Development of water-cooled solid breeder test blanket module in JAEA

One of the most important missions of ITER is to provide a test bed for breeding blanket modules, which are called as test blanket module (TBM). JAEA has been extensively developing a water-cooled solid breeder test blanket module (WCSB TBM) for ITER. JAEA developed fabrication technology of F82H rectangular cooling tubes, and has successfully fabricated the near-full scale first wall mock-up of WCSB TBM by hot isostatic press (HIP) technique, which is fully made of F82H. The mock-up has been high-heat flux tested in the DATS facility in JAEA, which is an ion beam test facility. The inlet temperature of the cooling water is about 280 °C with 15 MPa, which is almost the same as the WCSB TBM design conditions. The mock-up has endured a heat load of 0.5 MW/m², 30 s for 80 thermal cycles. Neither hot spots nor thermal degradation have been observed.     

Japanese contribution to the DEMO-R&D program under the Broader Approach activities

Several technical R&D activities mainly related to the blanket materials are newly launched as a part of the Broader Approach (BA) activities, which was initiated by the EU and Japan. According to the common interests for these parties in DEMO, R&Ds on reduced activation ferritic/martensitic (RAFM) steels as structural material, SiC_f/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology are carried out through the BA DEMO R&D program, in order to establish the technical bases on the blanket materials and the tritium technology required for DEMO design. This paper describes overall schedule of those R&D activities and recent progress in Japan carried out by JAEA as the domestic implementing agency on BA, collaborating with Japanese universities and other research institutes.     

Inactive commissioning of a micro channel catalytic reactor for highly tritiated water production in the CAPER facility of TLK

In future DT fusion machines, several events will generate highly tritiated water (HTW). Among potential techniques for HTW processing, isotopic swamping in a catalytic membrane reactor (PERMCAT) appears promising. The experimental demonstration of PERMCAT for HTW processing has started in the CAPER facility at the Tritium Laboratory of Karlsruhe (TLK). Without any HTW source, such water has to be produced on purpose.Catalytic HT oxidation would ensure clean operation but could be critical for operation due to possible occurrence of explosive mixture. A tritium compatible micro-channel catalytic reactor (μCCR) has been designed and manufactured to produce up to 10 mL min^?1 of HTW with very high specific tritium activity (stoichiometric DTO: 5.2 × 10¹⁶ Bq kg^?1). Prior to its integration in CAPER for tritium operation, this reactor has been commissioned at different feed flow rates, gas composition (air or Helium), and temperature. The results demonstrate the good performances of the μCCR in producing water.The combination of the μCCR with the O₂ sensor represents a reliable system able to produce HTW in a safe way and without radioactive waste. Accordingly, the CAPER facility can be upgrade in order to continue the R&D activity on HTW processing with PERMCAT.     

Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.     

Processing highly tritiated water desorbed from molecular sieve bed using PERMCAT

Tritium handling facilities use molecular sieve beds (MSB) to collect and recover tritiated water. After reaching the capacity limit of the MSB, the water is desorbed and decontaminated in a water detritiation system (WDS). In the case of highly tritiated water (HTW) absorbed into a MSB, an inherent safe option for processing is necessary due to the HTW specific properties. Ideally, HTW should be processed immediately in a continuous mode. With this in consideration, the water desorption process from a zeolite bed was developed and optimized in a dedicated non active facility. The results of this experiments were applied into the regeneration of a MSB previously loaded with HTW containing an activity of 1.9 × 10¹⁴ Bq kg^?1. The water was desorbed, by step increasing the temperature bed and fed by helium carrier gas into the PERMCAT for detritiation and tritium recovery. The processed water was collected in a dry MSB downstream of the PERMCAT. These initial studies successfully demonstrate the viability of the process. The obtained results of the preliminary study and the subsequent tests with tritium, will provide useful information for the design of tritium processes relying on MSB, such as the water processing foreseen for the test blanket modules in ITER.     

Overview of tritium safety technology at the Tritium Process Laboratory of JAERI

The Tritium Process Laboratory of the Japan Atomic Energy Research Institute is the only laboratory in Japan where grams of tritium can be handled to carry out R&D on tritium processing and tritium safety handling technologies for fusion reactors. The tritium inventory is approximately 13 grams. Since 1988, basic research has been performed using gram-level tritium quantities. During the past 5 years, approximately 1 kilogram of tritium has been handled in experimental apparatus. The total amount of tritium released through the stack of TPL was controlled to less than 1 Ci without any accidents. In order to establish more complete tritium safety for DT fusion reactors, main R&D areas on tritium safety technology at TPL were focused on a new compact tritium confinement system, reliable tritium accounting and inventory control, new tritium waste treatments, and tritium release behavior into a room.     

TBR comparison of water-cooled blanket based on PWR and SCWR water conditions

Nuclear analysis results were compared for water-cooled blanket based on PWR (pressurized water reactor) and SCWR (sub-critical water reactor) water conditions. The local TBR (tritium breeding ratio) in outboard zone was discussed in the range of P_n (neutron wall load) from 1 MW/m² to 5 MW/m². It was found that water fraction has little impact on TBR, which is an important factor related to blanket tritium efficiency. It indicated that TBR value of each P_n would be similar under the two kinds of water conditions, but PWR case is a little higher than that of SCWR's. In addition, it was found that beryllium is the dominant factor leading a higher TBR inside blanket. As a result, TBR is an insensitive value with the water condition variation. The results would be important to water condition choice for solid blanket in the future.     

