Test Blanket System Research Articles

Neutronics analysis and assessment of shielding options of pipe forest and Bioshield-Plug design for ITER TBSs

In ITER [1], it is foreseen to have two Test Blanket Systems (TBS) installed in two dedicated Equatorial Ports (EP) for the purpose of testing and validating different concepts of tritium breeding blankets [2]. Each TBS consists of a Test Blanket Module Port Plug (TBM PP) and its associated systems; the Pipe Forest (PF) in the Port Interspace (PI) area and the corresponding systems in the Port Cell (PC) area. While analyses have been conducted before [3], the design has since gone through major evolution with a redesign of the PF, a new BP based on the butterfly doors concept, and an optional activated water delay tank. In this paper, the results of a comprehensive nuclear analyses that aim at assessing the neutronics performance of the revised PF and BP, will be discussed. These include calculations of neutron flux distributions and Shutdown Dose Rate (SDDR), following SA2 [4] scenario, in a baseline configuration and exploring the option of adding additional shielding plates in the PF. It has been found that the additional shielding plates reduces the neutron flux in the PI by ∼40–60% and the SDDR in the PI, due to activated components, by ∼50%.

Microstructure and mechanical properties of LBW and EBW weld joints of CLF-1 steel: A comparative analysis

This study was focused on welding joints of reduced-activation ferritic/martensitic (RAFM) CLF-1 steel medium-to-thick plates produced via laser beam welding (LBW) and electron beam welding (EBW). Such joints were used in the ITER project, in particular, in the back plate mockup of Helium Cooled Ceramic Breeder Test Blanket System developed in China in 2015. The microstructural evolutions and the mechanical properties of fusion zones (FZs), weld zones (WZs), and heat-affected zones (HAZs) of LBW and EBW joints were analyzed and compared using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). Moreover, the microstructural differences between the two types of joints and their effect on the impact properties were discussed. Results showed that the well-formed welds without defects, such as pores, incomplete fusion, and cracks, could be obtained via LBW and EBW. Different heat inputs led to the differences in the grain size gradients, the boundary angles, the content of carbides, and the distribution of dislocations. Despite obvious differences in morphology among FZ, WZ, and HAZ, both types of joints exhibited excellent tensile properties and impact toughness. The tensile strengths of the LBW and EBW joints were 633 and 634 MPa at room temperature, respectively, versus 370 and 366 MPa at 550 °C. The mean impact-absorbing energy values of LBW and EBW joints were 251 and 284 J. The further analysis revealed that the low δ-ferrite content, small effective grain size, large grain size gradient, high grain boundary angle, diminutive size of M23C6 carbides, high-Ta content in MX carbides and low dislocation density led to high impact toughness of the LBW joint, while the microstructure of EBW joint failed to show the improvement of impact toughness, and the extraordinary impact toughness could be attributed to the high scattering band.

Conceptual design of the main Ancillary Systems of the ITER Water Cooled Lithium Lead Test Blanket System

The Water Cooled Lithium Lead Test Blanket System (WCLL TBS) is one of the EU Test Blanket Systems candidate for being installed and operated in ITER. In view of its Conceptual Design Review by F4E and ITER Organization (IO), planned for mid-September 2020, several technical activities have been performed in the areas of WCLL TBS Ancillary Systems design.In this article the outcomes of the conceptual design phase of the four main Ancillary Systems of WCLL TBS, namely the Water Cooling System (WCS), the Coolant Purification System (CPS), the PbLi loop and the Tritium Extraction System (TES), are reported and critically discussed.In particular, for each Ancillary System hereafter are reported: i) a short design description, including the conceptual design of their main components together with their operative conditions under the so-called Normal Operational State (NOS), ii) the ESP-ESPN classification for their main components, and iii) their arrangement and integration in the assigned ITER areas (PC#16, Vertical Shaft, TCWS Vault, Galleries and Tritium Process Room).

