Test Blanket System Research Articles

Modified cryogenic distillation for tritium enrichment in the tritium accountancy system of helium cooled ceramic breeding test blanket system

Given the prevalence of scenarios involving low-abundance and large-volume hydrogen isotope batch separation in the tritium extraction and enrichment for fusion reactors, as well as the deuterization of tritiated water in nuclear power plants or fusion reactors, it is imperative to develop a more efficient enrichment and separation process to meet this demand. Consequently, this paper proposes a “quasi-continuous distillation” coupled with batch distillation for the batch enrichment and separation of low-tritium-concentrated hydrogen isotope gases, involved in the tritium accountancy system (TAS) of Helium-Cooled Ceramic Breeding blanket (HCCB-TBS), as an input. The design of this separation process addressed a feed with a tritium concentration of 18 ppm and approximately 5.8 Nm3 of hydrogen isotope gas. The feasibility of this separation process was successfully verified through tritium verification experiment.

Tritium management in ITER test blanket systems port cell for maintenance operations

Four Test Blanket Systems (TBS) will be tested in the International Thermonuclear Experimental Reactor equatorial ports #16 and #18 to verify tritium breeding and heat extraction technology. A significant quantity of tritium would be produced in TBM, and partly released into the port cell from the pipework of TBS or other high-temperature components due to its strong mobility and high permeation. The port cell should be accessible during equipment maintenance and human intervention. This work built a multi-dimensional geometric model to characterize HTO transport in the port cell, absorption/desorption, and diffusion in walls and discussed the effect of paint thickness, ventilation rate, source term, and epoxy properties on detritiation efficiency. The results suggest that a 0.1–0.16 mm paint with the lowest HTO solubility is optimal from the compromise between quick cleanup and tritiated waste decommission. A higher ventilation rate could accelerate detritiation while minimizing the radioactive source by a tritium-resisting layer is the most direct method. The optimized design options for reducing the time required to reach 1 DAC in 12 h still need further discussion because of the delayed HTO source from epoxy paint and dead zone of the flow field.

Overview of progress on water cooled ceramic breeder blanket in Japan

National institutes for Quantum Science and Technology (QST) are the implementing body of ITER project in Japan and takes the lead in fusion DEMO reactor development in Japan. The primary candidate of the breeder blanket for JA DEMO is Water Cooled Ceramic Breeder (WCCB) blanket. ITER-Test Blanket Module (TBM) program is an opportunity for the demonstration of DEMO breeding blanket concept under actual fusion environment. QST has led the development of the WCCB TBM Concept in the program since the signing of the TBM Arrangement in 2014. The preliminary design of test blanket system (TBS) is on-going. Recently, the status of the preliminary design of TBS was assessed between QST and the ITER Organization (IO). So, the status of the latest TBS design and the DEMO relevancy is presented.

Major improvements in the WCLL TBM design towards next review gates

The European domestic agency (Fusion for Energy) is developing the WCLL-TBS (Water Cooled Lithium Lead Test Blanket System) concept for its implementation and testing in ITER. The system is based on the use of the eutectic Pb–15.7Li enriched at 90 % in 6Li as tritium breeder and tritium carrier. Water is used as coolant with current target nominal inlet/outlet temperatures of 295/328 °C and 15.5 MPa pressure (similar to PWR conditions). The Test Blanket Module –TBM– is the in-vessel component where tritium generation will take place and whose structural material is EUROFER 97. Its design has been significantly improved in the last years. These design improvements have focused on three major areas: (i) overcoming some weaknesses of the previous design in the rear manifold area associated to the stiffening rod concept, (ii) strengthening the link with manufacturing developments and (iii) optimizing the amount of EUROFER 97 in the TBM box, so that the design converges to the limits imposed by the ITER project requirements. The work presented in this paper describes the latest design evolutions achieved in the WCLL TBM design with respect to that presented at CDR and, in particular, the major re-design of the TBM in the manifold area based on the ‘crossing vertical’ stiffening plate concept proposed and developed by the ESTEYCO Mechanics design team in collaboration with F4E.

