Journal of Nuclear Engineering

This paper presents the results obtained from numerical evaluations for the possibility of large-scale ²³⁸Pu production in the light-water VVER-1000 reactor and the reactivity effect caused by the loss-of-coolant accident in the central fuel assembly of the reactor core. This fuel assembly containing the Np–Am-component of minor actinides was placed in the center of the reactor core and intended for intense production of ²³⁸Pu. Optimal conditions were found for large-scale production of plutonium with an isotope composition suitable for application in radioisotope thermoelectric generators. The reactivity effect from the loss-of-coolant accident in the central Np–Am fuel assembly was evaluated, and the perturbation theory was used to determine the contributions of some neutron processes (leakage, absorption, and moderation) to the total variation of the effective neutron multiplication factor. Full article

Fluid-fueled molten-salt reactors (MSRs) are actively being developed by several companies, with plans to deploy them internationally. The current IAEA inspection tools are largely incompatible with the unique design features of liquid fuel MSRs (e.g., the complex fuel chemistry, circulating fuel inventory, bulk accountancy, and high radiation environment). For these reasons, safeguards for MSRs are seen as challenging and require the development of new techniques. This paper proposes one such technique through the observation of the reactor’s off-gas. Any reactor design using low-enriched uranium will build up plutonium as the fuel undergoes burnup. Plutonium has different fission product yields than uranium. Therefore, a shift in fission product production is expected with fuel evolution. The passive removal of certain gaseous fission products to the off-gas tank of an MSR provides a valuable opportunity for analysis without significant modifications to the design of the system. Uniquely, due to the gaseous nature of the isotopes, beta particle emissions are available for observation. The ratios of these fission product isotopes can, thus, be traced back to the relative amount and types of fissile isotopes in the core. This proposed technique represents an effective safeguards tool for bulk accountancy which, while avoiding being onerous, could be used in concert with other techniques to meet the IAEA’s timeliness goals for the detection of a diversion. Full article

The use of tungsten fiber-reinforced tungsten composites (W_f/W) has been demonstrated to significantly enhance the mechanical properties of tungsten (W) by incorporating W-fibers into the W-matrix. However, prior research has been restricted by the usage of single fiber-based textile fabrics, consisting of 150 µm warp and 50 µm weft filaments, with limited homogeneity, reproducibility, and mechanical properties in bulk structures due to the rigidity of the 150 µm W-fibers. To overcome this limitation, two novel textile preforms were developed utilizing radial braided W-yarns with 7 core and 16 sleeve filaments (R.B. 16 + 7), with a diameter of 25 µm each, as the warp material. In this study, bulk composites of two different fabric types were produced via a layer-by-layer CVD process, utilizing single 50 µm filaments (type 1) and R.B. 16 + 7 yarns (type 2) as weft materials. The produced composites were sectioned into KLST-type specimens based on DIN EN ISO 179-1:2000 using electrical discharge machining (EDM) and subjected to three-point bending tests. Both composites demonstrated enhanced mechanical properties with pseudo-ductile behavior at room temperature and withstood over 10,000 load cycles between 50–90% of their respective maximum load without sample fracture in three-point cyclic loading tests. Furthermore, a novel approach to predict the fatigue behavior of the material under cyclic loading was developed based on the high reproducibility of the composites produced, especially for the composite based on type 1. This approach provides a new benchmark for upscaling endeavors and may enable a better prediction of the service life of the produced components made of W_f/W in the future. In comparison, the composite based on fabric type 1 demonstrated superior results in manufacturing performance and mechanical properties. With a high relative average density (>97%), a high fiber volume fraction (14–17%), and a very homogeneous fiber distribution in the CVD-W matrix, type 1 shows a promising option to be further tested in high heat flux tests and to be potentially used as an alternative to currently used materials for the most stressed components of nuclear fusion reactors or other potential application fields such as concentrated solar power (CSP), aircraft turbines, the steel industry, quantum computing, or welding tools. Type 2 composites have a higher layer spacing compared to type 1, resulting in gaps within the matrix and less homogeneous material properties. While type 2 composites have demonstrated a notable enhancement over 150 µm fiber-based composites, they are not viable for industrial scale-up unlike type 1 composites. Full article

