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Highlights • Investigation of the influence of homogenization on nuclear heating and shielding. • Heterogeneous model with detailed description of the BSM cooling system. • Homogeneous model simplified by removing pipes and introducing mixed materials. • Nuclear heating was calculated for several components and on a fine mesh for both models. • Homogenization tends to underestimate the local heat deposition and overestimates shielding. Abstract The electron cyclotron-heating upper launcher (ECHUL) will be installed in four upper ports of the ITER tokamak thermonuclear fusion reactor. Each ECHUL is able to deposit 8 MW power into the plasma for plasma mode stabilization via microwave beam lines. These beam lines and other components needed for the guidance of the microwaves are mounted into the upper port plug. In order to protect these components and to mitigate radiation leaving the reactor, several shield blocks are also included in the port plugs design. The ECHUL components which are closest to the fusion plasma are heated by the gamma and neutron radiation and therefore will need to be actively cooled. This paper reports on the influence of material homogenization by comparing results obtained in the neutronic analyses performed for the cooling system of the ECHUL blanket shield module (BSM). A small reduction in shielding performance of the BSM was found in the case of a heterogeneous description of the BSM compared to the standard approach of homogenizing the materials of the BSM by removing pipes and cooling circuits for the neutronic analyses. [ABSTRACT FROM AUTHOR]

Abstract Accurate measurement of the nuclear environment within a test tritium breeding-blanket module of a fusion reactor is crucial to determine tritium production rates which are relevant to self-sufficiency of tritium fuel supply, tritium accountancy and also to the evaluation of localised power levels produced in blankets. This requires evaluation of the time-dependent spectral neutron flux within the test tritium breeding-blanket module under harsh radiation and temperature environments. The application of an activation foil-based spectrometer system to determine neutron flux density using a pneumatic transfer system in ITER has been studied, deployed and tested on the Joint European Torus (JET) machine in a recent deuterium - deuterium campaign for a selection of high purity activation foils. Deployment of the spectrometer system has provided important functional and practical testing of the detector measurement system, associated hardware and post processing techniques for the analysis of large data sets produced through the use of list mode data collection. The testing is invaluable for the optimisation of systems for future planned testing in tritium - tritium and deuterium - tritium conditions. Analysis of the time and energy spectra collected to date and the status of the development of methods for post-processing are presented in this paper. [ABSTRACT FROM AUTHOR]

Highlights • Generation of a DAGMC calculation model of the HELIAS power reactor. • First neutronic analyses performed for the HELIAS power reactor. • Results for Neutron Wall Loading (NWL), Tritium Breeding Ratio and Neutron Flux. • Two different NWL approaches, results show agreement within statistical uncertainty. Abstract A first neutronics analysis of the Helical-Axis Advanced Stellarator (HELIAS) power reactor is conducted in this work. It is based on Monte Carlo (MC) particle transport simulations with the Direct Accelerated Geometry Monte Carlo (DAGMC) approach which enables particle tracking directly on the CAD geometry. A suitable geometry model of the HELIAS reactor is developed, including a rough model of a breeder blanket based on the Helium Cooled Pebble Bed (HCPB) breeder blanket concept. The resulting model allows to perform first neutronic calculations providing a 2D map of the neutron wall loading, a 3D distribution of the neutron flux, and a rough assessment of the tritium breeding capability. It is concluded that the applied methodology, making use of MC particle transport simulations based on the DAGMC approach, is suitable for performing nuclear analyses for the HELIAS power reactor. [ABSTRACT FROM AUTHOR]

Abstract Different design modifications of a Double Null (DN) DEMO were analysed by means of 3D Monte Carlo calculations using the MCNP code to assess the effect of a DN divertor configuration and various in-vessel components (IVC) on the tritium breeding performance. A simplified DN DEMO model was set up to this end. A Tritium Breeding Ratio (TBR) of 1.14 was obtained with this model for the HCPB breeder blanket concept without taking into account any auxiliary equipment such as limiters or port plugs. The inclusion of such design modifications results in a significant reduction of the TBR for different breeder blanket concepts. To compensate the loss of breeder space in the reactor either the integration of breeder materials in the divertor body or the enlargement of the available breeder zone could be considered. Both of these options can provide an increase of the TBR and ensure a sufficient margin for a DN DEMO with a HCPB breeder blanket. Meeting the TBR requirement with a Pb-Li based breeder blanket concept would be more challenging and requires significant changes to the overall IVC configuration. [ABSTRACT FROM AUTHOR]