Maintenance study on WCLL-TBS ancillary systems in process room

The Water-Cooled Lithium Lead Test Blanket System (WCLL-TBS) is one of the European TBS candidates for being installed and operated in ITER. The ongoing activities toward the Preliminary Design Review of the WCLL-TBS Ancillary Systems include the reallocation of their units in the reserved space, also in the view of the proper execution of the maintenance activities. Starting from the available documentation (System Design Description, Process Flow Diagrams, Process and Instrumentation Diagrams), detailed 3D CAD models have been developed for two Ancillary Systems to be installed in the Process room: the Tritium Extraction System (TES) and the Tritium Accountancy System (TAS). In the Process room (PR), the largest part of the TES process units is placed inside a Glove-box, as enclosure to avoid the contamination of the room in case of radioactive release due to failure of units. Based on the expected radioactivity content, all the TAS units are placed inside the Glove-box. Maintainability studies have been performed, focusing on the preventive (i.e. planned) and corrective (i.e. required by failures) maintenance tasks on the units installed inside the Glove-box. They were based on the 3D CAD models and supported using Virtual Reality as intuitive and immersive human-computer interface. They provided insights for the iterative improvement and consolidation of the Systems design and information for the evaluation of their availability.

Overview of tritium management in WCLL test blanket system of ITER

A key aspect for the achievement of the fuel self-sufficiency of future fusion power plants is the management of tritium in the different stages of the reactor fuel cycle. In particular, one of the greatest challenges for the success of liquid metal breeder concepts such as the Water-Cooled Lithium-Lead (WCLL) is the extraction of tritium from the liquid LiPb eutectic alloy (15.7 at.%Li), which fulfils the main functions of tritium breeder, tritium carrier and neutron multiplier, and the processing in a form suitable for an accurate tritium accountancy. Within this frame, significant advancements have been carried out in the last years as regards the WCLL Test Blanket System (TBS) of ITER. Tritium is mainly produced inside the LiPb and extracted through the Tritium Extraction Unit (TEU). The reference technology for the TEU of the WCLL-TBS is the Gas-Liquid Contactor (GLC) in the packed-column configuration. Tritium is then concentrated in ZrCo getter beds before being sent to the accountancy system where different technologies for both static and dynamic accountancy are adopted in order to provide precise and reliable tritium measurements, which are an essential need in view of an application to the European DEMO reactor. In this paper, the present status and the design solutions foreseen for tritium management in WCLL-TBS are presented and discussed. The implementation and optimization of the conceptual design of the lithium-lead loop (LLP), the Tritium Extraction System (TES) and the Tritium Accountancy System (TAS) have been performed in order to manage tritium concentration in LiPb and in the ancillary systems.

Preliminary Experimental Quantification of Helium Leakages from Flanged Connections at HCPB TBS Operative Conditions

The HCPB TBS (Helium-Cooled Pebble Bed Test Blanket System) is one of the two European TBSs that will be installed and tested in the ITER reactor. The use of flanged connections in the Helium Coolant System and the Tritium Extraction System of the HCPB TBS would make the remote maintenance operations easier and faster. Therefore, investigating the helium leakage from flanges becomes a fundamental step toward the control of the tritium activity in the Port Cell, as the helium flow will contain a variable but not negligible amount of tritium. The first set of experiments on helium leakages from flanged connections is described in this paper. The experiments were performed in a HeFUS3 facility, an eight-shaped helium loop designed to work at HCPB-TBS-relevant conditions. The facility can provide a helium mass flow rate in the range of 0.27–1.4 kg/s and can reach a pressure as high as 80 bar and a temperature up to 530 °C. Two types of gaskets were tested in this campaign: a spiral-wound gasket and an oval ring joint. The gasket/flange assemblies are described in detail in this paper, together with the test section that hosts them and the performed commissioning tests. The tests were carried out at 500 °C and 80 bar. In these conditions, the leak rate from the flange with the oval ring joint resulted in being, on average, 1.42·10−6 mbar∙L/s, while the leak rate from the flange with the spiral-wound gasket resulted in being, on average, 3.73·10−3 mbar∙L/s.

