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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Development for fissile assay in recycled fuel using lead slowing down spectrometer

A future nuclear energy system is under development to turn spent fuels produced by PWRs into fuels for a SFR (Sodium Fast Reactor) through the pyrochemical process. The knowledge of the isotopic fissile content of the new fuel is very important for fuel safety. A lead slowing down spectrometer (LSDS) is under development to analyze the fissile material content (Pu{sup 239}, Pu{sup 241} and U{sup 235}) of the fuel. The LSDS requires a neutron source, the neutrons will be slowed down through their passage in a lead medium and will finally enter the fuel and will induce fission reactions that will be analysed and the isotopic content of the fuel will be then determined. The issue is that the spent fuel emits intense gamma rays and neutrons by spontaneous fission. The threshold fission detector screens the prompt fast fission neutrons and as a result the LSDS is not influenced by the high level radiation background. The energy resolution of LSDS is good in the range 0.1 eV to 1 keV. It is also the range in which the fission reaction is the most discriminating for the considered fissile isotopes. An electron accelerator has been chosen to produce neutrons with an adequate target through (e{sup -},γ)(γ,n) reactions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Dynamic System Simulation of Fissile Solution Systems

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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Accelerated Basin De-Inventory Maximum Fissile Estimate

The Accelerated Basin De-inventory (ABD) program increases processing of Spent Nuclear Fuel (SNF) in H-Canyon to accelerate the closure of L Basin. The dissolved SNF will be dispositioned to Sludge Batches that will be processing in the Defense Waste Processing Facility and converted to glass. The fissile glass loading directly impacts the amount of glass canisters made. This memorandum provides an estimated maximum quantity of total fissile expected to be transferred to a Sludge Batch to support increasing the fissile glass loading and thus reducing the number of canisters created across the lifetime of the ABD mission.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Policy and Technical Issues Facing a Fissile Material (Cutoff) Treaty. Chapter 14

We report the largest obstacle to creating nuclear weapons, starting with the ones that destroyed Hiroshima and Nagasaki, has been to make sufficient quantities of fissile materials – highly enriched uranium (HEU) and plutonium – to sustain an explosive fission chain reaction.1 Recognition of this fact has, for more than fifty years, underpinned both the support for and the opposition to adoption of an international treaty banning at a minimum the production of more fissile materials for nuclear weapons, commonly referred to as a fissile material cutoff treaty (FMCT).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2011

Development of an Air Transport Type A Fissile Package

This paper presents the summary of testing by the Savannah River National Laboratory (SRNL) to support development of a light weight (<140 lbs) air transport qualified Type A Fissile Packaging. The package design incorporates features and materials specifically designed to minimize packaging weight. The light weight package is being designed to provide confinement to the contents when subjected to the normal and hypothetical conditions required of an air transportable Type A Fissile radioactive material shipping package. The objective of these tests was to provide design input to the final design for the LORX Type A Fissile Air Transport Packaging when subjected to the performance requirements of the drop, crush and puncture probe test of 10CFR71. The post test evaluation of the prototype packages indicates that all of the tested designs would satisfactorily confine the content within the packaging. The differences in the performance of the prototypes varied significantly depending on the core materials and their relative densities. Information gathered from these tests is being used to develop the final design for the Department of Homeland Security.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2015

Fissile material holdup measurement systems: an historical review of hardware and software

The measurement of fissile material holdup is accomplished by passively measuring the energy-dependent photon flux and/or passive neutron flux emitted from the fissile material deposited within an engineered process system. Both measurement modalities--photon and neutron--require the implementation of portable, battery-operated systems that are transported, by hand, from one measurement location to another. Because of this portability requirement, gamma-ray spectrometers are typically limited to inorganic scintillators, coupled to photomultiplier tubes, a small multi-channel analyzer, and a handheld computer for data logging. For neutron detection, polyethylene-moderated, cadmium-back-shielded He-3 thermal neutron detectors are used, coupled to nuclear electronics for supplying high voltage to the detector, and amplifying the signal chain to the scaler for counting. Holdup measurement methods, including the concept of Generalized Geometry Holdup (GGH), are well presented by T. Douglas Reilly in LA-UR-07-5149 and P. Russo in LA-14206, yet both publications leave much of the evolutionary hardware and software to the imagination of the reader. This paper presents an historical review of systems that have been developed and implemented since the mid-1980s for the nondestructive assay of fissile material, in situ. Specifications for the next-generation holdup measurements systems are conjectured.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Accelerating fissile material detection with a neutron source

