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In the prototype fast breeder reactor Monju, reaction rate distributions of the fission reaction rates of 239Pu, 235U, and 238U and the capture reaction rate of 238U were measured using activation foils during its system startup test. The measurements in the core and radial blanket regions were evaluated in detail, and their reliability and usefulness as the validation data for fast reactor neutronics design methodologies were examined through a comparison with calculations. The reaction rate data measured in Monju were confirmed all reliable and useful as the validation data. The fission reactions of 239Pu, 235U, and 238U can be validated with an accuracy of a few percent in the core and blanket regions. The capture reaction of 238U in the core region also can be validated with a similar accuracy, whereas a precise calculation of the foil cross section is necessary to consider the resonance shielding effects of the surrounding fuel pins and a foil. [ABSTRACT FROM AUTHOR]

This work presents neutronics models of a small and large fast-spectrum molten chloride-salt reactor. The models are similar to designs being pursued by industry, and they may serve as generic preconceptual and simplified neutronics models that provide information for decision making in licensing-related areas. The two models were created using Serpent, a Monte Carlo neutron transport code, and Moltres, a neutron diffusion core simulator tool. Specifically, this study focused on exploring the applicability of diffusion theory to fast molten salt reactor (MSR) models, the capabilities of an open-source, MSR-oriented simulation tool (Moltres), and optimal energy-group structures. The proposed two-step method involves group-constant generation with Serpent and a multigroup diffusion solution by Moltres. Three energy-group structures were applied. The accuracy of the solutions was determined through comparisons between the two-step and Monte Carlo flux and multiplication factor solutions. The findings indicated diffusion theory captures neutronics with minimal error for the large MSR and yielded best results with the 27-group structure. The 27-group structure yielded an average group flux error below 2% and keff agreement between diffusion and transport solutions within 30 pcm. The accuracy of the two-step method decreased for the very small (high-leakage) fast chloride MSR, but the neutronics were captured acceptably well with the 33-group structure. In addition to exploring the capabilities of Moltres, this work contributes to the sparse literature involving open-source models of fast-spectrum MSRs. Future work is noted as expanding the capabilities of the neutronics models to incorporate thermal hydraulics. [ABSTRACT FROM AUTHOR]

ITER is of key importance in the European fusion roadmap as it aims to prove the scientific and technological feasibility of fusion as a future energy source. The EUROfusion consortium of labs within Europe is contributing to the preparation of ITER scientific exploitation and operation and aspires to exploit ITER outcomes in view of DEMO. The paper provides an overview of the major progress obtained recently, carried out in the frame of the new (initiated in 2021) EUROfusion work-package called ' Pr eparation of I TER O peration' (PrIO). The overview paper is directly supported by the eleven EUROfusion PrIO contributions given at the 29th Fusion Energy Conference (16–21 October 2023) London, UK [www.iaea.org/events/fec2023]. The paper covers the following topics: (i) development and validation of tools in support to ITER operation (plasma breakdown/burn-through with evolving plasma volume, new infra-red synthetic diagnostic for off-line analysis and wall monitoring using Artificial Intelligence techniques, synthetic diagnostics development, development and exploitation of multi-machine databases); (ii) R&D for the radio-frequency ITER neutral beam sources leading to long duration of negative deuterium/hydrogen ions current extraction at ELISE and participation in the neutral beam test facility with progress on the ITER source SPIDER, and, the commissioning of the 1 MV high voltage accelerator (MITICA) with lessons learned for ITER; (iii) validation of neutronic tools for ITER nuclear operation following the second JET deuterium–tritium experimental campaigns carried out in 2021 and in 2023 (neutron streaming and shutdown dose rate calculation, water activation and activated corrosion products with advanced fluid dynamic simulation; irradiation of several materials under 14.1 MeV neutron flux etc). [ABSTRACT FROM AUTHOR]

Nuclear thermal propulsion is an evolving technology that can be utilized for long-distance space travel. This technology yields the advantage of a high thrust and specific impulse, but requires an examination of the potential design adjustments necessary to enhance its feasibility. The development of nuclear thermal propulsion requires a comprehensive understanding of the system-level behavior during transient and steady-state operation. This paper extends our previous research by including the proper handling of turbomachinery with multi-channel thermal hydraulic simulations only for steady-state solutions. The system-level approach presented here enables the treatment of the turbopump components through non-dimensional analysis that eliminates the assumption of constant efficiencies. All the other components within the system (e.g., reflector and core) can be discretized to multiple channels and layers, in which the full thermal hydraulic solution is established. The approach chosen here enables the realistic modeling of the propellant flow within the expander cycle by capturing the pressure losses, mass flow rate splits, and enthalpy gain for various operational conditions. The verification of the package is completed through point comparisons of previous investigations into similar system designs. Furthermore, sensitivity studies are used to benchmark the capabilities of the package and investigate solution variations due to the perturbation of operational conditions and regimes. The sensitivity studies performed here are important to capture variation in flow characteristics (e.g., temperature, pressure, mass flow rates) for different design objectives such as the thrust and specific impulse. This work demonstrates that system-level simulations lacking multi-channel capability and proper turbomachinery treatment may yield higher uncertainties in understanding the engine's response and characteristics to changing various requirements. This is extremely important when screening the design space of such propulsion systems and when transient simulations are required. [ABSTRACT FROM AUTHOR]

This paper introduces a multiphysics code, named nemoFoam, designed within the OpenFOAM environment to address the coupling of neutron and photon transport with thermal-hydraulics in nuclear reactor simulations. The code, conceived both for fusion and fission applications, employs a modular approach where the neutronic module is currently based on multi-group neutron diffusion equations, including a mono-kinetic treatment for photons. The thermal-hydraulic module is built on the standard OpenFOAM solver. The paper focuses on presenting the key features of nemoFoam and discussing the structure of the code, the nuclear and thermal-hydraulic models, and the coupling strategy between them.In order to assess the performance of the neutronic module, the code is applied to a fusion case study, indeed a benchmark against the Serpent Monte Carlo code for neutron and photon transport is performed, applying nemoFoam to the Affordable, Robust and Compact (ARC) fusion reactor design. Simulation results demonstrate good agreement with Monte Carlo benchmarks, emphasizing the potential of the code for multiphysics simulations. The modular design of nemoFoam allows users to extend the implemented model equations to study additional phenomena, focus only on selected aspects independently and, potentially, to add new solvers in each module.

In this study, we address the impact of irradiation conditions in a tokamak on the engineering properties of materials, leading to potential degradation of in-vessel components over their lifecycle. Our approach involves a predictive model for irradiation-induced damage, employing a multiscale computational framework. This framework integrates various simulation techniques, including Monte Carlo-based neutronics (OpenMC), dislocation dynamics (DD) using MoDELib, and finite element analysis (FEA) with Code_Aster. This integration offers a versatile solver capable of analysing tokamak components exposed to different irradiation doses and temperature conditions.To showcase the utility of this multiscale computational framework, we present a case study focused on tungsten monoblock designs. We assess the failure probabilities of these designs at different stages of their lifecycle. Neutron heating and damage energy values are obtained from OpenMC neutronics simulations. The neutron heating values serve as volumetric heat sources for the FEA thermal simulation. We calculate the displacement per atom (dpa) across the monoblock at various full power days (day 0, day 100, and day 1000) using the damage energy values. The irradiation-induced defect densities, dependent on temperature and dpa, are inputs to DD microstructural simulations performed on the representative volume element (RVE) using MoDELib. This allows us to obtain the yield stress of the material. Subsequently, the thermal fields from the FEA thermal simulation, along with the dpa and temperature-dependent yield stress from the DD simulation, are implemented for FEA mechanical simulations.To evaluate the failure probability of the monoblock designs at different stages of their lifecycle, we conduct an SDC-IC assessment, incorporating a plastic flow localization rule within the current framework. This comprehensive approach provides insights into the thermo-mechanical behaviour of in-vessel components subjected to neutron irradiation, offering a predictive capability for assessing their performance over time.

