Neutron Leakage Research Articles

External Moderation of Reactor Core Neutrons for Optimized Production of Ultra-Cold Neutrons

The ultra-cold neutron (UCN) source being commissioned at North Carolina State University’s PULSTAR reactor is uniquely optimized for UCN production in the former graphite-filled thermal column outside of the reactor pool. The source utilizes a remote moderation design, which is particularly well suited to the PULSTAR reactor because of its high thermal and epithermal neutron leakage from the core face. This large non-equilibrium flux from the core is efficiently transported to the UCN source through the specially designed beam port in order to optimize UCN production at any given reactor power. The increased distance to the source from the core also greatly limits the heat load on the cryogenic system. A MCNP (Monte Carlo N-Particle) model of this system was developed and is in good agreement with gold foil activation measurements using a test configuration as well as with the real UCN source’s heavy water moderator. These results established a firm baseline for estimates of the cold neutron flux available for UCN production and prove that remote moderation in a thermal column port is a valuable option for future designs of cryogenic UCN sources.

Advancements to generalized equivalence theory for preserving cross-sections of whole core simulations

The conventional two-step approach, without the use of an equivalence method, can introduce significant error for the simulation of non-LWRs (Light Water Reactors), especially fast reactors, due to the large leakage and high anisotropic neutron distribution. To improve upon the accuracy of the two-step approach, a 3D whole core model is simulated with a transport method to generate region-wise homogenized cross-sections (XS). These XS can then be used in a diffusion whole core solver during the second step to extend the application to cycle and transient analysis. Discontinuity factors (DFs) are then introduced to improve the accuracy during the simulation of the second step. With properly generated DFs from Generalized Equivalence Theory (GET), the region-averaged solutions from the first transport step can then be reproduced by the second step diffusion solver. The Monte Carlo method was selected to perform the whole core transport simulation to generate region-wise XS. However, simulating whole core problems with Monte Carlo may result in poor statistics near the peripheral region especially for partial current tallies. This paper introduces an advancement to GET to reproduce region-wise solutions for select regions when the reaction rates and surface currents have good statistics in these regions but poor statistics in other regions. A reference high-fidelity model was constructed using the Serpent 2 Monte Carlo code based on the EBR-II benchmark evaluation and verification was carried out using the TriPEN-4 method in PARCS. The results show that it is possible to reproduce the exact eigenvalue and power distributions of whole core problems in a feasible manner.

The Bimodal Epithermal Astronuclear Reactor (BEAR): A Long-Lived, Universal Space Nuclear Powerplant

A neutronic model of a Nuclear Thermal Propulsion reactor based on the NERVA Peewee design using High Assay Low Enriched Uranium (HALEU) fuel in a Zirconium Carbide matrix has been thoroughly characterised and continuously modified over the last few years at the CSNR. The initial design (Emu, for the flightless bird) created in the summer of 2020 and presented at the NETS 2021 conference, has undergone significant adjustments in this time. Two major investigations have been undertaken and evolved over time; one into the requisites for the provision of electrical power to the propelled spacecraft via the circulation of a noble gas mixture in a closed Brayton cycle, and another into the use of planetary regolith as a reflector, reducing neutron leakage at what would otherwise be the reactor’s end of life in vacuo, rekindling it for an extended period of time as well as providing a very easy ultimate disposal solution, replacing the control drums with rods. It was determined that there are no insurmountable obstacles to performance in either mode.

