Test Blanket Module Research Articles

Systematic approach to identify the dominant neutron leakage paths in the WCCB-TBM port

In the design activity of fusion reactors, neutronic analysis is one of the most necessary analyses. In this paper, the systematic approach to estimate the paths of neutron transport is introduced. In Monte Carlo code MCNP, users can set 'kill option' to arbitrary region, and then, the neutrons going into the region have their histories in MCNP terminated. With the combination of the calculations which, users set the kill options to different regions, the neutron transport paths can be estimated. In this paper, this approach is applied to the #18 equatorial port of ITER with Water Cooled Ceramic Breeding (WCCB) Test Blanket Module (TBM), and the shutdown dose rate (SDDR) at Pipe Forest (PF) due to the activated components is calculated. The result shows the main contributor for SDDR is not the neutrons going through TBM, and even if TBM has ideal shielding performance, SDDR can't be reduced less than half, and the most effective way to reduce SDDR at PF is to add the surrounding shield to PF. Due to the additional shield, SDDR can be reduced from 180 μSv/h to 79 μSv/h, and this is under the target value 100 μSv/h.

Research on thermal creep behavior of Chinese low-activation ferritic/martensitic steel

A series of thermal creep experiments are conducted on Chinese low-activation ferritic/martensitic steel (CLF-1 steel) at a temperature of 550, 575, and 600 °C with various normal stress (from 150 to 300 MPa), respectively. The results are analyzed by creep curves, microstructure, and creep properties. Finished samples display adequate fracture surface in short creep rupture tests (within 3000 h). The comparative creep curve shows no difference related to the thickness of CLF-1 steel plate (10 and 30 mm). Microstructure images before and after creep reflect a tempered lath martensite structure and carbides decorating both in prior austenitic grain boundaries and lath boundaries. Comparison is discussed between 30 mm plate and 10 mm plate. All current results reflect that all CLF-1 steel samples shows good creep behavior and proves that a series excellent progress of CLF-1 steel R&D work has been successfully achieved. Besides, collected data will contribute a lot to the design and R&D work of the Chinese Helium Cooled Ceramic Breeding Test Blanket Module (CNHCCB-TBM).

Towards the simulation of MHD flow in an entire WCLL TBM mock-up

In the frame of the EUROfusion breeding blanket research activities, the water cooled lead lithium (WCLL) blanket is considered as reference liquid metal blanket design to be tested in ITER and to be used in a DEMO reactor. For the study of pressure and flow distribution in a WCLL blanket module, magnetohydrodynamic (MHD) phenomena, which result from the motion of the electrically conducting breeder in the intense plasma-confining magnetic field, have to be taken into account. They are due to the induction of electric currents that generate strong electromagnetic forces, which modify the velocity distribution in the blanket compared to hydrodynamic conditions and increase pressure losses. In the WCLL test blanket module (TBM) for ITER a basic geometry, consisting of a rectangular box, representing the breeding zone, is repeated along the poloidal direction. The manifold that distributes and collects the liquid metal features two long poloidal ducts, electrically connected across a common wall. The final goal of the present work is to simulate liquid metal MHD flows in strong magnetic fields in a complete column of a WCLL TBM including poloidal manifolds and 8 breeder units. With this aim, the complexity of the model geometry has been progressively increased and various topologies of computational grids have been considered. First simulations have been performed in a single breeder zone connected with a portion of manifolds in order to test the suitability of the generated grid and the performance of the numerical code for this type of problem. In a second step, the MHD flow at moderate magnetic fields in an entire TBM mock-up has been simulated to get pressure distribution along the manifolds and flow partitioning among breeder units.

The CHIMERA facility development programme

The CHIMERA fusion technology facility will enable testing of large in-vessel component modules under reactor-like conditions of combined in-vacuum thermal power density and magnetic field. With an integral large superconducting magnet and a PWR-like water loop, CHIMERA is also ideally placed for experiments on liquid metal breeding blanket prototypes. Facility construction is underway at the UKAEA site in South Yorkshire. The superconducting magnet is fully wound and terminated, and the bespoke pulsed magnet power supply has been delivered. Even before construction and commissioning is complete, a parallel and supporting research and development programme is in progress and is reported here. A bespoke infrared heating system has been developed, capable of applying 0.5 MW/m2 to component surfaces up to the size of the ITER test blanket module first wall. The modules of this heater are highly specialised and designed to endure the high magnetic forces from the CHIMERA static and pulsed magnets. CHIMERA will feature a range of diagnostics, including load cells to measure static and pulsed magnetic forces, and induced current sensors, all of which have been tested to confirm acceptable operation in the pulsed magnetic field. Manufacture of the test mock-up to be used for facility commissioning is underway. Testing will be enhanced with simulation ‘digital twin’ capability, and a development project is now producing first virtual test results, informing commissioning of the CHIMERA device.

