Neutron Moderation Research Articles

A Possibility for Large-Scale Production of 238Pu in Light-Water Reactor VVER-1000

This paper considers the possibility for large-scale production of plutonium isotope 238Pu in the light-water nuclear power reactor VVER-1000. 238Pu is a unique source of long-term autonomous energy supply in various devices for remote regions of the Earth and in outer space. The design of the irradiation device with 237NpO2 as a starting material is proposed, which is placed in the central zone of the VVER-1000 reactor core and makes it possible to achieve 8% of the specific Pu production (Pu/237Np) by optimizing the pitch of NpO2-rod lattice. The computations showed that the scale of 238Pu production in the irradiation device was remarkably larger (2 ÷ 7 times more) than similar values in research reactors. At the same time, the use of heavy neutron moderators with low neutron absorption (natural lead or lead isotope 208Pb) around the NpO2 fuel assembly (FA) made it possible to obtain high-purity 238Pu with the content of 236Pu below 2 ppm. The paper also shows that if the irradiation device is placed in central zone of the VVER-1000 reactor core, then the displacement damage dose in the reactor vessel remains low enough to conserve its strength properties throughout the entire period of the reactor operation (60 years).

Improvement and benchmarking of the deterministic transport calculation code CBZ to design beam shaping assembly for BNCT

The quick and accurate neutron and photon transport calculations are desired in the optimization calculations for the neutron source design, and in the present work, we establish the deterministic neutron and photon transport calculation procedure with the nuclear reactor physics calculation code system CBZ. Numerical calculation conditions are carefully chosen, and the efficient and practical condition is determined. Test calculations are carried out for the simple cylindrical systems with various kinds of neutron moderators, and the results are compared with the reference solutions obtained by the continuous-energy Monte Carlo code PHITS. Generally good agreements are obtained for all the benchmark problems. In addition, another problem with the detailed geometry for the neutron source is prepared. In this realistic problem also, good agreement is obtained between CBZ and PHITS. These results demonstrate the high accuracy of CBZ in the application to the design optimization calculations for the neutron source.

Introduction of Graphene/h-BN Metamaterial as Neutron Radiation Shielding by Implementing Monte Carlo Simulation.

In this paper, graphene/h-BN metamaterial was investigated as a new neutron radiation shielding (NRS) material by Monte Carlo N-Particle X version (MCNPX) Transport Code. The graphene/h-BN metamaterial are capable of both thermal and fast neutron moderator and neutron absorber process. The constituent phases in graphene/h-BN metamaterial are chosen to be hexagonal boron nitride (h-BN) and graphene. The introduced target was irradiated by an Am–Be neutron source with an energy spectrum of 100 keV to 15 MeV in a Monte Carlo simulation input file. The resulting current transmission rate (CTR) was investigated by the MCNPX code. Due to concrete’s widespread use as a radiation shielding material, the results of this design were also compared with concrete targets. The results show a significant increase in NRS compared to concrete. Therefore, metamaterial with constituent phase’s graphene/h-BN can be a suitable alternative to concrete for NRS.

Integrated analysis to investigate the viability of using Thorium-based fuel instead of traditional fuel in CANDU reactor

This work gives the nuclear operator an integrated insight into the possibility of operating a CANDU reactor with thorium-based fuel types. MCNPX code was used to design 3D model of CANDU bundle. Thorium mixed with 235U, 233U and rgPu were investigated as a fuels in separate fuel cycles and their results were compared with the traditional fuel (UO2). The main neutronic and safety parameters such as infinity multiplication factor (kinf), neutron generation time (Λ), breeding ratio, Doppler reactivity coefficient (DRC), effective delayed neutron fraction (βeff) and moderator temperature coefficient (MTC) have been studied for the investigated fuels. Due to the significant impact of the rgPu and minor actinides (MAs) on the environment, their concentrations have been investigated with burn-up for the proposed fuels to reach the optimum one from the neutronic and economic point of view.

Feasibility of Low-Enriched Uranium Fueled Nuclear Thermal Propulsion in the Low-Thrust Region Below 16klbf

This paper establishes the feasibility of Low-Enriched Uranium fueled Nuclear Thermal Propulsion (LEU-NTP)reactors in the low-thrust region below 16klbf (71kN). A reference, 7.5klbf, High-Enriched Uranium (HEU) design is converted to 19.75% enriched LEU and shown to be capable of reaching criticality and meeting other performance requirements with a minimalmass increase. At this smaller scale, historical LEU-NTP conversion techniques that focus on maximizing neutron moderation or minimizing leakage within the active core are insufficient. Thus, the 7.5klbf preliminary design requires several unique modifications, including a reduction in the number of control drums and adjustments to selected materials. To verify the design’s feasibility, several keyneutronic and thermal-hydraulic performance parameters including burnup, Xenon worth, and submersion criticality are characterized with Serpent 2.0 and the Space Propulsion Optimization Code. The methods applied in this workreveal an opportunity for further LEU-NTP optimizationthat may directly translate to increased scalability and efficiency for future designs.

