Blanket System Research Articles

Preliminary steady and transient analysis for the CFETR helium cooled solid blanket system with RELAP

Abstract The helium cooled solid breeder (HCSB) blanket, which is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR), has been developed in University of Science and Technology of China (USTC). And the blanket cooling system (HCS) is also designed for HCSB. The steady-state results show that the blanket system has a good relevance to the design parameters. The transient analysis of possible accidents should proceed to evaluate its safety performance. The accident analysis in this paper adopts the ITER-like conditions and the ex-vessel loss of coolant accident (LOCA) case is investigated with RELAP5 code. The results show that the ex-vessel LOCA with detection couldn't cause serious consequence, however the ex-vessel LOCA without detection has potential to melt the blanket that we should take measures to avoid it.

Tritium release from molten salt FLiNaK under low flux neutron irradiation with an AmBe neutron source

For examination of tritium transport in a self-cooled liquid blanket system using fluoride molten salt breeder/coolants, a series of neutron irradiation experiments have been conducted under the low flux neutron environment with an AmBe neutron source at the OKTAVIAN facility of Osaka University in Japan. Tritium release from molten salt FLiNaK (LiF-NaF-KF) at 773 K is evaluated with the neutron source, energies of which are up to 10.7 MeV and the intensity is ∼2 × 106 n/s. To measure the amount of tritium released from FLiNaK and permeating through an Inconel crucible, TF and/or T2O, and T2 are recovered separately using water bubblers and CuO oxidation beds. The amount of tritium recovered for one week in a steady state, 12.0 Bq, has closed to the calculated value, 10.6 Bq, obtained with MCNP6. More than 98% of the recovered tritium from FLiNaK has been TF and/or T2O. The overall mass transfer coefficients for tritium released from a FLiNaK free surface have been evaluated from the transient changes in recovered tritium and those are comparable to the value for HF released from FLiBe previously reported in the literature.

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.

Current Status and Progress on the Shielding Blanket of CFETR

The shielding blanket (SB) system is to provide neutron shielding for Chinese fusion engineering test reactor in-vessel components and the external environment. The SB concept design activities have been carried out during the 2009–2011 campaign. Current design on the SB system is concentrated on neutronics analysis, SB modules design, and mechanics analysis using finite-element method. The neutronics analysis is made progress on the neutron shielding performance of shield materials and the SB thickness estimation in radial direction. A new SB module structure with prewater boxes and radial cooling tubes is proposed in the breeding area, which has the benefit to neutron moderating and heat removal issues. The advantage lies in the increasing coolant volume ratio inside the SB module. The heat removal capability of the SB module and the divertor blanket are explored using computational fluid dynamics (CFD) simulations. In addition, the preliminary mechanics analysis is completed on the electronmagnetic force including structure static stress under core plasma disruption. This paper summarizes SB component design activities and the progress at present status.

Neutronics Analysis of Helium Cooled Ceramic Breeder Blanket with S-shaped Lithium Zone and Cooling Plate for CFETR

China fusion engineering test reactor (CFETR) is a tokamak fusion experimental device under design to bridge the research and development gaps between international thermonuclear experimental reactor and demonstration. Helium cooled ceramic breeder (HCCB) blanket is the candidate blanket concepts for CFETR. A blanket with S-shaped lithium zone and cooling pipes reduces the space of helium manifold and is beneficial to tritium breeding. The neutronics analyses of the HCCB blanket have been performed with the Monte Carlo code monte carlo N particle transport code and nuclear cross section data from the fusion evaluated nuclear data library-3.1b data library. The tritium breeding ratio without port is 1.22 and meets the requirements of tritium self-sufficiency, as well as, the nuclear heating power of 214 MW is achieved and the global energy multiplication factor is 1.33. The shielding performance was shown to meet the design limit. The aim of the analysis is to provide the reference for the design and optimization of CFETR blanket system.

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.