The production of prismatic dislocation loops in nuclear reactor core materials results in hardening because the loops impede dislocation motion. Yielding often occurs by a localised clearing of the loops through interactions with gliding dislocations called channeling. The cleared channels represent a softer material within which most of the subsequent deformation is localized. Channeling is often associated with hypothetical dislocation pileup and intergranular cracking in reactor components although the channels themselves do not amplify stress as one would expect from a pileup. The channels are often similar in appearance to twins leading to the possibility that twins are sometimes mistakenly identified as channels. Neither twins nor dislocation channels, which are bulk shears, produce the same stress conditions as a pileup on a single plane. At high doses, when cavities are produced (either He-stabilised bubbles at low temperatures or voids at high temperatures), there can be reduced ductility because the material is already in an equivalent advanced stage of microscopic necking. He-stabilised cavities form preferentially on grain boundaries and at precipitate or incoherent twin/ε-martensite interfaces. The higher planar density of the cavities, coupled with the incompatibility at the interface, results in a preferential failure known as He embrittlement. Strain localisation and inter- or intragranular failure are dependent on many factors that are ultimately microstructural in nature. The mechanisms are described and discussed in relation to reactor core materials. Full article

This article studies the optimal design approach for a portable nuclear magnetic resonance (NMR) system for use in non-destructive flow measurement applications. The mechanical and electromagnetic design procedures were carried out using the Ansys Maxwell finite-element analysis (FEA) software tool. The proposed procedure considered homogeneity and strength constraints while ensuring the desired functionality of the intended device for a given application. A modified particle swarm optimization (MPSO) algorithm was proposed as a reference design framework for optimization stages. The optimally designed NMR tool was prototyped, and its functionality was validated via several case studies. To assess the functionality of the prototyped device, Larmor frequency for hydrogen atom was captured and compared with theoretical results. Furthermore, the functionality and accuracy of the prototyped NMR tool is compared to the off-the-shelf NMR tool. Results demonstrated the feasibility and accuracy of the prototyped NMR tool constrained by factors, such as being lightweight and compact. Full article

Machine learning methods allow the prediction of material properties, potentially using only the elemental composition of a molecule or compound, without the knowledge of molecular or crystalline structures. Herein, a composition-based machine learning prediction of the material properties of V–Cr–Ti alloys is demonstrated. Our machine-learning-based prediction of the stability of the V–Cr–Ti alloys is qualitatively consistent with the composition-dependent experimental data of the ductile–brittle transition temperature and swelling. Furthermore, our computational results suggest the existence of a composition region, Cr+Ti ~ 60 wt.%, at a significantly low ductile–brittle transition temperature. This outcome contrasts with a reportedly low Cr+Ti content of less than 10 wt.% in conventional V–Cr–Ti alloys. Machine-learning-based numerical stability prediction is useful for the design and analysis of metal alloys, particularly for multicomponent alloys such as high-entropy alloys, to develop materials for nuclear fusion reactors. Full article

Neutron thermalization leads to the complexity of the scattering cross-section calculation, which influences the accuracy of the neutron transport calculation in the thermal energy range. The higher precision of thermal scattering data is demanded in the small modular reactors (SMRs) design, especially for small-sized PWRs and SCWRs. Additionally, the thermal neutron scattering problems in supercritical water have not yet been solved. In this study, the thermal neutron scattering problems in subcritical water are tested. Based on thermal neutron scattering theory, the GA model and IKE model were analyzed. This work selected the corresponding input parameters, such as the frequency spectrum, the discrete oscillator energy, weight parameters and so on, as well as preliminary studies on how to calculate the thermal scattering data for HinH₂O to accomplish the calculation at various temperatures by developing LIPER code. The deviation between the calculated and reference results, which were both obtained by the Monte Carlo code, COSRMC, was below 0.2 pcm. The deviation of the scattering cross-section between the calculation results and reference was below 0.1%, indicating the reasonability of this study’s thermal scattering data calculation. Full article