A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly to count neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. The system includes a Poisson neutron generator for in-beam interrogation of a possible fissile neutron source and a DC power supply that exhibits electrical ripple on the order of less than one part per million. Certain voltage multiplier circuits, such as Cockroft-Walton voltage multipliers, are used to enhance the effective of series resistor-inductor circuits components to reduce the ripple associated with traditional AC rectified, high voltage DC power supplies.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Accelerating fissile material detection with a neutron source

A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly to count neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. The system includes a Poisson neutron generator for in-beam interrogation of a possible fissile neutron source and a DC power supply that exhibits electrical ripple on the order of less than one part per million. Certain voltage multiplier circuits, such as Cockroft-Walton voltage multipliers, are used to enhance the effective of series resistor-inductor circuits components to reduce the ripple associated with traditional AC rectified, high voltage DC power supplies.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Minimum critical masses and their uncertainties for several fissile nuclides and reflectors - 14296

A systematic effort was undertaken to reevaluate the minimum critical mass for several fissile nuclides and reflectors. This was done using MCNP6 in combination with several modern nuclear data libraries, viz. ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1.1. This combination was benchmarked extensively against criticality safety benchmarks. The procedure to extract the minimum critical mass from the MCNP6 calculations was carefully implemented, so that it would have a negligible impact on the accuracy of the end results. All parts of the calculation, i.e. nuclear data processing, MCNP6 models, and post-processing, were applied consistently for all cases, involving varying fissile nuclides and reflectors. Results for the minimum critical mass were generated for {sup 233,235}U, {sup 236,239,241}Pu, {sup 242m}Am, {sup 243,245,247}Cm, and {sup 249,251}Cf, for the case of water moderation and water reflection. For {sup 235}U, the minimum critical mass was also calculated for cases with other reflectors, such as beryllium, heavy water, graphite, lead, and concrete. The results for the water reflection cases are in good agreement with most of the literature values, but for the cases with reflectors such as graphite, lead, or concrete, there are sizable differences. The accuracy of the results was assessed in two ways. First, a comparison was made with criticality safety benchmark results, by selecting benchmarks that resemble the minimum mass systems the most. It was clear that for {sup 233,235}U and {sup 239}Pu with water reflection, the differences in minimum critical masses between the libraries are consistent with the differences obtained for benchmark cases. The difference between the benchmark calculations and the benchmark values is then an indication of the accuracy with which one can simulate comparable systems. Secondly, sensitivity runs were performed with the new MCNP6 sensitivity option, based on ENDF/B-VII.1 nuclear data. For each fissile nuclide, the sensitivity of k{sub eff} to cross sections such as capture, fission, elastic, inelastic was calculated for the minimum critical mass case. The same was done for the {sup 1}H and {sup 16}O cross sections in the moderator and reflector. The uncertainty in k{sub eff} was calculated by combining the sensitivity profiles with the appropriate ENDF/B-VII.1 covariance matrix, using the SUSD code. As a final step, the uncertainty in k{sub eff} can be translated to an uncertainty in the minimum critical mass. (author)

42 ENGINEERING↗
DOE OSTI2014

Effect of the fissile bead's and thermocouple wires’ sizes on the response time of a fission couple

The fission couple is proposed as a fast response miniature neutron detector in the measurement of time dependent energy depositions within the fissile material based on theoretical analysis, but the response time of a fission couple is relatively slow in practice. The time lag originated from heat transfer process was demonstrated to be the dominating factor by theoretical simulations and experimental verification in this paper. The response of a fission couple as a function of the bead size and the thermocouple wires’ sizes are simulated using ANSYS workbench. The decrease of wires’ diameter results in the decrease of response time, and the increase of bead's diameter leads to a slight increase of response time. During a pulse heating transient in the fuel of Chinese Fast Burst Reactor II with a FWHM of 181μs, the time lag originated from heat transfer process is about tens of microseconds for the peaks of the change rate of temperature, and is of the order of milliseconds to achieve 85% of the temperature rise for a typical fission couple with a Φ 1 mm fissile bead and two Φ 0.05 mm thermocouple wires. The results obtained provide foundation for the optimization of fission couples.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Measurement of the Delayed Photon Yield of the Photofission Reaction as a Method for Identifying Fissile Materials

