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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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DOE OSTI2024

ANS Winter 2024 Summary: MCCAFE: The Monte Carlo Constructor for ATR Fuel Elements

The Irradiation Experiment Neutronics Analysis Department at Idaho National Laboratory (INL) has implemented a new analysis workflow for experiments in the Advanced Test Reactor (ATR). One key piece of this workflow is the Monte Carlo Constructor for ATR Fuel Elements, or MCCAFE. For each ATR operating cycle, the Reactor and Nuclear Safety Engineering (RNSE) Department first solves the core in eigenvalue mode and depletes the driver fuel materials. In a separate calculation, neutronics analysts model and deplete the materials of one or more irradiation experiments, usually in a series of fixed-source Monte Carlo N-Particle (MCNP) models of the ATR for neutron transport calculations. It was desirable to use the results of the former calculations to inform the models of the latter. MCCAFE is a Python program developed using American Society of Mechanical Engineers Nuclear Quality Assurance-1 procedures at INL. Its purpose is to take the calculated results from the RNSE depletion solutions and the measured or projected operating parameters from the Nuclear Data Management and Analysis System (NDMAS) to generate fixed-source models of the ATR core at given points in time across one or more cycles.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Absolute Dosimetry Model of the University of Washington Clinical Neutron Therapy System (CNTS)

The University of Washington (UW) Clinical Neutron Therapy System (CNTS) generates a fast neutron beam using 50.5 MeV protons incident on a beryllium target. The CNTS is mainly used for fast neutron therapy in the treatment of cancer but is also used for radiation effects testing and the production of rare isotopes, or isotopes with a short half-life, for medical imaging and therapy. The accelerator and treatment head is modeled using the MCNP6 Monte Carlo code to enable beam characterization (neutron and photon fluence, linear energy transfer, secondary particle production), study the behavior of neutron dosimeters and ionization chamber (IC) responses, and to conduct first principle studies of the relative biological effectiveness (RBE) of fast neutrons compared to {sup 60}Co γ-rays. Measurements of neutron dose in selected CNTS fields were made using a model-17 ion chamber (IC-17) flowing methane-based, tissue-equivalent (TE) gas. As a first approximation, we compared IC-17 measurements to the MCNP calculated dose in a thin layer of water at selected locations along the central axis of the beam. To examine the impact of the composition of the chamber fill-gas and the chamber design characteristics, we also developed a high-resolution (∼ 50 μm) MCNP model of the IC-17 from a computerized tomography (CT) scan. Initial results indicate that the dose response of the IC-17 ion chamber is up to 30% larger (smaller?) compared to the dose to water for neutrons with energies >20 MeV. Differences in chamber and water response as large as 800% occur for neutrons with kinetic energies below 1 keV. The simulations of IC-17 response (and dose to water) for CNTS fast neutron spectra match measured doses within 4.6% for all field sizes and irradiation geometries considered. This reported studies provide additional evidence establishing use of the IC-17 ion chamber model or a thin layer of water as a reasonable surrogate for CNTS fast neutron dose calculations.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Practical Insights on Applying Simulation-Based Control Methods in Experimental Studies