Review of Fourth-Order Maximum Entropy Based Predictive Modeling and Illustrative Application to a Nuclear Reactor Benchmark: II. Best-Estimate Predicted Values and Uncertainties for Model Responses and Parameters

This work continues the review and illustrative application to energy systems of the “Fourth-Order Best-Estimate Results with Reduced Uncertainties Predictive Modeling” (4th-BERRU-PM) methodology. The 4th-BERRU-PM methodology uses the Maximum Entropy (MaxEnt) principle to incorporate fourth-order experimental and computational information, including fourth (and higher) order sensitivities of computed model responses with respect to model parameters. The 4th-BERRU-PM methodology yields the fourth-order MaxEnt posterior distribution of experimentally measured and computed model responses and parameters in the combined phase space of model responses and parameters. The 4th-BERRU-PM methodology encompasses fourth-order sensitivity analysis (SA) and uncertainty quantification (UQ), which were reviewed in the accompanying work (Part 1), as well as fourth-order data assimilation (DA) and model calibration (MC) capabilities, which will be reviewed and illustrated in this work (Part 2). The applicability of the 4th-BERRU-PM methodology to energy systems is illustrated by using the Polyethylene-Reflected Plutonium (acronym: PERP) OECD/NEA reactor physics benchmark, which is modeled using the linear neutron transport Boltzmann equation, involving 21,976 imprecisely known parameters. This benchmark is representative of “large-scale computations” such as those involved in the modeling of energy systems. The result (“response”) of interest for the PERP benchmark is the leakage of neutrons through the outer surface of this spherical benchmark, which can be computed numerically and measured experimentally. The impact of the high-order sensitivities of the response with respect to the PERP model parameters is quantified for “high-precision” parameters (2% standard deviations) and “typical-precision” parameters (5% standard deviations). Analyzing the best-estimate results with reduced uncertainties for the 1st—through 4th-order moments (mean values, covariance, skewness, and kurtosis) produced by the 4th-BERRU-PM methodology for the PERP benchmark indicates that, even for systems modeled by linear equations (e.g., the PERP benchmark), retaining only first-order sensitivities is insufficient for reliable predictive modeling (including SA, UQ, DA, and MC). At least second-order sensitivities should be retained in order to obtain reliable predictions.

Review of Fourth-Order Predictive Modeling and Illustrative Application to a Nuclear Reactor Benchmark. I. Typical High-Order Sensitivity and Uncertainty Analysis

This work (in two parts) will review the recently developed predictive modeling methodology called “4th-BERRU-PM” and its applicability to nuclear energy systems as exemplified by an illustrative application to the Polyethylene-Reflected Plutonium (acronym: PERP) OECD/NEA reactor physics benchmark. The acronym 4th-BERRU-PM designates the “Fourth-Order Best-Estimate Results with Reduced Uncertainties Predictive Modeling” methodology, which uses the Maximum Entropy (MaxEnt) principle to incorporate fourth-order experimental and computational information, including fourth (and higher) order sensitivities of computed model responses to model parameters, while yielding best-estimate results with reduced uncertainties for the first fourth-order moments (mean values, covariance, skewness, and kurtosis) of the optimally predicted posterior distribution of model results and calibrated model parameters. The 4th-BERRU-PM methodology encompasses the scopes of high-order sensitivity analysis (SA), uncertainty quantification (UQ), data assimilation (DA) and model calibration (MC), as will be illustrated in this work by means of the above-mentioned OECD/NEA reactor physics benchmark. This benchmark is modeled using the neutron transport Boltzmann equation involving 21,976 imprecisely known parameters, the solution of which is representative of “large-scale computations”. The model result (“response”) of interest is the leakage of neutrons through the outer surface of this spherical benchmark, which can be computed numerically and measured experimentally. Part 1 of this work illustrates the impact of high-order sensitivities, in conjunction with parameter standard deviations of various magnitudes, on the determination of the expected value and variance of the computed response in terms of the first four moments of the distribution of the uncertain model parameters. Part 2 of this work will illustrate the capabilities of the 4th-BERRU-PM methodology for combining computational and experimental information, up to and including forth-order sensitivities and distributional moments, for producing best-estimate values for the predicted responses and model parameters while reducing their accompanying uncertainties.