New insights into enhanced anaerobic degradation of coal gasification wastewater (CGW) with the assistance of graphene

The up-flow anaerobic sludge blanket (UASB) system with graphene assisted was developed for coal gasification wastewater (CGW) treatment. Short-term results showed that optimal graphene addition (0.5 g/L) resulted in a more significant enhancement of methane production and chemical oxygen demand (COD) removal compared with that of the optimal activated carbon addition (10.0 g/L). Long-term results demonstrated that COD removal efficiency and methane production rate with graphene assisted achieved 64.7% and 180.5 mL/d, respectively. In addition, graphene could promote microbes accumulation and enzymes activity, resulting in higher extracellular polymeric substances (EPS) and coenzyme F420 concentrations. X-ray Diffraction (XRD) analysis indicated that chemical of graphene changed insignificantly during the experiment. Meanwhile, with graphene assisted, cells were attached together to form microbial aggregates to facilitate sludge granulation process. Furthermore, the enriched Geobacter and Pseudomonas might perform direct interspecies electron transfer (DIET) with Methanosaeta via biological electrical connection, enhancing the anaerobic degradation of CGW.

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.

Study of EU DEMO WCLL breeding blanket and primary heat transfer system integration

The Water-Cooled Lithium Lead breeding blanket is a candidate option for the realization of European DEMO power plant. One of the main functions of the breeding blanket is to recover the thermal power from the first wall and the breeding zone and to drive it to the Primary Heat Transfer System. Moreover, due to the DEMO pulsed operation, an Energy Storage System is foreseen in order to ensure thermal energy availability and reduce cycling loading during dwell time. The blanket design must interface with the Primary Heat Transfer System and the Intermediate Heat Transfer System in an integrated solution with the aim of delivering feasible power plant, with high conversion efficiency.A thermal-hydraulic model of the main components of the water cooled Primary Heat Transfer System is developed, using RELAP5 system code, to support the design and sizing. A RELAP5/Mod3.3 extended version has been set-up to implement PbLi and molten salt (HITEC) fluid proprieties, as well as the relevant heat transfer correlations. The results of the preliminary applications provide a comprehensive set of thermo-hydraulic parameters (i.e. coolant inventory, pressure drops), relevant for the assessment of the design integration.

Studies of the plasma vertical instability and its stabilized concepts in JA and EU broader approach, DEMO design activity

Vertical instability of an elongated plasma and its stabilized concepts by in-vessel components and vacuum vessel (VV) design have been studied intensely in JA and EU Broader Approach, DEMO Design Activity. The vertical stabilization of the plasma represents one of the key issues for EU and JA DEMO, due to the large distance of the active control coils for the presence of thick breeding blanket system. A feasible DEMO reactor that maintains plasma vertical stability was proposed from an engineering viewpoint. The vertical stability performances are acceptable, without considering a passive stabilizer, if a maximum elongation of κ95 = 1.6 is chosen. For the higher-elongated plasmas (κ95 > 1.70), additional inboard passive stabilizer is effective.

Overview of the Design Status of the Superconducting Magnet System of the CFETR

The concept design of the China Fusion Engineering Test Reactor (CFETR) was started in 2012 in two stages. For CFETR-Phase I, the major radius is 5.7 m, the minor radius is 1.6 m, and the magnetic field at the plasma region, BT , is 4–5 T. There were 16 toroidal field coils and six central solenoid coils designed by a Nb3Sn cable-in-conduit conductor (CICC) with the maximum operation current of 64 and 50 kA, respectively. Three types of plasma equilibrium configuration (ITER-like single null, super-X, and snowflake) were designed. The maximum flux provided by the central solenoid is set at 180 V⋅s. However, to get much higher operation parameters such as steady-state operation, particle and heat exhaust, disruption mitigation and avoidance, ELM control, and material damage by high heat flux and neutron, the superconducting magnet system of CFETR-phase II has been updated based on the larger machine with R = 6.7 m−7.0 m, a = 2.0 m, and BT = 6.5–7 T. Such a new design makes over 1-GW fusion power and better plasma performance possible. Hybrid TF coils are designed by a high-performance Nb3Sn conductor, since the maximum magnetic field can reach to 14–15 T. Besides, in order to save the space for the blanket system and get higher flux, a high-temperature superconducting Bi-2212 magnet with better current-carrying performance under high field is supposed to be employed for the CS coils of CFETR-phase II. The Bi-2212 CICC sample has been tested at 4.2 K with a critical current of 26.6 kA under its self-field.