ITER assembly phase: Progress toward first plasma

By the end of 2019, ITER had completed about 67% of the overall work required to achieve First Plasma. Significant progress has been achieved since SOFT 2018, most visible in the completion of key buildings and infrastructure on the worksite. The tokamak building will be ready for equipment as of Spring 2020. Commissioning of the connection to the EU grid and the substation for the steady-state electric network is complete. Planning for commissioning of other key plant systems (e.g., secondary cooling water, cryoplant, pulsed power for magnet systems) is well underway. On the manufacturing front, progress is equally impressive. The base and lower cylinder of the cryostat were completed and turned over to ITER in July 2019. The first two poloidal field coils will be turned over to ITER in early 2020, and two more are being wound. Factory tests of the first central solenoid module will be completed in 2020, and five more modules are in the manufacturing process. The first toroidal field coil is to arrive onsite in early 2020, and progress is rapid on additional TF coils. The first sections of the thermal shield are already onsite. The first vacuum vessel sector is complete and will arrive onsite in early 2020. This progress means that, for most of the major first-of-a-kind components, the capability has now been demonstrated to fabricate according to ITER’s demanding specifications. With this progress achieved, ITER will enter Assembly Phase in 2020, remaining on schedule for First Plasma in 2025. The sequence of ITER operation from First Plasma (FP) to the achievement of the Q = 10 project goal has been consolidated in a Staged Approach, with all systems to be installed before the start of full fusion power operation in 2035. The ITER Research Plan, issued publicly in 2018, is consistent with the systems available in each phase. An adjustment of the machine baseline configuration was made in 2019, with two equatorial ports allocated to the Test Blanket Systems and the development of a better performing Disruption Mitigation System. Highlights from each of these areas (manufacturing, commissioning, tokamak assembly) will be presented along with the updated status and plans.

Experimental study of neutron irradiation effect on elementary semiconductor devices using Am-Be neutron source

An experiment has been conducted to evaluate the lifetime, reliability and operational performance of elementary semiconductor devices in the neutron radiation environment which supports to reduce the fatal in measurements and plan the preventive actions in nuclear facilities. It will also support the enhancement of electronics for nuclear facilities. The elementary semiconductor devices used in the experiment are Diode (1n4007), Zener Diode (5.1v), Light Emitting Diode, Transistor (BC547, 2n3904), Voltage controlling IC (7805), Operational Amplifier (LM741) and Optocoupler (4n35). The selection of devices has been made by keeping in mind their application in transmitting devices (i.e. Temperature transmitter, pressure transmitter, flow transmitter, monitors and controllers) for Indian test blanket system in ITER. Such devices are also used in general nuclear electronics. The devices have been irradiated in the Am-Be neutron source environment. The maximum fluence has been given up to 1011 n/cm2. The neutron source has energy range from low to high. All semiconductor devices have been characterized before and after irradiations. The deviation of 5 - 10% is observed in diodes I-V characteristics whereas transistors show a bit higher deflection in basic functionality. Optocoupler shows more than 50% deviation in its basic characteristics whereas voltage-controlling IC is not even functioning after the irradiation of 1011 n/cm2. The paper describes the details of the experiment and the behavior of semiconductor devices after irradiation. The experiment supports the selection and further research of the Indian test blanket system instruments.

Numerical simulations with RELAP5-3D of the first experimental campaign on In-box LOCA transient for HCLL TBS

• Results of RELAP5-3D simulations of an experimental campaign in a mock-up of the HCLL TBS LiPb loop. • RELAP5-3D is able to accurately reproduce the pressure wave propagation in the facility. • Two main general differences between numerical and experimental results have been highlighted. • Average discrepancy of about 3.5 bar between experimental data and numerical results on 5 tests. • The code accurately predicts the first pressure peak, the most relevant input for the designers. In this work the capabilities of RELAP5-3D are tested against the experimental data gathered in the first experimental campaign on THALLIUM (Test HAmmer in Lead LIthiUM). THALLIUM, a LiPb facility, was built in 2015 at ENEA R.C. Brasimone with the aim to study the In-box LOCA by reproducing the LiPb loop of the HCLL TBS and the pipe forest at 1:1 scale. HCLL TBS (Helium Cooled Lithium-Lead Test Blanket System) was one of the Breeding Blanket concepts selected for ITER. The ITER classification system identified the In-box LOCA as a critical accidental transient for this type of TBS. For this reason, codes able to reproduce the consequences of this transient are needed to support the design of the LiPb loop. The simulations with RELAP5-3D used LBE (Lead-Bismuth Eutectic) as system fluid, as the thermophysical properties of LiPb are tabulated only up to a maximum value of 40 bar. The results of the simulations suggest that RELAP5-3D is able to reproduce the pressure wave propagation in the facility, with an average discrepancy of about 3.5 bar (i.e. an average percentage discrepancy of about 9.7 %), but also highlight two general discrepancies between computations and experiments.

