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At least 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Moderator design studies for a new neutron reference source based on the $\mathrm{D}$–$\mathrm{T}$ fusion reaction

The radioactive isotope Californium-252 ( 252 Cf) is relied upon internationally as a neutron calibration source for ionizing radiation dosimetry because of its high specific activity. The source may be placed within a heavy-water (D 2 O) moderating sphere to produce a softened spectrum representative of neutron fields common to commercial nuclear power plant environments, among others. Due to termination of the U.S. Department of Energy loan/lease program in 2012, the expense of obtaining 252 Cf sources has undergone a significant increase, rendering high output sources largely unattainable. On the other hand, the use of neutron generators in research and industry applications has increased dramatically in recent years. Neutron generators based on deuterium-tritium (D-T) fusion reaction provide high neutron fluence rates and, therefore, could possibly be used as a replacement for 252 Cf. To be viable, the 14.6 MeV D-T output spectrum must be significantly moderated to approximate common workplace environments. Here this paper presents the results of an effort to select appropriate moderating materials and design a configuration to reshape the primary neutron field toward a spectrum approaching that from a nuclear power plant workplace. A series of Monte-Carlo (MCNP) simulations of single layer high- and low-Z materials are used to identify initial candidate moderators. Candidates are refined through a similar series of simulations involving combinations of 2 to 5 different materials. The simulated energy distribution using these candidate moderators are rated in comparison to a target spectrum. Other properties, such as fluence preservation and/or enhancement, prompt gamma production and other characteristics are also considered.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

Boron Filled Siloxane Polymers for Radiation Shielding

The purpose of the present work was to evaluate changes to structure-property relationships of 10 B filled siloxane-based polymers when exposed to nuclear reactor radiation. Highly filled polysiloxanes were synthesized with the intent of fabricating materials that could shield high neutron fluences. The newly formulated materials consisted of cross-linked poly-diphenyl-methylsiloxane filled with natural boron and carbon nanofibers. This polymer was chosen because of its good thermal and chemical stabilities, as well as resistance to ionizing radiation thanks to the presence of aromatic groups in the siloxane backbone. Highly isotopically enriched 10 B filler was used to provide an efficient neutron radiation shield, and carbon nanofibers were added to improve mechanical strength. This novel polymeric material was exposed in the Annular Core Research Reactor (ACRR) at Sandia National Labs to five different neutron/gamma fluxes consisting of very high neutron fluences within very short time periods. Thermocouples placed on the specimens recorded in-situ temperature changes during radiation exposure, which agreed well with those obtained from our MCNP simulations. Changes in the microstructural, thermal, chemical, and mechanical properties were evaluated by SEM, DSC, TGA, FT-IR NMR, solvent swelling, and uniaxial compressive load measurements. In conclusion, our results demonstrate that these newly formulated materials are well-suitable to be used in applications that require exposure to different types of ionizing conditions that take place simultaneously.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

An approach towards plastic scintillators from thermally activated delayed fluorescent dyes and cross-linkable bismuth compounds

Cross-linked polyvinyltoluene (PVT) based plastic scintillators loaded with dyes exhibiting thermally activated delayed fluorescence (TADF) and various amounts of (caproate)di-(methacrylate) bismuth (CMB) were developed. The co-polymerization of vinyl toluene monomer loaded with different weight ratios of CMB and 1 wt% of TADF dye yielded highly transparent cross-linked plastic scintillators with intense green photoluminescence. The thermal and mechanical properties of the scintillators were investigated using thermogravimetric analysis (TGA), differential scanning calorimetry (DSC) and micro indentation techniques. Here, their performance for radiation detection was assessed by performing gamma spectroscopy using a Cs-137 source and benchmarked to commercial plastic scintillators. The loading of CMB was guided by Monte Carlo N-Particle (MCNP) simulations. The results of this study show that the cross-linking approach and the use of TADF dyes yield scintillators with good mechanical properties with a uniaxial yield strength up to ~66 MPa for compositions loaded with 40 wt% of CMB and high optical quality. However, the light yield in gamma spectroscopy experiments is found to be reduced with the addition of CMB, suggesting a trade-off between mechanical yield strength and light yield in gamma spectroscopy.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on “fingerprinting” casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17 x 17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4 He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244 Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