Radial reflector macroscopic cross sections preparation impact on the low neutron leakage core Beavrs

Full core nodal calculations are essential for safe nuclear power plant operation. The accuracy and compliance with the calculation and the experiment are based on the proper macroscopic data preparation model. While fuel assembly data preparation is usually based on the simple infinite lattice model simulation, the reflector data preparation is not standardised and the influence of the reflector macroscopic data on full core simulation is not quantified. The paper evaluates the large core with low neutron leakage behaviour with the different reflector cross section preparation models with respect to criticality and power distribution prediction. However, in opposition to the first insight, the different reflector model cannot increase the compliance with the power prediction near the fuel assembly reflector boundary. However, the usage of a special rehomogenization model can be helpful for pin power reconstruction compliance near the edge of the core.

Possible criticality scenario and its mechanism of the Windscale Works criticality accident in 1970 analyzed by computational fluid dynamics and Monte Carlo neutron transport

The Windscale Works criticality accident in 1970 resulted from mixing an aqueous solution with an organic solvent with different plutonium concentrations and densities. Although this accident has been studied using improved computer capabilities in recent years, a precise criticality scenario has not yet been identified. This study aims to clarify a possible criticality scenario of the accident—the time variation of reactivity and its mechanism. The accident was simulated by combining the multiphase computational fluid dynamics solver of OpenFOAM and the delta-tracking-based Monte Carlo neutron transport code Serpent2. Consequently, the periodic uneven arrangement of fluids might have caused oscillations in neutron leakage and absorption, resulting in periodic wavy reactivity changes. Furthermore, the emulsion, which was thought to be the primary cause, might not be the dominant mechanism for reactivity change, although it contributed to the criticality of the accident.

Fast neutron leakage spectra of the EUCLID experiment

Special nuclear material in sub-critical and critical configurations measured in integral experiments are important for validation and adjustment of nuclear data. Many different evaluations of nuclear data exist, and these different evaluations can provide different values for individual cross sections that vary due to the uncertainties in differential experiments or lack of such data. For integral experiments, differences in these individual cross sections can have compensating errors, which lead to the same answer. One example of this is the Jezebel critical assembly, where keff of the system is correctly computed by both ENDF/B-VIII.0 and JEFF-3.3, despite having substantially different underlying evaluated values for specific reactions (such as elastic and inelastic cross sections).To reduce compensating errors in nuclear data, the Experiments Underpinned by Computational Learning for Improvements in Nuclear Data (EUCLID) project has utilized machine learning to design a set of sub-critical and critical experiments. These experiments include slab- and cube-like configurations of 239Pu in the form of the ZPPR plates. Six different responses were measured on a total of thirteen different configurations. One of these responses, the neutron leakage spectrum, was measured using an EJ301D detector. The results of the neutron leakage spectra show good agreement (within 1–2 σ) with the expected spectrum from simulations and will be used in the subsequent nuclear data adjustment done by the EUCLID team.

Practical Design of a Breed-and-Burn Lead-Cooled Fast Reactor with Rotational Fuel Shuffling Strategy

The purpose of this study is to demonstrate a practical core design for a lead-cooled, nitride fueled, rotational fuel shuffling breed-and-burn (RFBB) fast reactor. The core design is based on the Westinghouse Lead Fast Reactor (WH-LFR) and uses natural uranium nitride fuel with a sodium bond encased in oxide dispersion-strengthened steel cladding. Simulations confirmed the potential of the reactor to maintain criticality at the equilibrium state, with a reactivity swing of less than 200 pcm at every cycle interval and an average discharge burnup of 235 MWd/kg heavy metals (HM) for a 1050 effective full-power day refueling interval. Power profiles were maintained stable at the equilibrium state, while the cladding of the discharged fuel incurred over 650 displacements per atom over its entire residency in the core. From a nonproliferation perspective, the plutonium vector for the discharge fuel aligns with reactor-grade fuel standards, with over a 70% concentration of 239Pu and over 22% 240Pu, reducing the risk of weaponization. The adopted control rod system has been shown to offer sufficient negative reactivity of over 19 $ to bring the reactor into a subcritical state. Challenges such as the susceptibility of neutron balance to material thickness and neutron leakage have been addressed, emphasizing the necessity for meticulous design improvements. A steady-state thermohydraulic analysis confirmed the heat removal capacity from the hottest channel, ensuring operational safety. This study confirmed the feasibility of the RFBB strategy for a lead-cooled nitride-fueled fast reactor and sets a precedent for future research in enhancing fuel utilization and safety in nuclear reactors.

