Neutron Reflector Research Articles

NJOY+NCrystal: An open-source tool for creating thermal neutron scattering libraries with mixed elastic support

In this work we present NJOY+NCrystal, a tool to generate thermal neutron scattering libraries with support for coherent and incoherent elastic components for crystalline solid materials. This tool, which is a customized version of NJOY, was created by modifying the nuclear data processing program NJOY to call the thermal scattering software library NCrystal, and includes a proposed change in the ENDF-6 format to store both the coherent and incoherent elastic components. Necessary changes to enable this format in NJOY, as well as to sample it in the OpenMC Monte Carlo code, are detailed here. Examples of materials that are coherent–dominant, incoherent–dominant, and mixed elastic scatterers are presented, as well as the creation of novel libraries for MgH2 and MgD2, that are under consideration as advanced neutron reflectors in the HighNESS project at the European Spallation Source. NJOY+NCrystal greatly simplifies the process to generate thermal scattering libraries (TSL) and this is exemplified with 213 new and updated TSL evaluations.

Effect of Particle Sizes on the Efficiency of Fluorinated Nanodiamond Neutron Reflectors.

Over a decade ago, it was confirmed that detonation nanodiamond (DND) powders reflect very cold neutrons (VCNs) diffusively at any incidence angle and that they reflect cold neutrons quasi-specularly at small incidence angles. In the present publication, we report the results of a study on the effect of particle sizes on the overall efficiency of neutron reflectors made of DNDs. To perform this study, we separated, by centrifugation, the fraction of finer DND nanoparticles (which are referred to as S-DNDs here) from a broad initial size distribution and experimentally and theoretically compared the performance of such a neutron reflector with that from deagglomerated fluorinated DNDs (DF-DNDs). Typical commercially available DNDs with the size of ~4.3 nm are close to the optimum for VCNs with a typical velocity of ~50 m/s, while smaller and larger DNDs are more efficient for faster and slower VCN velocities, respectively. Simulations show that, for a realistic reflector geometry, the replacement of DF-DNDs (a reflector with the best achieved performance) by S-DNDs (with smaller size DNDs) increases the neutron albedo in the velocity range above ~60 m/s. This increase in the albedo results in an increase in the density of faster VCNs in such a reflector cavity of up to ~25% as well as an increase in the upper boundary of the velocities of efficient VCN reflection.

CFD Evaluation of Pressure Change Along Coolant Passages in Sodium-Cooled Fast Reactor with Nek5000

To support the design efforts of advanced sodium-cooled fast reactors (SFRs), a series of computational fluid dynamics (CFD) simulations are performed to investigate the pressure change along various flow passages in the proposed SFR system. The simulations are carried out with the state-of-the-art spectral element flow solver, Nek5000. Two specific case studies are presented in this paper: the flow exiting the axial neutron reflector channels and the flow entering the fuel pin bundle. Due to the high Reynolds numbers expected, a Reynolds-averaged Navier-Stokes (RANS) approach is necessary to model the turbulence. A newly developed regularized RANS model is adopted in the related CFD calculations. The first case study explores the effect of Reynolds number on the pressure change when flow exits the reflector channels. The pressure change in this case has two major contributors: the change due to wall friction and the Bernoulli effect. It is noted that the nondimensional pressure loss follows a log-linear trend up to Re = 105, and then the trend is flattened. In the second case study, the advanced NekNek coupling capability is tested where an integral domain can be divided into multiple subdomains with coupling interfaces, which would greatly ease the meshing process of complex engineering geometries and potentially save computational resources. The preliminary results obtained so far confirm the consistency between the NekNek results and those produced by regular Nek5000 simulation. The presented work demonstrates the readiness and flexibility of the related CFD techniques, which is part of the broader effort to leverage cutting-edge CFD to inform the advanced nuclear reactor designs.

