Neutron Transport Simulations Research Articles

Benchmarking and verification of the OpenMC code for accelerator-based neutron source analyses

OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. It is capable of performing fixed source particle transport simulation based on continuous-energy nuclear cross-section data. In this work, OpenMC has been benchmarked and verified for the application to accelerator-based neutron sources, e.g. the International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source facility (IFMIF-DONES). To this end, the McDeLicious code, which is an extension of the MCNP Monte Carlo code with the capability to simulate the deuterium-lithium neutron source on the basis of evaluated d + 6,7Li cross section data, has been migrated to the OpenMC code. The OpenMC-McDeLicious has been tested for the calculation of uncollided neutron and photon flux and spectra by comparing to results obtained with MCNP6-McDeLicious. SINBAD experimental benchmarks and d-Li experiments have been used for validating OpenMC. These tests and verifications show good agreement between OpenMC and MCNP code, as well as the experimental data. In addition, OpenMC-McDeLicious is used for obtaining the nuclear responses, e.g. Displacement Per Atom (DPA), gas production, nuclear heating, in the irradiation test module of the IFMIF-DONES neutron source. Most OpenMC-McDeLicious results show good agreement with MCNP6-McDeLcious within 5%. This indicates the successful benchmarking and validation of OpenMC code for the applications to accelerator-based neutron source facilities.

Nuclear performances of the water-cooled lithium lead DEMO reactor: Neutronic analysis on a fully heterogeneous model

The development of a conceptual design for the Demonstration Fusion Power Reactor (DEMO) is a key issue within the EUROfusion roadmap. The DEMO reactor is designed to produce a fusion power of about 2 GW and generate a substantial amount of electricity, relying on a closed tritium fuel cycle: it implies that the breeding blanket (BB) shall guarantee a suitable tritium production to enable a continuous operation without any external supply.The Water-Cooled Lithium Lead (WCLL) concept is a candidate for the DEMO BB: it uses liquid Lithium Lead as breeder and neutron multiplier and water in PWR condition as coolant. The neutronics analyses carried out in the past have been performed using a semi-heterogeneous representation of the BB, since the complexity of its structure makes the generation of a detailed MCNP model a very demanding and challenging task. Results highlighted good performances for the WCLL BB, both in terms of shielding effectiveness and tritium self-sufficiency.A recently updated assessment of the tritium breeding ratio (TBR) requirement for DEMO, considering margins for calculation uncertainties and incomplete models of the whole machine, led to the definition of a tentative 1.15 value for the TBR. Moreover, the implementation of an accurate BB neutronics model, consistent with the engineering design, is recommended for the evaluation of the tritium self-sufficiency. In order to tackle these issues, an MCNP model of the DEMO tokamak, integrating a fully heterogeneous WCLL BB has been developed for the first time, including an accurate description of the FW water channels, as well as a comprehensive definition of the breeding zone inner structure. A complete assessment of the WCLL BB nuclear performances, through 3D neutron and gamma transport simulations, has been carried out by means the MCNP Monte Carlo code and JEFF nuclear libraries.

Neutron Transport Simulations of RBMK Fuel Assembly Using Multigroup and Continuous Energy Data Libraries within the SCALE Code

The neutron transport simulations of RBMK-1500 fuel assembly were performed using both multigroup and continuous energy data libraries available within the SCALE code system in order to validate its suitability for the estimation of RBMK neutronic characteristics. The resonance processing of cross section, involved in the preparation of the multigroup data library, has a significant impact on neutron transport calculations. Standard Dancoff factors (DFs) used for the heterogeneous geometry of RBMK fuel assembly are insufficient for the accurate estimation of resonance self-shielding. Thus, the SCALE module MCDancoff was used in this study to determine location-specific DFs. The results of RBMK-1500 fuel assembly simulations using standard and user-defined DFs were compared. In addition, the continuous energy (CE) cross-section data library was applied for the benchmark calculations. The impact of different nuclear data libraries on neutron transport simulations was tested as well. It was found out that the usage of the multigroup data libraries generates some deviation from the reference simulations obtained with CE libraries. The CE library based on the estimated ENDF/B-VII.1 data proved to be the best alternative for neutron transport simulations of RBMK fuel assembly.