Fuel element microreactor integrating a square UO2 fuel rod with an internal heat pipe

In this study we propose the Fuel Element Micro Reactor (FEMR), with 308 square UO2 fuel rods, each one containing inside a heat pipe. We selected Mercury as the working fluid for the heat pipes and operate the microreactor at average temperature around 900 K or 637 °C. Heat pipes introduce empty regions into the core that increase neutron leakage and severely reduce core reactivity. Square fuel lattices allow greater fuel volume fractions and favor increasing core reactivity. The adoption of a 20 cm BeO reflector and Zircaloy as fuel cladding further increases core reactivity. The FEMR core thermal power is 2.3 MW, the power density distribution is rather flat with peaks occurring at the core-reflector interface and the peak fuel temperature is 1160 K. The reactor has a negative temperature isothermal coefficient of reactivity. The reactivity control system with 8 rotational drums and a central cruciform absorber rod meets the criteria of stuck rod and diversity of engineering principles. The average fuel discharge burnup is 8.96 MWd/kgU and the fuel cycle length without refueling is 8.7 years. The ratio between thermal power and total Uranium mass at beginning of life is 2.46 kW/kgU.

Simulation of Neutron Leakage Variations at Fuel Substitution in a Small Modular Reactor and Implications for Unattended Safeguards Monitoring

According to a recently proposed nuclear safeguards technique, monitoring the power-normalized, ex-core neutron detection rate over time could be used to detect undeclared changes to the fissile composition of a reactor core. In this study, Monte Carlo simulations have been used to verify some of the underlying assumptions of this technique and the possibilities of using it to detect undeclared fuel substitutions during the first 2-year cycle of a light water small modular reactor. Depletion calculations and neutron transport simulations were used to study the changes in the power-normalized neutron leakage rate J b / P core through the core barrel upon fuel substitutions and whether these changes are fully explained by changes in the core fissile composition. Several substitution scenarios have been studied, where partially depleted fuel assemblies were substituted with fresh fuel assemblies after 1 year of irradiation. The modeled substitution scenarios, which included substituting up to 4 out of 37 fuel assemblies in the core at a time, resulted in changes in J b / P core of up to 3.5% depending on which fuel assemblies were substituted. The results indicate that the ex-core neutron signatures are not only sensitive to core-averaged nuclide densities, fission cross sections, and neutron flux, but also the spatial distributions of these and other parameters throughout the core. Effects such as these could mean that monitoring the core fissile composition with the proposed technique might be more complex than previously suggested.

Generation of thermal neutron scattering data for GH3535 alloy

Recent studies related to the thermal neutron scattering (TNS) data in reactors were generally focused on fuels and moderators, while the TNS data for structural materials has received little attention. The TNS effect of the GH3535 alloy, a nickel-based alloy widely used in molten salt reactor (MSR), is studied in this work. The partial phonon densities of states (PDOS) for the main metals (Ni, Mo, Cr and Fe) in the alloy, which are required for the TNS data generation, are calculated using the density functional perturbation theory (DFPT) based on a special quasi-random structure (SQS) of the alloy. The TNS cross sections for the main metals in the alloy are generated using the modified NJOY code. To preliminarily validate the TNS data, the calculated scattering cross section, combined with the evaluated absorption cross section of the alloy, is compared with the experimental total cross section, which shows a good agreement. Moreover, an integral neutronics experiment of neutron leakage spectrum is designed to further validate the TNS data. The neutron leakage spectrum is simulated by the Monte Carlo method, which can be compared with the future experimental result. Based on the experiment geometry modeling, the TNS effects of the GH3535 alloy are analyzed at multiple incident neutron energies, sample thicknesses and temperatures, respectively. It is found that the TNS effect of the GH3535 alloy on the neutron leakage spectrum cannot be ignored, and the scattering cross section of the alloy with high precision should be included in designs of MSR.

