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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]

The Versatile Test Reactor (VTR) is a sodium-cooled fast reactor designed to accelerate the design and approval of new nuclear material and reactor concepts by providing a high neutron fast flux environment on U.S. soil. To ensure that the reactor simultaneously achieves the target irradiation environment while maintaining sufficient margin to safety limits, supporting design analysis of the VTR has been performed using MCNP and TRACE. High-fidelity MCNP calculations have been performed that confirm design parameters, such as control rod worth and neutron and photon flux distributions, and provide needed reactivity coefficients for TRACE analyses. The MCNP simulations additionally provide fuel rod power profiles of interest to fuel performance designers and provide an excellent model for experimental cartridge design within the VTR core. TRACE simulations of several postulated transients, such as station blackout, loss of heat sink, and transient overpower, have been performed (results included here are limited to the transient overpower), and the obtained results confirm the robust safety behavior of the VTR. The TRACE simulations provide a valuable confirmatory transient analysis capability using a U.S. Nuclear Regulatory Commission–developed safety analysis tool incorporating inputs from the high-fidelity neutronic simulations performed with MCNP. Taken together, the confirmatory analysis capability provided by MCNP and TRACE serves to further strengthen the understanding of and confidence in the VTR's performance. [ABSTRACT FROM AUTHOR]

In the prototype fast breeder reactor Monju, fuel reactivity worth was measured at six positions as the reactivity corresponding to the differences of critical control rod positions between cores with and without a dummy fuel subassembly. In this paper, the measurements are evaluated in detail, and their reliability and usefulness as validation data for fast reactor neutronics design methodologies are examined through a comparison with calculations. Calculated-to-experiment values and their uncertainty of fuel reactivity worth were 0.97 to 1.02 and 4% to 6%. Through this study, the consistency and reliability of the measurements and calculations were confirmed. [ABSTRACT FROM AUTHOR]

Abstract The Megawatt Implementation of a NuclEar ReActor using Low-enrichment uranium (MINERAL) is designed to deliver 2 MW(electric) of steady-state electricity to a colony established on the surface of Mars with a minimum lifetime of 10 years. The main challenge associated with a low-enrichment uranium fission surface power system is reducing the total mass, which will be higher than that of an equivalent high-enrichment uranium system. Optimizing the mass of the system is crucial to limit the amount of Earth-Mars cargo needed to deploy a MINERAL unit. The use of yttrium hydride as a moderator has shown promise in reducing the overall mass of the reactor. An in-house Python framework evaluates the neutronic, thermal-hydraulic, and heat rejection performance throughout the design process. The final design iteration uses a CO2 Brayton cycle cooled by a passive heat rejection system consisting of six panels with a total surface area of 4752 m2. The cylindrical core is fueled with low-enrichment uranium monocarbide with 0.83 wt% of pure 157Gd moderated with yttrium hydride and surrounded by a beryllium oxide reflector. The reactivity is controlled by ten control drums and a central control rod, which provide enough margin to operate the reactor and ensure its subcriticality in case of a submersion accident. The mass of the core with the reflector, reactivity control system, and shield is 7.2 tonnes. [ABSTRACT FROM AUTHOR]

The helium-cooled pebble bed (HCPB) breeding blanket (BB) is one of the two driver-blanket candidates for the European DEMO fusion reactor. Recent design activities were focused, among other objectives, on the achievement of an efficient shielding system to adequately protect the vacuum vessel (VV) and toroidal field coils (TFCs). Several shielding options have been studied in terms of architecture (e.g., in-BB shield and ex-BB shield) and materials (e.g., B4C, WC, WB, YHx, and ZrHx). In this study, the B4C material was selected as the most attractive option considering not only shielding performance but also availability, industrialization, experience, and cost factors. Subsequently, we performed a parametric study by implementing different thicknesses of a B4C external shield and reporting information of its effect on shielding performance, structural behavior, swelling and tritium breeding. Furthermore, a detailed structure for the VV was developed considering an internal layered configuration comprising steels/water with different boron contents. Corresponding shielding analyses were conducted regarding influence on neutron attenuation when implementing such a VV structure for both the baseline consolidated design of the HCPB and one of the previously developed and improved BSS configurations. The most critical responses (neutron flux and dpa) were fully established only using 10 cm B4C and an improved VV configuration. [ABSTRACT FROM AUTHOR]

BackgroundHeavy water moderated molten salt reactor (HWMSR) adopts heavy water as moderator and heavy metal elements dissolved in fluoride salt as the fuel, which has high neutron economy. But large temperature difference of fuel salt outlet will lead to the thermal fatigue of piping components at the top of the core.PurposeThis paper aims to optimize the core design of HWMSR to minimize the fuel salt outlet temperature.MethodsBy using the developed codes of neutron-thermal hydraulic coupling calculation and core critical search calculation, the power density distribution, outlet temperature distribution, initial 233U loading and breading ratio (BR) were analyzed for the cores with different size of molten salt channels.ResultsThe calculation results demonstrate that increasing the radius of molten salt channels in the inner core zone (correspondingly decreasing the radius of molten salt channels in the outer core zone) will decrease the peak of power density and the maximum outlet temperature of molten salt while the influence on BR and 233U initial loading.ConclusionsThis study provides a valuable foundation for core optimization of HWMSR.