Benchmark experiment for bismuth by slab samples with D-T neutron source

A benchmark experiment is carried out with slab sample to validate nuclear data of bismuth by measurement of leakage neutron spectra from 0.8–16 MeV. The measured experiments are performed at China Institute of Atomic Energy with the integral benchmark facility established on the D–T neutron generator. The neutrons leaking from the sample are recorded at the measuring angle of 60° and 120° by time-of-flight technique. To make understanding of the impact of accuracy of data files in evaluated library on the neutron transport simulation, the calculation is carried out by MCNP-4C code with a series of latest available evaluated libraries. The calculated spectra are analyzed by comparing with experimental ones in neutron spectrum shape and C/E values. The result indicates that the theoretical simulation with different data libraries overestimated or underestimated the measured ones in some energy ranges, and the secondary neutron energy distribution and angular distribution in data files have been present to explain it.

Comparative activation analyses for the HCPB breeding blanket in DEMO

The Helium Cooled Pebble Bed (HCPB) breeding blanket is one out of two concepts considered as driver blanket for the European DEMOnstration fusion power plant. The latest design, developed within the Breeding Blanket project of EUROfusion’s Power Plant Physics and Technology (PPPT) programme, assumes a fission-like breeder pin arrangement with breeder ceramics and Be12Ti as neutron multiplier.For the assessment of safety and waste management issues related to the use of the HCPB blanket in DEMO, a thorough activation analysis is required. Such an analysis should provide the total activity inventory and decay heat as well as the breakdown into components and materials as resulting from the irradiation over the assumed blanket lifetime in DEMO. An accurate calculation of the activity inventories requires the application of a suitable computational approach coupling the neutron transport simulation and the activation calculation with fully heterogeneous 3D geometry model of the DEMO.Precise materials specifications, including impurities and tramp materials, are essential prerequisites for adequate activation calculations. For the materials used in the HCPB blanket, the material compositions are provided with the material specifications set up for PPPT neutronics and activation calculations. These include the Eurofer steel, the breeder ceramics, tungsten and the beryllium in the Be12Ti used as neutron multiplier.Two difference tritium breeding material compositions of activation analyses were performed using each Li4SiO4+Li2TiO3 the latest material and previous material definition Li4SiO4. And beryllium titanate, the activation calculations were performed using the impurities specifications as provided by different potential beryllium suppliers. The most essential here is a uranium content in the ore used for the Be12Ti metallurgical engineering. This can permit a justification of the potential environmental risks associated with the application of the various materials options. The effect of the different Lithium compositions on the activity, both on the short- and long-term time scale, is evaluated in this work in the context of the radioactive waste management issue for DEMO.

Concrete Modeling for Neutron Transport and Associated Sensitivity Studies Performed at the AMANDE-MIRCOM Facility

Concrete plays a major role in nuclear facilities as protection against radiation. However, its chemical composition, which is fundamental, is often unknown. Several concrete samples, extracted from the AMANDE-MIRCOM Institute for Radiological Protection and Nuclear Safety (IRSN) facility, were analyzed. Various simulations were performed in order to evaluate the neutron fluence behind a 40-cm-thick concrete wall. These simulations were compared to experimental measurements performed with a Bonner sphere spectrometer and a neutron survey meter. No set of parameters tested was able to produce a simulation accurately matching all the experimental results, but sensitivity studies on several parameters highlight that the three most sensitive parameters are the hydrogen content, the density, and the concrete inhomogeneity. To improve the agreement between the simulations and the measurements, the concrete inhomogeneity modeling should be studied further. Nevertheless, using concrete compositions that are close to reality, especially for hydrogen content, is crucial to correctly simulate neutron transport.