The influence of source locations of different multi-beam concept on the power density distribution and nuclear waste transmutation performance for ADS–NWT

The multi-beam concept for the accelerator driven subcritical reactor (ADS) has the advantages in reducing the requirements for the accelerator and can flatten the power distribution of the core. However, the layout of the multi-beam should meet the goal of balancing the relationship between the total power density, the neutron leakage, and the nuclear waste transmutation performance. For the further study of the multi-beam concept for the ADS, this paper focuses on the influence of the source locations of different multi-beam concept on the power density distribution, the neutron leakage, and the nuclear waste transmutation performance for the ADS for a nuclear waste transmutation reactor (ADS–NWT). The different variations on the radial power density distribution and the nuclear waste transmutation performance for the three-, four-, six-, and seven-beam concept for ADS–NWT were analyzed. The results show that the related neutronics performance for the multi-beam concept for ADS-NWT is not only affected by the external source locations and the core geometry, but also affected by the keff and the interference effect between the spallation targets. Moreover, the fission reaction ratio and the fission reaction / (n, gamma) absorbtion ratio for minor actinides (MA) and plutonium (Pu) for the few beams concept are both higher than that for the many beams concept. Based on the analysis of the influence of source locations on the neutron spectrum, the total neutron flux, and the nuclear waste transmutation performance for MA and Pu, the related conclusions about how to arrange the best beam position from the perspective of nuclear waste transmutation and production capacity are provided.

The Bimodal Epithermal Astronuclear Reactor (BEAR): A LONG-LIVED, Universal Space Nuclear Powerplant

A neutronic model of a Nuclear Thermal Propulsion reactor based on the NERVA Peewee design using High Assay Low Enriched Uranium (HALEU) fuel in a Zirconium Carbide matrix has been thoroughly characterised and continuously modified over the last few years at the CSNR. The initial design (Emu, for the flightless bird) created in the summer of 2020 and presented at the NETS 2021 conference, has undergone significant adjustments in this time. Two major investigations have been undertaken and evolved over time; one into the requisites for the provision of electrical power to the propelled spacecraft via the circulation of a noble gas mixture in a closed Brayton cycle, and another into the use of planetary regolith as a reflector, reducing neutron leakage at what would otherwise be the reactor’s end of life in vacuo, rekindling it for an extended period of time as well as providing a very easy ultimate disposal solution, replacing the control drums with rods. It was determined that there are no insurmountable obstacles to performance in either mode.

Compacting the core of a commercial high-power lead-cooled fast reactor

Two options for the layout of a fast reactor with lead coolant are being considered. The design of the fuel assemblies and core fuel rods is identical to the BREST-OD-300 reactor. One of the layouts is traditional with a flattened core. The other layout is the most compact. It has an elongated core and is characterized by an optimal ratio of core diameter to height in terms of minimal neutron leakage. The electric capacity of the first of them is about 2.9 GW, the second - 13.3 GW. The second layout is a high-temperature reactor. When operating at rated power, the maximum coolant temperature reaches 1631 K. The use of cermets fuel based on UN-PuN mononitride micrograins and uranium metal nanopowder, coolant based on lead thorium ores and tungsten fuel rod coatings made it possible to ensure the safety of reactors in emergency modes such as "anticipated transients without scram". A distinctive feature of the layouts is to reduce the danger of emergency modes when they are applied simultaneously. In all emergency modes, large reserves are maintained up to the maximum permissible temperatures of fuel, coolant and structural materials.