An update on the latest and most relevant improvements achieved in the European helium-cooled ceramic breeder TBM design

The European TBM (Test Blanket Module) program under development will test two alternative concepts for tritium generation and extraction in ITER: a helium cooled / ceramic pebble breeder concept and a water cooled / lithium lead breeder concept.The former implements Advanced Ceramic Breeder pebbles that act as tritium breeding material, and Beryllium pebbles as neutron multiplier. A Reduced Activation Ferritic Martensitic steel, the EUROFER-97 (X10CrWVTa9-1), is used as structural material and pressurized Helium technology is implemented for efficient heat extraction (300-500°C, 8 MPa).The design of the TBM box involves challenges that must be tackled with a holistic approach to simultaneously consider a variety of aspects. Some of these challenges include: the thermal performance and the structural integrity of the TBM under ITER normal operations and accidental conditions, the implementation of pressurized equipment regulation(s), the convergence with manufacturing activities being developed in parallel and the high congestion in the reduced TBM volume –especially when considering instrumentation integration– without affecting too much the tritium breeding capability. It is also key to develop advanced methodologies based on design by analysis to virtually test alternative design solutions in an agile manner. These should be linked to the applicable nuclear design code (RCC-MRx) to consolidate design choices. In this way, a traceable evolution of the TBM design process can be built where alternatives are tested, compared and either discarded or selected in a consistent and comprehensive framework.This paper presents an overview of the very significant progress achieved in the design of the helium cooled concept TBM over the last two years, a period that corresponds to the first steps of the preliminary design phase towards the Preliminary Design Review (PDR) gate. This includes a significant change in the design of the back manifold area and the connection with the rest of the TBM-Set that solves long-standing historical problems inherited from pre-conceptual design stage. An update on the latest and most relevant methodological developments achieved is also provided in this paper, since these have been key to progress with the design evolutions in a sound and traceable manner. To end, the paper also presents some examples of the instrumentation definition and its complex integration within and outside the TBM box.

ITER Test Blanket Module—ALARA Investigations for Port Cell Pipe Forest Replacement

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.

Evaluation on Nonirradiated Thermo-mechanical Response of WCCB TBM Following RCC-MRx

Since the submodules as the most important structural component in Water-Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) contributes to tritium breeding and extracting, structural integrity assessment is necessary to keep normal operation and function of WCCB TBM-set. Recently, the nonirradiated material properties of F82H which is considered as the structural material of submodule has been summarized following the requirement of RCC-MRx, combined with EU directive, French order, and pressure vessel construction code (RCC-MRx), structural integrity assessment is investigated under the ITER relevant thermo-mechanical load in this study. Meanwhile, cooling water conditions of WCCB TBM-set with its operation scenario, relationship, and detailed values (temperature and pressure) are constituted. Based on the determined structural integrity assessment procedure from RCC-MRx and numerical analysis results, structural integrity of current design of submodule in WCCB TBM-set could be validated. According to the current achievements and established structural integrity assessment procedure, further evaluation on structural integrity combined with electromagnetic load shall be investigated.

Innovative Isostatic Processing Technologies for Defence and other Strategic Applications

Powder Metallurgy (P/M) involving Cold Isostatic Pressing (CIP) and Hot Isostatic Pressing (HIP) techniques is the best suited processing route for manufacture of critical components from ceramics and highly complex metallic materials. The CIP followed by sintering route was adopted for the development of fused silica radomes for application in high-speed target seeking missiles. On the other hand, the HIP, a single-step processing technique has been used for the development of First-Wall (FW) of Test Blanket Modules (TBMs) from India Specific Reduced Activation Ferritic Martensitic (INRAFM) steel. The FW is an important component required for International Thermo-nuclear Experimental Reactor (ITER) application.The details of innovative processes which have been established to overcome the specific technical challenges for the development of CIP and HIP technologies for manufacture of silica radomes and INRAFM steel first-wall respectively for intended applications are highlighted in this paper.