Development of tritium dynamic transport analysis tool for tritium breeding blanket system using Modelica

Demonstration of the engineering feasibility of tritium breeding, including tritium generation, tritium extraction, tritium control and tritium safety is one of the main objectives of ITER Testing Blanket Module (TBM) Program. As one of TBM concepts, the tritium transport assessment of the China Helium Cooled Ceramic Breeder TBM and its ancillary systems (called test blanket system, TBS) is absolutely vital to understand and analyze the dynamic tritium behaviors in the system in detail according to the ITER operation scenario. Considering the requirements of the tritium transport analysis for the HCCB TBS, an integrated dynamic tritium transport analysis tool has been developed by the system simulation language “Modelica”. The multiple dynamic physics effects are considered in the tool, including the aspects impacting the tritium transport, simplified thermal-hydraulics considering only the linear loss and local loss in fluid, 1-d thermal conduction in solid, hydrogen isotope transport in fluid, 1-d tritium permeation in solid, tritium solution and recombination on solid surfaces, tritium desorption from ceramic breeder etc. A simplified model of the HCCB TBS is modeled to verify the functions and accuracy of this tool. The results, such as dynamic change of tritium concentration/partial pressure in fluids, dynamic change of tritium inventory in solids, tritium permeation fluxes etc., have been obtained, which are preliminary verified at various aspects by comparing with analytical and hand calculations.

Overview of recent ITER TBM Program activities

Abstract The ITER Test Blanket Module (TBM) Program has significantly evolved since 2018. The number of equatorial ports allocated for operating the Test Blanket Systems (TBSs) has been reduced from three to two. As consequence, four TBSs can be simultaneously installed and operated, versus six previously. Since the dedicated space in the various rooms of the Tokamak Complex has been kept unchanged, the existing space constraints have been substantially relaxed. The paper addresses the possible selection of the four TBSs for the initial configuration and design of the involved TBMs and the main on-going R&D carried out by the ITER Members (IMs) in support of the TBS designs. It describes also the main progress of the design of the infrastructures needed for hosting the four TBSs (e.g., port plugs, port cell common components, common maintenance tools and equipment). The design of the TBS Connection Pipes (CPs) System has reached its final phase since, being captive, it needs to be installed before the First Plasma.

Experimental Qualification of New Instrumentation for Lead-Lithium Eutectic in IELLLO Facility

• results of a new experimental campaign in IELLLO facility, a LLE (Lead Lithium Eutectic) loop designed for transient simulation and instrumentation testing. • A type of differential pressure transducer and 3 flow meters were characterized in LLE: Vortex flow meter, Coriolis flow meter, Thermal flow meter. • The differential pressure transducers were used to characterize the pressure drops of some of the major components and the pressure head of the permanent magnet pump. • After a detailed analysis, the thermal flow meter was considered the most reliable for flow rate measurements in LLE. • A detailed uncertainty analysis was carried out to support the investigations. The experimental facility IELLLO was installed in ENEA Brasimone R.C. in 2007, aiming to support the design of liquid Test Blanket Modules that will be installed in ITER and to contribute to the development of Lead-Lithium Eutectic (LLE) technologies. IELLLO has been recently upgraded by installing instrumentation relevant for ITER application. Differential pressure transducers, a Coriolis and a thermal mass flow meters were installed in the facility. An experimental campaign was planned, setting two objectives. The first objective was to qualify the instrumentation for flowing LLE The installation of a differential pressure transducer across each flow meter made also possible to characterize the pressure drops across these instruments. The second objective of this activity was to improve the knowledge on the performances of the main components of the loop at lower mass flow rates (namely 0.5-1.2 kg/s) and to quantify their pressure drops. The investigated flow rates were chosen to be relevant for the LLE loop of the WCLL TBS (Water Cooled Lead-Lithium Test Blanket System). This work presents the results of the experimental campaign, paying particular attention to underline the lessons learned on how to correctly operate instrumentation for LLE.