A linear prolongation CMFD acceleration for two-dimensional discrete ordinate k-eigenvalue neutron transport calculation with pin-resolved mesh using discontinuous Galerkin finite element method

The development and extension of the lp-CMFD acceleration scheme to the 1D Discontinuous Galerkin Finite Element Method (DGFEM) SN neutron transport problem was recently reported. However, the extension to 2D DGFEM neutron transport problems with pin-resolved unstructured mesh for high fidelity neutron transport simulations has not been previously investigated. In this paper, a projection operator for lp-CMFD is developed to project the linear additive flux to the pin-resolved unstructured mesh for DGFEM SN. To demonstrate its effectiveness on different coarse mesh geometry shapes and sizes, two test problems are presented: the 2D C5G7 problem and a hexagonal type reactor problem. It was found that the projection operator developed for lp-CMFD is straightforward to implement in 2D DGFEM SN with pin-resolved unstructured mesh and is stable and more effective than the other CMFD variants for two test problems. The method developed here is shown to be effective and can be applied in realistic reactor geometries for existing DGFEM neutron transport codes.

INTRODUCTION OF OSCAR-4 AT THE HIGH FLUX REACTOR (PETTEN)

Since 2005 the nodal diffusion based code system, OSCAR-3, was used for reactor support calculations of operational cycles of the High Flux Reactor in Petten, The Netherlands. OSCAR uses a two-step deterministic calculation, in which homogenized cross sections are generated in lattice environments using neutron transport simulations, and then passed to a nodal diffusion core simulator to model the full reactor. Limitations in OSCAR-3 led to the need for improved modelling capabilities and better physics models for components present in the reactor core. OSCAR-4 offers improvements over OSCAR-3 in its approach to homogenization, and the new version of the diffusion core simulator allows for better modelling of movable components such as control rods. Fuel inventories calculated using OSCAR-4 can also easily be exported to MCNP, which allows the calculation of individual plate powers and local reaction rates amongst others. For these reasons OSCAR-4 is currently being introduced as a core support tool at the High Flux Reactor. In this work the steps that were followed to validate the reactor models are presented, and include results of validation calculations from both OSCAR-4 and MCNP6 over multiple reactor cycles. In addition differences in cross section library evaluations and their impact on the results are presented for the MCNP model.

COMPARISON OF WASTE DUE TO IRRADIATED STEELS IN THE ESFR AND DEMO

For either nuclear fusion or generation IV fission reactors to be viable as a commercial energy source the decommissioning and waste disposal solutions must be considered during the design. A multi-step simulation process combining Monte Carlo Neutron Transport simulations with inventory simulations have been performed to estimate the activation of steels in key reactor components of the European Sodium-cooled fast Reactor (ESFR). Waste classifications based on UK waste disposal regulations have been applied to the key components to estimate the expected masses of low level and intermediate level waste. The use of reduced activation steels, EUROFER and F82H, in reactor components external to the core results in a factor of 10 reduction in the percentage of waste classified as Intermediate Level Waste (ILW). Waste estimates are compared to existing waste estimates for the European Demonstration fusion reactor (DEMO). The ESFR has a lower percentage of ILW per total reactor mass due to irradiated steels compared to DEMO. However, there is no Higher Activity Waste (HAW) associated with DEMO, compared with arisings from the ESFR spent fission fuel.

Simulating neutron transport in long beamlines at a spallation neutron source using Geant4

The transport of neutrons in long beamlines at spallation neutron sources presents a unique challenge for Monte-Carlo transport calculations. This is due to the need to accurately model the deep-penetration of high-energy neutrons through meters of thick dense shields close to the source and at the same time to model the transport of low- energy neutrons across distances up to around 150 m in length. Typically, such types of calculations may be carried out with MCNP-based codes or alternatively PHITS. However, in recent years there has been an increased interest in the suitability of Geant4 for such types of calculations. Therefore, we have implemented supermirror physics, a neutron chopper module and the duct-source variance reduction technique for low- energy neutron transport from the PHITS Monte-Carlo code into Geant4. In the current work, we present a series of benchmarks of these extensions with the PHITS software, which demonstrates the suitability of Geant4 for simulating long neutron beamlines at a spallation neutron source, such as the European Spallation Source, currently under construction in Lund, Sweden.