Influence of modified neutron emission spectrum on tritium production performance in blanket systems with NBI-heated deuterium plasma

In high-temperature plasma sustained by neutral beam injection heating, the emission spectrum of the neutrons produced by fusion reactions is known to have a modified Gaussian distribution. In this study, neutron transport simulation is carried out for a plasma source that emits neutrons with a modified spectrum and the influence of this modified neutron emission spectrum on the tritium production performance is investigated. The results show that in tritium production using D(d,n)3He neutrons, the rate of the 9Be(n,2n)2α neutron multiple reactions, which has a threshold energy of 2.0 MeV, is significantly enhanced compared with that when a source with a Gaussian neutron emission spectrum was assumed. In addition, the influence of the blanket composition on the tritium production performance is discussed.

The effect of wall flow velocity on compatibility of high-purity SiC materials with liquid Pb-Li alloy by rotating disc testing for 3000 h up to 900 °C

The lead-lithium (Pb-Li) alloy is considered as a coolant and a tritium breeder for fusion reactor blanket systems. One of the critical requirements in realization of the systems is the compatibility of silicon carbide (SiC) and its composites, as structural and/or functional materials of the blanket, with liquid Pb-Li flow at high temperatures. This study aims to clarify the corrosion process of high-purity SiC materials under flow condition of liquid Pb-Li for 3000 h at 700 and 900 °C, respectively, by using the rotating disc corrosion equipment. Furthermore, the effect of wall flow velocity on the compatibility was specifically evaluated. For that purpose the decrease rate (about 32–37%) of the wall flow velocity on disc type’s sample by the dragging effect by fluidic viscosity was first calculated by numerical analysis of the rotating flow condition. The wall flow velocity dependence on the reaction layer’s thickness was then evaluated for both high purity SiC and SiC/SiC composites. The test results show that the formation process of the surface reaction layer mainly depends on both the time required for the chemical reaction and the wall flow velocity in the initial stage but the reaction layer thickness tended to approach to the constant regardless of the wall flow velocity with increasing exposed time. In short, no accelerated corrosion was found for about 3000 h. It was also found that there was no significant difference of the temperatures between 700 and 900 °C. Finally the key corrosion mechanism, i.e., the reaction of oxide impurities, e.g., Li2O, in liquid Pb-Li with the Si-oxide layer over SiC, was identified.

Study on lithium rod test module and irradiation method for tritium production using high temperature gas-cooled reactor

Large quantities of tritium are required to start up fusion reactors and conducts engineering tests using tritium for a fusion blanket system. However, tritium is very rare and kg orders of tritium must be produced artificially. Tritium production, by 6Li(n,α)T reaction using a high temperature gas-cooled reactor (HTGR) has been proposed. This method considers the loading of Li rods into burnable poison holes in the HTGR. In this paper, the Li rod suited for use in the High Temperature engineering Test Reactor (HTTR) was designed, and tritium production and leakage from Li-rod capsules were evaluated by adjusting the thicknesses of LiAlO2, alumina, and Zr layers. An irradiation test scenario to be conducted in the HTTR for demonstration of the Li rod’s tritium production and containment performance was presented.

Thermal Hydraulic Safety Assessment of LLCB Test Blanket System in ITER Using Modified relap/scdapsim/mod4.0 Code

This work attempts to investigate the thermal hydraulic safety of lithium lead ceramic breeder (LLCB) test blanket system (TBS) in International Thermonuclear Experimental Reactor (ITER) with the help of modified thermal hydraulic code relap/scdapsim/mod4.0. The design basis accidents, in-vessel and ex-vessel loss of coolant of first wall (FW) of test blanket module (TBM) are analyzed for this safety assessment. The sequence of accidents analyzed was started with postulated initiating events (PIEs). A detailed modeling of first wall helium cooling system (FWHCS) loop and lithium lead cooling system (LLCS) is presented. The analysis of steady-state normal operation along with 10 s power excursion before the accident is also discussed in order to better understanding of initial condition of accidents. The analysis discusses a number of safety concerns and issues that may result from the TBM component failure, such as vacuum vessel (VV) pressurization, TBM FW temperature profile, passive decay heat removal capability of TBM structure, pressurization of port cell and Tokomak cooling water system vault annex (TCWS-VA) and to check the capability of passive safety system (vacuum vessel pressure suppression system (VVPSS)). The analysis shows that in these accident scenarios, the critical parameters have reasonable safety margins.