Direct measurement of the scattering cross sections of liquid ortho-deuterium for ultracold neutrons and comparison with model calculations

Liquid deuterium is a fluid between the quantum and classical regimes. It attracts interest from fundamental research for the verification of quantum calculations but also from neutron physics as it is a widely used neutron moderator medium. We have measured the scattering cross sections of liquid ortho-deuterium in the ultracold-neutron range for four different temperatures (19-23 K) and compared them with calculations from a parameter-free analytical calculation model as well as with previous measurements at 19 K by another research group. All three show remarkable agreement, which establishes the validity of the calculation model and proves it is a reliable basis for the derivation of scattering kernels. The deconvolution of our measured transmission data changed the cross section results noticeably only for neutrons faster than 10 m/s. We found the total scattering cross section of liquid deuterium to be inversely proportional to velocity, as is predicted by theory.

Effect of changing the neutron moderator in a thermal subcritical nuclear reactor

Subcritical thermal nuclear reactor is the combination of nuclear fuel, moderator and external neutron source. Nevertheless, the amount of neutrons increases through nuclear fissions in the 235U in the fuel, the absorption and leakage of neutrons maintain the subcriticality condition. There are several subcritical reactors with different features and purposes. The open tank subcritical reactor Nuclear Chicago model 9000 is the heterogeneous combination of natural uranium, in hexagonal lattice, and light water, acting as moderator and reflector of neutrons, that uses a239PuBe neutron source. Worldwide there are several of these reactors mainly used in education, where due to the contact with water the Al tubes have corrosion. A possible solution is to use polyethylene as moderator. Aiming to evaluate the effect of changing the moderator, in this work Monte Carlo methods were used to model the reactor with polyethylene and water as moderators increasing the fuel load. For each model were estimated the keff and the μ parameters, and for Case 10a model was calculated the power, the Ambient dose equivalent and the fluence, of neutrons and γ-rays at five sites outside the reactor.

Stimulated perturbation on the neutron flux distribution in the mutually-dependent source-to-absorber geometry

The complexity of the neutron transport phenomenon throws its shadows on every physical system wherever neutron is produced or absorbed. The Monte Carlo N-Particle Transport Code (MCNP) was used to investigate the flux perturbations in the neutron field caused by an absorber. The geometry of the present experiment was designed to reach a simulation of an isotopic neutron field. The neutron source was a 241 AmBe with the production physics of neutrons being dependent only on alpha-beryllium interaction and is independent of what happened to the neutron after it was generated. The geometries have been designed to get a volume of uniform neutron densities within a spherical volume of radius 15 cm in every neutron energy group up to 10 MeV. Absorbers of different dimensions were placed within the volume to investigate the field perturbation. Different neutron absorbers were used to correlate the phenomenon to the integral cross-section of the absorber. Flux density inside and outside the absorber samples was determined, while the spatial neutron flux distribution produced by the AmBe source without an absorber was taken as a reference. This study displayed that absorbers of various dimensions perturb the neutron field in a way that is dependent on the absorption and scattering cross-sections, particularly in the neutron resonance region. Unlike the simple picture of reducing the number density of neutrons, the perturbation was found to influence the moderation of neutrons in the medium, significantly above 1 MeV.

Nested mirror optics for neutron extraction, transport, and focusing

Neutron scattering is a well-established tool for the investigation of the static and dynamic properties of condensed matter systems over a wide range of spatial and temporal scales. Many studies of high interest, however, can only be performed on small samples and typically require elaborate environments for variation of parameters such as temperature, magnetic field and pressure. To improve the achievable signal-to-background ratio, focusing devices based on elliptic or parabolic neutron guides or Montel mirrors have been implemented. Here we report an experimental demonstration of a nested mirror optics (NMO), which overcomes some of the disadvantages of such devices. While even simpler than the original Wolter design, our compact assembly of elliptic mirrors images neutrons from a source to a target, reducing geometric aberrations, gravitational effects and waviness-induced blurring. Experiments performed at MIRA at FRM-II demonstrate the expected focusing properties and a beam transport efficiency of 72% for our first prototype. NMO seem particularly well-suited to (i) extraction of neutrons from compact high-brilliance neutron moderators, (ii) general neutron transport, and (iii) focusing and polarizing neutrons. The phase space of the neutrons hitting a sample can be tailored on-line to the needed experimental resolution, resulting in small scattering backgrounds. As additional benefits, NMO situated far away from both the moderator and the sample are less susceptible to radiation damage and can easily be replaced. NMO enable a modular and physically transparent realization of beam lines for neutron physics similar to setups used in visible light optics.