Abstract The role of nuclear data for neutronics analyses of Fusion Technology (FT) facilities is presented in this paper. Nuclear data evaluations of high quality are required for transport simulations, uncertainty assessments, activation and radiation damage calculations affecting the design and performance of the FT facilities, as well as safety, licensing, waste management and decommissioning issues. A corresponding programme on the nuclear data development (NDD) and qualification is conducted within the Power Plant Physics and Technology (PPPT) programme of EUROfusion supporting the development of the DEMO fusion power plant and the IFMIF-DONES neutron source. The programme builds on the achievements of pre-ceding NDD activities within the European fusion programme as addressed in the paper. Further needs for design, shielding, activation and radiation dose calculations are discussed, and recommendations are given to further improve and qualify the nuclear data base for PPPT nuclear analyses. This includes dedicated experimental activities tailored to the needs of DEMO and IFMIF-DONES. [ABSTRACT FROM AUTHOR]

The deterministic solution of the neutron transport problem entails the coupled solution of several partial differential equations, one for each energy group, direction and/or spherical harmonic. Several techniques have been devised for accelerating the solution of this set of equations, both for time dependent and eigenvalue calculations. This paper describes an acceleration technique based on reduced order models and applicable to the segregated solution of time dependent solutions. In this work the technique is applied to the simple case of multi-group diffusion and tested on two cases of practical interest. It shows performances that are comparable to some commonly employed acceleration techniques. Some potential advantages have been observed for transients with significant flux deformations. In addition, possibly interesting features of the proposed technique are: a relatively easy implementation in general PDE solvers and numerical libraries; its potential applicability to any kind of problem requiring the iterative solution of a system of equations; a flexible implementation with a wide margin for possible modifications. [ABSTRACT FROM AUTHOR]

Abstract In nuclear system design, the neutron diffusion equation (NDE) is always used for steady and transient neutron behaviors simulation and in many types of nuclear reactor systems, the whole reactor core consists of several to hundreds of fuel assembles (FAs). In this paper, the Block Radial Basis Function Collocation Meshless (BRBFCM) method is specially developed based on local and global RBFCM method to solve the NDE in nuclear core scale. The whole calculation domain is divided into several blocks according to the FA arrangement and node collocation is used for spatial discretization. Five neutronics benchmark problems are analyzed, including the 2D2G, 2D4G, 2D-IAEA, 2D-TWIGL and 4G-LMFBR problem. The calculation results and the reference values are agreed well, which demonstrates that the more convenient BRBFCM method can solve neutronic problems accurately. Besides, the comparison between BRBFCM, LRBFCM and GRBFCM shows that the BRBFCM has the advantages both on accuracy and efficiency. [ABSTRACT FROM AUTHOR]

The csg2csg tool is a minimal dependency Python3 based program to facilitate the comparison of radiation transport Monte Carlo codes. The csg2csg can currently parse MCNP and OpenMC geometry descriptions, and it can subsequently export MCNP, FLUKA, PHITS, OpenMC and Serpent geometry files. This translation is achieved by reading the input geometry into a class based system of objects corresponding to surfaces, cells, materials and so on. The objects are then translated into a ‘generic’ form. This form once generalised can be converted to any of the supported code formats and then radiation transport calculations can be performed. The code has been tested on several geometries; most notably a complex JET model and a 70k cell ITER C-Model geometry.

In the D-T fueled tokamak, the neutrons not only carry the approximately 80% energy released in the per fusion reaction, but also are the source of radioactivity in the fusion system. Therefore, high-fidelity neutronics simulation is required to support such reactor design and safety analysis. In the present work, taking European HCPB DEMO (Helium Cooled Pebble Bed demonstration fusion plant) developed by KIT (Karlsruhe Institute of Technology) as an example, the preliminary neutronics analysis covering the assessments of NWL (neutron wall loading), TBR (tritium breeding ratio), nuclear power generation, radiation loads on PFCs (plasma-facing components) and TFCs (toroidal field coils) has been carried out by using SuperMC in the case of both unbiased and biased simulations. The preliminary results indicate that the blanket scheme could satisfy the design requirements in terms of TBR and shielding of inboard blankets. Specially, a speed-up by ~164 times in the calculation for thick shielding region (TFC region) is achieved by using global weight windows generated via GWWG in SuperMC. In addition, compared to MCNP, SuperMC shows advantages in accurate and efficient modeling of complex system, efficient calculation and 3D interactive visualization.

Engineering design studies are underway to assess the feasibility of converting the High Flux Isotope Reactor (HFIR) to operate with low-enriched uranium silicide dispersion (LEU3Si2-Al) fuel. These studies are supported by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization. A systematic approach employing neutronic and thermal-hydraulic analyses has been performed with the ORNL Shift and HFIR Steady State Heat Transfer Code tools, respectively, to predict reactor performance and thermal safety margins for proposed LEU3Si2-Al fuel designs. The design process was initiated by generating an optimized design with fabrication features identified from previous studies that result in excellent performance and safety metrics. The approach continued by substituting a single fabrication feature anticipated to be difficult to manufacture with another feature expected to perform an analogous function to that of the removed feature. Four conceptual fuel element design candidates, with various fabrication features, for conversion of HFIR with 4.8 gU/cm3 LEU3Si2-Al fuel have been generated and shown to meet pre-defined performance and safety metrics. Results to date indicate that HFIR could convert with the subject fuel system and meet performance and safety requirements if, among other considerations, fabrication of the specific design features are demonstrated and qualification of the fuel is complete under HFIR-specific conditions.