Blanket is of vital importance for engineering application of the fusion reactor. Nuclear heat deposition in materials is the main heat source in blanket structure. In this paper, the three-dimensional method for thermal-hydraulics/neutronics coupling analysis is developed and applied for the full-scale module of the helium-cooled ceramic breeder tritium breeding blanket (HCCB TBB) designed for China Fusion Engineering Test Reactor (CFETR). The explicit coupling scheme is used to support data transfer for coupling analysis based on cell-to-cell mapping method. The coupling algorithm is realized by the user-defined function compiled in Fluent. The three-dimensional model is established, and then the coupling analysis is performed using the paralleled Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC). The results reveal the relatively small influence of the coupling analysis compared to the traditional method using the radial fitting function of internal heat source. However, the coupling analysis method is quite important considering the nonuniform distribution of the neutron wall loading (NWL) along the poloidal direction. Finally, the structure optimization of the blanket is carried out using the coupling method to satisfy the thermal requirement of all materials. The nonlinear effect between thermal-hydraulics and neutronics is found during the blanket structure optimization, and the tritium production performance is slightly reduced after optimization. Such an adverse effect should be thoroughly evaluated in the future work.

There is widespread use of nuclear radiation for medical imagery and treatments. Worldwide, almost 40 million treatments are performed per year. There are also applications of radiation sources in other commercial fields, e.g., for weld inspection or steelmaking processes, in consumer products, in the food industry, and in agriculture. The large number of neutrons generated in a fusion reactor such as DEMO could potentially contribute to the production of the required radioactive isotopes. The associated commercial value of these isotopes could mitigate the capital investments and operating costs of a large fusion plant. The potential of producing various radioactive isotopes was studied from material pieces arranged inside a DEMO equatorial port plug. In this location, they are exposed to an intensive neutron spectrum suitable for a high isotope production rate. For this purpose, the full 3D geometry of one DEMO toroidal sector with an irradiation chamber in the equatorial port plug was modeled with an MCNP code to perform neutron transport simulations. Subsequent activation calculations provide detailed information on the quality and composition of the produced radioactive isotopes. The technical feasibility and the commercial potential of the production of various isotopes in the DEMO port are reported. [ABSTRACT FROM AUTHOR]

Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup neutron diffusion model for MSR analysis, with the primary aim for fast and accurate calculations for long transients, as well as sensitivity and uncertainty analysis of the reactor. A fictitious radial leakage cross section is introduced in the model to properly account for the radial leakage effects of the reactor. The leakage cross section and other consistent neutronics parameters are generated with the Monte Carlo code Serpent using high-fidelity three-dimensional (3D) models. The accuracy of the 1D consistent model is verified by the reference solution from the Monte Carlo model on the Molten Salt Reactor Experiment (MSRE) configuration. The 1D consistent model successfully reproduced the integrated flux from the 3D model and the reactor multiplication factor keff with the error in the range of 95 to 397 pcm (per cent mille), depending on discretized energy group structures. The developed model is also extended to estimate the reactivity loss due to fuel circulation in MSRE. The estimate of reactivity loss in dynamics analysis is in great agreement with the experimental data. This model functions as the first step in the development of a 1D fully neutronics/thermal-hydraulics coupled model for short- and long-term MSRE transient analysis. [ABSTRACT FROM AUTHOR]

Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup neutron diffusion model for MSR analysis, with the primary aim for fast and accurate calculations for long transients, as well as sensitivity and uncertainty analysis of the reactor. A fictitious radial leakage cross section is introduced in the model to properly account for the radial leakage effects of the reactor. The leakage cross section and other consistent neutronics parameters are generated with the Monte Carlo code Serpent using high-fidelity three-dimensional (3D) models. The accuracy of the 1D consistent model is verified by the reference solution from the Monte Carlo model on the Molten Salt Reactor Experiment (MSRE) configuration. The 1D consistent model successfully reproduced the integrated flux from the 3D model and the reactor multiplication factor keff with the error in the range of 95 to 397 pcm (per cent mille), depending on discretized energy group structures. The developed model is also extended to estimate the reactivity loss due to fuel circulation in MSRE. The estimate of reactivity loss in dynamics analysis is in great agreement with the experimental data. This model functions as the first step in the development of a 1D fully neutronics/thermal-hydraulics coupled model for short- and long-term MSRE transient analysis.

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.

Nuclear thermal propulsion is an evolving technology that can be utilized for long-distance space travel. This technology yields the advantage of a high thrust and specific impulse, but requires an examination of the potential design adjustments necessary to enhance its feasibility. The development of nuclear thermal propulsion requires a comprehensive understanding of the system-level behavior during transient and steady-state operation. This paper extends our previous research by including the proper handling of turbomachinery with multi-channel thermal hydraulic simulations only for steady-state solutions. The system-level approach presented here enables the treatment of the turbopump components through non-dimensional analysis that eliminates the assumption of constant efficiencies. All the other components within the system (e.g., reflector and core) can be discretized to multiple channels and layers, in which the full thermal hydraulic solution is established. The approach chosen here enables the realistic modeling of the propellant flow within the expander cycle by capturing the pressure losses, mass flow rate splits, and enthalpy gain for various operational conditions. The verification of the package is completed through point comparisons of previous investigations into similar system designs. Furthermore, sensitivity studies are used to benchmark the capabilities of the package and investigate solution variations due to the perturbation of operational conditions and regimes. The sensitivity studies performed here are important to capture variation in flow characteristics (e.g., temperature, pressure, mass flow rates) for different design objectives such as the thrust and specific impulse. This work demonstrates that system-level simulations lacking multi-channel capability and proper turbomachinery treatment may yield higher uncertainties in understanding the engine’s response and characteristics to changing various requirements. This is extremely important when screening the design space of such propulsion systems and when transient simulations are required.

A core design optimization process is developed as part of the design optimization support tool named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) for an efficient and innovative core design process. The process comprises analyses integrated by neutronics, thermal-hydraulics in a fuel assembly, fuel integrity, and plant dynamics for safety assessment. The optimal design parameters are explored using the Bayesian optimization (BO) algorithm to reduce the number of iterative calculations and solve the optimization problem. This study defines a representative problem by identifying the objective functions, constraints, and design parameters for an actual core design based on previous core design experiences to efficiently develop the core design optimization process. Next, a constrained single-objective optimization problem, the simplified representative problem, is solved by the integrated analyses only with neutronics and plant dynamics using the BO algorithm to confirm the applicability of the optimization process for the representative problem. Consequently, the design parameters that optimized the objective function within constraints can be obtained. The optimal solution correlates well with the reference solution. Furthermore, the effectiveness of the optimization process is discussed by comparing an ordinary and a defined core design process.