Temperature Effect on Short—Term Strength of Lithium Hydride with Tensile and Three—Point Bend Specimens

Strength is one important mechanical property of lithium hydride (LiH) for application as a neutron moderator and absorbent at elevated temperatures. Using the digital image correlation (DIC) method for deformation measurement, the short—term strength of LiH was investigated with both tensile and three—point bend (3PB) specimens in the range of room temperature (RT) to 600 °C. The results show that the temperature dependence of the strength from tensile and 3PB tests shows similarity. As the temperature increases, the strength changes slightly, then rises to the maximum and finally decreases due to the softening effect. The fracture surfaces of specimens show the main characteristics of trans—granular fracture at different temperatures, except for 600 °C. With increasing temperature, the fracture surfaces become rough and numerous tortuous micro—cracks are found. Especially for 3PB specimens, there are dimple—like structures with distorted grains and obvious intergranular cleavage fracture.

First principles calculation for generating new thermal neutron scattering data for Zirconium hydride

Lattice parameters of materials have the same magnitude as the energy of thermal neutrons in reactors, which directly affects the neutron cross section and its energy. While they are thermalized, incident neutrons can lose or gain energy during their interactions with materials components. Since several decades, methods and models were developed in the aim to generate nuclear data sub-libraries required in correcting neutrons interactions cross sections at thermal energies. However, very few experimental works were dedicated to this field. In this paper we focus our efforts on reviewing the theoretical models and their adequacy in describing thermal scattering events in the aim of proposing new formalisms to calculate the density of states (DOS) and phonon responses of zirconium hydride material, which constitutes an important moderator of neutrons in TRIGA reactors fuel elements. Generally the effects of thermal scattering are provided in nuclear data evaluations by a thermal sub-library ENDF file 7. Data in file 7 are described by the known thermal scattering law S(α,β) which is a function of momentum transfer and energy transfer parameters α and β respectively. The thermal scattering law has been used to calculate the double differential cross sections and the corresponding results are presented. Although the comparison with other models shows satisfactory results, no previously personalized use of data may be the raison of its usefulness in some cases and not in others.

Designs and neutronic characteristics of an epithermal neutron moderator at ambient temperature for neutron time-of-flight measurements

ABSTRACT Compact time-of-flight (TOF) measurement systems are desirable for easy installation in various facilities. To achieve such compact systems, one of the key considerations is the design of the neutron moderator. Therefore, systematic studies on moderators considering both neutron energy resolution and neutron intensity are important. In this paper, to design an epithermal neutron moderator at ambient temperature for short-distance TOF measurements, the neutron intensity and energy resolution of epithermal neutrons have been studied using the Monte Carlo simulation code PHITS with JENDL-4.0 for various types of moderators. The neutronic characteristics of a moderator comprising several components were evaluated. The relationships between the moderator dimensions and both the energy resolution and intensity of the epithermal neutrons were determined. From the obtained results, the appropriate high-resolution moderator designs for the short-distance TOF measurements were proposed.

Ultracold neutron storage and transport at the PSI UCN source

Efficient neutron transport is a key ingredient to the performance of ultracold neutron (UCN) sources, important to meeting the challenges placed by high precision fundamental physics experiments. At the Paul Scherrer Institute’s UCN source we have been continuously improving our understanding of the UCN source parameters by performing a series of studies to characterize neutron production and moderation, and UCN production, extraction, and transport efficiency to the beamport. The present study on the absolute UCN transport efficiency completes our previous publications. We report on complementary measurements, namely one on the height-dependent UCN density and a second on the transmission of a calibrated quantity of UCN over a sim 16 m long UCN guide section connecting one beamport via the source storage vessel to another beamport. These allow us quantifying and optimizing the performance of the guide system based on extensive Monte Carlo simulations.