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.

Digital assistances in remote operations for ITER test blanket system replacement: An experimental validation

Removal/reinstallation of all Test Blanket System (TBS) equipment present in the Port Cell is required during the ITER Long Term Shutdown. TBS 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. The expected level of radiation in this area still allows performing maintenance tasks hands-on. However, the cumulated dose could be significant for operators. Classical Dose Reduction Measure (DRM) is to deploy Remote Handling systems. Consequently, use of robotized equipment, remotely operated means or collaborative robotics, have been investigated. Taking advantage of new digital technologies such as digital assistances, is expected to help operators during complex remote operations under limited vision conditions. Experiments were performed on a set of three TBS maintenance representative tasks: remote visual inspection of DN80 pipe, dye penetrant testing operation on pipe and dexterity test. A panel of remote handling equipment operators of different skill level was selected and involved onto these tasks. The results prove without ambiguity that for all operators the quality of the task execution is significantly improved when using digital assistances

Modeling the Transport of Activated Corrosion Products in the WCLL PbLi Loop for ITER and the EU DEMO With the GETTHEM Code

This work presents the results of the implementation in the GEneral Tokamak THErmal-hydraulic Model of available models for the generation and transport of any dispersed material in the PbLi eutectic mixture circulating in the Water-Cooled Lithium-Lead Breeding Blanket of the EU DEMO fusion reactor, with main focus on Activated Corrosion Products as solid suspension. A simple test case is used to show that the distribution of the concentration of activated corrosion products at any point of the PbLi loop, both in the water-cooled lithium-lead breeding blanket and in the related ITER Test Blanket System, can be determined by the model. The results obtained with this tool can be useful not only for radiological safety purposes, but also because activated corrosion products may affect the PbLi flow itself and the efficiency of the tritium removal system, with consequences on the achievable Tritium Breeding Ratio. A rigorous verification of the model is also performed.

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.

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.

Dynamic Event Tree Analysis as a Tool for Risk Assessment in Nuclear Fusion Plants Using RAVEN and MELCOR

In the broad framework of the nuclear power plants (NPPs) industry, the dynamic probabilistic risk assessment could answer the time dependence deficiency of the event tree and fault tree analysis. The basic event tree approach relies on experts’ pre-constructed accident sequences without exploring the time-dependent nature of an accident scenario, which could strongly affect the accident sequence. Conversely, the effects of events timing can be studied by adopting a dynamic event tree (DET) approach. Developing a DET methodology requires integrating a system code capable of replicating an accident scenario and a logic-driver code able to generate the event tree sequence, trigger plant safety systems, and manage other relevant events throughout the simulation. For this purpose, Methods for Estimation of Leakages and Consequences of Release (MELCOR) and reactor analysis and virtual control environment (RAVEN) have been coupled through a Python script developed by the Sapienza University of Rome, Rome, Italy, to perform DET studies during accident transients in fusion and fission reactors. RAVEN is a software tool developed at the Idaho National Laboratory (INL), Idaho Falls, ID, USA, to act as a control logic driver and post-processing tool for different applications. MELCOR for fusion is a fully integrated design basis and severe accident code that simulates thermal-hydraulic behavior and self-consistently accounting for aerosol transport in nuclear facilities and reactor cooling systems for the evaluation of the source term in fusion reactors. The coupling between these codes will provide a wide range of NPP risk assessment analyses, establishing new best practices. In this work, a preliminary DET study has been performed, selecting as initiating event an ex-vessel loss of coolant accident (LOCA) in the water-cooled lithium lead test blanket system (WCLL TBS) to be tested in International Thermonuclear Experimental Reactor (ITER). Time-dependent parameters such as the intervention of the plasma shutdown system and the closure of the main system isolation valves have been sampled to study evolving system scenarios.