CHIMERA Fusion Technology Facility: Testing and Virtual Qualification

The Combined Heating and Magnetic Research Apparatus (CHIMERA) fusion technology test facility is under construction. The facility will be uniquely capable of semi-integral testing of fusion materials and component modules up to the size of the ITER test blanket module box, under combined conditions of in-vacuum high heat flux, static and pulsed magnetic fields, and high-temperature/high-pressure water cooling. This paper reports the high-level capabilities of the CHIMERA baselined design and the planned program of testing and describes the proposed strategy for use of simulations for virtual testing, qualification, and in-situ monitoring. The first step in testing of a component mock-up is to take data from as-built geometry and other measurements and transmit them to an integrated computational model that can closely mimic the physical asset and form a digital replica. Not only can this digital replica be queried in advance of physical testing in the facility, allowing optimization of the test program, but combined with subsequent test data, it also can deliver much greater insight into experimental results than can be obtained using test data alone. The digital replica is used as the basis for a digital twin, which is live coupled to the running experiment, and is under development as a proposed key facet of fusion reactor surveillance in-service. Physical mock-ups for testing can be subjected to in-vacuum heat flux up to 0.5 MW/m2 over the entire surface while within a strong horizontal magnetic field. The central field can be up to 4 T with a peak in the test region of 5 T. The same component mock-ups can also be subjected to repeated magnetic field pulses with ramp rate 12 T/s, which can simulate loading conditions of a plasma disruption. Facility upgrades are underway to include a liquid metal circulation loop to allow the study of magnetohydrodynamics effects and to add a high-heat-flux system using a very high-power continuous-wave laser to achieve divertor-relevant heat fluxes of 20 MW/m2 over the area of a small-scale mock-up. Four examples are given to illustrate the physical testing program that is currently foreseen.

Design of a Prototypical Mock-Up for the Experimental Investigation of WCLL First-Wall Performances

A large research effort is currently ongoing within the framework of the EUROfusion consortium for the study and design of a water-cooled lithium–lead (WCLL) breeding blanket (BB). This concept will be tested in ITER through the installation of a test blanket module (TBM) and it is one of the two candidates adopted as driver BBs in DEMO. In this framework, at the ENEA research centre of Brasimone, the realization of the experimental platform, W-HYDRA, is envisaged. The platform is dedicated to the support of the development of WCLL BB and ITER TBM and the investigation of the DEMO balance of plants. One of the most important experimental infrastructures is the water-loop facility, the aim of which is to provide water at a high pressure and temperature (PWR conditions), with a sufficient mass-flow rate and power for the experimental testing of BB and TBM components. The facility will be equipped with a vacuum chamber and an electron beam gun for the reproduction of high surface heat flux on plasma-facing components. In the present work, the design of a prototypical mock-up (MU) of the WCLL BB first wall is described. The MU is used to investigate the thermal, hydraulic and structural behavior of the current first-wall design under relevant heat loads at the expected operational conditions. The delineation of the main experimental test’s features and the instrumentation needed is assessed in the paper. A preliminary CFD calculation on the prototypical MU and the computational results are also presented.

Overview of recent ITER TBM Port Plug R&D activities

The objective of ITER project is the demonstration of the scientific and technological feasibility of fusion energy. Tritium fuel is required for the fusion reaction in DEMOnstration reactors and future commercial reactors but it is not available in nature because of its instability. To investigate tritium breeding self-sufficiency, two equatorial ports in ITER are dedicated to operating test blanket modules (TBMs) using the TBM Port Plug (TBM-PP). The TBM-PP design should be investigated with the physical mock-ups to validate the leak tightness in the primary vacuum boundary and to ensure TBM replacement and refurbishment compatible with Remote Handling (RH) operations in the ITER Hot Cell Facility during the ITER lifetime. This paper summarizes the achievements on the two most relevant R&D topics as described herewith. Several TBM-PP components (i.e., back parts of the TBM-Frame, the Dummy-TBMs, the TBM-Shields, and the TBM feeding pipes) are part of the vacuum boundary implying that these components are Vacuum Quality Classification (VQC) 1A. To test and develop high vacuum sealing for the TBM-Frame flange and the TBM-Set or Dummy-TBM flange, the ITER Large Seal Test Rig (LSTR), and metallic gasket seals for the TBM application have been manufactured and installed in the ITER site. The helium leak tests were performed at different operating temperatures to investigate the vacuum performance of metallic gasket seals with VQC 1A. The TBM-PP design shall guarantee the feasibility of a rapid replacement and refurbishment compatible with ITER Remote Handling (RH) operations by using RH tools in the Hot Cell Facility during the ITER's lifetime. Therefore, an experimental program has been performed to demonstrate the feasibility of the critical RH refurbishment tasks including the insertion and removal of TBM-Set/Dummy-TBM and the RH bolting.