Development of neutron scattering kernels for cold neutron reflector materials

The newest neutron scattering applications are highly intensity-limited techniques that demand reducing the neutron losses between source and detectors. In addition, the nuclear industry demands more accurate data and procedures for the design and optimization of advanced fission reactors, especially for the treatment of fuel and moderator materials. To meet these demands, it is necessary to improve the existing calculation tools, through the generation of better models that describe the interaction of neutrons with the systems of interest. The Neutron Physics Department at Centro Atomico Bariloche (CNEA, Argentina) has been developing over the time new models for the interaction of slow neutrons with materials, to produce scattering kernels and cross section data in the thermal and cold neutron energy region. Besides the studies carried out on neutron moderators, we have recently begun looking at materials that could serve as efficient neutron reflectors over those energy ranges. In this work we present the results of transmission and scattering experiments on diamond nanopowder and magnesium hydride, carried out simultaneously at the VESUVIO spectrometer (ISIS, UK), and compare them with newly generated cross-section libraries.

ПРОГРАММА СПЕКТРОМЕТРИИ НЕЙТРОННЫХ ПОЛЕЙ В РЕАКТОРЕ СМ ПОСЛЕ МОДЕРНИЗАЦИИ

A program for refining the energy spectra of neutron radiation in the neutron trap, core and reflector of the SM-3 reactor after modernization using the certified equipment and verified methods is presented in the article. The relevance of the article lies in familiarization with the experimental implementation of the methods in order to compare the experimental neutron-physical characteristics of the reactor before and after the last reconstruction. During the program creation, the previous experience of spectrometric measurements was taken into account, as well as experiments were carried out under conditions close to those of spectrometry after reconstruction. The results of the research are necessary for the planning of irradiation experiments and verification of calculation methods for determining the neutron-physical characteristics. The results of the described researches cover most science and technology areas, many aspects of human practice, including healthcare. The authors presented results of accompanying experiments, which made it possible to work out the regulations and features of short-term measurements. The developed and optimized program of neutron spectrometry in the SM-3 reactor channels will allow to obtain reliable information on the density of neutron fluxes for the first time after modernization to verify analytical data and to optimize the use of the SM-3 reactor facility to increase the production of transplutonium elements and radionuclide products.

Neutronic analysis of effects of inboard materials on the size of a tokamak fusion reactor

• Inboard materials play a key role in determining the size of the reactor. • The reactor parameters were determined via coupled tokamak systems analysis. • Optimal combination of multiplier and reflector can lead to small reactor. The effects of inboard materials on the size of a tokamak fusion reactor were studied based on reactor system parameters determined through a coupled analysis of tokamak systems and neutron transport. With an inboard blanket, the shielding capability was the best for tungsten carbide (WC), showing a smaller reactor size than cases using other shield materials. Shield materials such as WC, borated steel (BS), and ferritic/martensite steel (FMS) showed improved neutron economy for tritium breeding in the blanket, which was attributed to neutrons reflected by tungsten (W) and steel (Fe). In a configuration with only an outboard breeding blanket, it was found that the tritium self-sufficiency requirement could be satisfied with an inboard blanket replaced by a neutron multiplier and/or reflector materials. Titanium carbide (TiC) showed the best neutron reflecting capability. With lead (Pb) as a neutron multiplier, neutron spectrum softening by a Pb multiplier enhanced the reflector performance and increased tritium breeding at the outboard blanket. Thus, an optimum combination of multiplier and reflector materials can lead to smaller reactor sizes than reactors with an inboard blanket.

Clustering of Diamond Nanoparticles, Fluorination and Efficiency of Slow Neutron Reflectors.

