DEMO Blanket Research Articles

Current progress on fabrication processes of breeding functional materials for JA DEMO fusion applications

A JA DEMO blanket requires Be12V pebbles with 2 mm (or bigger) and Li2TiO3 with 0.2 mm as a mixed pebbles packing concept for achieving an 80 % packing fraction. The authors' group has achieved a significant milestone by successfully manufacturing Be12V pebbles through a pioneering rotating electrode process, marking a world-first achievement. Simultaneously, Li2TiO3 pebbles have been produced utilizing an emulsion process developed by the same research group. It is worth noting that most prior studies focused on pebbles with a 1 mm diameter, as this size was initially recommended for the ITER TBM before the introduction of the new packing concept. The rotating electrode process has successfully yielded Be12V pebbles measuring 2.5 mm or larger, and their characteristics have been thoroughly examined, offering a comprehensive overview of the current progress in Be12V pebble production. Furthermore, to meet the specific requirements of the JA DEMO blanket, experiments were conducted to fabricate smaller Li2TiO3 pebbles. It was discovered that Li2TiO3 pebbles with 0.3 mm could be successfully produced by adjusting the oil viscosity and flow ratio within the existing emulsion process system.

Testing needs for the development and qualification of a breeding blanket for DEMO

The purpose of this paper is to critically review the testing and qualification strategy for the DEMO breeding blanket in Europe, identifying the potential risks and weaknesses and recommending, where necessary, the changes required to strengthen or accelerate the programme. Based on the information presented in this paper, a risk mitigation strategy is required to reduce the technological, licencing, and regulatory risks associated with the performance and reliability of the breeding blanket technologies for DEMO. An attractive mitigation option, which was originally proposed more than 30 years ago but not implemented, is to perform testing and qualification of breeding blanket technologies and design concepts in a dedicated nuclear plasma device that serves as a 14 MeV volumetric neutron source.

Activation Analyses of Disposal Options for Irradiated Be12Ti

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.

Neutronic Activity for Development of the Promising Alternative Water-Cooled DEMO Concepts

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.

Design of a Test Section for the Experimental Investigation of the WCLL Manifold Hydraulic Features

A scaled-down test section representative of an Outboard Segment manifold of the Water-Cooled Lithium Lead Breeding Blanket for the European DEMO has been designed for installation and test in a high- mass flow branch of the W-HYDRA facility, under construction at the premises of ENEA Brasimone Research Center. The test section should confirm the flow repartition recently computed in the different breeding units on the full-scale manifold, validating at the same time the computational tools used for the design and analysis. The detailed objectives and requirements of the test section, as well as the scaling rationale and procedure adopted for its design, are presented in the paper. The final design of the test section is discussed. The preliminary analyses of the developed design are also presented and show that it is compliant with the initial objectives.

(Digital Presentation) Li Monitoring in Fusion Reactors: A Lithium Probe Based on Li6BaLa2Ta2O12 Solid-State Electrolyte for Molten Pb-Li