A POD reduced-order model for resolving the neutron transport problems of nuclear reactor

Due to the high-dimensional integro-differential properties of the neutron transport equation (NTE) and the large-scale property of the nuclear reactor system, the detailed neutron transport simulation can be very time-consuming, which urges researchers to develop a fast and accurate technique. To improve this condition, this work develops a set of reduced-order models for the NTE (NTEROM) by combining the proper orthogonal decomposition (POD) and the Galerkin projection techniques to efficiently solve the neutron transport problems. Results of some typical benchmarks show that the proposed POD based NTEROM can estimate the neutron transport behaviors with high fidelity and outstanding computing efficiency. Furthermore, the prediction function of the proposed NTEROMs can be extended to approximately calculate the neutron transport problems with the parameters outside of the snapshots’ parameter range. This work can provide new perspectives and ideas on the issue of improving the efficiency of numerical calculations in nuclear reactor engineering.

Benchmark experiment of large-angle scattering reaction cross section of iron at 14 MeV using two shadow bars – Comparison of experimental results with ENDF/B-VIII –

ABSTRACT It is important to perform neutron transport simulations with accurate nuclear data in the neutronics design of a fusion reactor. However, absolute values of large-angle scattering cross sections vary among nuclear data libraries even for well-examined nuclide of iron. Benchmark experiments focusing on large-angle scattering cross sections were thus performed to confirm the correctness of nuclear data libraries. The series benchmark experiments were performed at a DT neutron source facility, OKTAVIAN of Osaka University, Japan, by the unique experimental system established by the authors’ group, which can extract only the contribution of large-angle scattering reactions. This system consists of two shadow bars, target plate (iron), and neutron detector (niobium). Two types of shadow bars were used and four irradiations were conducted for one experiment, so that contribution of room-return neutrons was effectively removed and only large-angle scattering neutrons were extracted from the measured four Nb reaction rates. The obtained experimental results were compared with calculations for five nuclear data libraries including JENDL-4.0, JEFF.-3.3, FENDL-3.1, ENDF/B- VII, and recently released ENDF/B-VIII. It was found from the comparison that ENDF/B-VIII showed the best result, though ENDF/B-VII showed overestimation and others are in large underestimation at 14 MeV.

Nuclear analysis of the Water cooled lithium lead DEMO reactor

In the frame of the EUROfusion roadmap, the development of a conceptual design for the Demonstration Fusion Power Reactor (DEMO), beyond ITER, is a key issue. The DEMO reactor shall guarantee the tritium self-sufficiency, generate electricity and operate as a test facility for the fusion power plant relevant technologies, such as the breeding blanket (BB). The Water Cooled Lithium Lead (WCLL) concept has been chosen as a candidate for the DEMO BB: it relies on liquid Lithium Lead as breeder and neutron multiplier, Eurofer as structural material and pressurized water as coolant.A detailed MCNP model of the latest WCLL BB layout has been generated and integrated in the DEMO MCNP generic model suitably designed for neutronic analyses. Three-dimensional neutron and gamma transport simulations have been carried out by means the MCNP Monte Carlo code and JEFF nuclear data libraries in order to assess the WCLL-DEMO performances in terms of tritium self-sufficiency and shielding effectiveness to protect the vacuum vessel and the toroidal field coils. Moreover, the impact on the Tritium production of the water content in the first wall (FW) and the effect of its pressure/temperature has been addressed.The outcomes of the present study provide guidelines for the optimization of the WCLL DEMO reactor nuclear performances through the assessment of the loads on sensitive components and the estimation of its potential tritium generation capabilities.