Development of new analytical tools for tritium transport modelling

Tritium technologies, in particular tritium extraction from lithium-lead (LiPb, 15.7 at. % Li) and tritium concentration measurement in the eutectic alloy, are among the most challenging aspects of the R&D activities envisaged for the development of ITER and the European DEMO reactor. For instance, to efficiently design the systems devoted to the extraction of tritium, such as Gas-Liquid Contactors (GLC), Permeators Against Vacuum (PAV) or Liquid-Vacuum Contactor (LVC), theoretical models for the evaluation of the permeation flux are strictly necessary. In general, the same needs arise for the description of tritium permeators, which can find their application as Hydrogen isotopes Permeation Sensors (HPS) for the measurement hydrogen/tritium solubilized in the LiPb of either the Test Blanket Systems (TBS) or the Breeding Blanket. In this paper, new mathematical tools to describe the different permeation regimes both in the gas phase and in the presence of hydrogen isotopes monoatomically dissolved in a liquid phase, thus substantiating the theoretical background of hydrogen isotopes transport modelling throughout a membrane, is presented. For the sake of completeness, theoretical models in case of absence of a membrane (LVC) are also reported.

Supporting analysis for WCLL test blanket system safety

The Water-Cooled Lithium Lead Breeding Blanket is one of the most promising concepts to be used as a key component in fusion power devices. It provides tritium breeding and nuclear power conversion and extraction. Before its integration in a DEMO plant, a testing phase is required to gain data on both thermal and neutronic performances. For this purpose, a Test Blanket System is currently under conceptual design and will be integrated into dedicated ITER equatorial ports. Early insight on the safety performance of the Test Blanket System remains an essential element for integrating into ITER, and accident analysis is one of its critical components. Preliminary safety studies have been performed in the framework of Safety And Environment (SAE) work package of the EUROfusion consortium program to support the design and the integration of the Water-Cooled Lithium Lead Test Blanket System into ITER. At the present conceptual design phase, these studies are tailored to give insight on the effects of some parameters, such as postulated initiating event timing, plasma termination system or valves intervention, and possible safety provision implementation. A MELCOR model of the Water-Cooled Lithium Lead Test Blanket System has been developed, and two accident scenarios have been studied: an ex-vessel LOCA inside the port cell and a loss of flow accident because of pump seizure. The impact of mentioned parameters on accident evolution and consequences is investigated in support of the safety logic definition.

R&D Activities for Fusion DEMO in the QST Rokkasho Fusion Institute

The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokkasho Fusion Institute is reported. In order to advance the Japanese DEMO activity, not only Japanese domestic activity but also international collaborations of Broader Approach activity and ITER-related activities are conducted in the QST Rokkasho Fusion Institute. Activities for DEMO design and relevant R&D; design of a fusion neutron source and development of an accelerator, ITER Test Blanket System; tritium handling technology; and information technology infrastructures, including a supercomputer system and a remote experimentation system, are carried out for a Fusion DEMO reactor.

Simulation of loss of off-site power accident of CN HCCB TBS

Helium Cooled Ceramic Breeder Test Blanket System (HCCB TBS) is proposed by China, which will be tested in ITER facility to validate the tritium breeding blanket technology. Test Blanket Module (TBM) is the core component associated with its helium cooling system (HCS) and other ancillary systems. The accident analysis is necessary during the preliminary design process, and the results obtained provide safety feedback to the following design work. Loss of Off-site Power (LOSP) is one of the design basis accidents (DBA), and is simulated using the best estimated code RELAP5 MOD3.4. For the safety simulation, besides the thermal conductive and convective model, the radiation model is adopted as well. The influence of the natural circulation of the HCS primary loop and the necessity of the Class III power supply (safety class) are discussed. The simulation results show that the peak temperature of the structure material exceeds the design temperature (550 °C) for short duration, which has almost no impact on the structure. In the meantime, the duration of the temperature over 550 °C drops sharply while the natural circulation and the Class III Power are considered. The results verify the integrity and the safety of the HCS primary loop during the LOSP accident.