The neutron transport equation can be used to model the physics of the nuclear reactor core. Its solution depends on several variables and requires a lot of high precision computations. One can simplify this model to obtain the SPN equation for a generalized eigenvalue problem. In order to solve this eigenvalue problem, we usually use the inverse power iteration by solving a source problem at each iteration. Classically, this problem can be recast in a mixed variational form, and then discretized by using the Raviart-Thomas-Nédélec Finite Element. In this article, we focus on the steady-state diffusion equation with heterogeneous coefficients discretized on Cartesian meshes. In this situation, it is expected that the solution has low regularity. Therefore, it is necessary to refine at the singular regions to get better accuracy. The Adaptive Mesh Refinement (AMR) is one of the most effective ways to do that and to improve the computational time. The main ingredient for the refinement techniques is the use of a posteriori error estimates, which gives a rigorous upper bound of the error between the exact and numerical solution. This indicator allows to refine the mesh in the regions where the error is large. In this work, some mesh refinement strategies are proposed on the Cartesian mesh for the source problem. Moreover, we numerically investigate an algorithm which combines the AMR process with the inverse power iteration to handle the generalized eigenvalue problem.

Analyses of radiation fields resulting from a deuterium-tritium (DT) plasma in fusion devices is a critical input to the design and validation of many aspects of the reactor design, including, shielding, material lifetime and remote maintenance requirements/scheduling. Neutronics studies, which perform in-depth analysis are typically performed using radiation transport codes such as MCNP, TRIPOLI, Serpent, FLUKA and OpenMC. The Serpent 2 Monte-Carlo code, developed by VTT in Finland, is the focus of this work which seeks to benchmark the code for fusion applications. The application of Serpent 2 in fusion specific analysis requires validation of the codes performance in an energy range, and a geometrical description, which significantly differs to conventional nuclear fission analysis, for which the code was originally developed. A Serpent model of the Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) mock up experiment has been prepared and the calculated results compared against experimental data, as well as the reference Monte Carlo code MCNP. The analysis is extended to a model of DEMO with HCPB blanket concept. For this model, the flux, nuclear heating, tritium production and DPA are calculated, all of which are integral nuclear responses in fusion reactor analysis. In general, a very good agreement is demonstrated for both of the benchmarks, with any discrepancies pinpointed to different physics models implemented.

EU DEMO should be the first demonstrative fusion power plant, and it is expected to show that net electric energy can be produced and supplied to the grid. The availability and lifetime of the plant will be comparable to that of a commercial fusion power plant. The capacity to provide considerable net electric power safely and continuously to the grid (from 100 MW to 500 MW) is one of DEMO's significant purposes. The neutrons cause defects, leading to the transmutation of elemental atoms when interacting with the divertor component materials. These reactions produce gases, notably helium, which can induce material swelling and embrittlement. In this paper, the DPA and helium production in divertor has been assessed, mainly focusing on the structural material: EUROFER. Helium production rates reach around 117 appm/fpy as the peak, while DPA values reach about 4 dpa/fpy as the maximum value. These calculations were performed using the MCNP6 code with FENDL 3.2 nuclear data and the ADVANTG tool. • High energy neutron fluxes are one of the essential issues limiting the lifespan of the EU DEMO. • MCNP and ADVANTG - tools have been used. • Helium production rates reach up to 117 appm/fpy in the divertor Eurofer. • DPA rate in the divertor Eurofer reaches 4 dpa for a full power year. [ABSTRACT FROM AUTHOR]

• The preliminary design of helium-cooled mixed bed breeder (HCMB) blanket for CFETR is performed. • The neutronics and thermal-hydraulics analyses of CFETR are performed. • The global TBR of HCMB is larger than that of HCCB. Chinese Fusion Engineering Test Reactor (CFETR) is a tokamak device designed with different power levels to validate fusion engineering technologies in China. Helium-cooled ceramic breeder (HCCB) blanket and Water-cooled ceramic breeder (WCCB) blanket are the main candidates. Compared with WCCB, HCCB generally has a higher tritium breeding ratio (TBR) because helium has smaller neutron absorption cross sections than water. However, the TBR of the HCCB decreases with the increase of the fusion power. In this paper, the preliminary neutronics and thermal-hydraulics design of the Helium-Cooled Mixed bed Breeder (HCMB) blanket is proposed. The blanket structure with a fusion power of 1.5 GW is designed based on the current HCCB design of the CFETR. The HCMB adopts helium as the coolant and the mixed Li 4 SiO 4 /Be 12 Ti as the tritium breeder. The neutronics and thermal-hydraulics performances of HCMB blanket have been analyzed in the paper. The results show that the HCMB blanket design has a TBR up to 1.25 with the thermal-hydraulics requirements satisfied. [ABSTRACT FROM AUTHOR]

