J. Nucl. Eng., Volume 4, Issue 1 (March 2023) – 22 articles

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

On a Wendelstein 7-X (W7-X), an alkali metal beam (AMB) diagnostic system was installed in order to measure the plasma edge electron density profiles and turbulence transport. A sodium beam was injected in the plasma, and the light emission was observed by an optical system. During the last operation phase, OP1.2b campaign trial spectral measurements were performed with a dedicated optical branch. The results showed the emergence of potential CX lines in the light spectra during sodium injection. The lines were identified as Carbon III, which were the dominant lines observed by other diagnostics at the edge plasma. Based on these results, an additional dedicated optical system was developed and installed in 2021 for the upcoming operational phase, OP2. The optics were designed for multiple purposes: spectral measurements for the AMB system and for a He/Ne gas jet. The system was designed to allow implementation of further diagnostics on this port later (e.g., coherence imaging system). The details of the implementation of the design requirements and the main challenges of the manufacturing process and installation are discussed in this paper. Full article

In this study, we present the effects of 0.004~0.098 wt% Zr and thermo-mechanical processing (TMP) on the microstructure and mechanical properties of the China RAFM steel, CLAM, as a feasibility study for improving mechanical properties. The inclusions in ingots were characterized using optical microscope (OM) and scanning electron microscope (SEM), which could be classified as fine simple particles and large complex particles. The complexity of the alloy’s inclusion composition increases with the increasing Zr concentration. The higher the Zr content, the more complex the composition of inclusions in the alloy. The average diameter of inclusions in 0.004Zr steel was the smallest, which was 0.79 μm and the volume fraction was 0.018%. The highest yield strength, tensile strength, elongation, and impact energy of 0.004Zr alloy at room temperature were 548.3 MPa, 679.4 MPa, 25.7%, and 253.9 J. The structure of the TMPed steels was all tempered martensite. With the increase in tempering temperature, the yield and tensile strength of the experimental steel gradually decreased, while the elongation and impact energy gradually increased. The 0.004ZrD and 0.004ZrH alloys had the best yield strength and impact energy, which were 597.9 and 611.8 MPa and 225.9 and 243.3 J, respectively. In addition, the alloys showed good thermal stability during the aging at 600 °C for 1500 h. It was discovered that TMP is a simple and practical industrial technique that could successfully enhance the mechanical properties of CLAM steel without sacrificing impact toughness. Full article

Maintenance of the DEMO breeding blanket includes the removal and replacement of plasma-facing components. To access the breeding blanket, multiple coolant pipes need to be removed to allow access to the tokamak. As an option to reduce downtime and increase maintenance speed, the pipe-connection concept is developed to allow the removal of multiple pipes at the same time using a remotely operated mechanical connection. The remotely operated multi-pipe Mechanical Pipe Connection (MPC) needs to fulfil multiple requirements, such as high operating temperature and high external forces while at the same time maintaining an acceptable level of sealing between the high-pressure fluid and vacuum surroundings. In addition to the external conditions, the pipes of multiple sizes and fluids are connected in a manifold configuration. Although this will reduce the overall time required to operate the mechanical pipe connection when compared to multiple single-pipe connections, this will introduce additional forces and stresses due the interaction between pipe flow (e.g., simultaneous high- and low-temperature fluid pipes on the same manifold) through the manifold flange. The requirements and the boundary conditions of the multi-pipe MPC are taken into consideration during the design process of MPC. The design process is carried out to find the optimum form and size to allow the mechanical function of the pipe connection during the maintenance phase while withstanding the extreme operating conditions that the MPC will face the during operational phase. The resulting design will then be analyzed using numerical methods to assess the capability of the MPC designs. Full article

Fuel retention monitoring in tokamak walls requires the development of remote composition analysis methods such as laser-induced breakdown spectroscopy (LIBS). The present study investigates the feasibility of the LIBS method to analyse the composition and fuel retention in three samples from WEST divertor erosion marker tiles after the experimental campaign C3. The investigated samples originated from tile regions outside of strong erosion and deposition regions, where the variation of thin deposit layers is relatively small and facilitates cross-comparison between different analysis methods. The depth profiles of main constituents W, Mo and C were consistent with depth profiles determined by other composition analysis methods, such as glow-discharge optical emission spectroscopy (GDOES) and secondary ion mass spectrometry (SIMS). The average LIBS depth resolution determined from depth profiles was 100 nm/shot. The averaging of the spectra collected from multiple spots of a same sample allowed us to improve the signal-to-noise ratio, investigate the presence of fuel D and trace impurities such as O and B. In the investigated tile regions with negligible erosion and deposition, these impurities were clearly detectable during the first laser shot, while the signal decreased to noise level after a few subsequent laser shots at the same spot. LIBS investigation of samples originating from the deposition regions of tiles may further clarify LIBS’ ability to investigate trace impurities. Full article