Carbon-based energy sources stand for more than 85% of the total energy consumed. This energetic situation will no longer be sustainable in a near future. Fusion power can be a good option to face this energy challenge as they are expected to provide clean, safe, and low-cost effective energy. In future fusion reactors, deuterium and tritium are suggested to be used as fuel, releasing helium, neutrons, and large amounts of energy [1]. However, since there are no natural tritium sources, it must be in situ generated. Different concepts for tritium breeding modules (TBM) are considered to be constructed in nuclear fusion reactors [2]. One of the proposals is the so-called Helium-Cooled Lithium-Lead breeding blankets (HCLL). This TBM suggests the use of molten eutectic Pb-Li [3]. However, the transformation of lithium to tritium due to the breeding reaction will cause a decrease in the lithium concentration of the alloy [4]. 6Li + 1n → 4He + 3H + 4.8MeV (1)If the composition of the alloy changes, undesirable alterations in physical and chemical properties, such as the melting point or tritium diffusivity, will happen. For that reason, lithium concentration is a highly relevant parameter that must be controlled to ensure the correct operation of future fusion reactors. Therefore, new analytical tools for online Li monitoring in harsh chemical environments are required.In this work, potentiometric sensors for lithium monitoring in molten Pb-Li alloys were constructed using Li6BaLa2Ta2O12 (LBLTO) Li-ion conducting solid-state electrolyte. Sensors were constructed by binding LBLTO sintered disks to alumina tubes using a glass binder. Pb-Li alloys used for the working (WE) and reference electrodes (RE) were prepared by mixing the required quantities of lithium and lead inside a glovebox with controlled oxygen and moisture concentrations (O2 and H2O < 0.1 ppm).Experiments were performed inside a stainless-steel reactor placed inside a glovebox. In the RE, molten Pb-Li with 3 at% Li content was used. In contrast, Pb-Li alloys with concentrations close to the eutectic composition (between 13 at% Li and 19 %at Li) were used in the WE. Potentiometric measurements were performed at 500 ºC to simulate breeding blankets operation conditions. Experimental data was in good agreement with calculations using the Nernst equation.[1] J.Kenneth. Shultis, R.E. Faw, Fundamentals of nuclear science and engineering, Marcel Dekker, 2002.[2] L.M. Giancarli, M. Abdou, D.J. Campbell, V.A. Chuyanov, M.Y. Ahn, M. Enoeda, C. Pan, Y. Poitevin, E. Rajendra Kumar, I. Ricapito, Y. Strebkov, S. Suzuki, P.C. Wong, M. Zmitko, Overview of the ITER TBM Program, Fusion Engineering and Design. 87 (2012) 395–402. https://doi.org/10.1016/j.fusengdes.2011.11.005.[3] J.-C. Jaboulay, G. Aiello, J. Aubert, A. Morin, M. Troisne, Nuclear analysis of the HCLL blanket for the European DEMO, Fusion Engineering and Design. 124 (2017) 896–900. https://doi.org/10.1016/j.fusengdes.2017.01.050.[4] T. Giegerich, K. Battes, J.C. Schwenzer, C. Day, Development of a viable route for lithium-6 supply of DEMO and future fusion power plants, Fusion Engineering and Design. 149 (2019) 111339. https://doi.org/10.1016/j.fusengdes.2019.111339.

Application of Inelastic Method and Its Comparison with Elastic Method for the Assessment of In-Box LOCA Event on EU DEMO HCPB Breeding Blanket Cap Region

The Helium Cooled Pebble Bed (HCPB) breeding blanket, being developed by the Karlsruhe Institute of Technology (KIT) and its partners is one of the two driver blanket candidates to be selected for the European demonstration fusion power plant (EU DEMO). The in-box Loss of Coolant Accident (LOCA) is a postulated initiating event of the breeding blanket (BB) that must be accounted within the design basis. In this paper, the BB cap region is analyzed for its ability to withstand an in-box LOCA event. Initially, an assessment is performed using conventional elastic design codes for nuclear pressure vessels. However, it is thought that the elastic rules are not ‘equipped’ to assess the material damage modes which are essentially inelastic. Therefore, a non-linear inelastic analysis is further performed to better understand the damage in the material. Two predominant inelastic failure modes are thought to be relevant and addressed: exhaustion of ductility and plastic flow localization. While the design of HCPB BB has been predominantly based on the elastic design-by-analysis studies, results from the present study show that the elastic rules may be overly conservative for the given material and loading and could lead to inefficient designs. To our knowledge, this study is the first attempt to investigate the structural integrity of the European DEMO blankets under in-box LOCA conditions using the inelastic methods.