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.

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.

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.

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”.

Our aim is to study the neutronic behaviour of potential accident-tolerant fuel (ATF) claddings in a pressurised water reactor under normal operations. This work compares ATF silicon carbide composite (SiC/SiC) cladding to conventional ZIRLOTM cladding in APR-1400. Additionally, a "sandwich" cladding design developed by the CEA is used for SiC/SiC. The design structure includes a liner in between two layers of the composite to ensure leak tightness. The two proposed liners are Niobium (Nb) and Tantalum (Ta). Serpent 2, a Monte Carlo reactor physics lattice code, is employed to model both cladding materials in APR-1400 at three different levels: pin cell, fuel assembly, and core. The criticality, neutron spectrum, actinide inventory, and power distribution as a function of burnup are investigated. The simulations show that SiC/SiC with the Nb liner displays a far superior performance than the Ta liner across all examined characteristics. Ta leads to a harder neutron spectrum and increased Pu-239 content throughout the cycle, while Nb presents negligible effects. In fact, SiC/SiC with the Nb liner performs very similarly to ZIRLOTM at all model levels. The results indicate that, in terms of neutronics, the adoption of the SiC/SiC composite would entail little to no changes to current APR-1400 operations. [ABSTRACT FROM AUTHOR]

The operation of many nuclear pressurized water reactors is being extended beyond their design lifetime limit. From the perspective of possible further lifetime extension, safety requirements are a priority. Therefore, the quantification of the neutron irradiation embrittlement of the reactor pressure vessel (RPV) is an important issue, as this is a guiding parameter that influences the reactor lifetime. In this context, the Institut de Radioprotection et de Sûreté Nucléaire developed a calculation scheme for the analysis of RPV aging under neutron irradiation, named VACS (vessel aging calculation scheme). VACS couples a deterministic approach (CASMO5 and SIMULATE5) to evaluate the full-core fission neutron source term and a Monte Carlo modeling (MCNP6) approach to model the neutron attenuation from the core to sites of interest (RPV, surveillance capsules, etc.). To ensure the reliability of aging predictions, this paper describes a detailed analysis of the neutron H.B. Robinson-2 reactor pressure vessel dosimetry benchmark. The results indicate that VACS shows satisfactory accuracy when the ENDF-B/VII.1 or JEFF-3.3 nuclear data libraries are used in the attenuation calculation. However, the use of ENDF-B/VIII.0 leads to significantly worse results. [ABSTRACT FROM AUTHOR]

International Fusion Materials Irradiation Facility–DEMO-Oriented NEutron Source (IFMIF-DONES) is an accelerator based on deuterium–lithium (D-Li) neutron source designed to qualify materials at the irradiation conditions of the reactor at the DEMO nuclear fusion power station. This article presents the shut-down dose rate (SDDR) calculations performed with AutomateD VAriaNce reducTion Generator (ADVANTG) and Monte Carlo N-Particle (MCNP) codes, which were used for the variance reduction and particle transport accordingly. Application of such coupled computational method led to the SDDR map production, where the obtained IFMIF-DONES estimates verify the access cell (AC) containment building shielding after the relocation of irradiated high flux test module with target assembly. In addition, this article provides dose rate distributions for the safe operation of the device for the entire AC surroundings of DEMO-Oriented NEutron Source (DONES) building. SDDR calculations were performed using the R2Smesh code system. The dose rates were determined, where 1 Sv/h close to the source location, 1 mSv/h at the farthest wall, and 10–100- $\mu $ Sv/h leakage through the walls at certain positions at AC were identified, and dose rate maps were produced. [ABSTRACT FROM AUTHOR]