Experiments on the MHD Effect on the Drainage of a LiPb Channel and Supporting Numerical Computations with the Level Set Method

To analyze the impact of the magnetohydrodynamics (MHD) effect on the fast draining of a LiPb channel (lithium-lead eutectic, 15.7 at. % Li) for a liquid metal fusion blanket such as the water-cooled lithium-lead test blanket system of ITER or DEMO, an experimental campaign was carried out with the support of the Integrated European Lead Lithium LOop experimental facility (IELLLO), installed at the ENEA Brasimone research center, Italy. The experiments were carried out by measuring the drainage time of the internal permanent magnet pump channel, normally used to circulate the LiPb in the loop, with and without the magnetic field. Moreover, this paper proposes a new numerical methodology to study the time delay induced by the MHD by using the commercial software COMSOL Multiphysics. In this way, it was possible to evaluate the LiPb fraction present at each time step in the computational domain and to estimate the time necessary for the complete drainage of the channel. The level set method was used to describe the transient behavior of the MHD flow under low-Rm approximation. The developed code was compared with the experimental results and showed good agreement, and it constitutes the first step in model validation as a possible application to ITER and DEMO. The experimental and numerical analyses performed in this work can be used as a benchmark case for MHD code development.

Numerical Simulations with RELAP5-3D and RELAP5/mod3.3 of the Second Experimental Campaign on In-Box LOCA Transients for HCLL TBS

In-box LOCA was identified as one of the worst accidental scenarios for the HCLL TBS (Helium Cooled Lithium-Lead Test Blanket System). Aiming to experimentally analyze the consequences of this transient, ENEA designed and built THALLIUM (Test HAmmer in Lead LIthiUM), a facility that reproduces the LiPb loop of the HCLL TBS. Two experimental campaigns were carried out by simulating the rupture of a stiffening plate and the related helium injection in the LiPb loop. The obtained experimental data were used to check the capabilities of RELAP5 system code to reproduce the pressure wave propagation that follows this accident. The first simulations were made with RELAP5-3D using LBE (Lead–Bismuth Eutectic) as a system fluid, as the thermophysical properties of LiPb are tabulated only up to a maximum value of 40 bar in this version of the code. Then, LiPb properties were implemented in RELAP5/mod3.3, after selecting the proper correlations from a literature review. This work summarizes the numerical simulations of the second experimental campaign, which was simulated with both versions of the code. The simulations highlight that the code is able to accurately reproduce the experimental results and that RELAP5-3D is slightly more precise than RELAP5/mod3.3 in predicting the pressure trends.

Conceptual design overview of the ITER WCLL Water Cooling System and supporting thermal-hydraulic analysis

In this paper, the conceptual design of the International Thermonuclear Experimental Reactor (ITER) Water-Cooled Lithium-Lead (WCLL) Test Blanket Module (TBM) Water Cooling System (WCS) from Europe is presented. The system consists of two loops in series. This design feature allows the removal of heat from the TBM box avoiding at the same time the release of radionuclides into the ITER Component Cooling Water System (CCWS), that acts as WCLL Test Blanket System heat sink. For this purpose, the WCS primary loop deals with the direct heat removal from the ITER TBM and the secondary one implements physical separation between the contaminated primary loop coolant and the CCWS. The insertion of an economizer into the primary loop determines the characteristic “eight” shape of the circuit. This choice was done in order to reduce the temperature difference on the intermediate heat exchanger. Hairpin type and steam bubble are the technologies selected for heat exchangers and pressurizers, respectively. Pressure and temperature control systems are foreseen to limit excursions from rated values in normal operational states and abnormal transients.A computational activity was promoted to assess the WCLL-WCS conceptual design, using a modified version of the RELAP5 Mod3.3 system code. A detailed thermal-hydraulic model was developed on the basis of design outcomes. The nodalization scheme includes the TBM, the WCS, a portion of the CCWS and the lithium-lead circuit. The computational campaign involved both the normal operational state and selected abnormal transients. In all the scenarios simulated, the conceptual design has highlighted the capability of operating the system respecting all the thermal-hydraulic requirements. The abnormal transient selected and presented is the loss of flow in the CCWS (loss of heat sink). In these conditions, TBM cooling function has been verified, keeping standard control strategies without any external action.