An Improved Energy Deposition Model in MPACT and Explicit Heat Generation Coupling with CTF

The default energy deposition model in the CASL neutronics code MPACT assumes all fission energy is deposited locally in fuel rods. Furthermore, equilibrium delayed energy release is assumed for both steady-state and transient calculations. These approximations limit the accurate representation of the heat generation distribution in space and its variations over time, which are essential for power distribution and thermal-hydraulic coupling calculations. In this paper, an improved energy deposition model is presented in both the spatial and time domains. Spatially, the energy deposition through fission, neutron capture, and slowing-down reactions are explicitly modeled to account for the heat generation from all regions of a reactor core, and a gamma smearing scheme is developed that utilizes the gamma sources from neutron fission and capture. In the time domain, the delayed energy release is modeled by solving an additional equation of delayed heat emitters, similar to the equation of delayed neutron precursors.To allow the explicit heat generation coupling, the interfaces between MPACT and CTF were updated to transfer separate heat sources for different material regions (fuel, clad, moderator and guide tube). The results show that the distributions of the energy deposition between MPACT and MCNP agree very well for various 2-D assembly and quarter-core problems without TH feedback. The MPACT/CTF coupled calculation for the hot full power quarter-core case exhibited a reduced peak pin power by 2.3% and a reduced peak fuel centerline temperature by 17 K when using the explicit energy deposition and heat transfer. The new model also shows a maximum 100 pcm keff effect on assembly depletion problems and an increased overall energy release by 7% in a PWR reactivity-initiated accident (RIA) problem.

Liu, Yuxuan↗
DOE OSTI2023

The Covariance of PFNS Results from the Chi-Nu Experiment

The prompt fission neutron spectrum (PFNS) from neutron-induced fission is a fundamental quantity for the behavior of nuclear reactors, and has been measured many times on a wide variety of nuclei and covering different ranges of incident and emitted neutron energies. However, results from past measurements are frequently called into question in modern nuclear data evaluations because of a lack of thorough experimental documentation and incomplete uncertainty analyses. The Chi-Nu experiment at Los Alamos National Laboratory was designed to produce high-precision measurements of the PFNS of major actinides over a wide range of incident and emitted neutron energies, and with the documentation and covariance analysis required to ensure that the results of this experiment maintain their impact long into the future, thereby avoiding this pitfall of past measurements. In this work we describe the Chi-Nu experiment along with summaries of the treatment of and methods developed to address two important components of the analysis of Chi-Nu data: random-coincidence backgrounds and MCNP simulations. Furthermore, we describe the first results for correlations not just between all data points collected on a single target nucleus, but also between all data points from separate Chi-Nu measurements on 235 U and 239 Pu. These correlations are important for accurately calculating ratios of the PFNS from one actinide to another, which are rare and can be informative for nuclear data evaluation efforts.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Resolved Resonance Region Analysis of 206 Pb, 207 Pb, and 208 Pb for Next Generation Lead-Cooled Fast Systems

Recently, great interest has been generated in using lead as a coolant for fast neutron systems and as a result it is important to investigate the ENDF/B-VIII.0 isotopic evaluations that comprise stable le ad. To this end, resonance parameters for 206 Pb, 207 Pb, and 208 Pb were re-evaluated because their resolved reso­nance regions extend beyond 0.5 MeV meaning resonance parameters used in reconstructing cross sections and elastic scattering angular distributions impact fast systems. The impact of resonance parameters is demon­strated by the differences between the evaluations in predicting experimental results from the fully modeled RPI Quasi-Differential Scattering Experiment via MCNPv6.2. In addition, MCNP KCODE calculations of lead-sensitive fast spectra critical benchmarks showed variations of keff on average of 400 pcm, caused solely from differences in elastic scattering angular distributions in 208 Pb. Re-evaluation entailed fitting data with the program SAMMY.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Detection of special nuclear materials with the associate particle technique