The affordable, robust, compact (ARC) reactor is a tokamak fusion reactor concept currently under development by Commonwealth Fusion Systems and Massachusetts Institute of Technology. There are three important neutronics considerations for the operation of the ARC reactor: (1) breeding of enough tritium in the blanket to sustain the Dâ€"T reaction in the plasma; (2) ensuring low fluence on the superconducting toroidal field coils; and (3) assessing neutron volumetric heating in structural components. This work aims to perform a validation of the neutronics analysis approach by code-to-code comparison. State-of-the-art software stacks are employed for the neutronics analysis of the ARC reactor, and a computer-aided design (CAD) model is used directly for Monte Carlo (MC) neutron transport calculations. Three software stacks, Attila-MCNP, OpenMC-DAGMC, and Shift-DAGMC, are used to perform neutronic analyses of a 90° sector CAD model of the ARC reactor. Results show that the flux tallies calculated by the three software stacks are very close. Also, the volumetric heating and tritium breeding values have less than 0.6% relative difference between codes. [ABSTRACT FROM AUTHOR]

• A micro lead-based reactor with annular flow channel fuel elements is proposed. • As the M/F ratio increases, k eff first increases and then decreases. • An optimal M/F ratio can minimize the size of the reactor for energy output. • The MTC is a positive value and gradually increases as the M/F ratio increases. As a crucial parameter for the design of micro reactors, the neutron spectrum directly impacts the k eff , the critical size, burnup characteristics, reactivity temperature coefficients and so on. When considering the optimization of the reactor size, designs utilizing thermal neutron spectrum and fast neutron spectrum each present their unique advantages and challenges. To investigate the impact of energy spectrum on the reactor performance, a micro lead-based reactor with annular channel fuel elements is proposed and the study of the influence of varying the volume ratio of moderator and fuel (the M/F ratio) on the k eff , the critical size, burnup characteristic and reactivity temperature coefficients by using the Reactor Monte Carlo code (RMC code) is conducted. The results show that for the micro lead-based reactor proposed, when the reactor energy output demand is greater than 350 MWt·years, the design with fast neutron spectrum (without moderator) exhibits the minimum size. While the reactor energy output demand is less than 350 MWt·years, the design with a thermal neutron spectrum (with moderator) achieves the minimum size. Moreover, to ensure the negative reactivity temperature effect, the M/F ratio should be maintained below 3.4. The insights presented in this paper will serve as valuable references for the design of the micro reactor. [ABSTRACT FROM AUTHOR]

• The neutronics theory of the code NECP-Panda for the calculation of the Pebble-Bed HTGR is introduced. • Based on HTR-PM, the initial fuel loading calculations and absorber worth calculations are conducted by NECP-Panda. • Numerical verifications show that NECP-Panda achieves both high accuracy and efficiency. As one of the fourth-generation nuclear energy systems, the Pebble-Bed High Temperature Gas-cooled Reactor (PB-HTGR) shows promising prospects. Aiming to address the emerging requirements in the design analysis and technological improvement of the PB-HTGR core, the Nuclear Engineering Computational Physics (NECP) Laboratory of Xi'an Jiaotong University is autonomously developing a neutronics and thermal–hydraulic analysis code named NECP-Panda. The development work pertaining to the steady-state neutronics calculations for the reactor core has been successfully completed. This paper presents the computational framework of NECP-Panda, focusing on the theories and methodologies of neutronics calculation. Based on the model of the High Temperature Reactor Pebble-Bed Module (HTR-PM) in China, NECP-Panda conducted initial fuel loading calculations and absorber worth calculations. Numerical results demonstrate a high consistency between NECP-Panda's critical loading height prediction for HTR-PM and the Monte Carlo (MC) continuous energy calculation results, with the number of loading pebbles at criticality closely approximating experimental result. Moreover, reactivity worth for various absorber combinations and control rod integral worth curves show satisfactory agreement. NECP-Panda demonstrates exceptional computational capabilities, accuracy and acceptable efficiency in neutronics calculations for the PB-HTGR, laying a solid foundation for the subsequent development of other modules. [ABSTRACT FROM AUTHOR]