Status of Scoping Nuclear Analyses for the Evolving Design of ITER TBM Port Cells

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.

Verification and validation by eXtended reality simulations of Test Blanket Modules replacement operations in Air-Fed suits

To study the feasibility of the Test Blanket Modules (TBM) replacement operations in ITER Port Cells, CEA develops a mixed (digital & physical) validation by simulations approach with immersive technology to verify hands-on operations. Main validation criteria concern operator accessibility, posture, detailed procedure sequence and time spent at workstations. Such complex replacement operations will assume series of hands on tasks for both normal and abnormal situations, sometimes with protection such as Air-Fed Suit (AFS). After initial validation of key operation feasibility on digital mock-up, the goal is to optimize the process with minimization of occupational radiation exposure under the As Low As Reasonably Achievable (ALARA) requirement.Virtual reality tools provide relevant means to develop a credible replacement scenario in a context of complex configurations. The process includes features such as ergonomics assessment tool coupled to multi operators’ collaboration, tangible tools and integration of advanced real time physics engine. Lessons learnt from the field of existing CEA facilities enable identifying key parameters related to workstation Humans factors. This first set of observations would help to define methodology and criteria for qualification of operation in AFS.The Digital simulation validation is deployed with a nuclear operator avatar in AFS to demonstrate Nuclear Maintenance Operation feasibility considering study of arduousness, accessibility and visibility thanks to real time motion capture. The recording of the whole simulation allows tracing and replaying for post-analysis data including body angulations, clash detection and operator field of view.The integration of this approach at an early phase of the design process enables validating technical options for replacement scenario. This leads to promising and beneficial results concerning design and integration of complex components in constraint working environments.This paper presents a review of the CEA studies regarding the extended virtual reality capabilities and results to support design TBM integration and replacement scenario within a nuclear environment framework.

ITER Test Blanket Module—Engineering Investigations for Port Cell Pipe Forest Replacement

The ITER Test Blanket Modules (TBMs) are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each of them includes two TBMs. After each plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs by new ones during the ITER long-term shutdown (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 called pipe forest (PF). The TBS equipment in PC shall be designed so that occupational radiation exposure can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER Policy. To implement ALARA process requirements, a design shall consider careful integration assessment since early phase to address all engineering aspects of the replacement sequence. The design of the TBS equipment in PC shall optimize the reconfiguration operation by including human factor considerations and providing adequate space. Provisions for dose reduction measures (DRMs) considering protection means and minimization of occupancy times in radiation fields are also investigated by assessing remote technics to reduce operator exposure. This article describes engineering investigations for optimized PF replacement operations in various engineering fields. Results on the design of the TBS equipment in the TBM-PC are discussed. The PF design shall implement nuclear engineering aspects. Design options shall demonstrate constructability and maintainability meaning compliance with ALARA process and adequate construction codes (RCC-MRx, EN 13480). TBM-PP replacement sequence drives several elements of the TBM-PC component design and associated integration. Concepts are discussed with regard to nuclear engineering and operation aspects. Use of prototyping demo, including virtual reality (VR), will support the preliminary validation during the system engineering process in order to prepare the full-scale mock-up validation phase.