Neutrons can be an instrument or an object in many fields of research. Major efforts all over the world are devoted to improving the intensity of neutron sources and the efficiency of neutron delivery for experimental installations. In this context, neutron reflectors play a key role because they allow significant improvement of both economy and efficiency. For slow neutrons, Detonation NanoDiamond (DND) powders provide exceptionally good reflecting performance due to the combination of enhanced coherent scattering and low neutron absorption. The enhancement is at maximum when the nanoparticle diameter is close to the neutron wavelength. Therefore, the mean nanoparticle diameter and the diameter distribution are important. In addition, DNDs show clustering, which increases their effective diameters. Here, we report on how breaking agglomerates affects clustering of DNDs and the overall reflector performance. We characterize DNDs using small-angle neutron scattering, X-ray diffraction, scanning and transmission electron microscopy, neutron activation analysis, dynamical light scattering, infra-red light spectroscopy, and others. Based on the results of these tests, we discuss the calculated size distribution of DNDs, the absolute cross-section of neutron scattering, the neutron albedo, and the neutron intensity gain for neutron traps with DND walls.

Simulation-based optimization of a thermal neutron detector with SERS method

A Monte Carlo simulation-based optimization of a SERS-based thermal neutron detector is aimed at improving detection efficiency for thermal neutrons. The moderator and collimator have been designed for a 241Am-9Be source to maximize the thermal neutrons flux in the irradiation area. Simulated thermal neutron detection efficiency can be increased to approximately 50.5% when polyethylene is added as neutron reflector and 10B enriched 4-MPBA is used. With Monte Carlo software GEANT4, the SERS-based thermal neutron detector has been proven to have almost no effect on original thermal neutron flux as well as the distribution and could be used for thermal neutron detection accurately.

Densification of A3-3 matrix graphite to inhibit the infiltration of liquid FLiBe salt in molten salt reactors

In this study, a novel graphite was prepared by adding MCMBs (mesocarbon microbeads) to densify matrix graphite A3−3, a typical fuel element and moderator material in high-temperature gas-cooled reactors. The densified graphite was evaluated as a potential fuel element material for molten salt reactors (MSRs). In particular, the effects of MCMB loading on physical properties and the corresponding effects on the infiltration of liquid fluoride salt in MSRs were studied, as molten salt infiltration can impact safe reactor operation. MCMBs with a mean particle size of approximately 3 μm were added to matrix graphite A3−3 at loading levels of 1 wt%, 5 wt%, 10 wt%, and 15 wt%, and samples were prepared by a quasi-isostatic pressing process at 250 MPa. Mercury porosimetry data showed that MCMB-densified graphite (MDGs) had lower porosity (approximately 15 vol%) and median pore diameters (ranging from 105 to 306 nm) compared to A3−3 itself (22.9 vol% porosity and 573 nm median pore diameter). Importantly, the MDGs exhibited greatly reduced infiltration by FLiBe molten salt compared to unmodified A3−3. For instance, A3−3 modified with 1 wt% MCMB exhibited a 2.1 % weight increase after treatment at 1 MPa for 20 h, while unmodified A3−3 exhibited a weight increase of 11.6 %. This study revealed that MDGs could be promising candidate materials for the graphite matrix of fuel elements as well as a neutron moderator and reflector in MSRs.

Feasibility study on tritium recoil barrier for neutron reflectors of research and test reactors

Tritium release into the primary coolant of the JMTR and the JRR-3 M had been studied, and it is found that tritium recoil release from the chain reaction of beryllium neutron reflectors is dominant. To prevent the tritium recoil release, Al, Ti, V, Ni and Zr are selected as the candidate tritium recoil barrier materials in this feasibility study. It is clear that 20∼40 μm thickness is required depending on the material to reduce by 3 orders, and that an impact on the effective multiplication factor is about 0.2 % at most. Total evaluation including the activities, fabrication and usage experiences, suggests the selection of Al as the first candidate may have the least development risk as the tritium recoil barrier.