The control rod withdrawal (CRW) transient is one of the most important safety issues for the early design phase of sodium fast reactors (SFR). CRW transients can cause core melting, which can lead to severe accidents. To ensure the integrity of the core, it is crucial to study CRW transients in detail since early design phase of the reactor. Two approaches are used in numerical modeling of transients at CEA. Reference codes (APOLLO3 (Schneider et al., 2016), GERMINAL (Lainet et al., 2019) for example) are used to perform detailed and precise calculations, recently these tools are being coupled to form multiphysics platforms such as described in Pascal et al. (2021); Li et al. (2023). Despite being robust and validated, this type of simulation is not adapted to design phases due to the need in computing resources it requires. The second type of approach, selected in this paper, is to use simplified, parametric tools with a limited validity range. We present here a new numerical tool called MORPHEE, which is capable of simulating CRW transients. MORPHEE is a multiphysics and multiscale tool that includes neutronics, thermal-hydraulics, and fuel behavior. The tool has been developed to provide a comprehensive understanding of the CRW transient and to assess the safety of SFR. It is based on MACARENa (Droin et al. 2020) and OCARINa (Herbreteau et al. 2018) tools. They have been validated for simulation of thermal-hydraulics and fuel pin behavior, respectively. MORPHEE uses finite-differences methods to solve thermal-hydraulics equations, neutron kinetics equations and mechanical equations. A machine learning model (Jeannin et al., 2022) is also used to predict fuel cladding heat transfer through the gap. The simulations have demonstrated the importance of considering the multiphysics and multiscale aspects of the CRW transient to predict the behavior of the reactor during an accident. In conclusion, MORPHEE is a numerical tool that can simulate the CRW transient in SFR. It can be used to study the safety of SFR during the early design phase and to optimize the design to ensure the safety of the reactor. • Multiphysics methodology for Sodium Fast Reactors transients simulation. • Control Rod Withdrawal (CRW) phenomenological-based modeling with MORPHEE tool. • CRW modeling tool to guide design of SFR during the conception phase regarding safety. • Fuel cladding heat exchange coefficient and impact on transients over power. • Neutronics simplified simulation of flux shape deformation. [ABSTRACT FROM AUTHOR]

We introduce and demonstrate a method to mitigate ray effects in solutions of the discrete ordinates (S N) approximation of the transport equation through coupling the S N solver to the n -collided flux computed with a Monte Carlo simulation. This method is inspired by the classical approach of coupling the uncollided flux from a Monte Carlo simulator, or other semi-analytic estimator, that we extend in this work to compute an arbitrarily-high collision-count, n -collided flux that is free from ray effects by Monte Carlo. This flux is used to compute the (n + 1) -collided source that drives the S N solver to obtain the flux of particles that experience more than (n + 1) collisions. The sum of these two fluxes is the full solution of the transport equation with ray effects that diminish with increasing n. For demonstration purposes we employ a simple model of an experimental setup designed to compute the recorded neutron count rate of the BeRP ball measured by the SNAP detector at a distance of 100 cm. Due to the long trajectory that the neutrons travel in air between the source and detector a straight solution with S N methods exhibits severe ray effects that do not decrease sufficiently with increasing angular quadrature order N. This model has reference results that were generated with OpenMC using 32 group isotropic-scattering cross-sections. We test our new method on this model by using the steady-state, multigroup, S N code THOR which uses the methods of short characteristics to discretize the spatial variable on tetrahedral cells, coupled to the special-purpose, multigroup Monte Carlo code Hammer. Hammer is designed to tally the neutron flux on THOR's tetrahedral cells by energy group for a user-selected collision-count n. Our results establish that for this model problem, ray effects in the coupled solution decrease with increasing n , and the neutron count rate approaches its asymptotic value faster. Specifically, the 6th or 7th collision source coupled with the S 2 discrete ordinates solver is sufficient to eliminate noticeable ray effects in this configuration. These results represent a verification of the method compared to reference multigroup Monte Carlo calculations. • We introduce the general collision-count source method for ray effects mitigation. • This method (shortened as the GCCSM) is formulated in this work. • The GCCSM is an extension of Alcouffe's first collision source method. • The GCCSM is tested and results converge to close to the Monte Carlo reference. [ABSTRACT FROM AUTHOR]

Modeling the run-in of a pebble-bed reactor (PBR) can be challenging as a result of changes in the power, fuel type, and temperatures that occur throughout the run-in period. Previous work utilized high-fidelity neutronics simulations or lower-fidelity coupled neutronics/thermal-hydraulics models to capture the general characteristics of the run-in process. The present work employs high-fidelity neutronics simulations (using Serpent) coupled with thermal-hydraulics simulations (using Griffin–Pronghorn) to capture the thermal feedback present during the run-in and approach to equilibrium for a PBR. Incorporating thermal feedback enables important distinctions to be made about conditions occurring inside the core,as the power distribution, discharge burnup, and isotopic compositions are all affected by the temperature distribution. • Utilized high-fidelity Monte Carlo simulations to capture the run-in of a pebble-bed reactor. • Incorporated thermal feedback into neutronics calculations by using a coupled Griffin–Pronghorn simulation. • Created a mapping framework to exchange isotopic and temperature information between Serpent and Griffin. [ABSTRACT FROM AUTHOR]