Advanced nuclear reactors are crucial to the future of energy both in the United States and around the globe. In contrast to the current operating fleet, they are characterized as being deployable in remote locations and able to operate in semi-autonomous or autonomous fashion. This leap forward necessitates a new reactor control paradigm. Because advanced nuclear reactors are still under development in the United States, the creation of new control methods to achieve autonomous operations has been based on systems modeling and simulation. However, an important factor in successfully deploying these new control methods is the ability to seamlessly transition from simulation environments to real-world settings. Control methods tested in both simulation and experimental settings need to be investigated in the context of advanced reactor applications. This work developed a series of simple controllers for Idaho National Laboratory (INL)’s Microreactor Applications Research Validation and Evaluation (MARVEL) microreactor operating in load-following scenarios. These controllers were tested in both simulation and experimental settings, and a comparative performance analysis was performed. The simulation tests leveraged the Control and Optimization Modular Modeling Application for Nuclear Deployment (COMMAND) software developed in a previous stage of the current effort, along with the MARVEL Reactor Excursion and Leak Analysis Program (RELAP5-3D) and Monte Carlo N-Particle (MCNP) models. The experimental tests leveraged the COMMAND software, MARVEL models, and the U.S. Department of Energy Microreactor Program’s Microreactor Automated Control System (MACS). MACS was developed to serve as a control method testbed. It was customized to mirror the MARVEL microreactor, and COMMAND enabled MACS to emulate the physics of MARVEL. The load-following controller was developed using the simulation platform, with efforts to emulate real systems by introducing actuator saturation and noise. These factors were incrementally accounted for in the controller design. After finalizing the controller design, it was implemented with the experimental setup. The experimental conditions tested included an initial test under conditions similar to the final simulation test, and two additional scenarios. The first scenario introduced additional actuator saturation to account for equipment aging over time, which was unknown to the controller. The second scenario introduced sensor delay, a phenomenon anticipated with the use of remote operations or wireless communication in advanced reactors. These tests revealed several notable differences. While the controller performed well in simulation, it exhibited several limitations when transitioning to hardware. The main challenges involved maintaining the steady-state target power, as evidenced by larger error values between the true reactor power and setpoint power, as well as persistent oscillations in controlled reactor power. These issues could lead to unacceptable transient conditions in real reactor testing. Introducing actuator aging and stochastic delays in the experimental setup significantly impacted controller performance, resulting in increased overshoot and undershoot, and exacerbated error and oscillations previously mentioned. These findings underscore the importance of experimental testbeds for testing and validating control methods, as controllers developed using only theory and/or simulation may perform unexpectedly when applied to actual hardware. This research emphasizes the need for an experimental testbed for achieving such validation.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2018

Alpha Air Sample Counting Efficiency Versus Dust Loading: Evaluation of a Large Data Set

Dust loading on air sample filters is known to cause a loss of efficiency for direct counting of alpha activity on the filters, but the amount of dust loading and the correction factor needed to account for attenuated alpha particles is difficult to assess. In this paper, correction factors are developed by statistical analysis of a large database of air sample results for a uranium and plutonium processing facility at the Savannah River Site. As is typically the case, dust-loading data is not directly available, but sample volume is found to be a reasonable proxy measure; the amount of dust loading is inferred by a combination of the derived correction factors and a Monte Carlo model. The technique compares the distribution of activity ratios [beta/(beta + alpha)] by volume and applies a range of correction factors on the raw alpha count rate. The best-fit results with this method are compared with MCNP modeling of activity uniformly deposited in the dust and analytical laboratory results of digested filters. Finally, a linear fit is proposed to evenly-deposited alpha activity collected on filters with dust loading over a range of about 2 mg cm -2 to 1,000 mg cm -2 .

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Description of and MCNP Input Files for a Duct-Shielding Benchmark

A brief description of a one-, two-, and three-legged duct shielding benchmark and the MCNP model input files are given. The duct-shielding benchmark measurements were taken by Maerker and Muckenthaler [1967] at the Oak Ridge Tower Shielding Facility. The ducts were constructed from reinforced concrete. The experimental values are found to fall between MCNP simulations with homogenized reinforced concrete and the basic concrete.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2019

Summing Corrections For La-140 On Counter 137 [Slides]

The counting of filters for gamma activity is often done with the sample situated very close or wrapped around the detector. This improves the counting statistics but can increase the probability of summing and affect the calculation of yield. An MCNP model of counter 137 (a 65% HPGe detector) was used to compute the total and peak efficiencies for two sample types, 40 and 20. The type 40 is a long filter strip wrapped in a Marinelli beaker which sits around the detector in an annular fashion. The type 20 is a 8.4 in radius filter, quarter-folded and placed flat on the detector face. A comparison of the peak efficiency computed in MCNP was compared to the measured efficiency to ensure consistency. Using the measured peak efficiency and MCNP-computed total efficiency, the TrueCoinc program was run to estimate the summing values for La-140 peaks. The results showed the summing out of two peaks (487 and 1596 keV) was around 6-8% in for both sample types.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Design Optimization of A generic Fissile Solution for Mo99 production using Electron Beam-based Neutron Generator using MCNP+CFD