Mixed hydrogen isotopes plasma-driven permeation through CLF-1 RAFM steel for ITER HCCB TBM

The first wall of the helium-cooled ceramic breeder test blanket module (TBM) will be subjected to charge exchange neutral from the mixed deuterium (D) and tritium (T) deuterium–tritium (D–T) mixture plasma irradiation. To understand the isotope effects on mixed plasma-driven permeation (PDP) through the TBM wall, which is typically made by reduced activation martensitic/ferritic steel (RAFM), mixed D and protium (H) PDP experiments for a Chinese RAFM CLF-1 have been performed. Permeation fluxes of H and D through ∼mm thick CLF-1 have been recorded at various sample temperatures and particle incident energies. The co-permeation of H and D is found to be diffusion-limited in the bulk and obeys classical mass effects of diffusion species. The H/D ratio of the permeation fluxes is found to be stable when changing sample temperature and is insensitive to ion incident energy. In addition, the mixed H can provide more channels for D recombination to reduce the D permeation.

Ambient, elevated temperature tensile properties and origin of strengthening in friction stir welded 6 mm thick reduced activation ferritic-martensitic steel plates in as-welded and post-weld normalised conditions

Reduced Activation Ferritic-Martensitic (RAFM) steels are currently being considered for test blanket modules of International Thermonuclear Experimental Reactor. This study is aimed at understanding the microstructure and mechanical properties of Indian RAFM steel during friction stir welding (FSW) of 6 mm thick plates. The full penetration bead-on-plate welds were fabricated employing PcBN tool at a rotational speed of 200 rpm. The FSW joint consisted of stir zone, thermomechanically affected zone and BM. Microstructure, hardness and tensile properties were evaluated in as-received (base metal, BM), as-welded and post weld normalised + tempered states. The as-received microstructure was composed of Cr-rich M23C6 on prior austenite grain and tempered lath martensite boundaries with intra-lath V-and Ta -rich monocarbides. Substantial changes occurred during welding leading to destruction and dissolution of M23C6 and precipitation of Fe3C in SZ. The microstructure was restored to the BM level by giving a post weld normalising and tempering treatment. Comparative study on tensile properties has been carried out at room and elevated temperatures (up to 550 °C) in all the three states and the observed variations have been explained on the basis of initial microstructure and evolving substructures. The as-welded samples showed higher strength and low percentage elongation values. The ductility was minimum in all the states at an intermediate temperature and ascribed to the dynamic strain ageing. All the states revealed falling strength with increasing temperature up to 550 °C. Failure occurred in AW and PWNT states at the interface between TMAZ and BM. Irrespective of material state and tensile test temperature, the transgranular ductile fracture prevailed. The contribution of the different strengthening mechanisms to the yield strength of the SZ region has been estimated theoretically and it matches with the experimental results. The influence of low angle grain boundaries (LAGB) and high angle grain boundaries (HAGB) on the tensile behaviour of the SZ was well explained using the Electron Back Scattered Diffraction (EBSD) maps of these regions.

Scoping Nuclear Analyses of Shielding Options and Shutdown Dose Rate Contributions in ITER TBSs

One of the advances in the test blanket module program within the ITER project in the last few years concerned the evolution of the pipe forest (PF) and bioshield plug (BP) designs. In support of the design phase, nuclear analyses to assess several responses in the fusion neutronics environment inside the port interspace (PI) with the existence of the evolved PF and BP are deemed essential. Nuclear analyses were commenced using the new PF and BP with developing the neutronics models and performing preliminary assessment of the radiation fields and shutdown dose rate (SDDR) in the PI. In this paper, the results of a full suite of nuclear analyses are discussed, which covers more configurations and radiation sources, in two plasma operational modes: on and off. For the plasma-on mode, different shielding options were examined. The results show a clear benefit of combining the installation of shielding panels on the PF enclosure with those in the BP “dogleg,” through which the pipes penetrate to the port cell area. For the plasma-off mode, the SDDR was assessed from different sources: activated components and residual LiPb layers in pipes after drainage. As maintenance operations are foreseen during the lifetime of the facility, the SDDR was also assessed for access conditions, open BP doors, and transport conditions, with PF extracted in the gallery.