• A reliable SSC system based on NU-SSCCRs has been introduced in the VVER-1000 FA. • Three different sequences of NU-SSCRs withdrawal strategies have been considered. • The discharge burnup and CR for the three different withdrawal strategies have been improved. • Comprehensive mass calculations of the isotopic compositions for FA have been performed. A self-sustaining reactor system with both a high conversion ratio (CR) and fuel discharge burnup (B d) is essential for enhancing the utilization of nuclear fuel and improving its natural resources. The mechanical spectral shift control (SSC) method can improve the efficiency of nuclear fuel utilization by removing, or at least greatly lowering, soluble and burnable neutron poisons during reactor operation. The implementation of the proposed mechanical SSC method involved the on-power variation of the moderator-to-fuel volume ratio (V M /V F). This was achieved by means of movable natural uranium spectral shift control rods (NU-SSCRs) distributed uniformly within the fuel assembly. In this work, the neutronic characteristics of the VVER-1000 fuel assembly benchmark with mechanical SSC regulation have been investigated, compared to the traditional poison method described in the OECD/NEA benchmark model. We adopted three different sequences of NU-SSCRs withdrawal strategies at various burnup points, each associated with different V M /V F , during one batch fuel loading pattern. The findings revealed a notable improvement of ≈40 % in the discharge burnup across the three different withdrawal strategies, alongside the attainment of a CR value of 0.82 compared to the benchmark poisoned model. Additionally, the mechanical SSC regulation has effectively suppressed the excess reactivity at the beginning of the cycle compared to boric acid (H 3 BO 3) and gadolinium oxide (Gd 2 O 3), which are incorporated in the benchmark model by 754 pcm. The results showed a higher accumulated mass of plutonium in the fuel assembly of 3.3, 5.2, and 6 kg for the three different sequences of withdrawing strategies, respectively, at the end of the cycle. Furthermore, the void reactivity coefficient has been investigated for the three different withdrawal strategies in both partially voided (20 % void) and completely voided (100 % void) systems. [ABSTRACT FROM AUTHOR]

• The feasibility of five different fuel configurations (6.11 wt.% enriched UO 2 , thorium, TRU, reactor and weapons grade MOX fuel) for the SCWR core was studied. • All fuels except thorium exhibited a cycle length exceeding 1000 EFPDs and a discharge burnup greater than 30 MWD/kg. • MOX fuel with 16 wt.% RG plutonium demonstrated the best performance with over 1700 EFPD cycle length and a 53.8 MWD/kg discharge burnup. • Negative reactivity coefficients were obtained for all of the fuels during their respective critical states with little differences in the radial power peaking factors. This study aims to investigate the feasibility of five different fuel configurations for the supercritical water reactor (SCWR) core. Thorium, MOX and transuranic (TRU) fuels were considered alongside conventional UO 2 , and burnup simulations were conducted using the Monte Carlo code SERPENT for an extended period of 1800 EFPDs. All of the fuels, except thorium, successfully demonstrate cycle lengths greater than 1000 EFPDs and discharge burnups exceeding 30 MWD/kg. The reactivity coefficients of the cores also remain negative for the entire operational period. MOX fuel with 16 wt.% reactor-grade plutonium exhibits the best fuel cycle performance with over 1700 EFPD cycle length, a staggering 53.8 MWD/kg discharge burnup, the greatest conversion ratio of 0.5725, as well as the most negative fuel, moderator and coolant temperature coefficients. However, it has a reduced effective delayed neutron fraction and a slightly higher xenon and samarium content. Despite the shortest cycle length and lowest burnup, thorium fuel exhibits a high conversion ratio of 0.5585, indicating its potential for 233U breeding with modifications to fuel composition and assembly design. MOX fuel using weapons-grade plutonium and TRU fuel demonstrate cycle lengths less than conventional UO 2 (6.11 wt.% enriched), with TRU fuel having the least β eff , but they offer the opportunity to reduce the plutonium stockpile generated from currently operational reactors. [ABSTRACT FROM AUTHOR]

Accurate prediction of the activation of fluids flowing under irradiation is important for the timely development of fusion technology. One of the major current issues is the cooling water of ITER, which becomes activated by plasma neutrons during nuclear operations, thus raising a number of radiological implications such as radiation effects on sensitive equipment, compliance with radiological protection zoning, and compliance with radioactivity limits in regulations of pressure equipment and effluents. Further significant applications are the activation of other service fluids in ITER and of the LiPb in both ITER and DEMO breeding blanket modules. To avoid systematic uncertainties, simulation of the activation of fluids flowing in arbitrarily complex 3D geometry, flow regimes and neutron fields requires full coupling of activation with fluid-dynamics physics models. No such tools were available until recently: the Radio-Species Transport Model (RSTM), based on the well-established ANSYS Fluent® UDS methodology, is one conceived and developed at F4E. Here we review the methodology and capabilities of the RSTM, as well as earlier development, validation, benchmarking and application activities. We then report currently ongoing further applications and benchmarking being performed in collaboration with specific tasks of the EUROFusion programme Preparation for ITER Operation. Computations, results and comparison with other methodologies for several cases of interest are presented and discussed. [ABSTRACT FROM AUTHOR]