Numerical simulation for operational state switching processes of CN HCCB TBS helium cooling system

Abstract Understanding the thermal hydraulic behaviors of switching processes between operational states is one of key issues in the design and optimization of the helium cooling system (HCS) of China helium cooled ceramic breeder test blanket system (CN HCCB TBS). The thermal hydraulic model of HCS considering the influence of heat structural and thermal insulation layer was established, and the main switching processes for the main operational states of HCS were simulated. The results present the dynamic changes of the thermal hydraulic parameters, including temperature, pressure, helium mass, power of heater, etc. The results indicate that the operation states and relative switching processes are reasonable: the designed operational states can be achieved by controlling the heater power and circulator mass flow rate; the time consumptions for state switching processes are far less than the reserved time according to the operation campaign schedule of international thermal-nuclear fusion reactor (ITER). The work presented in this paper demonstrates the operation of HCS, which provides an important foundation for further design and optimization of the HCS.

European TBM programme: First elements of RoX and technical performance assessment for DEMO breeding blankets

Abstract The strategy for the development of the European DEMO Breeding Blanket (BB) and the TBM (Test Blanket Module) Programme have been recently realigned with the objective to mitigate the risks associated with the development of tritium breeding and power extraction technologies in a DEMO Fusion Plant. In such a context, a clear definition of the key technical requirements of the DEMO Breeding Blanket (BB) is essential not only to drive a consistent BB design but also to allow the TBM Programme to provide its contribution to the preliminary demonstration of compliance with those requirements. Such contribution is substantiated through all phases of the TBM Programme life cycle, from the TBS (Test Blanket System) conceptual design up to its installation in ITER, operation and final dismantling. After a recall of the status of development of the European BB DEMO and TBM Programme, the paper describes the strategy and the status of the activities for the preliminary demonstration of compliance with a selected group of DEMO BB requirements. Through the elements given in this paper it is highlighted the importance to have a continuous and strict connection between the DEMO BB development and the execution of the TBM Programme.

An integrated hydrogen isotopes transport model for the TRIEX-II facility

Abstract At ENEA Brasimone Research Centre, Italy, a new experimental facility named TRIEX-II ( Tri tium Ex traction) was designed and installed. Its aim is to characterize, in the range of operating conditions foreseen for the European Test Blanket System WCLL-TBS, several extraction technologies for hydrogen isotopes (Q2) solubilized in the flowing metallic LiPb alloy (15.7 at. % Li). One of these technologies is the packed column, an example of Gas/Liquid Contactors (GLCs). This paper proposes a multiscale modelling tool, combining different scales through two computational tools. The extraction column mock-up is described by a component-detail level model, developed in COMSOL Multiphysics, and integrated into a system level code of the whole TRIEX-II circuit, developed using MATLAB/Simulink. The integration is carried out by implementing the COMSOL component into an S-function of MATLAB/Simulink, preserving the process flow diagram of the loop. In this way, it was possible to quantify the Q2 concentrations and the permeation fluxes, and to evaluate the theoretical extraction efficiency. The LiPb flow field inside the extractor was also derived. Finally, a comparison with the experimental results was performed. The results suggest that this tool could be adapted for analyses of complex systems, at a multiscale level, in view of design improvements and safety studies for the tritium cycle of ITER.