Metallurgical analysis of corroded bellows of bellow-sealed valve in lithium flow system

Liquid lithium has been one of the candidates of the tritium breeder and possibly a coolant for the blanket of fusion reactors. The observation of corrosion behavior was conducted on the bellows of the bellow-sealed valve used in a lithium circulation loop at 350–500 °C for about 1500 h. The results were obtained from observation of the surfaces and cross sections. The bellows was found to be fractured and detached. Selective elements were depleted on the surface.     

Measurement of tritium concentration in water by imaging plate

Concentration of tritium in water (4–400 kBq cm^?3) was measured by exposing an imaging plate without protection layer (Fujifilm, BAS-IP TR) to vapor for 2–48 h. It was found that tritium gradually penetrated into Eu-doped BaFBr phosphor and induced sufficiently intense photostimulated luminescence (PSL) even at the concentration of 4 kBq cm^?3. The intensity of PSL was proportional to tritium concentration in water. In addition, tritium absorbed in phosphor was reversibly released by keeping IP in air, and IP was able to be used repeatedly if total duration of exposure was ca. 24 h or less. The contamination of IP with tritium was not serious. It was concluded that IP technique has potential to measure tritium concentration in water without direct handling of tritiated water and with a minimum amount of radioactive waste.     

Progress of fusion nuclear technologies in the broader approach framework

In the Broader Approach framework, the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project, the International Fusion Energy Research Center (IFERC) project, and the Satellite Tokamak project are implemented. In the IFMIF/EVEDA project, engineering design of IFMIF and engineering R&D include the construction and tests of an IFMIF prototype accelerator system up with a 9 MeV and CW deuteron beam, a liquid lithium test loop with free surface flow, and full scale irradiation test module including temperature control instrumentation. The commissioning of the EVEDA lithium test loop was completed in March 2011, and a lithium flow of ～5 m/s was obtained. As a part of the IFERC project, R&Ds on reduced activation ferritic/martensitic steels as blanket structural material, SiC_f/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology are carried out. At the beginning of 2011, the integrated DEMO design team was established among the IFERC project team and EU/JA home teams, where the design criteria, other design basis are discussed as an initial work. A high performance supercomputer with the peak performance of 1.3 Pflops is under installation at the Rokkasho BA site.     

Conceptual design of a water cooled breeder blanket for CFETR

China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by ⁶Li(n,α)T reaction. Li₂TiO₃ pebbles and Be₁₂Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li₂TiO₃ and Be₁₂Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be₁₂Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into account to replace Be plate in viewpoint of safety. In this contribution, study on neutronics and thermal design for a water cooled breeder blanket with superheated steam is reported.     

Status of Indian LLCB TBM program and R&D activities

The Indian LLCB TBM, currently under development, will be tested from the first phase of ITER operation (H–H phase) in one-half of the ITER port no-2. The present LLCB TBM design has been optimized based on the neutronic as well as thermal hydraulic analysis results. LLCB TBM R&D activities are primarily focused on (i) development of technologies related to various process systems such as Helium, Pb–Li liquid metal and tritium, (ii) development and qualification of blanket materials viz., structural material (IN-RAFMS), tritium breeding materials (Pb–Li, and Li₂TiO₃), (iii) development and qualification of fabrication technologies for TBM system. The present status of LLCB TBM design activities as well as the progress made in major R&D areas is presented in this paper.     

Requirements for helium cooled pebble bed blanket and R&D activities

This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R&D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM.The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R&D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R&D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.     

Thermal-hydraulic design of water-cooled pressure tube blanket for a fusion driven subcritical reactor

A Water-cooled Pressure Tube Energy production blanket (WPTE) for fusion driven subcritical reactor has been designed to achieve 3000 MW thermal power with self-sustaining tritium cycle. Pressurized water has great advantages in energy production; however the high pressure may cause some severe structural design issues. This paper proposes a new concept of water-cooled blanket. To solve the problem of the high pressure of the coolant, the pressure tube was adopted in the design and in the meantime, the thickness of the first wall can be significantly reduced as result of adopting pressure tube. The numerically simulating and calculating of temperature, stress distribution and flow analyses were carried out and the feasibility of using water as coolant was discussed. The results demonstrated the engineering feasibility of the water-cooled fusion–fission hybrid reactor blanket module.     

Effect of outside tritium source on tritium balance of a D-T fusion reactor

Attainable tritium breeding ration in the blanket system must be larger than the required breeding ratio when no effective tritium resources from outside are expected. It is revealed recently that a considerable amount of tritium can be trapped to the re-deposition layer of the first wall materials and that the time constant of this phenomenon is rather long. Then, the tritium breeding ratio around 1.1 is required in the blanket system when 3 years is claimed for the tritium doubling time to prepare tritium for the initial inventory of a next reactor. Construction of an outside tritium supply is one of the possible ways to compensate the lack of tritium because it is generally considered that the attainable tritium breeding ratio in the solid breeder system is around 1.05. It is reported recently that a high-temperature gas-cooled reactor can produce 10 kg of tritium per year. The preferable amount of tritium production rate of the outer tritium supply is discussed in this study from the viewpoint of tritium balance in a D-T power reactor.     