REACTOR PERFORMANCE IMPROVEMENT OPTIONS TO SUSTAIN HIGH FLUX ISOTOPE REACTOR LEADERSHIP INTO THE FUTURE*

The mission of the Neutron Sciences Directorate (NScD) at the U.S. Department of Energy’s Oak Ridge National Laboratory (ORNL) is the undertaking of high-impact research into the structure and properties of materials across the spectrum of biology, chemistry, physics, materials science, and engineering. NScD operates two world-leading neutron scattering facilities: the High Flux Isotope Reactor (HFIR) and the Spallation Neutron Source. HFIR achieved full power in 1966, and over a half century later, it continues to serve a variety of national missions. HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world to support scientific missions including cold and thermal neutron scattering, isotope production, and materials irradiation research. To sustain leadership in neutron sciences into the future, ORNL is exploring areas in which HFIR can be improved to enhance its performance. Many improvement areas are being explored including upgrading the cold source and neutron scattering facilities. The improvement areas discussed herein include replacing the reactor pressure vessel, upgrading the neutron reflector, and ensuring that reactor performance is maintained or enhanced after converting from high-enriched uranium to low-enriched uranium fuel.

SENSITIVITY ANALYSIS OF STAINLESS STEEL REFLECTOR FOR VR-1 TRAINING REACTOR

Pressurized water reactors are typically surrounded in the radial direction by neutron reflectors made from stainless steel and water. These reflectors decrease neutron leakage and provide protection of pressure vessel from fast neutrons damaging its integrity. Such a radial reflector influences multiplication factor of the core and distribution of neutron flux and fission power inside the core. All these effects can be analyzed by full-core simulations using macroscopic constants. Methodology for generation of the macroscopic constants for non-fuel regions will be tested for new stainless steel reflectors at the VR-1 reactor. Rods from SS 304l material will be used for construction of radial reflectors for the VR-1 reactor. They will be design to generate sufficient measurable response in selected core characteristics. The study is focused on core power distribution and reactivity worth of absorbing rods in a VR-1 reactor core. The core typically consists of about 20 IRT-4M fuel assemblies and seven absorbing rods UR-70. Replacing water surrounding the core by several reflector assemblies containing stainless steel will influence leakage and distribution of neutrons inside the core. The current analysis deals with local effects and employs the sensitivity study to discover the nature of reflectors’ impact on the reactor core. These effects were studied even for several past VR-1 reactor core configurations. All calculations were carried out in Serpent2 Monte-Carlo code with various evaluated libraries: ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 data.

ПРЕИМУЩЕСТВА БЫСТРОГО РЕАКТОРА С ФИЗИЧЕСКИ ТОЛСТЫМ ОТРАЖАТЕЛЕМ ИЗ СВИНЦА-208

The paper analyzes the effects produced by the use of lead isotope 208Pb instead of natural lead as a coolant and neutron reflector on neutron-physical parameters of fast reactor. A possibility is demonstrated to reach substantial (up to three orders of magnitude) elongation of mean prompt neutron lifetime in the fast reactor with physically thick 208Pb-reflector. So large elongation of prompt neutron lifetime opens an opportunity for remarkable enhancement of the reactor safety under conditions of reactivity-induced accidents. Numerical analyses have shown that it became possible to form high flux of slowed down and epithermal neutrons in 208Pb-reflector even at large distances from the reactor core. Spacious area of high neutron flux out of the reactor core may be used to transmute long-lived fission products and to produce significant quantities of necessary isotopes. Weak neutron absorption in 208Pb makes it possible to reach remarkable (up to 20 %) reduction of plutonium fraction in mixed uranium-plutonium fuel composition. It is proposed to place a layer with neutron absorber in 208Pb-reflector above the reactor core for the reactivity control with no variations in neutron-physical properties of the reactor core. If reactivity of the control rods in the reactor core is lower than effective fraction of delayed neutrons, then the reactor is obviously subcritical on prompt neutrons of the reactor core. In this case the danger of the reactivity-induced accident on prompt neutrons of the reactor core disappears. If the reactivity-induced accident is caused by joint action of the neutrons produced in the reactor core and in 208Pb-reflector, then the accident propagation will be substantially slowed down because of the longer lifetime of neutrons in 208Pb-reflector.