Neural networks require a large quantity of training spectra and detector responses in order to learn to solve the inverse problem of neutron spectrum unfolding. In addition, due to the under-determined nature of unfolding, non-physical spectra which would not be encountered in usage should not be included in the training set. While physically realistic training spectra are commonly determined experimentally or generated through Monte Carlo simulation, this can become prohibitively expensive when considering the quantity of spectra needed to effectively train an unfolding network. In this paper, we present three algorithms for the generation of large quantities of realistic and physically motivated neutron energy spectra. Using an IAEA compendium of 251 spectra, we compare the unfolding performance of neural networks trained on spectra from these algorithms, when unfolding real-world spectra, to two baselines. We also investigate general methods for evaluating the performance of and optimizing feature engineering algorithms. Full article

A new module for the AZtlan Nodal HEXagonal (AZNHEX) code, which is part of the AZTLAN Platform, was recently developed based on the Simplified Spherical Harmonics ($S{P}_L$) scheme to deal with the challenges presented in small fast reactor cores, such as the China Experimental Fast Reactor (CEFR), with high leakage and significant scattering effects. For the verification and validation process, we generated nodal homogenized macroscopic cross-sections (XS) through a full heterogeneous core model using the stochastic code SERPENT and subsequently, these XS were employed in AZNHEX. To verify the $S{P}_L$ implementation, several mesh sensitivity exercises were performed demonstrating that the $S{P}_L$ module was implemented successfully. Furthermore, to validate the code with this new implementation, we modeled some exercises contained in the CEFR benchmark with AZNHEX and compared the results with the experimental data available. The final results show a great improvement compared with the original diffusion solver reducing the deviations significantly from experimental data. In conclusion, it is shown and discussed the relevance of improved numerical models (transport approximations instead of diffusion) for the deterministic calculations of small fast reactors. Full article

The international fusion materials irradiation facility-DEMO-oriented neutron source (IFMIF-DONES) is a facility that is designed under the framework of the EU fusion roadmap. It is going to be an essential irradiation facility for testing and qualifying candidate materials under severe irradiation conditions of a neutron field having an energy spectrum like the one present in a fusion power reactor. The material specimens are irradiated in a containment structure named the test cell (TC), which is part of the test systems (TS). The TC also houses a part of the other major system (lithium system, LS), which provides the liquid lithium for the reaction through a piping system. At a point, the lithium piping needs to exit the TS, but the primary safety boundary must be continuous around these penetrations. Therefore, a special barrier, called the test systems–lithium systems interface cell (TLIC), has been developed around the piping system to provide a safety-approved and remotely maintainable vacuum boundary envelope. In this paper, the integrated design development of the TLIC is described, consisting of the design development according to the RCC-MRx code, the remote-handling (RH) needs, and the procedures and safety-related special needs of the design. Full article

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. Full article

The fusion program in the Karlsruhe Institute of Technology (KIT) leads the R&D of the DEMO helium-cooled pebble bed (HCPB) breeding blanket within the work package breeding blanket (WPBB) of the Eurofusion Consortium in the European Union (EU). A new design of the HCPB breeder zone, with a layout inspired by a nuclear reactor fuel rod arrangement, was developed recently and called the fuel-breeder pin concept. In addition, a mock-up (MU) of this fuel-breeder pin was designed and manufactured at KIT in order to test and validate its thermal–hydraulic performance. This paper reports on the results of the first experimental campaign dedicated to the fuel-breeder pin MU testing that was performed in the Helium Loop Karlsruhe (HELOKA) facility. The paper presents: (i) the integration of the fuel-breeder pin MU into the HELOKA loop including considerations of the experimental set-up, (ii) an overview of the plan for the experimental campaigns, and (iii) a discussion of the experimental results with a focus on aspects relevant for the validation of the thermal–hydraulic design of the HCPB breeder zone. Full article

IFMIF-DONES, or the InternationalFusionMaterialsIrradiationFacility-DEMOOrientedNeutronSource, is a facility for investigations into foreseen fusion power plant materials using the relevant neutron irradiation of 14 MeV. This special n-irradiation is generated by the interaction of deuteron beams with liquid lithium. A critical issue during the operation of IFMIF-DONES is the enrichment of dissolved impurities in the Li-melt loops. The danger occurs as a result of hydrogen-induced corrosivity and embrittlement of the loop components, as well as the security hazards associated with the radioactive tritium. Hence, the application of liquid lithium in IFMIF-DONES requires a suitable impurity control system for reliable and low-level maintenance under the operating conditions of DONES. Regarding those requirements, an electrochemical sensor for hydrogen monitoring was developed in the frame of an international EUROFusion–WPENS task, to determine H-concentrations via the electro-motive force (EMF) of Li-melts and a suitable online-monitoring system. Long-term tests demonstrated that the sensor fulfills the requirements of chemical and mechanical stability and functionality under the harsh Li environment under the planned DONES conditions. Obtained results and operational experiences will be discussed in regard to application windows, reproducibility and calibration needs. Additionally, recommendations will be outlined for upgraded systems and future qualification needs. Full article