The objective of the ITER test blanket module (TBM) program is to provide experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. The ITER test blanket modules are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each PP includes two TBMs. After each 18-month-long plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs with new ones during the ITER long-term shutdown, called long-term maintenance (LTM). The replacement of a TBM requires the removal/reinstallation of all test blanket system (TBS) equipment present in the port cell (PC), including those in the port interspace (PI), called pipe forest (PF). TBSs shall be designed so that occupational radiation exposure (ORE) can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER policy. To implement ALARA process requirements, design activities shall consider careful integration investigations starting from the early phase to address all engineering aspects of the replacement sequence. The case study focuses on the PF replacement, in particular the port cell operations. This paper describes the investigations and findings of the ALARA optimisation process implementation in the early engineering phase of the PF. Full article

The use of heat pipes (HP) for the DEMO in-vessel plasma-facing components (PFCs) has been considered because of their high capacity to transport the heat from a heat source to a heat sink by means of the vaporization and condensation of the working fluid inside and their ability to enlarge the heat transfer area of the cooling circuit substantially. Recent engineering studies conducted in the framework of the EUROfusion work package Divertor (Wen et al, 2021) indicate that it is possible to design a heat pipe with a capillary limit above 6 kW using a composite capillary structure (wherein axial grooves cover the adiabatic zone and the condenser, and sintered porous material covers the evaporator). This power level would correspond to an applied heat flux of 20 MW/m², rendering such a design interesting with respect to a divertor target concept. To validate the results of the initial engineering analysis, several experiments have been conducted to evaluate the actual performance of the proposed heat pipe concept. The present contribution presents the experiment’s results regarding the examination of the operating limits of two different designs for an evaporator: one featuring a plain porous structure, and one featuring ribs and channels. Full article

People’s affective response in relation to radiation is important in their risk perceptions of low-dose radiation (LDR), medical interventions involving LDR, and acceptance of nuclear power production. Risk perception studies generally relate to the health field of LDR or nuclear power. This study combines risk perceptions and acceptance of both. While acceptance by those with an understanding of radiation is demonstrated in focus groups, survey results disproved this correlation. Emotional response to the word radiation together with greater perceptions of risk to X-rays, were predictors of acceptance of nuclear power production. Full article

Additive manufacturing techniques enable a wide range of possibilities for novel radiation detectors spanning simple to highly complex geometries, multi-material composites, and metamaterials that are either impossible or cost prohibitive to produce using conventional methods. The present work identifies a set of promising formulations of photocurable scintillator resins capable of neutron-gamma pulse shape discrimination (PSD) to support the additive manufacturing of fast neutron detectors. The development of these resins utilizes a step-by-step, trial-and-error approach to identify different monomer and cross-linker combinations that meet the requirements for 3D printing followed by a 2-level factorial parameter study to optimize the radiation detection performance, including light yield, PSD, optical clarity, and hardness. The formulations resulted in hard, clear, PSD-capable plastic scintillators that were cured solid within 10 s using 405 nm light. The best-performing scintillator produced a light yield 83% of EJ-276 and a PSD figure of merit equaling 1.28 at 450–550 keVee. Full article