In the frame of the French trans-governmental R and D program against chemical, biological, radiological, nuclear and explosives (CBRN-E) threats, CEA is studying the detection of Special Nuclear Materials (SNM) by neutron interrogation with fast neutrons produced by an associated particle sealed tube neutron generator. The deuterium-tritium fusion reaction produces an alpha particle and a 14 MeV neutron almost back to back, allowing tagging neutron emission both in time and direction with an alpha particle position-sensitive sensor embedded in the generator. Fission prompt neutrons and gamma rays induced by tagged neutrons which are tagged by an alpha particle are detected in coincidence with plastic scintillators. This paper presents numerical simulations performed with the MCNP-PoliMi Monte Carlo computer code and with post processing software developed with the ROOT data analysis package. False coincidences due to neutron and photon scattering between adjacent detectors (cross talk) are filtered out to increase the selectivity between nuclear and benign materials. Accidental coincidences, which are not correlated to an alpha particle, are also taken into account in the numerical model, as well as counting statistics, and the time-energy resolution of the data acquisition system. Such realistic calculations show that relevant quantities of SNM (few kg) can be distinguished from cargo and shielding materials in 10 min acquisitions. First laboratory tests of the system under development in CEA laboratories are also presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Heavy reflector experiments in the IPEN/MB-01 reactor: Stainless steel, carbon steel and nickel

New experiments devoted to the measurements of physical parameters of a light water core surrounded by a heavy reflector were performed in the IPEN/MB-01 research reactor facility. These experiments comprise three sets of heavy reflector (SS-304, Carbon Steel, and Nickel) in a form of laminates around 3 mm thick. Each set was introduced individually in the west face of the core of the IPEN/MB-01 reactor. The aim here is to provide high quality experimental data for the interpretation and validation of the SS-304 heavy reflector calculation methods. The experiments of Carbon Steel, which is composed mainly of iron, and Nickel were performed to provide a consistent and an interpretative check for the SS-304 reflector experiment. The experimental results comprise critical control bank positions, temperatures and reactivities as a function of the number of the plates. Particularly to the case of Nickel, the experimental data are unique of its kind. The theoretical analysis was performed by MCNP-5 with the nuclear data library ENDF/B-VII.0. It was shown that this nuclear data library has a very good performance up to thirteen plates and overestimates the reactivity for higher number of plates independently of the type of the reflector.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Reactivity worth measurements at the IPEN/MB-01 nuclear reactor

Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. The objective of this work was to perform a series of experiments of reactivity worth measurements, using a digital reactivity meter developed at IPEN. The experiments employed small metallic and ceramic samples inserted in the central region of the core of the experimental IPEN/MB-01 reactor. The theoretical analysis was performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, and the ENDF/B-VII.0 nuclear data library.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Feasibility study for measurement of insulation compaction in the cryogenic rocket fuel storage tanks at Kennedy Space Center by fast/thermal neutron techniques

The liquid hydrogen and oxygen cryogenic storage tanks at John F. Kennedy Space Center (KSC) use expanded perlite as thermal insulation. Some of the perlite may have compacted over time, compromising the thermal performance and also the structural integrity of the tanks. Neutrons can readily penetrate through the 1.75 cm outer steel shell and through the entire 120 cm thick perlite zone. Neutrons interactions with materials produce characteristic gamma rays which are then detected. In compacted perlite the count rates in the individual peaks in the gamma ray spectrum will increase. Portable neutron generators can produce neutron simultaneous fluxes in two energy ranges: fast (14 MeV) and thermal (25 meV). Fast neutrons produce gamma rays by inelastic scattering which is sensitive to Si, Al, Fe and O. Thermal neutrons produce gamma rays by radiative capture in prompt gamma neutron activation (PGNA), which is sensitive to Si, Al, Na, K and H among others. The results of computer simulations using the software MCNP and measurements on a test article suggest that the most promising approach would be to operate the system in time-of-flight mode by pulsing the neutron generator and observing the subsequent die away curve in the PGNA signal.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

Dosimetric study of surface applicators of HDR brachytherapy GammaMed Plus equipment

The cone type surface applicators used in HDR brachytherapy for treatment of small skin lesions are an alternative to be used with both electron beams and orthovoltage X-ray equipment. For a good treatment planning is necessary to know the dose distribution of these applicators, which can be obtained by experimental measurement and Monte Carlo simulation as well. In this study the dose distribution of surface applicators of 3 and 3.5 cm diameter, respectively of HDR brachytherapy GammaMed Plus equipment has been estimated using the Monte Carlo method, MCNP code. The applicators simulated were placed on the surface of a water phantom of 20 × 20 × 20 cm and the dose was calculated at depths from 0 to 3 cm with increments of 0.25 mm. The dose profiles obtained at depth show the expected gradients for surface therapy.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of the control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Correction of steel casing effect for density log using numerical and experimental methods in the slim borehole