As the operational period of the Joint European Torus (JET) nears its scheduled conclusion in 2023, accurate predictions of the nuclear waste inventory and end-of-life shutdown dose rates (SDRs) based on actual operations are required for the transition to the repurposing and decommissioning of the reactor. The novel nuclear waste analysis method based on the Rigorous Two-Step (R2S), Cell-under-Voxel (CuV) approach used in the newly updated Mesh Coupled R2S (MCR2S) code has been employed to give precise predictions for the waste inventory and SDRs in JET. From this analysis, SDRs from the reactor and individual in-vessel components after extraction were obtained to guide waste management considerations. Also, crucially, the CuV method increased the precision of waste inventory predictions, allowing for a more rigorous decommissioning of the machine. • JET nuclear analysis performed. • Nuclear waste analysis performed using CuV method in MCR2S. • SDR analysis performed for JET and isolated in-vessel components. • Improved analysis method allows improved decommissioning and waste management of JET. [ABSTRACT FROM AUTHOR]

• Nuclear loads on Cr-based monoblock divertor target for DEMO have been assessed. • The impact of WCLL and HCPB BB on the load's distribution has been quantified. • Nuclear loads peaks are achieved on the external units of the target baffle region. • Damage and heating loads are higher with WCLL than HCPB blanket. • Helium production appears to be not significantly affected by the blanket type. The Plasma Facing Components (PFCs) of DEMO divertor play a fundamental role for the heat removal and particle exhaust functions, during fusion plasma. This implies a very harsh operation environment characterized by the combination of complex loading and stress conditions, intense particle bombardment, high heat flux deposition and a significant neutron irradiation. In the framework of the European DEMO divertor project (WPDIV), several target design concepts alternative to the baseline ITER-like technology are under evaluation, such as Chromium block design. It consists of chromium block, tungsten armor tile and copper alloy cooling pipe, and it is very attractive from an activation reduction, waste management and safety point of view. Therefore, the assessment of the most impacting nuclear loads for the PFC design is a pivotal aspect for DEMO divertor R&D program and this study is fully devoted to the comprehension of the spatial distributions of the main nuclear quantities (radiation damage, He-production and nuclear heating density) on the overall divertor targets. Neutronics analyses have been performed with MCNP5 Monte Carlo code using both DEMO MCNP Water Cooled Lithium Lead (WCLL) and Helium Cooled Pebble Bed (HCPB) blanket models, with a fully heterogeneous representation of the divertor cassette and Chromium PFCs geometry. The results of this analysis show that nuclear heating density and radiation damage distributions are higher with WCLL blanket than HCPB, except for He-production which seems not significantly affected by the blanket configuration. For each nuclear quantity, the peak values are achieved in correspondence of the baffle region and especially on the external units of the targets, as expected, being this region the most exposed to irradiation. [ABSTRACT FROM AUTHOR]

In this paper, we present a comparison between measured and calculated decay values for spent nuclear fuel assemblies from the Clab, GE-Morris EMAD facilities. Two codes are applied: CASMO5 and SNF , using different nuclear data libraries (JEF-2.2, JEFF-3.1.1, JEFF-3.2, ENDF/B-VII.1 and ENDF/B-VIII.0) and different code versions. Additionally values from four standard and guides for calculation of decay heats (ANS 2014, DIN 2014, ISO 2022 and RG-3.54 2018) were also compared with the measurements. All simulations with different codes, code versions and methods are based on the same geometry and irradiation descriptions for a single assembly, providing more convenient comparisons of results. The impact of nuclear data libraries and computational methods are shown, compared to the spread of C/E (Calculations over Experiment) values between the 256 considered measurements. Finally, a comparison of the most important nuclides contributing to the decay heat is presented between measured assemblies and more recent cases corresponding to higher burnup, initial enrichment and longer cooling times. • Comparison between calculated and measured decay heat for spent nuclear fuels. • Use of different nuclear libraries. • Use of standard methods. • Determination of the main contributors. [ABSTRACT FROM AUTHOR]