Tritium permeability measurement in hydrogen-tritium system

Understanding of thermodynamic equilibria of multi-components hydrogen isotopes is required to accurately measure tritium permeability at the expected low tritium partial pressure and non-negligible high hydrogen partial pressure in a fusion blanket system. A gas-driven tritium permeation system that is capable of independently controlling hydrogen and tritium partial pressures was developed at Idaho National Laboratory to accurately measure low partial pressure tritium permeability. The thermodynamic equilibria for hydrogen (H) – tritium (T) permeation through metal are discussed to accurately measure tritium permeability, and the experimental conditions required for evaluating tritium permeability in H-T system are presented.

Non-linear assessment of critical failure modes in the first wall of the European TBM

The assessment of the different components of the Helium Cooled Pebble Bed (HCPB) and Helium Cooled Lithium Lead (HCLL) Test Blanket Modules (TBMs) based on elastic FE analyses has been performed considering main failure modes and respective structural design criteria. This study has been carried out for the operation at full power and for the so-called Approach 2 cooling that is set up at reduced heat extraction capability for compliance with temperature targets to be relevant to DEMOnstration power plant blanket system. Consequently this situation is the worst condition in terms of cooling performance as a result it reveals that many locations in the First Wall (FW) become critical in the sense that certain criteria are not fulfilled, particularly those for immediate plastic flow localization, progressive deformation (ratcheting) and creep-fatigue interaction. Of course one can better reach code compliance by improving the cooling performance (so called Approach 1) but it will degrade the DEMO relevancy.The critical failure modes are re-assessed performing non-linear analysis using an elasto-viscoplatic model, particularly developed for RAFM steels such as the structural material EUROFER97 by Aktaa & Schmitt, and considering the appropriate design criteria of the inelastic route of RCC-MRx and SDC-IC. Despite the conservatism in these criteria, the results show all-clear signal with respect to immediate plastic flow localization and at least much smaller breaches of the limits with respect to the other failure modes. Reporting details and results of the non-linear failure assessments demonstrates their potential and capability in supporting specific and efficient improvements to the design of highly loaded components. The conservatism in the considered design criteria and ways for its reduction are discussed in addition.

Tritium technologies and transport modelling: main outcomes from the European TBM Project

The tritium technologies and transport modelling play an essential role for the achievement of the TBM Programme scientific objectives. Tritium technologies deployed in TBS (Test Blanket System) can be divided in processing systems and measurement devices. In the current design of the European HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) TBS, tritium extraction from flowing Pb-16Li and tritium concentration in He are the most relevant tritium technologies under development. The optimization and integration of real time tritium concentration measurement devices in TBS is a complex task needed to keep under control the tritium concentration in the TBSs and to support their correct operation. For HCLL and HCPB-TBS application, the tritium concentration measurement devices have to operate in flowing Pb-16Li and in He in a wide range of operating conditions.In parallel with the activities on tritium technologies, an important effort has been spent in the last few years in developing tritium transport modelling capabilities, which is one of the most relevant objectives of the European TBM Programme.This paper first of all provides an overview on the most relevant and updated R&D and design results achieved in the above mentioned areas within the European TBM Project. Recent results on tritium inventory and permeation rate in HCLL-TBS, predicted by a 1-D dynamic system simulation code developed within the Ecosimpro platform, are presented and discussed in the part dealing with the TBS tritium transport modelling development.Considerations on DEMO relevancy are also given, particularly for the tritium processing systems as the change of scale from TBM to a breeding blanket for DEMO may impact their feasibility and performance.

Investigation of technical gaps between DEMO blanket and HCCR TBM

Korea has own programs toward DEMO and fusion reactors that will require further improved DEMO blanket and energy utilization systems. Primary goals of DEMO blanket development in Korea are to develop and verify the integrated blanket design tools; to develop blanket materials, cooling and tritium fuel cycle technologies; and to develop and evaluate fabrication and joining technologies. The concept of helium-cooled ceramic reflector (HCCR) blanket is adopted to be tested in ITER as Test Blanket Module (TBM). Currently, the design and R&D activities are mainly performed through the ITER TBM program in Korea. It is expected to demonstrate the major objectives of the breeding blanket: extraction of heat from burning plasma, tritium self-sufficiency and its integrity within the whole systems considering safety features. Although TBM program will provide very essential data and experience, there will be a considerable technical gap from DEMO blanket. This paper presents the technical gap in the main technical categories of the breeding blanket based on the experience of the development of HCCR TBM and breeding blanket. It is found that still about 60–70% of the DEMO blanket technology can be achieved through the HCCR TBM depending on the technology maturity level, and some ways to DEMO blanket from HCCR TBM are proposed.