In the present work, an assessment of the Neutronic Benchmark of the Molten Salt Fast Reactor (MSFR) was performed using mesh based Monte Carlo Neutron Transport (MCNT) calculations with numerical uncertainty quantification due to discretization in neutronic parameters. Calculations with Constructive Solid Geometry (CSG) models where made as a baseline for the developed mesh based models. The numerical uncertainty given by the mesh utilization is evaluated using an extended version of the Grid Convergence Index (GCI). The fuel salt reprocessing is evaluated regarding a constant reprocessing rate. The fuel salt inventory variation with time for the developed models (CSG and meshed) is presented. The differences caused by the discretization procedure are noticeable, which shows that mesh based MCNT require careful mesh sensitivity evaluation and further validation.

CABRI is an experimental reactor at the French Alternative Energies and Atomic Energy Commission (CEA) used to study fuel behavior during reactivity insertion transients. As these transients have a high level of multiphysics, it is important to develop suitable modeling and simulation tools to simulate them in order to be able to optimize testing and control of experimental conditions. This paper focuses on the development and validation of the neutron deterministic APOLLO3® calculation scheme that is included in the CABRI neutronic/thermal-hydraulic multiphysics coupled simulation tool; it represents the first stage of a stepwise validation process for the CABRI multiphysics simulation tool. The neutron calculation scheme is based on a classical two-step approach. The first step consists of a 281-energy-group flux calculation with the TDT-MOC (Method of Characteristics) solver for nuclear data space and energy collapsing for the different CABRI assembly clusters. The biases on a two-dimensional (2D) core neutron calculation due to self-shielding correction and collapsing on a restricted pattern are investigated by means of comparison with a direct full 2D calculation on a quarter core. The second step relies on a three-dimensional (3D) core calculation. Two approaches are presented. The first one consists of a best-effort approach corresponding to a 3D pin-scale description of the core, performing a transport calculation with the SN solver MINARET. And, the second one, a best-estimate approach, which will be implemented for kinetics calculations, relies on solving a simplified transport SPN equation in the solver MINOS with an exact 3D cell description of the core. The best-estimate calculation scheme is then used to analyze three experimental CABRI transients. A stepwise validation process is followed to quantify the calculation biases on physical parameters such as reactivity, reaction rates, and total core power at each step using static reference calculations with the stochastic code TRIPOLI4® or transient experimental data. The next development stage toward a multiphysics calculation scheme will be implementation and validation of coupling with a core thermal-hydraulic model. [ABSTRACT FROM AUTHOR]

Inline algorithms have been proposed for coupling Monte Carlo neutron transport solvers with several other physics, such as xenon and iodine densities and thermal hydraulics. This paper proposes a new inline algorithm that can be applied to burnup calculations. The algorithm is a modification of the predictor-corrector method, where the corrector-step nuclide densities are converged simultaneously with the fission source. This could, in principle, obviate the need for two full neutronics solutions per time-step while still allowing the accuracy of predictor-corrector methods with improved stability. This paper describes the algorithm and demonstrates its stability properties through a Fourier analysis. Although not unconditionally stable, judicious use of batching and relaxation are shown to greatly improve the algorithm's stability properties in realistic systems. [ABSTRACT FROM AUTHOR]

For evaluating control rod (CR) interaction, CR worth values with and without the insertion of interaction CRs were measured in six CR interaction patterns by the continuous period method and the compensation method in the prototype fast breeder reactor Monju. The measured CR worth was evaluated in detail by conducting corrections, such as the thermal contraction of CR drive shafts, and compared with calculations to examine the reliability. The calculation-to-experiment values (C/Es) of examined CR worths with and without the interaction were 0.96 to 0.98 ± 0.03. The values of the measured CR interactions were obtained from those examined CR worth measurements as a change in target CR worth by the insertion of interaction CRs: 5.0 ± 0.6% to 10.1 ± 0.1%. The C/Es of the interaction in those cases were 0.78 to 0.93 within 5% to 27% uncertainty. Through this study, the consistency and reliability of the measurements and the calculations of the CR interaction were confirmed, and the measurements were judged useful. [ABSTRACT FROM AUTHOR]

The assessment of radiation fields in the lithium loop pipes and dump tank during the operation were performed for International Fusion Materials Irradiation Facility – DEMO-Oriented NEutron Source (IFMIF-DONES) in order to obtain the radiation dose-rate maps in the component surroundings. Variance reduction techniques such as weight window mesh (produced with the ADVANTG code) were applied to bring the statistical uncertainty down to a reasonable level. The biological dose was given in the study, and potential shielding optimization is suggested and more thoroughly evaluated. The MCNP Monte Carlo was used to simulate a gamma particle transport for radiation shielding purposes for the current Li Systems’ design. In addition, the shielding efficiency was identified for the Impurity Control System components and the dump tank. The analysis reported in this paper takes into account the radiation decay source from and activated corrosion products (ACPs), which is created by d-Li interaction. As a consequence, the radiation (resulting from ACPs and Be-7) shielding calculations have been carried out for safety considerations.

The activity and disposal options for irradiated Be12Ti were assessed for the HCPB DEMO blankets making use of a code system that enables performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. The dedicated full-scale geometry MCNP model of the 11.25 degree HCPB DEMO torus was adapted to the requirements for the coupled 3D neutron transport and activation calculations. Special attention was paid to the use in the activation calculations of the commercial materials containing technological impurities. This has a crucial effect on the results and the impurities must be accounted for in any nuclear safety analyses. The short-term activity is formed by the radionuclides produced through the activation of Be and Ti nuclei and the long-term activity is formed by the products of the neutron irradiation taking place on the impurities. A prerequisite for the disposal or the recycling of the irradiated Be12Ti is its deep detritiation; otherwise, the very high-tritium activity would fully prevent any attempt for its treatment. The most preferable is the use of the Be12Ti with the composition including less material impurities, especially uranium. There could be the option to dispose the Be12Ti intermediate-level wastes in the French repository after 1 year of cooling, assuming the detailed control of the impurities that fulfil the French authority requirements. The USA near-ground repositories could be an alternative to the European sites. The recycling of the irradiated Be12Ti must first be elaborated and approved to ensure its treatment in a safe and efficient way. The remote handling technique must be developed for the re-fabrication of the Be12Ti blocks. [ABSTRACT FROM AUTHOR]

An emerging breeding blanket that fulfills performance criteria, meets the safety requirements, and is reliable enough to meet the plant availability is a challenging issue that assumes complex studies involving numerous neutronic analyses based on the Monte Carlo simulations with MCNP code. Two different concepts are now candidates to be implemented as a driver blanket for DEMO fusion reactor: WCLL (Water-Cooled Lithium Lead) and HCPB (Helium-Cooled Pebble Bed). The current R&D work within the EUROfusion DEMO project is concentrated on a search for the new water-cooled blanket layouts: a deep upgrade of the WCLL blanket to ensure a sufficient tritium breeding capability and an elaboration of the hybrid concept coupling technological advantages of water coolant, lead neutron multiplier, and ceramic breeder. To this end, very detailed, fully heterogeneous MCNP geometry models were developed for the newest designs of the WCLL-db (WCLL-double bundle) and WLCB (Water-cooled liquid Lead Ceramic Breeder) DEMO blankets to verify the new engineering solutions. This makes rigorous calculations possible to find an optimal breeder blanket layout. The basic response, tritium breeding ratio (TBR), was assessed for both concepts, and it appeared to be TBR = 1.16 for the WCLL-db and TBR ≤ 1.13 for the WLCB DEMOs, respectively. Several geometry layouts of the WLCB breeder blanket were investigated to reach the TBR sufficient for a sustainable tritium fuel cycle. Two promising novel solutions were suggested to enhance the tritium breeding performance of the WLCB blanket and to achieve TBR ≥ 1.16: heavy water coolant and an advanced breeder ceramic. Various nuclear safety aspects of the technologies utilized in both blanket concepts are addressed in this work to facilitate engineering decisions aimed at the consolidated blanket design for the upcoming DEMO reactor. [ABSTRACT FROM AUTHOR]