A method for assessing 3D decay heat and temperature considering accurate distributions of the decay gamma fields

The decay heat has a strong impact in the design of the fusion facility due to its importance on safety, remote handling and waste management among others. For these applications, the energy deposition due to alpha and beta particles can be considered local, since the mean free path of these particles is small. On the contrary, the photons can travel longer distances affecting to the distribution of the heating, and consequently, on the temperature. Nowadays, most of the advanced tools dedicated to simulate accurate decay photon fields in fusion facilities do not estimate the decay heat, because alpha and beta heating is not coupled to photon heating. This article proposes a methodology, suitable to advanced mesh based R2S, which enables the estimation of the decay heat in each material with high spatial resolution. The aim to develop this methodology is to consider the accurate 3D decay photon field in the estimation of the decay heat. Furthermore, this methodology incorporates ELMER FEM code to calculate the temperature field taking into account the decay heat. This methodology was applied to a pre-conceptual DEMO blanket to quantify the importance of the photon transport in the estimation of the decay heat distribution as well as the temperature field.

Status of HCCR TBM program for DEMO blanket

Development of the Helium Cooled Ceramic Reflector (HCCR) breeding blanket is underway according to the development strategy of core technology for K-DEMO under the fusion energy development roadmap in Korea. The main goals are to validate integrated design tools and to develop core technologies in the field of materials, manufacturing and joining technologies, helium cooling and tritium technologies, system integration, and safety technologies. A unique graphite reflector concept was adopted to save cost by reducing the amount of beryllium neutron multiplier. The main functions and design concepts of the HCCR breeding blanket are to be verified in ITER through the HCCR Test Blanket Module (TBM) program. Not only breeder modules are to be tested in ITER but also the technologies of cooling, coolant purification, tritium extraction, etc. are to be tested or proved in ITER before employed in DEMO. This paper introduces the current updates of the design and R&D activities of the HCCR TBM systems in the preliminary design stage.

A Water cooled Lead Ceramic Breeder blanket for European DEMO

The development of a demonstration fusion power plant (DEMO) is widely considered a crucial step beyond ITER leading toward fusion electricity. In a viable DEMO plant two major functions have to be achieved, namely to breed tritium for fuel self-sufficiency and to extract high-grade heat to allow an efficient electricity production. The breeding blanket is the key component to ensure these two functions. The solid breeder blanket is considered in the design of several ITER Test blankets and DEMO breeding blankets. In Europe, the helium cooled solid breeder blanket, so-called Helium Cooled Pebble Bed (HCPB) blanket, is one of the two driver blanket candidates selected for the European DEMO. In spite of the considerable progress and achievements, several potential issues inherent to the choice of the neutron multiplier and coolant remain with this concept. To overcome these challenges, research activities on an innovative concept (a Water cooled Lead Ceramic Breeder - WLCB), in which Lead/Lead-alloy is used as neutron multiplier, ceramic pebbles as tritium breeder, pressurized sub-cooled water as coolant, are on-going. This concept avoids the potential safety issue (exothermic hydrogen-producing Be-water reaction or PbLi-water reaction in case of water leakage) when Beryllium/Beryllide is used as neutron multiplier or lithium lead is used tritium breeder in a water cooled blanket concept. The neutronics, thermal hydraulics and mechanical performance figures of the above-mentioned innovative blanket concept are presented. This work concludes with possible future work and outlook towards the development of innovative alternative blanket concepts for the European demonstration fusion power plants.

Mechanical assessment of first wall / blanket for DEMO

As a critical part of a breeding blanket of fusion reactor or DEMO, it is a common method to design and optimize the first wall separately before the mechanical evaluation of the fully integrated blanket. To understand the difference between the simplified first wall models and the fully integrated first wall/blanket model, the differences of the mechanical behavior between the simplified and integrated models are studied in this paper. The results provide a reference for the conceptual design and optimization of the first wall. Based on the HCCB (helium-cooled ceramic breeder) DEMO blanket model, the mechanical analysis of the integrated model was carried out, and the relatively accurate mechanical characteristics of the first wall for M-type and C-type damage were obtained. Considering the results of the integrated model as a basis, the mechanical analysis of the M-type and C-type for the 2 typical simplified model was carried out. For the separate first wall model, the maximum differences of M-type and C-type are -19% and -17%, respectively. For the most simplified model, the maximum differences of M-type are -17%, the differences of C-type are seriously deviated from the results of the integrated model. The results indicate that one quarter first wall model with reasonable constraints is closer to the results of the fully integrated FW/blanket model. The analysis results of one quarter first wall model can be used as a reference for the conceptual design of the first wall and can be utilized to optimize the first wall design based on the stress limits for prevention of the M-type and C-type damage modes.