Measurement of leakage neutron flux from reactor tank surface of AHWR-critical facility

AHWR - Critical Facility (AHWR - CF) is a “zero power” reactor designed to carry out various reactor physics experiments for validation of AHWR physics design. AHWR-CF is a vertical tank type reactor. This paper describes the experiments carried out for neutron flux measurements on the reactor tank surface of AHWR-CF. A number of bare Au and cadmium covered Ni foils were used to measure thermal and fast neutron flux at the outer surface of the reactor tank. One cadmium covered Au foil was also irradiated at one of the locations on the tank surface, to estimate the epithermal component of neutron flux. Westcott thermal flux on the tank surface (at an elevation of 120 cm from tank bottom) was found to be (5.48 ± 0.57)E+6 n cm−2 s−1 on one of the sides of the reactor tank. Cadmium ratio for Au for this location was 40.28, indicating a highly thermalized neutron spectrum. No statistically significant activity was found in the irradiated Ni foils, indicating the absence of fast neutron flux above the measurable threshold. Low value of flux on the reactor tank surface implies that any irradiation damage in the AHWR-CF reactor tank material will be insignificant over its period of operation.

Measurement and simulation of the leakage neutron spectra from Fe spheres bombarded with 14 MeV neutrons

Measurement and simulation of the leakage neutron spectra from Fe spheres bombarded with 14 MeV neutrons

Development and verification of a one-dimensional collision probability based neutron transport code to model axially heterogeneous cylindrical vessels containing aqueous and organic plutonium nitrate

This paper presents the development and verification of a collision probability (CP) code, capable of modelling neutron transport in one-dimensional slabs and axially heterogeneous cylinders with varying radii. The CP code is used to model layered systems of aqueous and organic plutonium nitrate, as process criticality accidents are more likely to occur in these systems compared to dry systems. The use of the CP code is desired as it offers a computationally inexpensive method for calculating neutron transport when compared to higher fidelity codes such as MCNP. For slab geometries, the CP code can be used effectively, given they contain at least 0.7 g cm−2 plutonium. The approximation employed by the CP code to model heterogeneous cylinders overestimated the rate of radial neutron leakage such that vessels with radii of 30.0 cm could not reliably calculate reactivity to within 1 $ of MCNP. Increasing the radii to 40.0 cm improved the accuracy of the CP code to within 1 $ of MCNP for systems containing at least 2.75 kg plutonium. The error in the CP code increased when used to model cylindrical geometries with dished ends and complete vessels with dished ends, such that systems with large dished ends and low plutonium content should be avoided. As a simple, neutronics based model, the CP code could be used as part of rough order of magnitude calculations for criticality transients, where high levels of accuracy are not required, given that potential errors in results have been previously identified.

Measurement and analysis of leakage neutron spectra from Lead slab samples with D–T neutrons

The leakage neutron spectra from three different sizes of Lead samples were measured by a TOF technique at 60° and 120°. The essential characteristic properties of the experimental measurement spectra can be reproduced well by MCNP code simulations with the ENDF/B-VIII.0, CENDL-3.2, JENDL-5.0, JEFF-3.3 and TENDL-2021 evaluated nuclear data libraries. The calculated results of JENDL-5.0 and JEFF-3.3 libraries agree better with the experimental data in the whole energy range. The results from ENDF/B-VIII.0 and CENDL-3.2 are overestimated in the 4–9 MeV range at 60° and in the 4–12.5 MeV range at 120°. The differences of the leakage neutron spectra by MCNP simulations using five evaluated nuclear data libraries mainly originate from the differences of the spectrum distributions of neutron reaction channels in these libraries. And the secondary neutron energy distribution and angular distribution from the five libraries have been present to explain it.

Measurement of Neutron Leakage and Facility Survey of a Newly Commissioned Quadruple Photon Energy Medical Linear Accelerator Using a Neutron Survey Meter

Measurement of Neutron Leakage and Facility Survey of a Newly Commissioned Quadruple Photon Energy Medical Linear Accelerator Using a Neutron Survey Meter