel foil being added to an existing critical or nearly criticalmore » stack. The reactivity insertion rate is 0.05 $/s, based on the speed of the vertical-lift platform. It is assumed that none of the safety systems will function properly during the accident and that the operating crew is unable to mitigate the accident.« less