One of the main challenges in the development of a plasma diagnostic and control system for DEMO is the need to cope with unprecedented radiation levels in a tokamak during long operation periods. A list of diagnostics required for plasma control has been developed during the pre-conceptual design phase. Different approaches are proposed for the integration of these diagnostics in DEMO: in equatorial and upper ports, in the divertor cassette, on the inner and outer surfaces of the vacuum vessel and in diagnostic slim cassettes, a modular approach developed for diagnostics requiring access to the plasma from several poloidal positions. According to each integration approach, diagnostics will be exposed to different radiation levels, with a considerable impact on their design. This paper provides a broad overview of the radiation environment that diagnostics in DEMO are expected to face. Using the water-cooled lithium lead blanket configuration as a reference, neutronics simulations were performed for pre-conceptual designs of in-vessel, ex-vessel and equatorial port diagnostics representative of each integration approach. Flux and nuclear load calculations are provided for several sub-systems, along with estimations of radiation streaming to the ex-vessel for alternative design configurations. The results can be used as a reference by diagnostic designers. [ABSTRACT FROM AUTHOR]

The construction and operation of the first generation of magnetically controlled nuclear fusion power plants require the development of proper physics and the engineering bases. The analysis of data, recently collected by the actual largest and most important tokamak in the world JET, that has successfully completed his second deuterium and tritium campaign in 2021 (DTE2) with a full ITER like wall main chamber, has provided an important consolidation of the ITER physics basis. Thermonuclear plasmas are highly nonlinear systems characterized by the need of numerous diagnostics to measure physical quantities to guide, through proper control schemes, external actuators. Both modelling and machine learning approaches are required to maximize the physical understanding of plasma dynamics and at the same time, engineering challenges have to be faced. Fusion experiments are indeed extremely hostile environments for plasma facing materials (PFM) and plasma-facing components (PFC), both in terms of neutron, thermal loads and mechanical stresses that the components have to face during either steady operation or off-normal events. Efforts are therefore spent by the community to reach the ultimate goal ahead: turning on the first nuclear fusion power plant, DEMO, by 2050. This editorial is dedicated at reviewing some aspects touched in recent studies developed in this dynamic, challenging project, collected by the special issue titled "New Challenges in Nuclear Fusion Reactors: From Data Analysis to Materials and Manufacturing". [ABSTRACT FROM AUTHOR]

ITER is an international collaborative effort towards the realization of fusion energy via the magnetic confinement concept. Two of the equatorial ports in the facility are dedicated to the testing of tritium breeding concepts, which is essential for the tritium self-sufficiency of future fusion reactors. The concerned Test Blanket System (TBS) consists of a Test Blanket Module (TBM) residing inside the TBM–Port Plug (TBM-PP) and its associated ancillary systems in the Tokamak facility. In this paper, the results of a full suite of nuclear analyses concerning the shielding performance of the Pipe Forest (PF) and Bioshield Plug (BP), to reflect on the evolution of their designs, are discussed. On the BP side, the design of the peripheral part has been reviewed considering the ventilation openings and butterfly doors, to assure the design compliance with the Radiation Map (RadMap) requirements for the neutron flux in the Port Cell (PC), behind the BP. On the PF side, the pipes routing and maintenance corridor door have been redesigned, by taking into account results from previously concluded nuclear analyses. The neutronics model was developed from CAD and was used to perform transport simulations in two plasma modes: on and off. For plasma-on mode, the plasma neutron field in the Port Interspace (PI) as well as behind the BP was assessed and few shielding options were explored. The responses due to decay neutrons from 17N in activated cooling water were also considered. For the plasma-off mode, the focus was shifted to further refine the ShutDown Dose Rate (SDDR) maps, which is of importance for maintenance operations that are foreseen to take place at various stages of ITER operation, in particular following the FPO-1, FPO-2, and Short operation scenarios. In addition, detailed activation analyses were carried out to provide a provisional waste classification.

Integration of safety aspects in IFMIF-DONES design is a main objective of EUROfusion and European Commission projects. IFMIF-DONES will be a radioactive facility of the first category, and stringent safety objectives must be achieved and demonstrated. A very low acceptable risk for the worker, the public and the environment is the main principle in the design phase. The progress of safety activities is performed iteratively as detailed engineering develops, taking into account the uniqueness of the facility: a high-power deuterons accelerator (125 mA, 40 MeV), a target of flowing liquid lithium, traps for activation products, a dedicated-design module for irradiated samples, a massive shielding cooled room with confinement function, and a number of conventional systems with safety functions. Several phases are developed: (i) identification of sources and materials at risk, radioactive and nonradioactive, subject to potential mobilization, (ii) failure mode analysis and effects of systems, starting at the functional level, and support with probabilistic analysis, (iii) identification of scenarios leading to unacceptable risk if unmitigated, (iv) proposal of layers of defense by means of safety-credited components and design features, (v) deterministic analysis of scenarios in support of requirements, and (vi) definition and demonstration of safety requirements charged to components.

OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. The Weight Window Mesh (WWM) function and an automatic Global Variance Reduction (GVR) method was recently developed and implemented in a developmental branch of OpenMC. This WWM function and GVR method broaden OpenMC's usage in general purposes deep penetration shielding calculations. However, the Local Variance Reduction (LVR) method, which suits the source-detector problem, is still missing in OpenMC. In this work, the Weight Window Generator (WWG) function has been developed and benchmarked for the same branch. This WWG function allows OpenMC to generate the WWM for the source-detector problem on its own. Single-material cases with varying shielding and sources were used to benchmark the WWG function and investigate how to set up the particle histories utilized in WWG-run and WWM-run. Results show that there is a maximum improvement of WWM generated by WWG. Based on the above results, instructions on determining the particle histories utilized in WWG-run and WWM-run for optimal computation efficiency are given and tested with a few multi-material cases. These benchmarks demonstrate the ability of the OpenMC WWG function and the above instructions for the source-detector problem. This developmental branch will be released and merged into the main distribution in the future.

Nuclear data is the basis of reactor physics analysis. This paper aim at studying the influence of major evaluated nuclear data libraries, CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0u, on the neutronics modelling of CEFR start-up tests. Results show these four libraries have a good performance and consistency in the modelling CEFR start-up tests. The JEFF-3.3 results exhibit only an 8 pcm keff difference with the measurement. The difference in criticality is decomposed by nuclide, which shows the large overestimation of CENDL-3.2 is mainly from the cross-section of 52Cr. Except for few cases, the calculation results are within 1σ of measurement uncertainty in control rod worth, sodium void reactivity, temperature reactivity, and subassembly swap reactivity. In the evaluation of axial and radial reaction distribution, there are about 65% of relative errors that are less than 5% and 82% of relative errors that are less than 10%.