• Multiscale and multiphysics are vital to simulate the behavior of nuclear reactors. • Coupling are needed to simulate interactions between physical phenomena. • The development of simulation platforms streamlines code integration. • Spain hosts research groups dedicated to multiscale, multiphysics and coupling. In the field of reactor physics analysis, the coupling of multiple physics phenomena plays a crucial role in enhancing the accuracy of predictions and understanding complex systems. Multiphysics refers to the study and analysis of physical phenomena that involve multiple interconnected physical processes occurring simultaneously in a nuclear system to ensure essential aspects of reactor design, safety analysis, and optimization, among others. Multiphysics encompasses various domains of physics, such as Thermal-Hydraulics, Neutronics, Structural Mechanics, Radiation Transport, Chemistry, and Materials Science. On the other hand, the Multiscale approach involves combining high-fidelity models with lower-fidelity ones to accurately capture the relevant physical phenomena in a Nuclear Power Plant (NPP). Multiscale involves three main scales: system level, core level, and detailed analysis scale. This paper presents an overview of the main research performed by Spanish groups in the field of multiphysics and multiscale calculations, particularly on thermal–hydraulic analysis. This revision is focused mainly on two key aspects: thermal–hydraulic multiscale coupling and thermal–hydraulic / neutronic coupling. The paper also includes the expected future trends in this field. [ABSTRACT FROM AUTHOR]

This paper provides neutronics analyses of the Xcimer Energy Corporation (XEC) HYLIFE-III Inertial Fusion Energy Power Plant concept. This design is based on the thick-liquid-wall HYLIFE-II reactor, but with much larger fusion yield, due to enhanced driver energy. Although HYLIFE-II neutronics was extensively studied, the differences between the two concepts suggested new analyses are required. Further, computational advances in neutronics calculations also motivate updating results from more than 25 years ago. The neutron spectra emitted from the much larger yield hybrid Inertial Confinement Energy (IFE) target is presented. Selected breeding materials are compared by tritium breeding ratio (TBR), activation, and first wall protection where FLiBe is proposed as the overall best choice. The first wall neutron activation and structural damage, including gas generation in the wall, is presented for various FLiBe protective wet-wall thicknesses. Final optic neutron damage is also examined and results in optics long enough lifespan with moderate annealing temperatures. Some limited comparisons of first wall damage for ICF and Magnetic Confinement Fusion relevant conditions is presented. HYLIFE-III with FLiBe as the breeding material and first wall protection provides a very robust TBR above 1.2. [ABSTRACT FROM AUTHOR]

• EU-DEMO Limiter System Baseline. • First integrated design concept of the EU-DEMO Upper Limiter to withstand disruptive loads. • Assessment workflow based on neutronic, thermal-hydraulic, electro-magnetic and structural assessment. • Design and assessment workflow useful for driving the design of the other limiters. The EU-DEMO first wall protection relies on a system of limiters. Although they are primarily designed for facing the energy released by a limited plasma during transients, their design should safely withstand a combination of loads relevant for in-vessel components (IVCs) during steady-state operation. They are not meant to breed tritium, nor to provide plasma stability. However, sitting in place of blanket portions, they should ensure an adequate shielding function to vacuum vessel and magnets while withstanding both their dead weight and the electro-mechanical loads arising from the interaction between current induced in the conductive structure and magnetic field. During plasma disruptions they will be subjected to halo currents flowing from/to the plasma and the grounded structures, whose effects must be added to the eddy current ones. Disruption-induced electro-mechanical loads are hence IVC design-driving, despite the uncertainties in both eddy and halo currents' magnitude and distribution, which depend on IVC design, electrical connectivity, plasma temperature and halo width. The integrated design of the limiter is made of two actively water-cooled sub-components: the Plasma-Facing Wall (PFW) directly exposed to the plasma, and the Shielding Block (SB) devoted to hold the PFW while providing neutronic shielding. The PFW design is driven by disruptive heat loads. Disruption-induced electro-magnetic loads are instead SB design drivers, meaning that the design details (i.e. geometry, electrical connections, attachments) affect the loads acting on it, which, in turn, are affected by the mechanical response of the structure. The present paper describes the design workflow and assessment of the Upper Limiter (UL), resulting from a close and iterative synergy among different fields. Built on static-structural and energy balance hand calculations based on, respectively, preliminary electro-magnetic and neutronic loads, the UL integrated design performance has then been verified against electro-magnetic, neutronic, thermal-hydraulic and structural assessment under the above-mentioned load combination. The outcome will be taken as reference for future limiter engineering designs. [ABSTRACT FROM AUTHOR]