Current design and R&D progress of the Chinese helium cooled ceramic breeder test blanket system

As the testing mockup of tritium breeding blanket for DEMO, the Chinese helium cooled ceramic breeder test blanket system (HCCB TBS) is under development by China and will be tested in ITER to verify key tritium breeding blanket technologies. After the approval of the conceptual design of the HCCB TBS by the ITER Organization in 2015, the design optimization and more research and development activities for the HCCB TBS have begun for the preliminary design. This paper gives an overview of the recent progress of the HCCB TBS development in China, including the design optimizations, material development, mockup fabrication and test facility construction.

Hydrogen adsorption performance for large-scale cryogenic molecular sieve bed

Cryogenic hydrogen adsorption using molecular sieve beds is considered to be one of the main candidate processes for recovery of produced tritium from purge gas in breeding blankets and it has been chosen for separating hydrogen isotopes in Tritium Extraction System (TES) of Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS). Various adsorbents and their performance have been studied for the cryogenic adsorption using small-scale experiments. However, large-scale experiments comparable to TBS-relevant scale are required to have sufficient confidence for component design and performance prediction of Cryogenic Molecular Sieve Bed (CMSB) for the TBS and beyond. To properly evaluate hydrogen adsorption performance of a large size CMSB, a series of experiments have been performed using PGLoop facility which is constructed and operated in the National Fusion Research Institute. The experimental conditions were set to include breeding-blankets-relevant parameters. As such, the effects of swamping ratio, total pressure, and flow rates on the performances of CMSB were studied in the range of hydrogen partial pressures from 100 to 700 Pa. While a slight reduction in hydrogen adsorption performance is observed in comparison to the small isotherm experiments, which can be attributed to scale-up effects, it shows that the experimental results agree reasonably well with existing literature data.

Conceptual design of the enhanced coolant purification systems for the European HCLL and HCPB test blanket modules

The Coolant Purification Systems (CPSs) is one of the most relevant ancillary systems of European Helium Cooled Lead Lithium (HCLL) and Helium Cooled Pebble Bed (HCPB) Test Blanket Modules (TBMs) which are currently in the preliminary design phase in view of their installation and operation in ITER.The CPS implements mainly two functions: the extraction and concentration of the tritium permeated from the TBM modules into the primary cooling circuit and the chemistry control of helium primary coolant.During the HCLL and HCPB-TBSs (Test Blanket Systems) Conceptual Design Review (CDR) in 2015 it was recognized the need of reducing the tritium permeation into the Port Cell #16 of ITER. To achieve this and, then, to lower the tritium partial pressure in the Helium Cooling Systems in normal operation, the helium flow-rate treated by each CPS has been increased of almost one order of magnitude.In 2017, to satisfy the CDR outcomes and the new design requirements requested by Fusion for Energy (F4E, the European Domestic Agency for ITER), ENEA performed a preliminary design of the “enhanced” CPSs.This paper presents the current design of the “enhanced” CPSs, focusing on design requirements, assumptions, selection of technologies and preliminary components sizing.

Updated design and integration of the ancillary circuits for the European Test Blanket Systems

The validation of the key technologies relevant for a DEMO Breeding Blanket is one of the main objectives of the design and operation of the Test Blanket Systems (TBS) in ITER. In compliance with the main features and technical requirements of the parent breeding blanket concepts, the European TBM Project is developing the HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed)-TBS, focusing in this phase on the design life cycle and on R&D activities in support of the design. The TBS ancillary systems are mainly circuits devoted to the removal of thermal power and to the extraction and recovery of the tritium generated in the Test Blanket Modules. They are:•The Helium Cooling System (HCS);•The Coolant Purification System (CPS);•The Tritium Extraction System (HCLL-TRS, HCPB-TES);•The Lead Lithium Loop.Their conceptual design was deeply analyzed during the ITER Conceptual Design Review (CDR) in 2015. The assessment of the CDR was overall positive and recommendations for improvements were made. The present design takes into account the main CDR recommendations as well as the implementation of the requirements related to ITER operation, safety principles application and physical space constrains.