High-fidelity nuclear data libraries and neutronics simulation tools are essential for the development of fast reactors. The IAEA coordinated research project on “Neutronics Benchmark of CEFR Start-Up Tests” offers valuable data for the qualification of nuclear data libraries and neutronics codes. This paper focuses on the verification and validation of the CEFR start-up modelling using OpenMC Monte-Carlo code against the experimental measurements. The OpenMC simulation results agree well with the measurements in criticality, control rod worth, sodium void reactivity, temperature reactivity, subassembly swap reactivity, and reaction distribution. In feedback coefficient evaluations, an additional state method shows high consistency with lower uncertainty. Among 122 relative errors in the benchmark of the distribution of nuclear reaction, 104 errors are less than 10% and 84 errors are less than 5%. The results demonstrate the high reliability of OpenMC for its application in fast reactor simulations. In the companion paper, the influence of cross-section libraries is investigated using neutronics modelling in this paper.

ITER is an international collaborative effort towards the realization of fusion energy via the magnetic confinement concept. Two of the equatorial ports in the facility are dedicated to the testing of tritium breeding concepts, which is essential for the tritium self-sufficiency of future fusion reactors. The concerned Test Blanket System (TBS) consists of a Test Blanket Module (TBM) residing inside the TBM--Port Plug (TBM-PP) and its associated ancillary systems in the Tokamak facility. In this paper, the results of a full suite of nuclear analyses concerning the shielding performance of the Pipe Forest (PF) and Bioshield Plug (BP), to reflect on the evolution of their designs, are discussed. On the BP side, the design of the peripheral part has been reviewed considering the ventilation openings and butterfly doors, to assure the design compliance with the Radiation Map (RadMap) requirements for the neutron flux in the Port Cell (PC), behind the BP. On the PF side, the pipes routing and maintenance corridor door have been redesigned, by taking into account results from previously concluded nuclear analyses. The neutronics model was developed from CAD and was used to perform transport simulations in two plasma modes: on and off. For plasma-on mode, the plasma neutron field in the Port Interspace (PI) as well as behind the BP was assessed and few shielding options were explored. The responses due to decay neutrons from 17N in activated cooling water were also considered. For the plasma-off mode, the focus was shifted to further refine the ShutDown Dose Rate (SDDR) maps, which is of importance for maintenance operations that are foreseen to take place at various stages of ITER operation, in particular following the FPO-1, FPO-2, and Short operation scenarios. In addition, detailed activation analyses were carried out to provide a provisional waste classification. [ABSTRACT FROM AUTHOR]

The Plasma Facing Components (PFCs) of the divertor target contribute to the fundamental functions of heat removal and particle exhaust during fusion operation, being subjected to a very hostile and complex loading environment characterized by intense particles bombardment, high heat fluxes (HHF), varying stresses loads and a significant neutron irradiation. The development of a well-designed divertor target, which represents a crucial step in the realization of DEMO, needs the assessment of all these loads as accurately as possible, to provide pivotal data and indications for the design and structural performance prediction of the PFCs. In a particular way, this study is fully devoted to the comprehension of the distributions on the divertor target of the main nuclear loads due to neutron irradiation, performed for the first time using an extremely detailed approach. This work has been carried-out considering the latest configuration of the DEMO reactor, including the updated design of the divertor and ITER-Like PFCs geometry, varying the blanket layout (Water Cooled Lithium Lead—WCLL and Helium Cooled Pebble Bed—HCPB), thus evaluating the impact of the different blanket concept on the above-mentioned distributions. Neutronics analyses have been performed with MCNP5 Monte Carlo code and JEFF3.3 nuclear data libraries. 3D DEMO MCNP models have been created, focusing in particular on a thorough representation of the divertor and PFCs, allowing for the assessment of the distributions of the main nuclear loads: radiation damage (dpa/FPY), He-production rate (appm/FPY) and nuclear heating density (W/cm3) and total nuclear power deposition (MW). These results are presented by means of 2D maps and plots for each PFCs sub-component both for WCLL and HCPB blanket case: W-monoblocks, Cu-interlayers\CuCrZr-pipe and PFC-CB (Cassette Body) supports made of Eurofer steel. [ABSTRACT FROM AUTHOR]

During the commissioning and start-up phases of IFMIF-DONES, the 40 MeV deuteron beam will be stopped in a conical beam stopper made of copper, giving rise to intense neutron fields. Accurate coupled deuteron-neutron transport simulations are required in order to prevent the potential risks to the workers and equipment involved. However, the experimental data concerning the neutron emission under these circumstances is scarce and limited to a particular range of energies. Different deuteron nuclear data libraries such as JENDL-5 and special files of TENDL-2021 exhibit some differences on the spectrum of the emitted neutrons. The differences are present in the high energy tails, beyond the available experimental data. Calculations of the prompt neutron fields have been performed employing both deuteron data libraries, as well as employing an estimation of the neutron energy-angle distribution derived from experimental data at different irradiation energies. Even though the high energy tails represent a small part of the flux, they have a major impact on the dose rates. The disparity of the results evinces large uncertainties in the IFMIF-DONES beam dump design process. Further experiments are required in order to resolve the differences among nuclear data libraries and reduce the uncertainty.

In water-cooled nuclear reactors, the issue of neutron-activated products transport along the primary heat transfer system (PHTS) is very demanding, as it is a coupled neutronic/fluid-dynamic problem requiring a challenging balance between accuracy and reasonable computational time. This work addresses the transport of water activation products in large hydraulic circuits. Regarding the nuclear calculations, the assessment of the production rates of the radioisotope concentrations has been performed by Monte Carlo analyses adopting the MCNP5.1.6 code, while for the transportation calculations, an innovative method has been expressly developed. It foresees a one-dimensional nodalization, in a MATLAB-Simulink environment, of the hydraulic circuit considered with a computational fluid-dynamic (CFD) characterization (by ANSYS CFX code) of the nodes under neutron flux, that is the components where radioisotopes are formed, and the highest gradients of concentration are present. The method was compared with one-dimensional models not supported by fluid-dynamic analysis. The results of this comparison showed that in cases involving fairly complicated geometries and radioisotopes with a small half-life, CFD analyses are necessary to achieve adequate accuracy. The procedure was applied to very large and rather complex hydraulic circuits like the divertor PHTSs of DEMO fusion reactor to obtain the concentrations of the activation products of the water constituents ( ^16 N, ^17 N, ^19 O, ^14 C, ^41 Ar) along such systems.

ITER is of key importance in the European fusion roadmap as it aims to prove the scientific and technological feasibility of fusion as a future energy source. The EUROfusion consortium of labs within Europe is contributing to the preparation of ITER scientific exploitation and operation and aspires to exploit ITER outcomes in view of DEMO. The paper provides an overview of the major progress obtained recently, carried out in the frame of the new (initiated in 2021) EUROfusion work-package called ‘ Pr eparation of I TER O peration’ (PrIO). The overview paper is directly supported by the eleven EUROfusion PrIO contributions given at the 29th Fusion Energy Conference (16–21 October 2023) London, UK [ https://www.iaea.org/events/fec2023 ]. The paper covers the following topics: (i) development and validation of tools in support to ITER operation (plasma breakdown/burn-through with evolving plasma volume, new infra-red synthetic diagnostic for off-line analysis and wall monitoring using Artificial Intelligence techniques, synthetic diagnostics development, development and exploitation of multi-machine databases); (ii) R&D for the radio-frequency ITER neutral beam sources leading to long duration of negative deuterium/hydrogen ions current extraction at ELISE and participation in the neutral beam test facility with progress on the ITER source SPIDER, and, the commissioning of the 1 MV high voltage accelerator (MITICA) with lessons learned for ITER; (iii) validation of neutronic tools for ITER nuclear operation following the second JET deuterium–tritium experimental campaigns carried out in 2021 and in 2023 (neutron streaming and shutdown dose rate calculation, water activation and activated corrosion products with advanced fluid dynamic simulation; irradiation of several materials under 14.1 MeV neutron flux etc).