• Nuclear analyses for the ITER diagnostic equatorial port 2 integration have been performed. • Solutions to improve the shielding in port interspace area are proposed and discussed. • This study has contributed to DSM#2 in-vessel shielding block design optimisation. The present work is devoted to nuclear analyses in support of the ITER diagnostic Equatorial Port 2 (EP#2) integration. ITER EP #2 is a diagnostic port based on the long-modular Diagnostic Shielding Module (DSM) housing the following systems: Disruption Mitigation System (DMS) in DSM#1 and #3 and X-Ray Crystal Spectroscopy Core (XRCS-Core) in DSM#2. Ensuring adequate radiation shielding is a major challenge since the diagnostic systems require several apertures from the Vacuum Vessel (VV) through the Port Interspace (PI) and up to the Port Cell (PC). In the present study, a three-dimensional MCNP model of EP#2 has been developed, starting from the latest design available from Preliminary Design Review stage (PDR), and successively integrated into the reference 40° ITER C-Model. Comprehensive nuclear analyses have been carried out employing the D1SUNED v3.1.4 code based on the MCNP Monte Carlo transport code. Relevant nuclear quantities during and at the end of plasma operations have been evaluated: i.e., neutron and gamma fluxes and energy spectra along the port from the Diagnostic First Wall (DFW) up to the Bio-Shield Plug (BP), nuclear heating, neutron damage, helium and tritium production, and shutdown dose rate. This analysis allowed the identification of potentially critical areas, and therefore, the implementation of additional shielding options aimed at reducing the neutron streaming and mitigating of the radiation field in the PI region. In this work, the results of the analyses are presented and discussed. Some solutions to mitigate nuclear loads and to improve the shielding in PI area are proposed and their impact has been assessed. Finally, some recommendations for the optimization of the design of EP#2 are provided as well. [ABSTRACT FROM AUTHOR]

The Radial Neutron Camera (RNC) is a diagnostic system located in the ITER Equatorial Port 1 (EP01) composed by two sub-systems (i.e.: in-port and ex-port RNC) probing a poloidal section of the plasma through a set of fan-shaped Lines of Sight (LOS). The RNC is designed to provide a time resolved measurement of the neutron and alpha particles source profiles and of the total neutron source strength, through the application of reconstruction techniques to the line-integrated neutron fluxes. The Ex-port sub-system is composed by 16 LOS distributed in two different toroidal planes and enclosed in a massive shielding unit, extending from the EP01 closure plate through the Port Interspace, up to the Bioshield Plug. Neutrons, generated in the plasma core, stream through dedicated optical paths hollowed out in the central EP01 diagnostic shielding module and reach the detectors units located at the end of collimating structures. Each detector unit contains one 4He gas scintillator, one plastic scintillator as well as one single Crystal Diamond (sCD) matrix. The performed nuclear analysis allowed the selection of the SWX-277Z-5 castable borated hydrogenated mix as suitable material for the RNC shielding block; moreover, the evaluation of the nuclear heating on the ex-port RNC subcomponents, provided as input for structural analyses, highlighted that the impact of the radiation streaming is negligible with respect to the environmental thermal loads. [ABSTRACT FROM AUTHOR]

The operation of many nuclear pressurized water reactors is being extended beyond their design lifetime threshold. From the perspective of possible further lifetime extension, satisfying safety requirements is a priority. Characterization of the structural integrity of the reactor pressure vessel (RPV) is an important issue as it is a guiding parameter that influences the reactor lifetime. Embrittlement of RPV material is primarily induced by the bombardment of fast neutrons (with energies greater than 1 MeV). Consequently, fast neutron fluence is one of the quantities used by safety authorities to characterize the structural integrity of RPV. However, future RPV aging assessments might lean on new variables with respect to current laws, such as neutron fluence considering the whole neutron spectrum or displacements per atom (dpa) since the latter is more representative of overall damage generated in the RPV. In order to meet these challenges, a versatile calculation scheme for RPV aging assessments is proposed in this paper. The developed methodology allows one to compute (fast and non-fast) neutron fluence as well as dpa rate, using the Norgett-Robinson-Torrens dpa model and the Athermal Recombination Corrected dpa model, for a wide azimuthal and axial range on the RPV and in the capsules of the aging monitoring program (which contain dosimeters and vessel material samples). This methodology is based on a coupling between deterministic (CASMO5 and SIMULATE5) and Monte Carlo (MCNP6) numerical approaches. First, the deterministic approach is used to evaluate the full-core fission neutron source term. Second, Monte Carlo modeling is used to perform the neutron attenuation from the core to sites of interest, such as the RPV. The computational efficiency, accuracy, and potential benefits of the methodology are presented. Moreover, the frequency at which neutron transport calculations should be performed in order to obtain sufficiently accurate time-integrated data over a reactor cycle is discussed. Finally, the validity of the fast neutron fluence as an indicator of RPV aging is compared against the use of dpa. [ABSTRACT FROM AUTHOR]