CdO-rich quaternary tellurite glasses for nuclear safety purposes: Synthesis and experimental gamma-ray and neutron radiation assessment of high-density and transparent samples

We present the preparation phase and comprehensive analysis of nuclear radiation shielding characteristics of novel melt-quenched 20P2O5·30TeO2.(50-x)ZnO.xCdO (x = 0, 15, 20, 30, and 40 mol percent) quaternary-tellurite glasses. The primary objective is to compare the changes in nuclear radiation absorption qualities that come from maintaining high transparency and increasing the CdO contribution rate to maximum values, such as 40 mol%. Consequently, experimental gamma-ray and neutron transmission systems are used to investigate the monotonic impacts of increasing CdO reinforcement on the functioning of synthesized glasses utilizing the well-known melt-quenching method. For the determination of attenuation coefficients, a standard gamma-ray setup is used with an Ultra germanium detector and 133Ba radioisotope. In addition, using the Canberra NP-100B BF3 gas proportional detector, glass shields are bombarded with a 241Am/Be neutron source (10 mCi/4.5 MeV) using a gas proportional detector. The addition of 40% mole CdO to the basic composition of glass significantly improved the transition resistance to gamma and neutron radiation. Furthermore, it was shown that the degree of transparency in the C40 sample synthesized with a 40% CdO additive ratio was equivalent to that of an ideal transparent glass sample. Moreover, C40 sample had better gamma-ray attenuation properties than all other shielding materials (except for RS-520). According to the findings, C glasses have a greater capacity for neutron attenuation than investigated conventional moderators. It can be concluded that C glass family is an effective gamma-shield and neutron moderator for research and medical radiation applications.

Development of personal neutron dosimeter with energy resolution ability

The prototype neutron personal dosimeter consists of three passive implanted planar silicon detectors. Polyethylene, B4C-containing polyethylene, and 6LiF act as neutron moderators, absorbers, and converters in front of each detector. This personal dosimeter has two modes of operation: real-time mode and unfolding mode. In real-time mode, it shows three neutron personal dose equivalent values, which represent the personal neutron equivalent of high energy (1–15 MeV), intermediate energy (1 × 10−3 MeV –1MeV), and low-energy (2.53 × 10−8 MeV −1 × 10−3 MeV), respectively. Unfolding method based on Genetic Algorithm was used to deduce neutron spectra from pulse height spectra of recoil proton. The dosimeter was calibrated in monoenergetic neutron fields and was tested in an 241Am–Be neutron field.

Heavy-water-based moderator design for an AB-BNCT unit using a topology optimization algorithm

Objective. The design of neutron moderators for BNCT treatment units currently relies on parametric approaches, which yield quality results but are ultimately limited by human imagination. Efficient but non-intuitive design solutions may thus be missed out. This limitation needs to be addressed. Approach. To overcome this limitation, we propose to use a topology optimization algorithm coupled with a state-of-the-art Monte-Carlo transport code. This approach recently proved capable of finding complex optimal configurations of particle propagators with limited human intervention. Main results. In this study, we apply this algorithmic solution to optimize some heavy-water neutron moderators for a specific AB-BNCT treatment unit. The moderators thus generated are compact yet succeed in limiting the exposure of patient’s healthy tissues to levels below recommended limits. They present subtle, original geometries inaccessible to standard parametric approaches or human intuition. Significance. This approach could be used to automatically fit the design of a BNCT moderator to the location and shape of the tumor or to the morphology of the patient to be treated, opening a path for more targeted BNCT treatment.

Synthesis and experimental characterization on fast neutron and gamma-ray attenuation properties of high-dense and transparent Cadmium oxide (CdO) glasses for shielding purposes

We present synthesis and thorough characterization phases of newly developed 20P2O5·30TeO2.(50-x)ZnO.xCdO (x = 0, 2, 4, 6, 8, 10 mol%) glasses. Experimental gamma-ray and neutron transmission systems are used in terms of exploring the monotonic effects of increasing CdO reinforcement on behavioural changes of synthesized glasses through well-known melt-quenching method. A conventional gamma-ray setup along with a high purity germanium detector as well as 133Ba radioisotope is used for determination of attenuation coefficients. Glass shields are irradiated with a241Am/Be neutron source using the Canberra NP-100B BF3 gas proportional detector. The addition of 10% mole CdO to the basic glass composition had a significant favorable impact on the transition resistance to gamma and neutron radiation. In comparison to all available shielding materials, (except RS-520), the C10 sample demonstrated superior gamma-ray attenuation capabilities. According to results, C glasses generally exhibit superior neutron attenuation capabilities than conventional moderators. It can be concluded that C glass family may be considered as suitable gamma-shield and neutron moderator in different types of applications from research to medical radiation fields.