Molybdenum-99 ( 99 Mo) is a medical isotope used in 80% of all medical imaging procedures today. However, 99 Mo is not directly administered to patients for imaging; its decay product, technetium-99m ( 99 mTc), is a metastable isotope with a half-life of ~6 hours. Such a short half-life does not allow for production, separation, and application to a patient before decaying further. Therefore, the solution is to produce the precursor to 99 mTc, 99 Mo. 99 Mo has a half-life of 66 hours, long enough to produce the material and send it to the desired location before it fully decays. Under National Nuclear Security Administration’s (NNSA) highly enriched uranium (HEU) minimization mission, the Material Management and Minimization (M3) office leads the Molybdenum-99 (Mo-99) program. A major goal of the Mo-99 program is to develop methods of producing 99 Mo using low enriched uranium (LEU) opposed to the traditionally used HEU domestically in the U.S. Internationally, LEU is the new standard for use in research reactors and isotope production facility. In this study, we introduce a generic fissile solution system for 99 Mo production facility meeting the design requirement of 1) LEU usage as fissile material and 2) subcritical fissile system coupled with a horizontal electron beam (E-beam) accelerator. The focus of this research is to optimize a fissile solution configuration for maximum 99 Mo production yield using MCNP model, and ultimately implement the optimized design with computational fluid dynamics (CFD) model to understand thermal hydraulic behavior and solution convection characteristic. The report consists of two parts: 1) isotope yield calculation and criticality analysis with a wide range of parametric test matrix, and 2) thermal hydraulic analysis for solution mixing and cooling assessment.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Characterizing the Thermal Column Beam in the University of Massachusetts Lowell Research Reactor

The University of Massachusetts Lowell operates a 1 MW, light water cooled and moderated Research Reactor (UMLRR) fueled by low-enriched uranium to generate neutron flux that is used at a variety of in-core and beam port facilities within the reactor to perform irradiation experiments for academic and industrial research. One of the irradiation sites is a graphite-moderated thermal column designed to deliver a collimated beam of thermal neutrons to a neutron radiography imaging system. A detailed characterization of the neutron beam is essential for the design and interpretation of experiments at the facility and for radiation safety. This work aimed to characterize the neutron flux and energy spectrum in the UMLRR thermal column beam by employing experimental and computational methods known as the foil activation and spectrum unfolding technique. A previously existing Monte Carlo N-Particle transport (MCNP) model of the UMLRR reactor core has been extended to the thermal column where the neutron flux and energy spectrum have been computed and experimentally validated. This work employed the foil activation and spectrum unfolding technique along with neutron transport computation to extend the knowledge of the neutron fluence and energy spectrum from the UMLRR core to the adjacent thermal column beam. Radiography systems are used for materials characterization, studying dynamic behavior of engineering systems, and for nuclear forensics applications. Foil activation is an experimental method where thin foils of highly pure materials are placed in the beam where interactions between the neutrons and the foil material activate radioisotopes in the foils whose activities can be measured after the irradiation. Downstream computation with MCNP and the spectrum unfolding code SAND can deconvolute the neutron fluence and energy spectrum that must have been necessary to induce the production rates of radioisotopes that were obtained through foil activation. The necessary ingredients for successful spectrum unfolding are the experimentally measured production rates and some a priori (initial guess) knowledge of the flux spectrum. The thermal column is equipped with a neutron radiography imaging system, so any experimenters who desire to use neutron radiography at UMass Lowell will benefit significantly from knowing the characteristics of the neutron beam. Comprehensive knowledge of these characteristics of the beam is essential to compare and benchmark results obtained for similar experiments at other facilities. The Integrated Nuclear Security and Safeguards Laboratory (INSSL) is pursuing the investigation of the performance of self-healing materials under stress-strain environments and the characterization of radiation detectors and special containers for storing and transporting nuclear material. This work contributes to the improvement of the current radiography system, expands the utility of the system for nuclear safeguards and security application and provides information that will enhance the general safety of the UMLRR experimental facilities. The main objective of this work has been achieved; an appropriate neutron flux spectrum for the UMLRR thermal column beam has been determined using the foil activation method followed by spectrum unfolding. There now exists an experimentally validated computational model of the neutron flux spectrum and the thermal column's neutron environment is characterized. We can conclude that the thermal column is well characterized; an appropriate neutron flux spectrum has been determined through an experimentally validated computational model. Future experimenters interested in the thermal column may now expand on the computational model or use the spectral information for their own purposes. The knowledge gained herein will be helpful for the INSSL's future work using neutron radiography to analyze the effectiveness of self-healing materials and special containers. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Experimental demonstration and modeling of a robotic neutron detector with spectral and directional sensitivity for treaty verification