The paper presents results from neutron flux calculations in the divertor cassette body in cases of Helium Cooled Pebble Bed, and Water-Cooled Lithium Lead concepts are used as a breeding blanket. The same divertor model configuration and specified DEMO neutron source were used for both cases studied. MCNP6 code was used for neutron transport calculations and employing the FENDL-3.1 nuclear data library. |In addition, this paper presents the neutron-induced dose rate calculations performed with support by ADVANTG (AutomateD VAriaNce reducTion Generator) tool and MCNP code, which were used for the variance reduction and particle transport accordingly. The application of such a coupled computational method led to the dose rate and neutron flux maps production, which was obtained for the divertor of the DEMO reactor. [ABSTRACT FROM AUTHOR]

Blankets are of practical importance in the fusion reactor due to their main functions for engineering application. As the fusion power designed for the Chinese Fusion Engineering Test Reactor (CFETR) is increased from 200 MW to 1.5 GW during its phase II operation, more challenges are issued for blankets from the points of view of thermal-hydraulics, neutronics, mechanics, and safety. In this paper, the Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC) was developed using the user-defined function (UDF) in the CFD software FLUENT. The explicit coupling scheme based on cell-to-cell mapping approach was adopted to transfer data between neutronics calculation and thermal-hydraulics simulation. Three-dimensional thermal-hydraulic model and neutronics model were established using Fluent and MCNP4C, respectively. The helium-cooled solid breeder (HCSB) blanket of CFETR was simplified as a slice model to reduce the computing resources. With the slice model of CFETR HCSB blanket, the coupling analysis was carried out using the developed CATNIC. The calculation results from coupling and non-coupling analyses were compared to verify the feasibility of the coupling method in this study. As the helium coolant was considered as the transparent medium for neutron, the thermal-hydraulic and neutronics analyses were both converged after two iterations. However, using the developed coupling method in this study, the detailed change of tritium production ratio (TPR) and nuclear heat deposition in different parts of the blanket was revealed, which could provide important guidance for the design of the distribution for both coolant and purge gas.

An accelerator driven subcritical system (ADS) is a new nuclear energy system which could not only produce clean energy but also incinerate nuclear waste. In this paper, inherent safety analysis of an ADS is performed with neutronics and thermal-hydraulics coupled code named ARTAP. Five typical accidents are carried out, including the cases of proton beam interruption, transient overpower, reactivity insertion, loss of flow, and loss of heat sink. The transient simulations are performed in the average channel and the hottest channel of the fuel pin in the ADS core. The simulation results for beam interruption show that the highest temperature of the pellet is in the middle of the fuel element in the average channel, while the peak temperature of the cladding is in the top of the fuel element. After the beam is interrupted for 20s, the maximum temperature drops at the fuel center, the cladding inner surface, and the outlet coolant in the hottest channel are 644.46K, 162.27K, and 136.42K respectively. For transient overpower accidents with the increase of beam intensity, the maximum temperature of the fuel and the cladding are below the safety limit. Concerning the reactivity insertion accident, it is found that the ADS has good inherent safety and its margin of criticality safety is large. The calculation results for loss of flow show that the power drop is small due to low sensitivity of the subcritical core to negative reactivity feedback, and the maximum temperature of the cladding reaches 1726K, which means the fuel element would rupture. However, the power and the temperatures of fuel, cladding, and coolant could decrease quickly to the safety level after the proton accelerator is cut off under a loss of flow accident. The results also show that the peak temperature of the cladding is lower than the safety limit under a loss of heat sink accident. The present simulation results reveal that the ADS has a remarkable advantage against severe accidents. It also implies that its inherent safety characteristics could ensure reactor shutdown by cutting off the proton beam after accidents occur.

In the prototype fast breeder reactor Monju, fixed absorber worth was measured as the difference of core reactivity measured by control rod worth between cores with and without a single fixed absorber or three fixed absorbers. In this paper, the measurements are evaluated in detail, and their reliability and usefulness as validation data are investigated through a comparison with calculations using the latest neutronics design methodology developed at Japan Atomic Energy Agency. Calculated-to-experiment values and their uncertainties of fixed absorber worth were 1.00 ± 0.05 and 1.02 ± 0.04, respectively. Through this study, the measurements and calculations were found consistent and reliable. [ABSTRACT FROM AUTHOR]