Conceptual design progress of advanced fusion neutron source

Based on the results of the IFMIF/EVEDA project in the broader approach (BA) activities, the conceptual design of the Advanced Fusion Neutron Source (A-FNS) to obtain irradiation data of DEMO blanket materials is under consideration in Japan.This paper describes the progress of the A-FNS conceptual design, which consists of a 40 MeV, 125 mA deuteron beam, a liquid lithium target system, and a test and post-irradiation test facilities. the main objectives of the A-FNS are the acquisition of RAFM irradiation data up to 2035 and the tritium recovery test on the blanket. In addition, it is expected to be used for various neutron researches.The technical design activities of the A-FNS are focused on improvement of the lithium target design, review of impurity criteria, remote handling, and irradiation application plan.

Progress of water cooled ceramic breeder test blanket module system

A Water-Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) System is being developed as one of the most important steps toward DEMO blanket in Japan. As for the ITER-TBM program, the conceptual design of WCCB-TBM system has been approved by the ITER organization (IO). And two years have already passed after the start of the preliminary design activity. WCCB TBM Team had a concern about TBM box structure withstanding over 15 MPa of the coolant pressure, and decided to change the design as the result of study on structural integrity. Recently, the design change of the TBM structure has been approved domestically. So, WCCB TBM Team has applied the design change to the IO, and agreed with the IO to receive the conceptual design review which is limited the scope to the TBM structure. This paper provides an overview of the recent achievements of the development of the WCCB Test Blanket Module System in Japan.

Development and characterization of advanced neutron multiplier materials

Neutron multiplier materials are essential for self-sufficient tritium production and are closing the fuel cycle of fusion reactors. Until now, a concept of pebble bed consisting of interchanging layers of beryllium and lithium ceramic pebbles was considered for the Helium-Cooled Pebble Bed (HCPB) tritium-breeding module of the first experimental fusion reactor ITER as well as for the next demonstration fusion reactor DEMO. However, this concept depends on the availability of large amounts of pure beryllium pebbles and is also limited by its material properties like for example the tritium accumulation under irradiation.The results of tritium retention and analytical microstructural studies of beryllium pebbles obtained within the framework of the HIDOBE irradiation campaign suggest that a significant fraction of generated tritium (up to 100% below 500°C) is trapped within helium bubbles. Being negligible in the ITER tritium-breeding module (TBM), the total accumulated tritium inventory imposes severe safety issues and exceeds acceptable limits for the DEMO blanket. Therefore, advanced neutron multiplier materials such as beryllides have to be well characterized for their applicability in the HCPB blanket of DEMO and beyond. The usage of an advanced material with lower volumetric swelling, lower tritium retention, increased irradiation and chemical resistance as well as with higher melting temperature allows to switch from the pebble bed concept to a solid hexagonal block-based one. For the fabrication of titanium beryllide samples both the semi-industrial fabrication route utilizing the hot extrusion of rods and the industrial approach using vacuum hot pressing of a Be-Ti powder mixture were explored. In this contribution, we discuss reasons for the transition from pure beryllium to beryllides, respective changes of the HCPB blanket design and a successful demonstration of the feasibility of a beryllide block fabrication by an industrial method.