Development of a technical scale PERMCAT reactor for processing of highly tritiated water

In view of future fusion rectors fueled by deuterium–tritium mixtures, highly tritiated water (HTW) of up to 5.2·10¹⁶ Bq kg^?1 will be produced, during routine operation and scenarios as an accidental release of tritium into a glove box. Also in the solid breeder blanket concept, a non-negligible fraction of the tritium produced will be in the tritiated water fraction. To decontaminate HTW the PERMCAT using isotope swamping in a Pd/Ag membrane reactor has been identified as a robust and reliable solution. In order to investigate the decontamination of HTW at flow rates relevant for future fusion power plants, a technical scale, fully tritium compatible PERMCAT consisting of a bundle of finger-type membranes inserted in a single catalyst bed was developed. Nevertheless, it represents only one part of a PERMCAT cascade necessary to achieve the required performance to process HTW on technical scale. By improving the existing PERMCAT geometry using experimental data obtained from isotopic exchange between D₂O and H₂, the performance of the existing PERMCAT reactor was optimised. Based on the optimised geometry a new fully tritium compatible technical scale PERMCAT cascade comprising of two PERMCAT reactors in series was designed, manufactured and commissioned as presented in this paper.     

Evaluation of tritium diffusion through the Neutral Beam Injector calorimeter panel

The Neutral Beam Test Facility (NBTF) to be realized in Padoa will test the Neutral Beam Injection (NBI), one of the Heating and Current Drive Systems foreseen for ITER. The NBI is based on the acceleration of hydrogen or deuterium negative ions up to 1 MeV. This work has been aimed at assessing the tritium release from the NBTF in order to provide data for the safety analysis. In particular, the diffusion of the tritium through the neutral beam target material (the CuCrZr alloy calorimeter panels) has been assessed by using literature data of the diffusion coefficient. The tritium generated inside the calorimeter panels moves into both the vacuum and water side: the tritium diffusion flux has been evaluated during the beam-on (200 °C) and the beam-off (20 °C) phases of the NBTF experiments consisting of an interim campaign and a final test. The penetration depth of the tritium through the 2 mm thick CuCrZr alloy material has been also evaluated by using a Monte-Carlo code. As main result, the assessed diffusion flux of tritium during both the beam-on and the beam-off phases are modest. In fact, at the end of the interim campaign (100 days), about the 96% of the all generated tritium (626.5 MBq) exits the calorimeter while the residual tritium inventory (25 MBq) leaves the copper alloy with a diffusion time of about 1 month. At the end of the final test (14 days) about the 99% of the total generated tritium (1.023 × 10⁴ MBq) leaves the copper alloy and the remaining tritium inventory (152.2 MBq) is released by about 32 days. In both the interim campaign and the final test, more than the 99% of the total tritium is transferred into the vacuum side of the calorimeter panel while negligible tritium amounts enter the cooling water system thus showing a very low impact on the environment.     

Progress and achievements of the BA activities in Rokkasho

In Rokkasho Japan, the International Fusion Energy Research Center (IFERC) project and the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project are on going under the Broader Approach framework. The IFERC project consists of three sub-projects; a fusion demonstration reactor (DEMO) Design and R&D Coordination Center, a Computational Simulation Center (CSC), and an ITER Remote Experimentation Center (REC). DEMO Design activity has been conducted by the IFERC project team in Rokkasho and home teams in EU and JA. In the DEMO R&D activity, five R&D tasks mainly of the blanket materials are carried out intensively. A supercomputer with 1.23 Pflops of LINPAC performance has been installed in December 2011, the operation started in January 2012. Discussion of overall plan of REC has started in 2012 between EU and Japan. In the IFMIF/EVEDA project, an IFMIF prototype accelerator system up to 9 MeV with 125 mA CW deuteron beam will be installed and tested in Rokkasho. Major components of the accelerator are under development or fabrication in EU. The first component of the accelerator, an injector with an ECR ion source, will be delivered to Rokkasho in March 2013.     

Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.     

Overview of TBM R&D activities in India

In India, development of Lead–Lithium Ceramic Breeder (LLCB) blanket is being performed as the primary candidate of Test Blanket Module (TBM) towards DEMO reactor. The LLCB TBM will be tested from the first phase of ITER operation (H-H phase) in one-half of an ITER port no. 2. The Indian TBM R&D program is focused on the development of blanket materials and critical technologies: structural material (IN-RAFMS), breeding materials (Pb–Li, Li₂TiO₃), development of technologies for Lead–Lithium cooling system (LLCS), helium cooling system (HCS), tritium extraction system (TES) and TBM related fabrication technologies. This paper will provide an overview of LLCB TBM R&D activities under progress in India.