Future fusion reactors using deuterium–tritium fuel will exhibit high fluences of high-energy neutrons inside and around the reactor vacuum vessel (VV). As well as causing material damage, fusion neutrons will activate materials, the decay of which leads to radiation fields in and around the reactor after shutdown. Gamma-ray emission from activated materials is a particular radiological hazard during periods of reactor shutdown. This must be accounted for in the design of the reactor shielding to ensure that risks are reduced as low as reasonably achievable. Recent neutronics work has evaluated the shutdown dose rates (SDDRs) in the EU DEMOnstration power plant (DEMO) around the ports and throughout the cryostat, incorporating prospective shielding improvements to the VV and ports. Prior to the proposed shielding design improvements, calculations for the model including the helium-cooled pebble bed (HCPB) blanket showed that radiation leakage through the blanket and VV leads to biological-equivalent SDDRs (following 12 days' decay) above 1 0 3 μ Sv/h throughout the cryostat, ignoring additional contribution from radiation streaming through the port openings. Inclusion of the proposed VV changes reduces this dose rate to below 100 μ Sv/h. The work finds an approximate order-of-magnitude reduction in SDDR throughout the cryostat when all proposed shielding improvements are applied, leading to dose rates in the cryostat in the range of hundreds to thousands of μ Sv/h for the full model. The work shows that to further reduce dose rates inside the cryostat, improving the shielding performance of the ports is required, with particular emphasis on the lower port and the equatorial electron–cyclotron launcher which currently dominate the dose rates. • Proposed EU-DEMO shielding improvements are incorporated into neutronics models. • Gamma-ray dose rates following shutdown are calculated throughout the cryostat. • Approximate order-of-magnitude dose rate reductions result from the improvements. • Vacuum vessel design changes are found to be effective in reducing radiation leakage. • Targeted shielding around the equatorial and lower ports should be prioritised. [ABSTRACT FROM AUTHOR]

The standard modular Equatorial Port Plug (EPP) in ITER is at its final stage of design. It serves as common platforms for different diagnostics systems which are integrated together in three cassettes called Diagnostic Shielding Module (DSM). The equatorial port #12 will host an EPP with six diagnostics: the Visible-Infrared Camera and the H-α Monitor integrated in DSM #1, the Glow Discharge Wall Conditioning, the Radial X-rays Camera and the Hard X-rays Monitor, installed in DSM #2 and the Core Thomson Scattering in DSM #3. The EPP should not only serve as a diagnostic port, but also it has to provide the necessary shielding to allow for maintenance in port interspace and in port cell. This work is devoted to nuclear analyses performed to assess the nuclear loads and the shielding performance of the system. A very detailed neutronic model of EPP #12 has been developed and integrated in a 40° ITER sector MCNP model (C-model). The implemented model has been used to evaluate the neutron and gamma flux distributions, nuclear heating, radiation induced damage, tritium and helium concentration at the ITER end of life in the port plug area. [ABSTRACT FROM AUTHOR]

Fuel burnup analysis requires a high computational cost for full core calculations, due to the amount of the information processed for the total reaction rates in many burnup regions. Indeed, they reach the order of millions or more by a subdivision into radial and axial regions in a pin-by-pin description. In addition, if multi-physics approaches are adopted to consider the effects of temperature and density fields on fuel consumption, the computational load grows further. In this way, the need to find a compromise between computational cost and solution accuracy is a crucial issue in burnup analysis. To overcome this problem, the present work aims to develop a methodological approach to implement a Reduced Order Model (ROM), based on Proper Orthogonal Decomposition (POD), in fuel burnup analysis. We verify the approach on 4 years of burnup of the TMI-1 unit cell benchmark, by reconstructing fuel materials and burnup matrices over time with different levels of approximation. The results show that the modeling approach is able to reproduce reactivity and nuclide densities over time, where the accuracy increases with the number of basis functions employed.

Nuclear electric propulsion (NEP) offers unique advantages for the interplanetary exploration. The extremely high conversion efficiency of magnetohydrodynamics (MHD) conversion nuclear reactor makes it a highly potential space power source in the future, especially for NEP systems. Research on ultra-high temperature reactor suitable for MHD power conversion is performed in this paper. Cermet is chosen as the reactor fuel after a detailed comparison with the (U,Zr)C graphite-based fuel and mixed carbide fuel. A reactor design is carried out as well as the analysis of the reactor physics and thermal-hydraulics. The specific design involves fuel element, reactor core, and radiation shield. Two coolant channel configurations of fuel elements are considered and both of them can meet the demands. The 91 channel configuration is chosen due to its greater heat transfer performance. Besides, preliminary calculation of nuclear criticality safety during launch crash accident is also presented. The calculation results show that the current design can meet the safety requirements well.