Monte carlo analysis of various geometrical blankets in a fusion-fission hybrid reactor

• A dense plasma focus device design as a fusion neutron source. • Investigate the neutron diffusion equation in the cylindrical layer, the spherical layer and the frustum cone layer geometries. • Investigate the lowest probability of a neutron returning from the second blanket to the first blanket in the cylindrical layer, the spherical layer and the frustum cone layer geometries. • Investigate the neutron parameters in the One-Blanket Reactor and Two-Blanket Reactorand the ultimate energy of the hybrid reactor by MCNPX code. • To check the safety of staff and reactor facilities, the dose rate of neutrons and photons in the concrete shield is calculated by MCNPX code. Fusion - fission hybrid reactor consists of two main parts. The first part is the Dense Plasma Focus (DPF) where the fusion process is performed and the second part is the blankets which surround the DPF and the fission process is done in the blankets. The most of the fusion neutrons, near the fusion source, are directed upward, so the blanket over the DPF is designed in forms cylindrical, spherical and frustum cone of volumes 183,347 cm 3 , 136,378 cm 3 and 58,223.2 cm 3 respectively at the same effective multiplication factor. Then the thermal neutron diffusion equation, buckling and neutron parameters are studied. After selecting the first blanket, the second blanket is designed in forms cylindrical, spherical and frustum cone of volumes 4.500 × 10 6 cm 3 , 3.689 × 10 6 cm 3 and 6.441 × 10 5 cm 3 respectively. So the frustum cone is chosen for the first and second blanket. The neutron return probability in the frustum cone from the second blanket to the first blanket is low, so the first blanket's effective multiplication factor does not change and the reactor's total effective multiplication factor is stable. Also, MCNPX code has been used to checkout and select neutron multiplier and reflector cells for the most neutron yield.

Favorable Effects of an Inner Lead Cavity on Neutronic Physics of a Lead-Cooled Fast Reactor

The paper considers the effects produced by arranging a lead cavity in the central part of a lead-cooled fast reactor core. It is shown that the most favorable effect from the presence of an inner lead cavity at the center of the core can be obtained when radiogenic lead with a dominant 208Pb isotope content is used as the material of the cavity, coolant, and neutron reflector. This is explained by the extremely low neutron absorption of 208Pb in a wide range of neutron energies. If the thickness of the 208Pb inner cavity is chosen properly, then a spectral region with the prevailing share of resonance, epithermal, and thermal neutrons is created in the cavity. This results in a sharp increase in the mean prompt neutron lifetime (by few orders of magnitude) and enhancement of the stabilizing Doppler effect. The neutron migration from the inner cavity to the reactor core is able to intensify the fission chain reaction (FCR) and contribute positively to reactor reactivity. In addition, the high flux of slowed neutrons in the inner 208Pb-cavity makes it possible to expect efficient transmutation of long-lived fission products.

Advantages of a Fast Reactor Core Surrounded by a Physically Thick Neutron Reflector Made of Lead-208

In the paper, the influence of replacement of natural lead with lead-208 as a coolant and fast reactor reflector on its neutron-physical parameters is considered. The possibility of significantly increasing the mean prompt neutron lifetime by using lead-208 is shown. It is proposed to place a layer containing a neutron absorber over the core in its reflector to control the reactor reactivity. It is shown that when using lead-208, the efficiency of this layer is noticeably higher, and the Doppler effect is enhanced than when using natural lead.