This contribution presents neutron transport studies for the 5-period helical-axis advanced stellarator stellarator using the Serpent2 code. These studies utilize a parametric geometry model, enabling scans in neutronics modeling by varying the thickness of the reactor layers. For example, the tritium breeding ratio (TBR) can be determined by exploring various blanket material options and thicknesses to identify the threshold configuration that meets the TBR design criterion of 1.15. We found out that with the helium-cooled pebble ped candidate option, the TBR criterion is met with a breeding zone thickness of 26 cm, while with the dual-coolant lithium lead the threshold is exceeded at a thickness of 46 cm. Furthermore, the geometry includes non-planar field coils, allowing to study the fast neutron flux in these superconducting coils with a technological limit of $1\times10^{9}\,1/\mathrm{cm}^{2}\,\mathrm{s}$ . It is shown that the neutron fast flux is not constant at the coils, necessitating a neutron transport simulation to determine the distribution of the fast-flux at the coils. We show that the peak fast flux can be more than a factor of 2 higher than the average flux, and that the peak flux location rotates helically.

This work presents the results following the first irradiation of ITER materials samples in a tokamak D–T plasma environment operating at significant fusion power. The materials exposed to this nuclear environment at the Joint European Torus during the DTE2 experimental campaign that took place in 2021 include representative ITER samples from various components such as poloidal field coil jacket samples, toroidal field coil radial closure plate steels, EUROFER 97 steel, W and CuCrZr materials from the divertor, Inconel-718 and 316L stainless steel for blanket modules, as well as vacuum vessel forging samples. The experimental results discussed include high-resolution gamma spectrometry measurements and analysis conducted with the post-irradiated samples, of which there were 68 in total. These samples were exposed through different experimental campaigns, including deuterium, deuterium–tritium and tritium phases. Diagnostics that supported the analysis included 25 dosimetry foil-based neutron diagnostics and two ‘VERDI’ neutron spectrometry diagnostics. A further 12 samples for positron annihilation spectroscopy were also irradiated. The irradiation of all these samples took place in a long-term irradiation assembly located near the JET vacuum vessel. The post-irradiation analysis of the ITER material samples has yielded valuable insights into their material activation levels and radiation fields. Comparative assessments between experimental measurements and comprehensive neutronics simulations have demonstrated a significant level of agreement in this work, while also revealing some discrepancies in specific material instances. The data and interpretation from this work not only serve as a robust experimental foundation for enhancing the precision and predictability of neutronics simulation approaches for ITER and next-step devices but also present some opportunities for the refinement of simulation methodologies. In light of these findings, a series of recommendations have been proposed, aimed at improving confidence in nuclear predictions associated with materials that have been exposed to fusion nuclear environments and advancing understanding in this important domain.

There is widespread use of nuclear radiation for medical imagery and treatments. Worldwide, almost 40 million treatments are performed per year. There are also applications of radiation sources in other commercial fields, e.g., for weld inspection or steelmaking processes, in consumer products, in the food industry, and in agriculture. The large number of neutrons generated in a fusion reactor such as DEMO could potentially contribute to the production of the required radioactive isotopes. The associated commercial value of these isotopes could mitigate the capital investments and operating costs of a large fusion plant. The potential of producing various radioactive isotopes was studied from material pieces arranged inside a DEMO equatorial port plug. In this location, they are exposed to an intensive neutron spectrum suitable for a high isotope production rate. For this purpose, the full 3D geometry of one DEMO toroidal sector with an irradiation chamber in the equatorial port plug was modeled with an MCNP code to perform neutron transport simulations. Subsequent activation calculations provide detailed information on the quality and composition of the produced radioactive isotopes. The technical feasibility and the commercial potential of the production of various isotopes in the DEMO port are reported.

It is of practical significance to study the multi-physical processes of solid state nuclear systems for device design, safety analysis, and operation guidance. This system generally includes three multi-physical processes: neutronics, heat transfer, and thermoelasticity. In order to analyze the multi-physical field behavior of solid state nuclear system, it is necessary to analyze the laws of neutron flux, temperature, stress, and other physical fields in the system. Aiming at this scientific goal, this paper has carried out three aspects of work: (1) Based on Galerkin's finite element theory, the governing equations of neutronics, heat transfer, and thermoelasticity have been established; (2) a neutronics-thermal-mechanical multi-physical finite element analysis code was developed and verified based on benchmark examples and third-party software for multi-physical processes; (3) for a solid state nuclear system with a typical heat pipe cooled reactor configuration, based on the analysis code developed in this work, the neutronics-thermal-mechanical coupling analysis was carried out, and the physical field laws such as neutron flux, temperature, stress, etc., of the device under the steady-state operating conditions were obtained; and (4) finally, the calculation results are discussed and analyzed, and the focus and direction of the next work are clarified. [ABSTRACT FROM AUTHOR]

We analyse a posteriori error estimates for the discretization of the neutron diffusion equations with mixed finite elements. We provide guaranteed and locally efficient estimators on a base block equation, the one-group neutron diffusion equation. We pay particular attention to AMR strategies on Cartesian meshes, since such structures are common for nuclear reactor core applications. We exhibit a robust marker strategy for this specific constraint, the direction marker strategy. [ABSTRACT FROM AUTHOR]

Integration of safety aspects in IFMIF-DONES design is a main objective of EUROfusion and European Commission projects. IFMIF-DONES will be a radioactive facility of the first category, and stringent safety objectives must be achieved and demonstrated. A very low acceptable risk for the worker, the public and the environment is the main principle in the design phase. The progress of safety activities is performed iteratively as detailed engineering develops, taking into account the uniqueness of the facility: a high-power deuterons accelerator (125 mA, 40 MeV), a target of flowing liquid lithium, traps for activation products, a dedicated-design module for irradiated samples, a massive shielding cooled room with confinement function, and a number of conventional systems with safety functions. Several phases are developed: (i) identification of sources and materials at risk, radioactive and nonradioactive, subject to potential mobilization, (ii) failure mode analysis and effects of systems, starting at the functional level, and support with probabilistic analysis, (iii) identification of scenarios leading to unacceptable risk if unmitigated, (iv) proposal of layers of defense by means of safety-credited components and design features, (v) deterministic analysis of scenarios in support of requirements, and (vi) definition and demonstration of safety requirements charged to components. [ABSTRACT FROM AUTHOR]