The European Spallation Source bunker is a common shielding area that surrounds the target monolith, protecting the instrument areas from the high dose of ionizing radiation that accompanies the spallation reaction. A comprehensive neutronic work was performed, which took into account dose rates outside the bunker wall and roof, skyshine and activation. A wall with a thickness of 3.5 m and a stepped roof with a thickness of 1–1.5 m, both made of heavy concrete with a density of 3.8 g/cm3, were recognized as the best solution that satisfies the radioprotection requirements for the operation of the facility. [ABSTRACT FROM AUTHOR]

Abstract A neutron shielding assessment of an integrated ITER C-lite neutronics baseline model for the Remote Handling Lower Port rack and port interspace components has been performed. The aim of this paper is to decrease the neutron leakage as a mean to eventually reduce the activation relevant for shutdown dose rates. The performance of the neutron shielding assembled in the baseline model of the Remote Handling Lower Port rack has been assessed. As a mitigation strategy, neutronics calculations have been conducted to evaluate the radiation field during operation in the lower port with the Remote Handling Lower Port rack and the interspace components installed. A progressive improvement of the neutron shielding has been carried out with additional shielding elements including a large shielding block. As a result, the neutron streaming has been significantly reduced along the lower port with a neutron fluence rate in the closure flange in the range of 1·109 n/cm2s to 1·108 n/cm2s. [ABSTRACT FROM AUTHOR]

The Collective Thomson Scattering (CTS) will be the ITER diagnostic responsible for measuring the alpha-particle velocity distribution. Using mirrors, a powerful microwave beam is directed into the plasma via an opening in the plasma-facing wall. The microwaves will scatter off fluctuations in the plasma, and the scattered signal is recorded after transmission through a series of mirrors and waveguides. Several components of the CTS system will be directly exposed to neutron radiation from the plasma which can change the properties of the components and reduce their lifetime. In this paper, a neutronics analysis is presented for the CTS system. A study on the influence of different materials on the nuclear heat loads in the launcher mirror is also presented, along with the design of a simple cooling system. All the studies were conducted using the Monte Carlo program MCNP6. The outputs, in particular the nuclear heat loads, will be used to perform the thermal analysis of the system. [ABSTRACT FROM AUTHOR]

The ITER plasma position reflectometry system will be used to estimate the distance between the position of the magnetic separatrix and the first wall at four predefined locations, complementing the magnetic diagnostics system. The antennas of the system studied in this paper are to be installed between the blanket shield modules of rows 3 and 4, on the high-field side of the tokamak. As the antennas and part of the corresponding waveguides will be directly exposed to the plasma, they will be subjected to high radiation doses from neutrons and gammas, which may cause irradiation-induced changes in the material properties and compromise the integrity of the components. In this paper, the Monte Carlo simulation code MCNP6 and the most up-to-date ITER reference neutronics model were used to estimate the thermal loads in the system. The results, complemented with a finite-element analysis performed with ANSYS Mechanical, show that the nuclear heat loads, combined with the thermal radiation coming from the plasma, translate into operation temperatures that reach 480 °C at the tips of the antennas, slightly above the maximum recommended temperature of operation of stainless steel for ITER in-vessel components under irradiation (450 °C). Further, thermal analyses and design iterations are therefore required to improve the thermal and mechanical behavior of the system. [ABSTRACT FROM AUTHOR]

DONES (DEMO-Oriented Neutron Source) is to be constructed in order to investigate materials that are intended to be used for DEMO (DEMOnstration Power Plant) since they will be irradiated with very high fluxes of neutrons. A deuteron beam will bombard a lithium target situated in the so called Test Cell (TC) shielded with concrete. Monte Carlo simulations by means of the MCNP (Monte Carlo N-Particle) code were carried out to calculate both horizontal and vertical radiation dose maps in and around of the TC facility. Three geometrical configurations of the shielding and two kinds of concrete (ordinary and magnetite ones) were examined. For each case a spatial dose distribution and an accompanying errors map were provided. Advanced techniques of variance reduction were applied to compute doses in and outside of thick concrete walls. On the grounds of the obtained results a final configuration of the shielding was selected. [ABSTRACT FROM AUTHOR]

Molten salt reactors (MSRs) are selected as one of six generation-IV nuclear power systems due to its remarkable advantages in terms of inherent safety, nuclear non-proliferation and economy. Liquid fuel makes the neutronics and thermal-hydraulics characteristics of MSRs different from those of conventional solid fuel reactors. In this study, transients perturbed by control rod ejection, local perturbation and over feeding U235 were analyzed for a graphite-moderated MSR based on a 3D coupled neutronics/thermal hydraulics code MOREL2.0. The effective delayed neutron fraction decreased with the rise of fuel mass flow. The core power changed more fiercely and even occurred supercritical phenomenon for larger fuel mass flow when the same reactivity was introduced. Local blockage introduced negative reactivity, and MSR maintained a safety state. While local overcooling resulted in positive reactivity introduction, and core power increased first then tended to be stable due to the negative temperature feedback. Furthermore, the response of core power at the transient of over feeding fissile isotopes was simulated. Those numerical results provide valuable information for the research and design of this new generation reactor. [ABSTRACT FROM AUTHOR]