On a significant slowing-down of the kinetics of fast transient processes in a fast reactor

The kinetics of nuclear reactors is determined by the average neutron lifetime. When the inserted reactivity is more than the effective delayed neutron fraction, the reactor kinetics becomes very rapid. It is possible to slow down the fast reactor kinetics by increasing the neutron lifetime. The authors consider the possibility of using the lead isotope, 208Pb, as a neutron reflector with specific properties in a lead-cooled fast reactor. To analyze the emerging effects in a reactor of this type, a point kinetics model was selected, which takes into account neutrons returning from the 208Pb reflector to the reactor core. Such specific properties of 208Pb as the high atomic weight and weak neutron absorption allow neutrons from the reactor core to penetrate deeply into the 208Pb reflector, slow down in it, and have a noticeable probability to return to the reactor core and affect the chain fission reaction. The neutrons coming back from the 208Pb reflector have a long ‘dead-time’, i.e., the sum of times when neutrons leave the reactor core, entering the 208Pb reflector, and then diffuse back into the reactor core. During the ‘dead-time’, these neutrons cannot affect the chain fission reaction. In terms of the delay time, the neutrons returning from the deep layers of the 208Pb reflector are close to the delayed neutrons. Moreover, the number of the neutrons coming back from the 208Pb reflector considerably exceeds the number of the delayed neutrons. As a result, the neutron lifetime formed by the prompt neutron lifetime and the ‘dead-time’ of the neutrons from the 208Pb reflector can be substantially increased. This will lead to a longer reactor acceleration period, which will mitigate the effects of prompt supercriticality. Thus, the use of 208Pb as a neutron reflector can significantly improve the fast reactor nuclear safety.

The influence of neutron reflector with special properties on fast reactor kinetics

The paper substantiates the possibility of increasing the safety of fast reactors by surrounding the reactor core with a neutron reflector with special properties. The particular case of using lead isotope as a neutron reflector has been considered in the lead-cooled fast reactor in an exemplary manner. An asymptotic point kinetics model, which takes into account neutrons returning from the reflector to the reactor core and participating in the development of fission chain reaction, was proposed for assessment of the emerging effects. The in-hour equation for fast reactors, surrounded by a neutron reflector with special properties, has been developed. It was revealed that the neutron lifetime can be substantially increased. Therefore, the contribution of the neutron reflector would mitigate the consequences of accidents when inserted reactivity exceeds the effective fraction of delayed neutron. In addition, the maximum rates of power increase have been calculated for different cases of reactivity insertion. The obtained results have considerable practical value for scientific and advanced studies of reactor accidents.

PECULIARITIES OF MACHINERY CONDITION VIBRATION MONITORING WITH USING THE S-DISCRIMINANTS

The paper presents the results of the specific vibration condition monitoring approach for maintenance of unique critical equipment provided complete access impossibility to its components (due to radiation environment) throughout the entire operation period. The approach bases on the previously proposed S-discriminants of vibrations which are calculated with adaptation to the reference state of the equipment. The method showed its high efficiency in detecting the moment of machine operation process imbalance and in tracking the equipment components degradation, starting from the earliest stage. It is illustrated on the example of the movable neutron reflector PO-3 of the IBR-2 research reactor that it is possible not only to detect in time the moment when a fault occurs, but also to make a preliminary express diagnosis of its condition.

Defect topology and annihilation by cooperative movement of atoms in neutron-irradiated graphite

Graphite has been used as neutron moderator or reflector in many nuclear reactors. The irradiation of graphite in a nuclear reactor results in a complex population of defects. Heating of the irradiated graphite at high temperatures results in annihilation of the defects with release of an unusually large energy, called the Wigner energy. From various experiments on highly irradiated graphite samples from CIRUS reactor at Trombay and ab-initio simulations, we have for the first time identified various 2-, 3- and 4-coordinated topological structures in defected graphite, and provided microscopic mechanism of defect annihilation on heating and release of the Wigner energy. The annihilation process involves cascading cooperative movement of atoms in two steps involving an intermediate structure. Our work provides new insights in understanding of the defect topologies and annihilation in graphite which is of considerable importance to wider areas of graphitic materials including graphene and carbon nanotubes.