The results of measuring the photofission reaction yield for actinide nuclei in the near-threshold energy region are presented. At the LUE-8-5 accelerator of the Institute for Nuclear Research of the Russian Academy of Sciences, the energy and time spectra of the delayed gamma quanta were obtained by irradiating fissile isotopes by bremsstrahlung photons. The conclusion is drawn about the possibility of identifying fissile isotopes by the behavior of the photofission cross section in the near-threshold region.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

Operational Characteristics of an Accelerator Driven Fissile Solution System

Operational characteristics represent the set of responses that a nuclear system exhibits during normal operation. Operators rely on this behavior to assess the status of the system and to predict the consequences of off-normal events. These characteristics largely refer to the relationship between power and system operating conditions. The static and dynamic behavior of a chain-reacting system, operating at sufficient power, is primarily governed by reactivity effects. The science of reactor physics has identified and evaluated a number of such effects, including Doppler broadening and shifts in the thermal neutron spectrum. Often these reactivity effects are quantified in the form 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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Productivity Techniques and Quality Aspects in the Criticality Safety Evaluation of Y-12 Type-B Fissile Material Packages

The inventory of certified Type-B fissile material packages consists of ten performance-based packages for offsite transportation purposes, serving transportation programs at the Y-12 National Security Complex. The containment vessels range from 5 to 19 in. in diameter and from 17 to 58 in. in height. The drum assembly external to the containment vessel ranges from 18 to 34 in. in diameter and from 26 to 71 in. in height. The weight of the packaging (drum assembly and containment vessel) ranges from 239 to 1550 lb. The older DT-nn series of Cellotex-based packages are being phased-out and replaced by a new generation of Kaolite-based (&#x27;Y-12 patented insulation&#x27;) packages capable of withstanding the dynamic crush test 10 CFR 71.73(c)(2). Three replacement packages are in various stages of development; two are in use. The U.S. Department of Transportation (DOT) 6M specification package, which does not conform to the U.S. Nuclear Regulatory Commission requirements for Type-B packages, is no longer authorized for service on public roads. The ES-3100 shipping package is an example of a Kaolite-based Type-B fissile material package developed as a replacement package for the DOT 6M. With expanded utility, the ES-3100 is designed and licensed for transporting highly enriched uranium and plutonium materials on public roads. The ES-3100 provides added capability for air transport of up to 7-kg quantities of uranium material. This paper presents the productivity techniques and quality aspects in the criticality safety evaluation of Y-12 packages using the ES-3100 as an example.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

On the Feynman-alpha method for reflected fissile assemblies

The Feynman-alpha method is a neutron noise technique that is used to estimate the prompt neutron period of fissile assemblies. The method and quantity are of widespread interest including in applications such as nuclear criticality safety, safeguards and nonproliferation, and stockpile stewardship; the prompt neutron period may also be used to infer the k eff multiplication factor. The Feynman-alpha method is predicated on time-correlated neutron detections that deviate from a Poisson random variable due to multiplication. Traditionally, such measurements are diagnosed with one-region point kinetics, but two-region models are required when the fissile assembly is reflected. This paper presents a derivation of the two-region point kinetics Feynman-alpha equations based on a double integration of the Rossi-alpha equations, develops novel propagation of measurement uncertainty, and validates the theory. Validation is achieved with organic scintillator measurements of weapons-grade plutonium reflected by various amounts of copper to achieve k eff values of 0.83–0.94 and prompt periods of 5–75 ns. The results demonstrate that Feynman-alpha measurements should use the two-region model instead of the one-region model. The simplified one-region model deviates from the validated two-region models by as much as 10% in the estimate of the prompt neutron period, and the two-region model reduces to the one-region model for small amounts of reflector. The Feynman-alpha estimates of the prompt neutron period are compared to those of the Rossi-alpha approach. The comparative results demonstrate that the Feynman-alpha method is more precise than the Rossi-alpha method and more accurate for k eff < 0.92, whereas the Rossi-alpha method is generally more accurate for higher multiplications. Here, the uncertainty propagation developed in this work should be used for all Feynman-alpha measurements and will therein improve fitting accuracy and appropriate precision estimates.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2019