The development of next-generation nuclear reactors requires a careful investigation of their stability characteristics and of their overall dynamical behavior. In the current work, a stability analysis is carried out from the perspective of linear systems for all the Gen-IV reactor concepts. Linear, zero-dimensional models of the reactors are developed in MATLAB ® and the root locus criterion is applied to investigate the stability of the systems over their entire power range. The analysis is carried out for the stand-alone cores, assuming the inlet coolant temperature as a fixed parameter, and also considering the primary coolant circuit of each reactor, in order to evaluate the effect of out-of-core energy dynamics on the systems stability. All the reactors proved to be stable with a large margin over their nominal power. In the last part of the work, a comparison of the dynamic behavior of the Gen-IV reactors is presented, in order to point out the influence of the different geometrical features and of the different materials employed as coolant. [ABSTRACT FROM AUTHOR]

As one of the advanced reactors proposed by the Generation IV International Forum (GIF), Lead-bismuth eutectic (LBE) coolant cooled fast reactor has shown great advantages over the other reactors. To analysis this reactor, several steps are carried out covering the code development and the coupled analysis. Firstly, the physical property of lead-bismuth, heat transfer correlation, turbulent mixing and pressure drop models are implemented into the thermal-hydraulics code COBRA-YT, where the inlet flow rate boundary condition is calculated by Computational Fluid Dynamics (CFD). The neutronics model has seven neutron energy groups and six groups of delayed neutron precursors. The neutronics parameters such as cross-section and feedback coefficient are calculated by the DRAGON code with endfb7 library. Secondly, based on sub-channel code COBRA-YT and neutron code SKETCH-N, the coupled of neutronics and thermal-hydraulics code has been developed via Parallel Virtual Machine (PVM) software platform. COBRA-YT code performs the thermal-hydraulics calculation and transfers its results such as coolant density and fuel temperature to the neutronics code SKETCH-N to update the cross-section; then SKETCH-N carries out the neutron-physical simulation of the reactor and provides the power density to the thermal-hydraulics code COBRA-YT as boundary conditions. Finally, this coupled code platform is exploited to the lead-bismuth fast reactor design to simulate some transient and control rod withdrawal accidents. The results achieved so far indicates that an increase of the reactor power and cladding temperature occurs after the control rod withdrawal, but the thermal-hydraulics results exceed their design limits. Time step size effect and withdrawal velocity analysis are also discussed. [ABSTRACT FROM AUTHOR]

We have been developing an in-house deterministic neutron and gamma transport solver, AETIUS, which uses an unstructured tetrahedral mesh for spatial discretization such as Attila. We make the most of the spatial discretization of AETIUS for a shutdown dose rate analysis. This is similar to the MCR2S unstructured mesh method with the exception that the deterministic solver AETIUS is used for the neutron and gamma transport calculations instead of MCNP(X). AETIUS provides neutron spectra for every unstructured tetrahedral element belonging to the activation zones. Activation calculations are then carried out using FISPACT-2007, and the resulting gamma spectra of each element are used as decay gamma sources for the AETIUS gamma transport calculation to obtain a shutdown dose map. We calculated the shutdown dose rate on the ITER port plug computational benchmark, and our results are the most similar to those of the FDS R2S method. [ABSTRACT FROM AUTHOR]

This paper introduces the verification and validation (V&V) work on the newly developed fast reactor code SARAX. The verifications were done based on numerical benchmark calculations and code-to-code comparisons with the Monte-Carlo code. The validation works were performed and compared with evaluated reactor physics experiments. In this paper, results from different calculation schemes are discussed, and the uncertainty from nuclear data is quantified. The results show that SARAX has good accuracy in modeling a sodium fast reactor regardless of the integration parameters, such as k eff , or the distributed parameters like the sub-assembly power/reaction rate. The uncertainty quantification shows that, for a MOX-fueled SFR, the greatest uncertainty in contributed nuclear data is from the inelastic scattering cross section of U-238, which requires further improvement to reduce the uncertainty of k eff . [ABSTRACT FROM AUTHOR]