Preliminary assessment of tritium permeation and retention in the European Water Cooled Lithium Lead Test Blanket Module with TESSIM-X

• T inventory in the ITER EU Water-Cooled Lithium-Lead Test Blanket Module investigated. • T from vacuum vessel (gas, neutrals) and solute T in PbLi included in modeling. • Effect of neutron damaging on trap concentrations included in models. • T retention in the range of 0.75 g is expected over plasma operation period. • Comparison with the TMAP7 code performed. Tritium self-sufficiency is a key requirement for future fusion power plants. Therefore, it will be necessary to minimize tritium losses to the surrounding systems, such as the tritium breeding modules, plasma-facing components, cooling system, etc. These components and systems will act also as a tritium sink, as tritium will be retained in the metal walls in traps produced during manufacturing or induced by neutron irradiation. The design of the tritium breeding systems and plasma-facing components will therefore have a direct impact on the performance, operational safety and cost of any future power plant. Consequently, accurate modeling of tritium transport and retention in these systems is needed for future engineering designs of the tritium breeding systems and supporting safety requirements. In this work, tritium permeation and retention was modeled with the diffusion-transport code TESSIM-X for the current European WCLL (Water Cooled Lithium Lead) Test Blanket Module (TBM) for the ITER TBM Programme, with the inclusion of neutron-induced traps. Trap saturation is considered for dpa values of 0.25 and above, leading to an increase in trap concentrations relative to undamaged material of roughly a factor 3. The code was also benchmarked against the well-established TMAP7 code.

Practical considerations in developing nuclear detectors for tokamak harsh environments

Fusion tokamaks (FT) and hybrid fusion-fission reactors (HFFR) present harsh working conditions characterized by intense neutron and gamma fluxes (>1012 cm−2 s−1), high working temperatures (up to 600 °C) and corrosive environment. The breeding blanket region (BB) of these plants are resulting very hostile to the detectors used to monitor/measure fundamental nuclear parameters such as neutron/gamma fluxes and energy spectra, and tritium production. Presently no detectors are ready for being hosted in the harsh environment of the BB and R&D activity is needed to develop and test the candidate detectors. Some important lessons can be learned from past activities carried out in the EU and devoted to studying and realizing nuclear detector prototypes for the European Test Blanket Modules (TBM) of ITER. Amongst the other, these studies pointed out the need for intense neutron fields and calibration facilities closely reproducing the expected working environments to be used for reliable testing and calibration of the prototypes.Accurate simulation by Monte Carlo technique of the proposed detectors allows to mimic and foresee the response and performances of the detectors pointing out several fundamental and critical aspects on the physical response of the detector so helping in understanding the detectors response. This can help in selecting the best performing detector. The selection is based upon a multi-step procedure.The lesson learned for ITER-TBM can be helpful to study and develop nuclear detectors to be used in HFFR reactors and in next fusion machines like DEMO this because, despite the difference, the ITER-TBMs and the BB of fusion devices and HFFR reactors experience a number of similarities in terms of radiation level, temperature and nuclear quantities to be measured.In this paper, after discussing the requirements to be fulfilled by the nuclear detectors that must operate in the harsh environments we will discuss an example of detector development by considering the case of a self-power neutron detector (SPND) with chromium emitter studied and developed for ITER-TBM. The detailed Monte Carlo analysis is also reported and the many issues not yet solved are highlighted and the possible follow up to HFFR instrumentation discussed.

Heat treatment effects on the weld joint of CLF-1 fabricated by laser welding

In this paper, laser weld technology was applied to produce the welded joints of CLF-1 steel, which is the primary material for manufacturing test blanket module (TBM). The microstructures and mechanical properties of the joints were studied at as-welded state, post-weld heat treatment (PWHT) of 710 °C/2 h/air cooling (PWDT), and PWHT of 980 °C/1 h/air cooling (AC) followed by tempering at 710 °C/2 h/AC (PWNT). The results showed that abundant lath martensite phase with width 800 nm and small amount of δ-Fe phase were present in the weld at as-welded state. Its impact toughness was about 40 J. The microstructures of the weld at PWDT state included a large amount of tempered lath martensite with width 650 nm, few δ-Fe, M23C6 carbide (distributed along the boundary of martensite) and MX carbide (precipitated inside the martensite). Its impact toughness was about 237 J (equivalent to the base metal). The microstructures of the weld at PWNT state included a large amount of tempered lath martensite with width 350 nm, M23C6 carbide and MX carbide. Even though the δ-Fe phase disappeared, the impact toughness (about 156 J) was lower than that of the weld at PWDT state due to the aggregated growth of M23C6 and MX carbides.