Liquid plasma facing walls allow for increased neutron-wall loading expanding the design space of fusion power-plants and experimental devices towards compact high-field reactors. This study presents the design of a compact radial build blanket for fusion devices composed of variable quantities of Lead (Pb) and Lithium-Lithium Hydride (Li-LiH). A tank-like cylindrical neutronic model of the early design of the stellarator reactor proposed by Renaissance Fusion is implemented in OpenMC (neutron transport and dose rate analyses). The reactor's radial build composition and blanket layer thicknesses are varied to fulfill the requirements on tritium breeding ratio (TBR), nuclear heat extraction, radiation shielding (for the coils, internal structures and external environment) for a stellarator-based power-plant. The analyses suggest that a radial build lower than a meter thick between the plasma and coils would be sufficient to allow for a TBR ∼ 1.60, an energy multiplication factor of ∼ 1.07, to capture ≥ 90% of the nuclear heat, limit the neutron fluence at the coils below 1019 n/cm2, and limit the structural damage on the liquid metal vessel and magnet structure. In particular, a blanket composed of 32 cm of Pb and Li-LiH, 54 cm of a heavy metal hydride such as vanadium hydride (VH 2), along with a 1.3 m of concrete bioshield, would minimize the radial build of the stellarator reactor while fulfilling tritium breeding, shielding and heat extraction requirements. [ABSTRACT FROM AUTHOR]

• This study focuses on the neutronic and burnup assessment of the PWR-based ACP-100 SMR core and proposes a core model using differently enriched fuel assemblies. • Burnup analysis was carried out over 1570 EFPDs, evaluating several neutronic parameters (beta effective, reactivity coefficients) as well as fuel cycle characteristics (cycle burnup, discharge burnup and cycle length). • A strategic combination of soluble boron, IFBAs and control rods were employed to achieve exact criticality (unit k eff) and near unit radial power peaking factor in the proposed core. • The suggested core successfully remained critical for over 1.5 years and reached a discharge burnup of 23.37 GWd/t for a two-batch refueling scheme. This study focuses on the neutronic and burnup assessment of the PWR-based Chinese ACP-100 SMR core. Four different core models with varying fuel compositions and assembly configurations were analyzed using the Monte Carlo code SERPENT 2. Three of these models shared uniformly enriched fuel assemblies of 3 wt.%, 4 wt.%, and 4.45 wt.% 235U respectively, while the fourth model (the proposed design for the ACP-100 in this study) featured a mix of differently enriched fuel assemblies along with the inclusion of burnable absorbers. The burnup analysis of the cores spanned 1570 EFPDs, evaluating several neutronic parameters (beta effective, reactivity coefficients) as well as fuel cycle characteristics (cycle burnup, discharge burnup and cycle length). Changes in the isotopic concentration of fissile, fertile, and poison materials were observed and the assembly-wise linear power distribution, neutron flux spectra, and radial power peaking factors of the models were determined. In the mixed core, a combination of borated water, IFBAs and control rods were employed to achieve exact criticality (unit k eff) and uniform power distribution (unit radial power peaking factor). Insertion of control rods into 41 of the 57 assemblies, in conjunction with IFBAs placed strategically, attained a near unit radial power peaking factor (PPF) of 1.0586. Finally, the combination of IFBAs, sol-bor of 936 ppm concentration, and CRs with rod worth of 414.33 pcm per assembly brought down the k eff to 1.000, but increased the PPF to 1.117. While the achievement of uniform power distribution came at the cost of fuel burnup, the proposed mixed core successfully remained critical for over 1.5 years and reached a discharge burnup of 23.37 GWd/t for a two-batch refueling scheme. [ABSTRACT FROM AUTHOR]

In nuclear reactor safety assessment studies, void formation in the coolant, both in nominal operation and in accidental scenarios, must be considered since bubble nucleation can lead to fast reactivity changes, thus putting a strain on the reactor control. The evaluation of such a phenomenon is not trivial due to the non-linearities related to the competing phenomena (e.g. neutron absorption and scattering) and the spatial effects. This work aims to analyse the impact of void formation on the multiplication factor in the Pavia (Italy) TRIGA Mark II reactor, focusing on spatial effects. A model of the TRIGA has been developed using Serpent Monte Carlo code and validated against experimental measurements conducted at the Pavia reactor. Two approaches have been adopted. The first consists of directly evaluating the multiplication factor of several configurations, each featured by a different water density to mimic a homogeneous void formation. The second approach consists of a perturbative procedure, i.e., a first-order sensitivity analysis, which allows the gathering of more information on the aforementioned competing phenomena. Both cases subdivide the core into several radial and axial regions to recover the spatial effect. The two approaches appear to be complementary in the information they provide. The results show that the void coefficient is negative in the core and strongly dependent on both position and void fraction. Moreover, the results show the strong influence of the fuel elements type (101 and 103), their location, and the experimental methods adopted on the void coefficient. • Monte Carlo analysis of void feedback coefficient is performed. • The sensitivity tool has been exploited. • Experimental data have been used to validate the numerical model. • Experimental procedures for assessing void coefficient have been analysed. [ABSTRACT FROM AUTHOR]

• Fuel pattern optimization of a WWER1000 in order to extend reloading cycle length is done. • The Multi-Verse Optimization algorithm is applied and its performance is demonstrated by challenging functions. • Multi objective fitness functions in LPO are studied and results compared by several algorithms. • Neutronics and thermal-hydraulics are carried out by Parcs and Cobra-EN and effective parameters of the proposed core is reported. In the fuel management of nuclear reactors, which has a direct impact on power generation costs and environmental factors, it is essential to obtain an optimal arrangement of assemblies and core parameters. This complex task relies on both neutronics and thermal hydraulics. This research solves this by employing PARCS and COBRA-EN codes, alongside Multi-Verse Optimization algorithm (MVO) as a novel algorithm. Findings demonstrate the MVO algorithm has reliability and favorable convergence rate. Moreover, integrating the nodal expansion technique with MVO notably decreases the computational time for optimization in the loading pattern optimizations (LPOs). The optimized loading pattern aligns well with the findings in the literature. Also, using multi objective fitness function leads to improving core flatness, cycle length and safety parameters. [ABSTRACT FROM AUTHOR]

This paper presents the shielding analyses, which will be provided for supporting the design of the Test Cell Lithium System Interface Cell (TLIC) and safety assessment. The study considers the radiation decay source from IFMIF-DONES operation with beam-on. Radiation shielding simulations have been performed for the safety concerns. Detailed simulation geometry was created from CAD drawing for the TLIC. Neutron transport calculation was performed for all TLIC components concerned. The particle flux, biological dose and absorbed dose maps are provided in the study. A series of neutron transport are performed using the MCNP Monte Carlo McDeLicious neutron source code coupled with FISPACT for inventory calculations for TLIC design. • Shielding analysis for supporting the design of the Test Cell Lithium System Interface Cell of DONES facility. • Calculations of the absorbed dose rates in silicon maps. • Absorbed dose rates for each TLIC component for gamma photons. • The precise activities, decay heat and contact dose rates for each TLIC component were computed. • The most active radionuclides in irradiated insulation material were short living Si-31 and Al-28. [ABSTRACT FROM AUTHOR]