The present paper addresses the impact of the divertor option on the nuclear performances of the Demonstration fusion reactor (DEMO). As the effect of the number and size of the divertor has been already evaluated, in this work the focus has been posed on the composition in terms of amount of cooling inside the divertor cassette. Transport responses, as the Tritium breeding ratio (TBR), neutron and gamma fluxes and spectra inside the plasma chamber, as well as activation responses such as shutdown dose rate, decay gamma fluxes and heating have been evaluated for two different blanket concepts of the future European DEMO reactor: DCLL and WCLL. Three different divertor compositions have been tested demonstrating the importance of this component not only locally but in the global radiation field. The transport analysis has been performed with the Monte Carlo code MCNP5 and the JEFF3.1.1 and JEFF3.2 data libraries. The activation responses calculated using Advanced D1S method have been recently assessed and summarized in the present paper. [ABSTRACT FROM AUTHOR]

Design of modern sodium fast reactors such as the French SFR project ASTRID are axially heterogeneous in order to maximize neutron leakage and therefore improve their natural behavior during transients. This feature is however challenging for neutronic calculations that usually rely on a first step based upon the generation of averaged cross sections over planar geometries. It is in particular the case of the APOLLO3® reference calculation route that is currently considered for SFR applications. This paper investigates the corresponding biases in the case of a 3D fissile assembly calculation and highlights the consequences of the fundamental mode assumption. It is shown that even though an acceptable global accuracy can be achieved, the flux distribution presents a slight axial drift. The homogenization and energy collapsing of anisotropic cross sections over radial geometries is pointed responsible for it. Several proposals are made to correct this inconsistency, leading to new calculation schemes and enhanced performances. [ABSTRACT FROM AUTHOR]

Additive manufacturing is under investigation as a novel method of fabricating the control elements (CEs) of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory with greater simplicity, eliminating numerous highly complex fabrication steps and thereby offering potential for significant savings in cost, time, and effort. This process yields a unique CE design with lumped absorbers, a departure from traditionally manufactured CEs with uniformly distributed absorbing material. This study undertakes a neutronics analysis of the impact of additively manufactured CEs on the HFIR core physics, seeking preliminary assessment of the feasibility of their practical use. The results of the MCNP transport simulations reveal changes in the HFIR reactor physics arising from geometric and nuclear effects. Absorber lumping in the discrete CEs yields a large volume of unpoisoned material that is not present in the homogeneous design, in turn yielding increases in free thermal flux in the CE absorbing regions and their immediate vicinity. The availability of additional free thermal neutrons in the core yields an increase in fission rate density in the fuel closest to the CEs and a corresponding increase in neutron multiplication on the order of 100 pcm. The absorption behavior exhibited by the discrete CEs is markedly different from the homogeneous CEs due to several competing effects. Self-shielding arising from absorber lumping acts to reduce the effective absorption cross section of the discrete CEs, but this effect is offset by geometric and spectral effects. The operational performance of the discrete CEs is found to be comparable to the homogeneous CEs, with only limited deficiencies in reactivity worth that are expected to be operationally recoverable via limited adjustment of the CE positions and withdrawal rate. On the whole, these results indicate that the discrete CEs perform reasonably similarly to the homogeneous CEs and appear feasible for application in HFIR. The physical phenomena identified in this study provide valuable background for follow-up design studies. [ABSTRACT FROM AUTHOR]

The development of a fine mesh coupled neutronics/thermal-hydraulics framework mainly using free open source software is presented. The proposed contributions go in two different directions: one, is the focus on the open software approach development, a concept widely spread in many fields of knowledge but rarely explored in the nuclear engineering field; the second, is the use operating system shared memory as a fast and reliable storage area to couple the computational fluid dynamics (CFD) software OpenFOAM to the free and flexible reactor core analysis code milonga . This concept was applied to model the behavior of a TRIGA-IPR-R1 reactor fuel pin in steady-state mode. The macroscopic cross-sections for the model, a set of two-group cross-sections data, were generated using the Serpent code. The results show that this coupled system gives consistent results, encouraging system further development and its use for complex geometries simulations. [ABSTRACT FROM AUTHOR]

This paper describes the high-priority testing activities supporting the ITER radial neutron camera (RNC) design, performed by a consortium of European institutes within a framework contract placed by fusion for energy, the ITER European Domestic Agency. The main role of the RNC is to measure the uncollided 14- and 2.5-MeV neutrons from deuterium–tritium and deuterium–deuterium fusion reactions through an array of flux monitors/spectrometers located in collimated lines of sight viewing the plasma through the ITER equatorial port plug #1. The line-integrated neutron fluxes will be used to evaluate, through reconstruction techniques, the radial profile of the neutrons emitted per unit time and volume (neutron emissivity) and, therefore, the neutron yield and the alpha particles’ birth profile. The activity of high-priority testing is dedicated to the preparation and the design of experimental test environment, the conduction of appropriate tests and reporting of test results for the high-priority prototypes, clarifying or verifying the expected key function and system behavior, and enhancing learning on specific issues (potential showstoppers). [ABSTRACT FROM AUTHOR]