International safeguards and arms control agreements often require labor-intensive, intrusive onsite inspections to perform verification tasks. The ability to localize a neutron source and/or characterize a neutron field may be imperative for identifying anomalies. We are interested in the role of autonomous mobile robots, which, if designed properly, may be more effective and efficient and less intrusive than their human counterparts. Toward developing such a capability, we, for this study, have constructed the N-SpecDir Bot, comprised of three boron-coated straw detectors azimuthally-distributed within a cylinder of high-density polyethylene, which is mounted on an omni-directional robotic platform. Our N-SpecDir Bot is specifically designed to provide spectral and directional sensitivity, in addition to gross counts, by utilizing the signals from the three detectors. The detection system has been extensively characterized by MCNP modeling, which has been benchmarked to experiments conducted at the Princeton Plasma Physics Laboratory. We demonstrate the spectral and directional sensitivity experimentally and in simulation, and provide a simple yet robust method for directional measurements.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2026

Evaluation of Sandia NCS Benchmark Suite Updates

The Sandia Nuclear Criticality Safety (NCS) program’s benchmark suite was recently updated. This suite is used to ensure that NCS calculations using computer-based neutron transportation codes have an established baseline comparison of calculated versus known experimental results. The Evaluated Nuclear Data File (ENDF) version used for the MCNP models in the suite was changed from ENDF/B-VII.1 to ENDF/B-VIII.0. Additionally, relevant thermal scattering law data libraries (TSLs) were updated. The sensitivity of the calculational bias of each benchmark model to these changes is discussed. Implementation of the ENDF/B-VIII.0 library and updated TSLs results in improvements to bias distribution in the intermediate enriched uranium, plutonium, and mixed uranium–plutonium (IEU, PU, and MIX) fissionable material benchmark categories, but a small bias increase in low- and high-enriched uranium categories (LEU and HEU, respectively). The results also highlight the sensitivity of the benchmarks, with average lethargy of neutrons causing fission energies (EALF) in the intermediate energy range to ENDF/B library changes. The most numerous bias changes were observed in the thermal energy region when transitioning from ENDF/B-VII.1 to ENDF/B-VIII.0. In conclusion, most of the unique bias changes observed in MCNP 6.3.0 between the two nuclear data libraries were in the LEU-COMP-THERM evaluation subset.

ICSBEP↗
DOE OSTI2020

A Mercury Model of the Molly-G Fast Burst Reactor (FBR)

In order to design neutron irradiation experiments at a given nuclear reactor, one requires a representative neutron spectrum for the configuration of the reactor that will exist during the experiment. This usually implies modeling the operation of the reactor using a neutron transport code. Since the modeling of pulsed neutron reactors is quite complex, requiring coupled modeling of neutronics and thermo-structural response, the typical methodology employs analysis of the neutron spectrum which is obtained from a static keff calculation. While this time-independent Eigenfunction is not truly representative of that during a reactor pulse, we postulate that it is a reasonable representation for our purposes. The Nuclear Survivability (NS) Program at LLNL is considering fielding such experiments at the Molly-G (‘Molybdenum Godiva-II’) Fast Burst Reactor (FBR) at the White Sands Missile Range (WSMR). The goal of these experiments is to (a) test / proof / calibrate diagnostics, and (b) test the efficacy of neutron shield configurations, before fielding at other nuclear reactors, such as the Annular Core Research Reactor (ACRR) at SNL/NM. After attempting to secure either drawings or a copy of an existing MCNP model of the Molly-G FBR from various institutions, it was decided to develop a new model for use with the Mercury Monte Carlo transport code. The model described herein will be used to design these experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Thermal Flux Maximization from a DD Neutron Generator