Conceptual design of HCCR-TBM diagnostics

The design of diagnostics system for the Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) has now progressed from generic approaches to the conceptual design integration phases. It is proposed to monitor the temperature, strain, magnetic field strength and neutron flux in selected positions of the corresponding Test Blanket Module (TBM). The temperature can be measured by thermocouple sensors to be integrated for the first wall and breeding zones of TBM. The strain will be monitored by appropriate gauges on the first wall of TBM. Several magnetic pick-up coils could be installed to measure magnetic field strength inside the TBM structure. For the neutron flux monitoring, pairs of Neutron Activation System (NAS) probes and Micro-Fission Chambers (MFCs) would be installed in the TBM structure. In this paper, a brief description on the proposed diagnostics system as well as the key components is presented, and the integration challenges of HCCR TBS diagnostics system are outlined.

Breeding blanket system design implications on tritium transport and permeation with high tritium ion implantation: A MATLAB/Simulink, COMSOL integrated dynamic tritium transport model for HCCR TBS

An integrated, multi-physics, dynamic predictive tool to quantify tritium retention, removal, and permeation for HCCR Test Blanket System (TBS) is presented in this paper. The tool expands from detailed COMSOL component models developed previously at UCLA, into an integrated, system-level blanket model using MATLAB/Simulink. It aims at achieving self-consistent predictions in particular concerning dynamic tritium concentration built-up in the He coolant. The integration is achieved by implementing COMSOL component models in the discrete section of Simulink S-Functions. The model replicates HCCR TBS process flow diagram and preserves main tritium flow characteristics for both helium cooling and tritium extraction systems. Current results demonstrate importance of detailed component models as well as dynamic simulation for improved accuracy on answers to questions related to safety/licensing and designs.

Activation characteristics of candidate structural materials for a near-term Indian fusion reactor and the impact of their impurities on design considerations

Activation analyses play a vital role in nuclear reactor design. Activation analyses, along with nuclear analyses, provide important information for nuclear safety and maintenance strategies. Activation analyses also help in the selection of materials for a nuclear reactor, by providing the radioactivity and dose rate levels after irradiation. This information is important to help define maintenance activity for different parts of the reactor, and to plan decommissioning and radioactive waste disposal strategies. The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential, due to the presence of a high-energy neutron environment which makes decisive demands on material selection. This study comprises two parts; in the first part the activation characteristics, in a fusion radiation environment, of several elements which are widely present in structural materials, are studied. It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment. The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions. The structural materials selected for this study, i.e. India-specific Reduced Activation Ferritic−Martensitic steel (IN-RAFMS), P91-grade steel, stainless steel 316LN ITER-grade (SS-316LN-IG), stainless steel 316L and stainless steel 304, are candidates for use in ITER either in vessel components or test blanket systems. Tungsten is also included in this study because of its use for ITER plasma-facing components. The study is carried out using the reference parameters of the ITER fusion reactor. The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port. The presence of elements like Nb, Mo, Co and Ta in a structural material enhance the activity level as well as the dose level, which has an impact on design considerations. IN-RAFMS was shown to be a more effective low-activation material than SS-316LN-IG.

ITER TBM Program and associated system engineering

Six Test Blanket Systems (TBS) will be installed and operated in ITER in the framework of the ITER Test Blanket Module (TBM) Program. After a short description of each TBS and of the main operating parameters, this paper gives the main features of the various TBSs sub-systems, in particular, the cooling systems (helium, water and/or lithium-lead), the tritium system for extracting the tritium from the breeder, the coolant purification system, the tritium accountancy system, and several instrumentation & control systems. This paper addresses also the main services that are needed to operate a TBS such as the secondary water cooling system, the chilled water system to cool the helium circulators, the liquid and gaseous nitrogen used in the tritium extraction process, the compressed air needed to operate the various valves present on the circuits, and the electrical power supply for all the active electrical components such as pumps, circulators and heaters. Other important aspects are the required interfaces with other ITER plant systems such as those involving air and process gas detritiation, power supply, radiological and environmental monitoring, remote handling, measurements, and services.