Fissile material detection using neutron time-correlations from photofission

The detection of special nuclear materials (SNM) in commercial cargoes is a major objective in the field of nuclear security. In this work we investigate the use of two-neutron time-correlations from photo-fission using the Prompt Neutrons from Photofission (PNPF) detectors in Passport Systems Inc.’s (PSI) Shielded Nuclear Alarm Resolution (SNAR) platform for the purpose of detecting ~5 kg quantities of fissionable materials in seconds. The goal of this effort was to extend the secondary scan mode of this system to differentiate fissile materials, such as highly enriched uranium, from fissionable materials, such as low enriched and depleted uranium (LEU and DU). Experiments were performed using a variety of material samples, and data were analyzed using the variance-over-mean technique referred to as Y 2F or Feynman-α. Results were compared to computational models to improve our ability to predict system performance for distinguishing fissile materials. Simulations were then combined with empirical formulas to generate receiver operating characteristics (ROC) curves for a variety of shielding scenarios. We show that a 10 second screening with a 200 μA 9 MeV X-ray beam is sufficient to differentiate kilogram quantities of HEU from DU in various shielding scenarios in a standard cargo container.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2013

Neutron Diffusion Model for Prompt Burst Simulation in Fissile Solutions

A simulation model, based on neutron diffusion, has been developed to predict prompt burst excursions in fissile solutions. This model combines neutron diffusion with simple expressions for radiolytic gas generation and solution temperature change through a neutron cross-section model that accounts for reactivity changes. The simulation model tracks neutron flux as a function of time and position during transients. The model is intended to serve as a general survey tool for the behavior of unreflected aqueous fissile solutions commonly encountered at a variety of nuclear materials processing facilities. The model has been tested against experimental data from the KEWB and Silene research reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

An Evaluation of a Neutron Time Correlated Interrogation Method for Measurement of Fissile Content in Research Reactor Spent Fuel Assemblies

There are many research reactors worldwide that have been, or are being converted from the use of high enrichment uranium (HEU) to low enrichment uranium (LEU, < 20% enriched). The verification of fissile content and initial enrichment of the spent fuel is needed for the effective safeguards of the fuel. The advanced experimental fuel counter (AEFC) was developed for the measurement of spent fuel rods and assemblies from research reactors for safeguards verification. This measurement system contains components for active neutron interrogation, passive neutron totals counting, neutron coincidence counting, and gross gamma-ray counting. This report presents the first application of the time correlated interrogation technique for the measurement of the 235 U content in research reactor spent fuel assemblies. The technique, called time correlated induced fission (TCIF), uses a 252 Cf neutron source to irradiate the fuel assembly, and the subsequent induced fission events in the fissile material are measured by coincidence counting. The doubles rates are enhanced by having the neutron trigger events from both the 252 Cf source and the induced fission neutrons in the same time gate in the coincidence analysis. The average neutrons per fission of the 252 Cf source is 3.76 and the induced fission neutrons for 235 U is 2.44, so the number of neutrons that are produced is higher than for random neutron interrogation. This high effective neutron number increases the multiplicity counting rates and reduces the statistical error. The background coincidence counts from the 252 Cf are reduced by the water in the sample cavity and the polyethylene surrounding the 3 He detector tubes. This method of active neutron interrogation has been applied to the measurement of spent research reactor (IRT) fuel assemblies. The advanced experimental fuel counter (AEFC) was used to compare the TCIF method with the typically used AmLi neutron interrogation source that emits neutrons that are random in time. The statistical uncertainty for the use of the random neutron source (AmLi) and the time correlated source ( 252 Cf) for spent fuel interrogations was evaluated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