The University of Sharjah recently acquired a model DD-109.4M neutron generator from Adelphi Technology. The generator is moderated and has three neutron ports which are capable of providing fast and thermal neutrons. The neutron ports are plugged with polyethylene layers of different thickness in order to control the neutron flux at different regions in the port. Neutron Activation Analysis (NAA) is an important analysis technique to determine the amounts of isotopes in a sample at different concentrations. For many isotopes the ability to precisely determine their concentrations is enhanced by utilizing thermal neutrons in the activation process; where the capture cross section is greatest. It is thus important to subject the sample to the maximum thermal flux in order to minimize the time required to activate the sample and consequently extend the lifetime of the neutron generator. A detailed MCNP model of the neutron generator has been developed for the purpose of determining the moderator thickness required to maximize the thermal neutron flux in the neutron port. The maximization of the neutron flux must therefore take into account both the forward and backscattered neutrons reaching the sample. Also provided in this work is a mapping of the thermal flux at different moderator thicknesses in order to study the variation of the flux over the area of the sample. The developed model will be used as the basis for optimizing experiments and predicting their results. A study was performed to determine the moderator thickness needed to maximize the thermal neutron flux that can be obtained from a DD neutron generator. This has a significant role in the efficient utilization of the neutron generator. The maximum thermal neutron flux was found at approximately a 3.2 cm moderator thickness. The thermal neutron flux is resulting from both the directly moderated neutrons and the backscattered neutrons from the surrounding polyethylene. The results obtained in this work will be verified experimentally by irradiating samples at different moderator thickness and performing a neutron activation analysis for these samples. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Accurate activity determination of a californium neutron source

Accurately counting analog events requires constructing an electronics chain that produces one count for each input pulse. Here, in this work, we review the use of Nuclear Instrumentation Module electronic units for counting neutron capture events in a 3He tube. We identify two unique types of false trigger events in a leading-edge discriminator and show how a dual timer module can be used to produce a veto window to exclude these events. We use the constructed electronics chain to build an apparatus to measure neutron pulses from a 252 Cf neutron source. We compare the measurements with a Monte Carlo N-Particle (MCNP) model to determine the activity of the neutron source. Furthermore, by making additional measurements with borated polyethylene attenuators between the source and detector, we are able to determine the boron concentration of the polyethylene. This technique provides accurate determination of the source activity to a precision of 2.8% at the k=1 level. The method used is simple, inexpensive, and requires no additional calibrated instruments.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Impact of Higher Fidelity Design Iterations on Critical System Criteria

Nuclear criticality experiments are effective at informing the performance of nuclear data libraries across many applications. This work explores the implications of refining critical experiment MCNP models from their low fidelity optimization phase to penultimate neutronic models. Specifically, this work is focused on two series of plutonium fueled experiments funded through internal programs at Los Alamos National Laboratory building off previous efforts under the EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) collaboration. Thales, the first of the two collaborations, is a fast spectrum Ta-reflected plutonium experiment to support operations at PF-4. The second experiment are twin configurations designed to target the intermediate energy cross sections in 239 Pu. Motivation for this experiment stems from the PARallel Approach of Differential and InteGral Measurements (PARADIGM) collaboration which hopes to achieve a significant reduction in 239 Pu cross section uncertainties in the intermediate region.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2021

Performance Demonstration of Self-Powered Neutron Detectors for Steady-State Reactor Operations