During the maintenance period of a future fusion reactor power plant, called DEMOnstration Power Plant (DEMO), remotely handled casks are required to confine and handle DEMO in-vessel components during their transportation between the reactor and the active maintenance facility. In order to limit the dispersion of activated dust, a Contamination Control Door (CCD) is designed to be placed at an interface between separable containments (e.g., vacuum vessels and casks) to inhibit the release of contamination at the interface between them. The remotely operated CCD—technically, a double lidded door system—consists of two separable doors (the cask door and port door) and three different locking mechanisms: (i) between the cask door and cask, (ii) between the cask door and port door and (iii) between the port door and port. The locking mechanisms are selected and assessed according to different criteria, and the structure of the CCD is optimized using an Abaqus Topology Optimization Module. Due to the elastic properties of the CCD, deflections will occur during the lifting procedure, which may lead to malfunctions of the CCD. A test rig is developed to investigate the performance of high-risk components in the CCD in the case of deflections and also malpositioning. Misalignment can be induced along three axes and three angles intentionally to test the single components and items. The aim is to identify a possible range of operating in the case of misalignments. It is expected that the proposed CCD design should be able to operate appropriately in the case of ±3 mm translational misalignments and ±1° rotational misalignments. Full article

The SM-MSR (small modular molten salt reactor) has a good prospect for development with regards to combining the superiority of the molten salt reactor and modularization technologies, showing the advantages of safety, reliability, low economic cost and flexibility of site selection. However, because its internal structural parts are not easily replaced, and the outer shielding structure is limited, the lifespan of the reactor vessel and its in-reactor shielding design needs to be addressed. In order to find an optimal shielding model with both high fuel efficiency and strong radiation shielding capability, five different design schemes were proposed in this work, which varied in thickness and boron concentration in inner-shielding materials. The neutron/gamma flux and DPA (displacements per atom)/helium production rates were evaluated to obtain an appropriate scheme. Several beneficial results were obtained. Considering the above factors and the actual manufacturing process, 20 cm-thick boron graphite with a 5 wt% Boron-10 concentration combined with a 1 cm-thick Hastelloy barrel was chosen as the in-reactor shielding structure. Outside the reactor, the neutron flux was reduced to 8.33 × 10¹⁰ cm⁻² s⁻¹, and the gamma flux was decreased to 1.13 × 10¹¹ cm⁻² s⁻¹. The vessel/barrel material could maintain a lifespan of more than 10 years, while the burnup depth was 6.25% lower than that of a model without inner-shielding. The conclusions of this research can provide important references for the shielding design and parameter selections of small molten salt reactors in the future. Full article

Plasma-facing materials (PFMs) in the ITER divertor will be exposed to severe conditions, including exposure to transient heat loads from edge-localized modes (ELMs) and to plasma particles and neutrons. Tungsten is the material chosen as PFM for the ITER divertor. In previous tests, bubble formation in ITER-grade tungsten was detected when exposed to fusion relevant conditions. For this study, ITER-grade tungsten was exposed to simultaneous ELM-like transient heat loads and D/He (6%) plasma in the linear plasma device PSI-2. Bubble formation was then investigated via SEM micrographs and FIB cuts. It was found that for exposure to 100.000 laser pulses of 0.6 GWm⁻² absorbed power density (P_abs), only small bubbles in the nanometer range were formed close to the surface. After increasing P_abs to 0.8 and 1.0 GWm⁻², the size of the bubbles went up to about 1 µm in size and were deeper below the surface. Increasing the plasma fluence had an even larger effect, more than doubling bubble density and increasing bubble size to up to 2 µm in diameter. When using deuterium-only plasma, the samples showed no bubble formation and reduced cracking, showing such bubble formation is caused by exposure to helium plasma. Full article

In this work we report the characterization of thin metallic coatings of interest for nuclear fusion technology through the ns double-pulse LIBS technique. The coatings, composed of a tungsten (W) or tungsten-tantalum (W-Ta) mixture were enriched with deuterium (D), to simulate plasma-facing materials (PFMs) or components (PFCs) of the next generation devices contaminated with nuclear fuel in the divertor area of the vacuum vessel (VV), with special attention to ITER, whose divertor will be made of W. The double pulse LIBS technique allowed for the detection of D and Ta at low concentrations, with a single laser shot and an average ablation rate of about 110 nm. The calibration free (CF-LIBS) procedure provided a semi-quantitative estimation of the retained deuterium in the coatings, without the need of reference samples. The presented results demonstrate that LIBS is an eligible diagnostic tool to characterize PFCs with high sensitivity and accuracy, being minimally destructive on the samples, without PFCs manipulation. The CF-LIBS procedure can be used for the search for any other materials in the VV without any preliminary reference samples. Full article