Analysis of the multi-physics approach using the unified lattice Boltzmann framework

This work establishes a unified lattice Boltzmann method (LBM) framework to solve the nuclear reactor multi-physics processes involving the neutron transport and heat transfer processes, wherein all the SN, SP3 and P1 approximations are included for neutron transport simulation. The different physical fields can be solved using the general LBM format with different LBM variables definitions, while the same approach and boundary treatment can be applied due to the unified scheme. No complicated data interpolation is required, and the existing LBM codes can be easily extended. Results show that this unified framework can be used to solve the multi-physics problem in both transient and steady state accurately, and all the SN, SP3 and P1 approximations can be applied. This paper can provide a novel framework for the nuclear reactor multi-physics calculation, and due to the high localization, this framework can further be used to large-scale engineering calculation in parallel.

Neutronic analyses in support of the conceptual design of the DTT tokamak radial neutron camera

Abstract The Divertor Tokamak Test (DTT) facility, whose design phase is currently under finalization, is an Italian project aimed to investigate alternative power exhaust solutions for DEMO. It is designed to operate with significant power loads and enough flexibility to test innovative divertor configurations, different plasma edge and bulk conditions approaching, as much as possible, those planned for DEMO. Among the neutron diagnostics, a multi-channel neutron camera, most likely equipped with the liquid scintillators NE213, is foreseen to provide spatially resolved measurements of several plasma parameters needed for fusion power estimation, plasma control and plasma physics studies. This paper presents a preliminary study performed in support of the DTT neutron camera design. A detailed MCNP model representing a 20° sector of the machine integrating its main components and detectors assemblies has been developed and used for this study. Three-dimensional neutron transport simulations have been carried out by means of the MCNP Monte Carlo code coupled with the FENDL nuclear data libraries. The diagnostic design was optimized starting from the assessment of the expected detector performances obtained by using the calculated neutron fluxes and spectra and the NE213 response functions. The outcomes of this analysis provide the detectors requirements and guidelines for the development of the above-mentioned diagnostics, investigating its feasibility and suitability with the neutron emissivity foreseen for the DTT operational scenarios.

Neutron escape probabilities in 3-D reentrant geometries computed using ray tracing

Escape probabilities are fundamental inputs in the calculation of self-shielded cross sections for multigroup neutron transport models. Rational approximations are often used to compute escape probabilities in geometries commonly encountered in nuclear reactors. Although analytical approximations are simple and rapid, they are limited in applicability and accuracy. Here we show that ray tracing can be used to compute accurate escape probabilities in reentrant geometries with and without spherical symmetry. We use 3-D ray tracing to compute escape probabilities and their uncertainties in a spherical shell with a central void and a six-pointed cross shape. Results are benchmarked by comparison to Monte Carlo neutron transport simulations done in MCNP6.2 and analytical results. The escape probabilities in the spherical shell computed with ray tracing are within 0.51% of the analytical benchmark, and in the cross geometry the escape probabilities are within 1.4% of the MCNP6.2 benchmark.

Comparison of neutron induced single event upsets in 14 nm FinFET and 65 nm planar static random access memory devices