The irradiation testing of sensors in reactors is a crucial step towards calibrating and qualifying sensors prior to their deployment in experiments. This report details the process toward qualifying and calibrating custom-designed rhodium-based self-power neutron detectors (Rh-SPNDs) for steady-state reactor irradiations. This process serves to both demonstrate the performance capabilities of Rh-SPNDs as well as to provide experimental data for development of a sensor sensitivity model. Two designs of Rh-SPNDs were tested in various reactors to demonstrate: detection resolution in a low neutron flux environment, a delayed-response compensation technique, output linearity in a large range of neutron flux, and measurement accuracy verified with dosimetry. The detection resolution and compensation technique was demonstrated in the AGN-201m reactor at Idaho State University. The irradiation confirmed the sensors’ capability to perform steady-state operations in a low neutron flux of ~2E8 n/cm 2 -sec. Sensor output linearity coupled with the delayed-response compensation was investigated at the neutron radiography reactor at Idaho National Laboratory. A Rh-SPND was irradiated to neutron fluxes ranging from 2E8 to 2E13 n/cm2-sec range. The measured data demonstrated a wide and linear range of operation with a measured linear sensitivity of 1.0129 ×10 -13 A/W with a correlation-squared value of r 2 =0.9927. The measurement accuracy was investigated at the Advanced Test Reactor Critical reactor. The SPNDs were inserted into a test vehicle with collocated flux wires. Two irradiations with different flux levels were performed, and the SPNDs relative measurement between the two irradiations was calculated to be 1.2613 ± 0.0153 for the small SPND design and 1.1809 ± 0.0108 for the large SPND design. Both SPND measurements fell between the co-axial dosimetry result, which reported 1.218 ± 0.047. Additionally, the preliminary MCNP model for calculating SPND sensitivity was developed in parallel to this work. Modeled neutron spectrum with measured magnitude was used for inputs to determine the simulated SPND output. The results showed an overestimation of signal strength by a factor of 5, which was expected because of model simplification. This leads to future modeling work to account for signal losses from additional physical properties, including high temperature environments for FY-21.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

High pressure Marinelli for counting low activity compressed gas samples

Quantification of low-activity noble gases in air is typically accomplished through separation of the noble gas from air followed by radiometric assay. Here, this work is aimed at quantification of radioactive noble gas in air without extraction. A high pressure aluminum Marinelli counting vessel was designed and fabricated that can be placed on a coaxial high purity germanium detector for gamma counting. Characterization of the performance of this Marinelli using MCNP modeling, large excesses of activity, and low-activity noble gas in air is discussed. Minimum detectable concentrations achieved during a 24 h count are: 5, 10, 50, and 1 Bq/m 3 for 133 Xe, 133m Xe, 131m Xe, and 135 Xe, respectively.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Measurements of effective actinides neutron capture cross sections in a cadmium-filtered epithermal neutron spectrum

In this work, a total of 20 actinide samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL): 233 U (1), 235 U (1), 236 U (1), 238 U (1), 237 Np (2), 239 Pu (1), 240 Pu (1), 242 Pu (2), 244 Pu (2), 241 Am (2), 243 Am (2), 244 Cm (2), and 248 Cm (2). The samples were loaded in a basket containing a 1.1-mm thick cadmium filter. The presence of the filter drastically changes the neutron spectrum in the samples and accentuates the reaction rates in the resonance region. Isotopic ratios were determined using Multi-Collector Inductively Coupled Plasma Mass Spectrometry (MC-ICP-MS) that allowed inferring effective capture cross sections. Detailed best-estimate MCNP models of the experiment using ENDF/B-VII.0 nuclear data were developed to calculate the effective neutron cross sections and, subsequently, compare them with the cross sections inferred from experimental results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

The Prompt Fission Neutron Spectrum of 235 U(n,f) below 2.5 MeV for Incident Neutrons from 0.7 to 20 MeV

New prompt fission neutron spectrum measurements are reported for 235 U(n,f) reactions induced by neutrons with energies from 0.7 to 20 MeV. These measurements cover outgoing neutron energies from 2.5 MeV down to 10 keV, using an array of 6 Li-glass scintillators for neutron detection and a double time-of-flight technique. The neutrons were produced at the Weapons Neutron Research facility of the Los Alamos Neutron Science Center. A detailed MCNP ® model of the experimental equipment and the surrounding room was used to interpret the experimental results. Backgrounds were measured in situ, making use of the time-dependent singles rates of the various detectors with asynchronous readout from waveform digitizers. The results presented here have been included in a re-evaluation of the fission neutron spectra for this fissioning system, a description of which is presented elsewhere in this issue.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