Functionally graded tungsten/steel composites are attractive to be used as an interlayer to join tungsten (W) and steel for the first wall of future fusion reactor to reduce the thermally induced stresses arising from the different coefficient of thermal expansion (CTE) of W and steel. W/steel composites, with three W contents: 25, 50 and 75 vol% W, will serve as individual sublayers of this functionally graded material. Therefore, the present work exploits an emerging sintering technique, field-assisted sintering technology, to produce these composites. Firstly, a systematic parameter study was conducted aiming to reduce the residual porosity to a minimum while keeping the formation of intermetallic phases at the W/steel interface at a low level. The optimized composites 25, 50 and 75 vol% W achieved a relative density of 99%, 99% and 96%, respectively. Secondly, mechanical tests at elevated temperatures reveal that these composites are ductile above 300 °C, which is the minimum operating temperature of the first wall. Lastly, the measured CTE, specific heat capacity and thermal conductivity were consistent with the theoretically expected values. Full article

The outcome of the electromagnetic (EM) analyses carried out during the DEMO pre-conceptual phase demonstrated that EM loads are relevant for the structural assessment of the breeding blanket (BB) and, in particular, for the definition of the boundary conditions at the attachment system with the vacuum vessel. However, within the scope of the previous campaign, the results obtained using simplified models only give a rough estimation of the EM loads inside the BB structure. This kind of data has been considered suitable for a preliminary assessment of the BB segments, but it is not considered representative as input for structural analysis in which a detailed BB internal structure (that considers cooling channels, thin plates, etc.) is analyzed. Indeed, mesh dimensions and computational time usually limit EM models that simulate a whole DEMO sector. In many cases, these constraints lead to a strong homogenization of the BB structure, not allowing the calculation of the EM loads on the internal structure with high precision. To overcome such limitations, an EM sub-modeling procedure was investigated using ANSYS EMAG. The sub-modeling feasibility is studied using the rigid boundary condition method. This method consists of running a global “coarse” mesh, including all the conducting structures that can have some impact on the component under investigation and inputting the obtained results on the detailed sub-model of the structure of interest as time-varying boundary conditions. The procedure was tested on the BB internal structure, taking as reference a DEMO 2017 baseline sector and the helium cooled pebble bed (HCPB) concept with its complex internal structure made by pins. The obtained results show that the method is also reliable in the presence of non-linear magnetic behaviour. The methodology is proposed for application in future BB system assessments. Full article

The mechanical consequences of core disruptive accidents (CDAs) are a major safety concern in sodium-cooled fast reactors. Once core disruption occurs, liquefied core materials rapidly disperse vertically and radially. The dispersed materials penetrate the pin bundles and control rod guide tubes (CRGTs) before freezing at the edge of the penetration zone as heat is transferred to surrounding structures. Such freezing phenomena can suppress the negative reactivity feedback of fuel dispersion. The discharge of core materials can be impeded, resulting in a molten core pool formation when tight blockages occur inside CRGTs due to frozen material. Accordingly, freezing phenomena of core materials play a key role in governing the mechanical consequences of a CDA. To validate a freezing model implemented in our CDA analysis code, ASTERIA-SFR, a preliminary simulation of the THEFIS RUN#1 test, was performed. The calculation results show that freezing on the structural wall and crust formation were key phenomena affecting the penetration behavior, and the structural heat transfer is an important parameter. A remarkable reduction of the heat transfer coefficient was required to reproduce the penetration length observed in the experiment. This suggests that the momentum exchange and flow regime at the leading edge as well as heat transfer should be well modeled to predict the freezing phenomena in rapidly evolving CDAs. Full article