Zero-power reactors stand as indispensable tools for shaping the future of the nuclear industry. Addressing safety concerns, advancing reactor technology, mitigating proliferation risks, fostering education, and promoting economic viability, these reactors hold the key to unlocking the full potential of nuclear energy in a sustainable and responsible manner. As the world seeks cleaner and more efficient energy solutions, the importance of zero-power reactors cannot be overstated in charting the course for the nuclear industry's future. The paper presents a short history of the various zero-power/zero-energy experimental facilities constructed and used worldwide. Many of the names seemed to be lost to history and archives, which means that all the experimental data carried in the those facilities is lost as well. However, re-introducing the various names can spark an interest in "digging up" and revisit experiments of the past, which can help in the design of experiments and new systems in the future. It is clear that a new experimental facility should be built. The next frontier in zero-power reactor design envisions a design for versatility, this future concept addresses diverse energy needs while contributing to a sustainable and responsible nuclear energy landscape. This was demonstrated in the framework of the Zero-power Experimental PHYsics Reactor design proposed by French Atomic Energy Commission. • A historical overview of various zero-power facilities constructed and operated worldwide. • Emphasize the importance and various utilization of zero-power reactors. • Focus on future flexible system based on representativity and spectral flexibility. • Showcase the Zero-power Experimental PHYsics Reactor and an ambitious system to tackle future needs in reactor physcics. [ABSTRACT FROM AUTHOR]

The EU DEMO fusion reactor, which is part of the EUROfusion PPPT program and the successor to ITER, aims to facilitate the transition of fusion from the science laboratory to industrial use. Its key objective is net energy production. Its critical component, the divertor, manages power exhaust, ash, and helium removal. The key elements of the divertor include Plasma Facing Components (PFCs), shielding liners, reflector plates, divertor cassettes, and a vacuum vessel locking system. It is imperative that the shutdown dose rates are assessed to ensure workers' radiation protection during maintenance, plan the disposal of radioactive material, and ensure the overall safety of the plant. The shutdown dose rate tool employed in the present study is R2Smesh, which couples MCNP for radiation transport and FISPACT for neutron activation to calculate neutron-induced shutdown dose rates, utilizing an advanced, semi-heterogeneous EU DEMO divertor model, presenting a challenging task in neutron transport calculations. • R2Smesh was utilized for the evaluation of SDDR. • The analysis focused on SDDR values within a specific layer in the divertor. • 10 years after the shutdown, SDDR exhibits a significant reduction in Eurofer and Tungsten. • SDDR exhibits a significant reduction in EUROFER and Tungsten 10 years after the shutdown. • A four-order of magnitude drop in SDDR was noted in wishbone pins during a 100-year period. [ABSTRACT FROM AUTHOR]

Various Machine Learning (ML) techniques have seen recent and growing interest in the creation of surrogate neutronics models as a potential way to avoid the computational expenses associated with conventional high-fidelity modeling. Artificial Neural Networks (ANNs) have been shown to be particularly useful for single-assembly predictions involving pin-wise power distributions and multiplication factors. In this paper, the LatticeNet neural network framework is investigated as a method to predict Doppler and moderator temperature coefficients for Pressurized Water Reactor (PWR) fuel assemblies, as well as differential boron worth. This approach uses the built-in tools developed alongside LatticeNet to construct two fully-connected network architectures capable of predicting k -eigenvalues from two inputs such as fuel enrichment and temperature when trained with data generated with CASMO-4E. A single network taking in all study parameters as inputs was then used to predict k -eigenvalues for fuel temperature, moderator temperature, and boron perturbation cases. The calculated temperature coefficients and differential boron worth values were compared with a bank of reference values to validate the efficacy of this method. Overall, differences in k -eigenvalues were within 0.017% in the worst case. The temperature coefficients saw mean errors of 1.85% and 1.69% for the two-parameter networks, respectively. The all-parameter network was then shown to predict 1100 points in 236 ms compared to the 4.95 min CASMO-4E took to generate them. Additionally, the differential boron worth achieved the lowest mean error of 0.30%; each of these values were within our acceptance criteria. [ABSTRACT FROM AUTHOR]

FeCrAl alloys have excellent oxidation resistance and exhibit slower reaction kinetics with steam. Kanthal APMT is an advanced powder metallurgical FeCrAl alloy that shows promise as accident tolerant fuel cladding material. The neutronic properties of APMT as a fuel cladding material were evaluated in a Light Water Reactor (LWR) core and compared with zirconium-based alloys through the help of Monte Carlo N-Particle Transport (MCNP) code. The results show that there is neutronic penalty associated with the use of APMT as a cladding component which can be compensated by either modification in the fuel-cladding geometry or enrichment in the fuel. Creep tests were performed at 823–973 K and at stresses ranging from 100 to 300 MPa. The creep rupture data was found to obey Larson-Miller and Monkman-Grant relationships. The hardness of the crept specimens, measured after the creep test, shows significant reduction of mechanical strength near the fractured tip. Two types of second phase particles, YFeCr-rich and YAlO-rich, were observed in the fractured surface of the crept specimen through energy dispersive spectroscopy. • Neutronic penalty associated with APMT cladding can be compensated by fuel-cladding geometry modification or fuel enrichment. • Tensile creep tests were conducted at 823-723K and 100–300 MPa. • Larson-Miller and Monkman-Grant analyses showed alignment with other Al-containing ferritic alloys. • Two types of precipitates: YFeCr-rich and YAlO-rich observed in the fracture region of crept specimen. [ABSTRACT FROM AUTHOR]

• The BYU Molten Salt Micro-Reactor was geometrically optimized using machine learning and CFD. • Millions of data points populated point clouds from only 35 CFD model variations of the base design. • Quantity of moderator, fuel salt, structural material, plus the ML regressor and data quantities were varied. • The resulting optimal design was produced using less than 10% of the potential training data, showing significant improvement with limited required data acquisition. • The optimal design has a 220 K reduction in max. temp., 85 K in ave. temp., and 49 K in st. dev. of temp. Nuclear reactors are vital to the global economy due to the large amounts of baseload power they can provide with minimal intermittence. However, these reactors are extremely complicated to design and build. Therefore, any improvement in the design process and timeline of nuclear reactor development would have a significant impact on the reactor design community and the world. In recent years, machine learning (ML) algorithms have gained popularity in handling difficult problems that previously required too much data or computational power to effectively solve. These include natural language processing, computer vision, and advanced process optimization. ML algorithms present a potentially powerful tool in the realm of reactor design and have already been implemented in some specific reactor design and operational optimization applications by researchers around the world. While these previous uses of ML algorithms are effective, they are very specific and a more generalized, and easily applied framework could potentially be used to accelerate the design of Gen IV reactors, reducing the workload of reactor designers and researchers. The development of such a versatile framework for reactor optimization using ML, CFD, and neutronic codes, and the validation of such, was the purpose of this work. The baseline reactor to be optimized was the BYU Molten Salt Micro-Reactor (Larsen et al., 2023). To establish this ML optimization, 35 training runs were completed using STARCCM + CFD software, and OpenMC neutronic modeling to generate data. The 3-dimensional cartesian data were extracted for each run, providing up to several million data points per training run. Moderator thickness, salt thickness, and moderator housing (plate) thickness were variables in the geometric optimization, and regressor type, data quantity per run, and training run quantity were variables in the model optimization. The model was first optimized by weighing model variables against train and test accuracy. This optimal model was then given 4096 geometric reactor iterations, and a temperature profile was predicted for each. The lowest-temperature option was chosen as lower temperature results in longer structural material life and less required secondary cooling expense, as well as more favorable neutronics. This result was then validated in CFD. The optimal reactor design surpassed all training runs by at least 220 K in maximum temperature, 85 K in average temperature, and 49 K in standard deviation of temperature. These represent significant improvement over all training designs and the baseline design. [ABSTRACT FROM AUTHOR]