Conceptual design of DEMO blanket materials test modules for A-FNS

• A-FNS is a fusion-like neutron source for irradiation of fusion reactor materials. • We established conceptual designs of 3 test modules for DEMO blanket materials. • 1-BFMTM: Irradiation data for blanket functional materials are obtained. • 2-TRTM: In-situ tritium recovery properties for the materials are evaluated. • 3-ACPM: ACPs are evaluated under the fusion-like neutron irradiation. A conceptual design of A dvanced F usion N eutron S ource, A-FNS, has been conducted to achieve early realization of fusion-like neutron irradiation test for fusion reactor materials in Japan. A-FNS provides eight test modules to obtain irradiation data for fusion reactor materials. Conceptual design activities on Blanket Functional Materials Test Module (BFMTM), Tritium Release Test Module (TRTM) and Activated Corrosion Products Module (ACPM) were described among the A-FNS test modules. Also, basic concepts of the sub-system cells for the TRTM and the ACPM were also discussed.

Tritium release and retention in beryllium and titanium beryllide after neutron irradiation up to damage doses of 23-38 dpa

• For the first time, beryllium and titanium beryllide as massive pellets were irradiated in a nuclear reactor at temperatures of 710−1040 K up to production of 430–653 appm tritium and 4144–5992 appm helium, respectively. • The tritium retention in irradiated Be-7at.%Ti pellets is much lower than that in Be pellets, for example, 1 and 0.5 % to 48 and 34 % at 940 and 1040 K, respectively. • Open channels along grain and sub-grain boundaries are formed under neutron irradiation in beryllium, essentially contributing to the enhanced tritium release. Titanium beryllide is considered as an advanced neutron multiplier in the Helium Cooled Pebble Bed (HCPB) blanket of DEMO. Neutron irradiation of titanium beryllide together with beryllium as a reference material in material testing nuclear reactors can give essential data for the DEMO blanket design. Be-7at.%Ti (Be-7Ti) as well as Be were irradiated in the HFR, Petten, the Netherlands, at four temperatures of 710, 800, 940, 1040 K up to 23, 31, 36, 38 dpa, respectively. The post-irradiation examination (PIE) included thermal-programmed desorption (TPD) and Vickers hardness tests as well as microstructure study by optical metallography. Be and Be-7Ti pellets maintained their integrity after irradiation. Microstructure of Be-7Ti contains two phases, mainly, TiBe 12 , and also small amount of Be. Under irradiation, gas bubbles were formed in Be samples as well as in Be-phase in Be-7Ti samples. These bubbles contain helium and tritium produced in Be under irradiation. TPD tests showed a much lower tritium retention in Be-7Ti than in Be for all four irradiation temperatures. Vickers hardness of TiBe 12 phase is much higher than that of Be-phase. According to the obtained data, Be-7Ti could be considered more preferred than Be as a neutron multiplier material in future fusion reactors due to the enhanced radiation damage resistance.

Industrial-scale manufacturing experience of titanium beryllide block for DEMO blanket application

Abstract For the first time, the TiBe12 blocks have been manufactured by powder metallurgy using industrial equipment at the Ulba Metallurgical Plant. Such blocks will be used for neutron multiplication in the helium-cooled breeding blanket for the EU DEMO fusion reactor. The vacuum annealing of a cold-pressed composite of Be and Ti powders at 600–1275 °C approved the conditions for TiBe12 synthesis. After hot isostatic pressing at 1150 °C, a single-phase TiBe12 structure was formed, but the sample fragmented into several pieces. These beryllide pieces were used to manufacture TiBe12 powder. Vacuum hot pressing of the obtained powder resulted in a solid titanium beryllide workpieces of O150 mm × 170 mm. Electrical discharge machining and waterjet cutting were used to machine the external and internal surfaces in accordance with the design requirements. The obtained block has a density of 98.3 % of the theoretical density of TiBe12. Microstructural analysis revealed a bimodal grain size distribution with grains of 5–10 μm and 20–40 μm. Electrical conductivity and thermal expansion coefficient of TiBe12 are similar to other beryllides. To mimic the pulsed operating conditions in DEMO, a sample of titanium beryllide was rapidly heated to 900 °C and cooled with a rate of 12 K/s. After 50 cycles, the sample had no cracks or severe surface oxidation.