The impacts of charge transfer, localization, and metallicity on hydrogen retention and transport capacity

Solid state hydrides such as early transition metal hydrides are of inestimable importance for the future of hydrogen energy and are actively being investigated for energy conversion and storage applications such as fuel cells, solid-state batteries and neutron moderators. The retention and transport behavior of hydrogen in these hydrides has a huge role on the extended performance of components. While early transition-metal-based compounds exhibit many peculiar properties due to their unique correlated electronic signatures arising from d-orbital electrons, the fundamental chemistry and transport behavior of hydrogen in such hydrides is not well understood. In the present work, using density functional theory, a highly intricate bonding feature is revealed through the theoretical investigation of the electronic structure of early transition metal hydrides YH2 and ZrH2. In particular, a pronounced charge transfer from the transition element to H, results in localized electron densities at deep energy levels. The interplay between intrinsic charge transfer, charge localization, and metallicity in YH2 and ZrH2 leads to strong chemical bonding between metal and hydrogen atoms and large energy barriers for the migration of hydrogen vacancies. Specifically, hydrogen vacancies are found to be stable in the neutral state due to electron screening effects, accompanied by substantially high migration barriers between 0.8–1.2 eV along different crystallographic directions. In contrast, recent literature shows the migration barrier for charged H vacancies in insulating s-block metal hydrides lie between 0.1–0.4 eV, which is suitable for fast conduction applications. This pivotal electron structure difference exploited between early transition metal hydrides and alkali/alkaline earth metal hydrides determines extended hydrogen retention in these early transition metal hydrides. This work explains fundamental differences between the electronic structure of s-block and d-block metal hydrides, and its impact on the mobility of hydrogen vacancies.

The structure evolution in neutron-Irradiated nuclear graphite and post-annealing

The neutron irradiated microstructure and performance of nuclear graphite as a neutron moderator and structural materials are directly related to the safety and longevity of reactors. Nuclear-grade isotropic graphite IG-110U and ETP-10 were irradiated at 200 °C with low-dose (1.38 × 1023 n/m2, 0.02dpa) and high-dose (1.92 × 1024 n/m2, 0.25dpa) neutron and high-temperature annealed at high temperatures up to 1400 °C, with the whole process analyzed using X-ray diffractometer and CARBONX program. After neutron irradiation of both doses, the c-axis lattice constant <c> became significantly larger, while the a-axis lattice constant <a> only changed a little. Although the crystallite size along c-axis Lc and the number of internal atom layers M both increased significantly, the crystallite size along a-axis La changed slightly for the low-dose samples. However, all three parameters decreased significantly after high-dose irradiation. During the subsequent high-temperature annealing, there were "recovery" points where both the microscopic parameters and macroscopic dimensions were restored to the initial unirradiated value and kept mostly constant. The movement and recombination of interstitial atoms, vacancies and their clusters played a major role in this annealing process. However, the crystallite sizes of two graphites irradiated with high-dose neutrons did not reached their initial values even with annealing at 1400 °C, mainly due to the unrecoverable defects, such as interstitial loops, pores, and microcracks.

Improved beam extraction at compact neutron sources using diamonds nanoparticles and supermirrors

Aimed to investigate potential benefits from mounting reflecting layers of diamond nanoparticles or supermirrors in neutron extraction channels, a conceptual compact neutron source was designed using Geant4 and MCNP. The scattering of low-energy neutrons with nanodiamonds was implemented in Geant4 and validated with existing experimental data. The source design was based on a 13 MeV proton beam hitting a beryllium target, with neutron moderation facilitated by a water premoderator, supplemented by a quasi 1-dimensional parahydrogen moderator, before emission towards a few (2–5) beamlines. In an attempt to increase the neutron flux available for experiments, reflecting layers of diamond nanoparticles or supermirrors were placed in the neutron extraction channel, and results for different implementations were assessed and compared.The results show that the cold neutron flux is increased by using the nanodiamond powder, in particular for large divergence neutrons (greater than 2°), but a significant effect is observed also for low-divergence neutrons, which are of particular interest to neutron scattering experiments. The effect is dependent on the neutron wavelength and the size of nanodiamonds, with larger effects in general obtained for longer wavelengths, and particles of larger size. The same configuration was studied, where instead of nanodiamonds, supermirrors were used. The use of supermirrors at the beam extraction predicts even higher potential gains. These results are very promising to increase the efficiency of compact sources and can be applied to various configurations of such sources.