<sec> Based on the wide-spectrum neutron beam (covering thermal neutrons and <i>E</i> > 10 MeV neutrons, with maximum energy of 1.6 GeV) provided by the China Spallation Neutron Source (CSNS), this paper focuses on the single event effect study of 14 nm FinFET large-capacity SRAM and 65 nm planar process SRAM device, using combined techniques of irradiation experiment, reverse analysis, and Monte-Carlo neutron transport simulation. The aim is to reveal the effect of integrated circuit process changing on the sensitivity of neutron induced single-bit and multiple-bit upsets (MBU), and to analyze the inner mechanisms, including the distribution of secondary particles in the sensitive volume, the characteristics of deposited charges, etc. </sec><sec> The results show that compared with the 65 nm device, single event upset (SEU) cross section of the 14 nm FinFET device, induced by <i>E</i> > 10 MeV neutrons, is reduced by about 40 times, while the MBU ratio increases from 2.2% to 7.6%, which is due to the reduction of sensitive volume size of the 14 nm FinFET device (80 nm × 30 nm × 45 nm), pitch, and critical charge (0.05 fC). The main forms of MBU are double-bit upset, triple-bit upset and quadruple-bit upset. Unlike the phenomenon that the 65 nm device is immune to thermal neutrons, the use of the <sup>10</sup>B element near M0 in the 14 nm FinFET device causes it to present the thermal neutron sensitivity to a certain extent. The SEU cross section induced by thermal neutrons is about 4.8 times smaller than that induced by <i>E</i> > 10 MeV neutrons. </sec><sec> Based on the device cross-section and memory area images obtained from the reverse analysis, a device model is established and neutron transport simulation based on Geant4 toolkit is carried out. The <i>E</i> > 10 MeV neutrons result in abundant secondary particle distribution in the sensitive volume of the device, covering n, p into even W. The neutron energy and presence or absence of the W plug near the sensitive volume have an importantinfluence on the type and probability of secondary particles in the sensitive volume. The analysis and calculations show that a large number of high-<i>Z</i> secondary particles with long range and large LET values generated by high-energy neutrons in the sensitive volume of the device are the inducement of MBU, and SEUs mainly result from the contribution of light ions such as p, He, and Si. </sec>

Time Implicit Unified Gas Kinetic Scheme for 3D Multi-Group Neutron Transport Simulation

Time Implicit Unified Gas Kinetic Scheme for 3D Multi-Group Neutron Transport Simulation

On the impact of nuclear data uncertainties on LWR neutron dosimetry assessments

In this work, an overview on the relevance of the nuclear data (ND) uncertainties with respect to the Light Water Reactors (LWR) neutron dosimetry is presented. The paper summarizes results of several studies realized at the LRT laboratory of the Paul Scherrer Institute over the past decade. The studies were done using the base LRT calculation methodology for dosimetry assessments, which involves the neutron source distribution representation, obtained based on validated CASMO/SIMULATE core follow calculation models, and the subsequent neutron transport simulations with the MCNP® software. The methodology was validated using as reference data results of numerous measurement programs fulfilled at Swiss NPPs. Namely, the following experimental programs are considered in the given overview: PWR “gradient probes” and BWR fast neutron fluence (FNF) monitors post irradiation examination (PIE). For the both cases, assessments of the nuclear data related uncertainties were performed. When appropriate, a cross-verification of the deterministic and stochastic based uncertainty propagation techniques is provided. Furthermore, the observations on which particular neutron induced reactions contribute dominantly to the overall ND-related uncertainties are demonstrated. The presented results should help with assessing the overall impact of the various nuclear data uncertainties with respect to dosimetry applications and provide relevant feedback to the nuclear data evaluators.

Development of the real-time double-ring fusion neutron time-of-flight spectrometer system at HL-2M

A real-time double-ring neutron time-of-flight (TOFII) spectrometer system has been proposed to achieve plasma diagnosis on HL-2M tokamak with a relatively high count rate and sufficient energy resolution. The TOFII system is in its development stage, and this work describes its characteristics in terms of design principle, system structure, electronic system design, preliminary tests, and neutron transport simulation. The preliminary test results illustrate that the TOFII system can demonstrate the real-time dynamic spectrum every 10 ms. The results also show that based on the support vector machine method, the n–γ discrimination algorithm achieves the discrimination accuracy of 99.1% with a figure of merit of 1.30, and the intrinsic timing resolution of the system is within 0.3%. The simulated flight time spectrums from 1 to 5 MeV are obtained through the Monte Carlo tool Geant4, which also provide the reasonable results. The TOFII system will then be calibrated on mono-energetic neutron sources for further verification.