Density log is widely applied for a variety of fields such as the petroleum exploration, mineral exploration, and geotechnical survey. The logging condition of density log is normally open holes but there are frequently cased boreholes. The primary calibration curve by slim hole logging manufacturer is normally the calibration curves for the variation of borehole diameter. In this study, we have performed the correction of steel casing effects using numerical and experimental methods. We have performed numerical modeling using the Monte Carlo N-Particle (MCNP) code based on Monte Carlo method, and field experimental method from open and cased hole log. In this study, we used the FDGS (Formation Density Gamma Sonde) for slim borehole with a 100 mCi 137Cs source, three inch borehole and steel casing. The casing effect between numerical and experimental method is well matched.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

Radiography benchmark 2014

The purpose of the 2014 WFNDEC RT benchmark study was to compare predictions of various models of radiographic techniques, in particular those that predict the contribution of scattered radiation. All calculations were carried out for homogenous materials and a mono-energetic X-ray point source in the energy range between 100 keV and 10 MeV. The calculations were to include the best physics approach available considering electron binding effects. Secondary effects like X-ray fluorescence and bremsstrahlung production were to be taken into account if possible. The problem to be considered had two parts. Part I examined the spectrum and the spatial distribution of radiation behind a single iron plate. Part II considered two equally sized plates, made of iron and aluminum respectively, only evaluating the spatial distribution. Here we present the results of above benchmark study, comparing them to MCNP as the assumed reference model. The possible origins of the observed deviations are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Effect on the tritium breeding ratio for a distributed ICRF antenna in a DEMO reactor

The paper reports results of MCNP-5 calculations to assess the effect on the Tritium Breeding Ratio (TBR) when integrating a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna in the blanket of DEMO fusion power reactor. The calculations consider different parameters such as the ICRF covering ratio and the type of breeding blanket including the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) concepts. For an antenna with a full toroidal circumference of 360°, located poloidally at 40° with a poloidal extension of 1 m, the reduction of the TBR is −0.349% for the HCPB blanket and −0.532% for the HCLL blanket. The distributed ICRF antenna is thus shown to have only a marginal effect on the TBR of the DEMO reactor.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Molybdenum carbide nanoparticles as catalysts for oil sands upgrading: Dynamics and free-energy profiles

There is no doubt that a huge gap exists in understanding heterogeneous catalysis between a cluster model of a few atoms and a bulk model of periodic slabs. Nanoparticles, which are crucial in heterogeneous catalysis in industry, lie in the middle of the gap. We present here our work on the computational modelling of molybdenum carbide nanoparticles (MCNPs) as the catalysts for the upgrading of oil sands in the in-situ environment, using benzene hydrogenation as a model reaction. With a cluster model, efforts were first made to understand the mechanism of the reaction with a density functional theory (DFT) study on the adsorption of benzene and its hydrogenation product – cyclohexane, as well as the cyclic hydrogenation reaction intermediates on the Mo{sub 2}C(0001) surface. From the thermodynamic data, along with literature information, it was found that the benzene hydrogenation reaction on molybdenum carbide happens most likely through a Langmuir-Hinshelwood mechanism with the gradual lifting up of the benzene molecule. The electron localization function (ELF) was then used to help understand the nature of the interactions between the MCNPs, identifying strong multi-center interactions between the adsorbates and the MCNPs. To enable the treatment of larger nanoparticles, a fast semi-empirical density functional tight-binding (DFTB) method was parameterized. With this method, the potential energy profiles of benzene hydrogenation reactions on different sizes of MCNPs are calculated. The study was then extended to consider a MCNP embedded in solvent (benzene), using a quantum mechanical (DFTB) / molecular mechanical approach. Calculations on the free energies profiles with the umbrella sampling method show that the entropy of the MCNPs and the solvent are essential in understanding the catalytic activity of the transition metal related nanoparticles for solid/liquid heterogeneous catalysis.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