Effect of HIP at 800 and 900 °C on microstructure and properties of extruded Be-Ti composites

• Various beryllides are formed during HIP of the extruded Be-Ti composite. • TiBe 12 has a fine grained microstructure with a mean grain size of 190 nm. • Beryllides have a very high microhardness up to 1680 HV. • In the extruded composite, beryllium and titanium phases interact mainly in the temperature range of 670–740 °C. Hot isostatic pressing (HIP) has been proposed for manufacturing large hexagonal TiBe 12 blocks for neutron multiplication in the new reference design of the DEMO blanket. This paper investigates the effect of HIP at 800 and 900 °C on the microstructure and the properties of extruded Be-Ti composites with the main aim of optimizing HIP parameters. Be-Ti composites produced by powder extrusion at 650 °C consist exclusively of Be and Ti phases. During HIP, they interact forming titanium beryllides with a volume fraction of 70–79%. These beryllides have a very high microhardness of 1130–1680 HV. X-ray diffraction showed that the main beryllide phase is TiBe 12 . After HIP at 900 °C, it has a very fine-grained microstructure with a mean grain size of 190 nm. Auger electron spectroscopy revealed that the TiBe 2 beryllide forms a thin layer surrounding the remaining Ti phase. Ti 2 Be 17 can be found at the prior titanium phase locations in the form of small particles. The titanium and beryllium phase do not dissolve completely during the used HIP process. In addition, the beryllium phase exhibited a higher porosity after HIP. This results in densities as low as 94.4% and 95.9% of the theoretical density of TiBe 12 after heat treatment at 800 °C and 900 °C, respectively. Differential scanning calorimetry showed that beryllides are mainly synthesized in the temperature range of 670–740 °C, which can be monitored by the observation of a heat release. To accomplish diffusion processes in the Be-Ti composite, HIP should be carried out at temperatures exceeding 900 °C. Heating rates should be less than 10 K/min to avoid excessive overheating.

The issue of Tritium in DEMO coolant and mitigation strategies

A certain fraction of the tritium produced in DEMO blanket permeates into the primary coolant due to the high temperatures and the large metallic areas. From primary coolant, tritium can reach the working areas and the external environment via permeation and/or leaks. The mitigation of the tritium migration phenomenon can be accomplished by using appropriate anti-permeation barriers and by an opportune dimensioning of the coolant purification system. This work describes the technological status of the mitigation strategies under development for HCPB and WCLL blankets. The results of the tritium permeation analysis at system level reveal that the tritium permeation rate from blanket into primary coolant is much higher in WCLL than in HCPB. Therefore in HCPB the tritium concentration into primary coolant can be controlled exclusively with a Coolant Purification System (CPS) able to treat a by-pass of about 3 kg s−1. Two possible Helium-CPS processes are proposed and their pre-conceptual design is presented. Instead in WCLL due to the larger tritium permeation rate into coolant and the limit in water CPS size, the presence of appropriate coatings is required. Possible solutions are proposed and discussed based on the combination of feasible permeation reduction factors and water CPS size.

Conceptual design of test modules for DEMO blanket, diagnostic device, and RI production for A-FNS

In the conceptual design activity of advanced fusion neutron source A-FNS, a variety of test modules for fusion DEMO reactor are planned. In these modules, progresses of design activities on Blanket Nuclear Property Test Module (BNPTM) and Diagnostic and Control Device Test Module (DCDTM) were reported. The BNPTM is a module in order to evaluate accuracies of nuclear analyses of the DEMO blanket such as tritium production rate. The influences of the cooling water and test cell wall were evaluated in order to decide its design. The DCDTM is one to achieve irradiation data of functional materials on diagnostic and control devices such as mirror and window. It was clarified from its nuclear analysis that the neutron fluence obtained in the DCDTM was enough to investigate the accumulated irradiation effects on the materials. In addition to the irradiation tests on the fusion reactor materials, various neutron applications are planned at A-FNS. One of the applications is medical isotope production. We clarified that enough amount of the medical isotope molybdenum-99 could be produced in comparison with the demand in Japan by using Radio-Isotope Production Module (RIPM). The RIPM affects little influence on the fusion reactor material irradiation tests.