Tritium breeding control within liquid metal blankets

A key requirement for DEMO is the on-site breeding of tritium. In order to do this, a robust control system must be employed to ensure enough tritium is being bred to sustain the fusion reactor, whilst not breeding an amount which would exceed the plant's tritium inventory license. A tritium breeding method which is cost effective and reduces radioactive waste for disposal is that of the liquid metal breeder such as those based around LiPB and FLiBe. This paper focuses on the modeling of a simplified fusion reactor design with a LiPb blanket with linked radiation transport, nuclide burn-up and control theory. Two simple models were simulated using the FATI code which incorporated a PID (proportional integral derivative) controller that adjusted the Li6/Li7 ratio in order to increase/decrease tritium production based on the difference between the measured excess tritium inventory and the desired excess inventory. The modelling has initially demonstrated that a linear PID controller has the capability to manage tritium production within a LiPb liquid blanket.     

Effect of Lithium Enrichment on the Tritium Breeding Characteristics of Various Breeders in a Fusion Driven Hybrid Reactor

Selection of lithium containing materials is very important in the design of a deuterium–tritium (DT) fusion driven hybrid reactor in order to supply its tritium self-sufficiency. Tritium, an artificial isotope of hydrogen, can be produced in the blanket by using the neutron capture reactions of lithium in the coolants and/or blanket materials which consist of lithium. This study presents the effect of lithium-6 enrichment in the coolant of the reactor on the tritium breeding of the hybrid blanket. Various liquid–solid breeder couples were investigated to determine the effective breeders. Numerical results pointed out that the tritium production increased with increasing lithium-6 enrichment for all cases.     

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

The Indian test blanket module(TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the RD activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices(ITER relevant and DEMO).The Indian Lead–Lithium Cooled Ceramic Breeder(LLCB) blanket concept is one of the Indian DEMO relevant TBM,to be tested in ITER as a part of the TBM program.Helium-Cooled Ceramic Breeder(HCCB) is an alternative blanket concept that consists of lithium titanate(Li_2TiO_3) as ceramic breeder(CB) material in the form of packed pebble beds and beryllium as the neutron multiplier.Specifically,attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions.These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.     

Tritium management and safety issues in ITER and DEMO breeding blankets

Safe, reliable and efficient tritium management in the breeder blanket faces unique technological challenges. Beside the tritium recovery efficiency in the tritium extraction and coolant purification systems, the tritium tracking accuracy between the inner and outer fuel cycle shall also be demonstrated. Furthermore, it is self-evident that safe handling and confinement of tritium need to be stringently assured to evolve fusion as a reliable technique. The present paper gives an overview of tritium management in breeder blankets. After a short introduction into the tritium fuel cycle and blanket basics, open tritium issues are discussed, thereby focusing on tritium extraction from blanket, coolant detritiation and tritium analytics and accountancy, necessary for accurate and reliable processing as well as for book-keeping.     

Engineering studies for integration of the test blanket module (TBM) systems inside an ITER equatorial port plug

The European test blanket module (EU-TBM), first prototype of the breeding blanket concepts under development for the future DEMO power plant to produce the tritium, will be developed to be tested in three equatorial ports of ITER dedicated to this. The CEA Cadarache under the contract of Association EURATOM/CEA and in close relation with Association EURATOM/HAS works on the integration of the EU-TBM inside ITER tokamak.The installation of the TBM into the vacuum vessel is made with the help of a port plug, constituted with two components: the Shield module and the Port-Plug frame. The Shield module provides the neutron shielding inside the Port-Plug frame, which maintains in cantilever position the TBM and its shield module and closes the vacuum vessel port.This paper will describe the EU-TBM design and integration activities on the cooled shield module and on its interface with the TBM component. A particular attention, in term of thermal and mechanical studies, is dedicated to the design of the shield and test blanket module attachment, and also to the shield design and its internal cooling system.     

R&D status on Water Cooled Ceramic Breeder Blanket Technology

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li₂TiO₃ pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.     

Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons

At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different ⁶Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation.     

Preliminary tritium safety analysis on China DFLL-TBM for ITER

The dual-functional lithium-lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R&D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM.     

Measurement of tritium production rate distribution in natural LiAlO2/HDPE assembly irradiated by D-T neutrons

A neutronics experiment was performed to measure the tritium production rate (TPR) profile in the breeder assembly with LiAlO₂ as breeder and high density polyethylene (HDPE) as neutron reflector. The breeder assembly was irradiated with 14 MeV neutrons from DT neutron generator at IPR Neutronics Laboratory. The objective of the experiment was to validate the tritium production prediction capability of the Monte-Carlo code MCNP and FENDL 2.1 data library. The tritium production rate profile in the breeding assembly was measured by irradiating Li₂CO₃ pellets kept at various locations and then tritium counting liquid scintillation technique. Experiment was analyzed with 3D Monte-Carlo code MCNP with FENDL 2.1 cross-section data library. The calculation results were found to agree with the measured tritium production rates except one point near to the source. This experiment is a starting experiment in the series of benchmark experiments for the Indian Demo breeding blanket.     

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

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

氦气、水和熔盐冷却的混合堆中子学性能研究

氦气、水、熔盐(Flibe)在强磁场中流动不存在严重的MHD问题,因此适合在基于磁约束的聚变-裂变混合堆中作为冷却剂.针对氦气、水、Flibe这3种冷却剂对混合堆包层中子学性能的影响进行研究,分析包层中能谱特点及燃料增殖特性.通过燃耗计算,研究氚增殖率(TBR)、能量倍增因子(M)、keff等随运行时间的变化.中子学输运采用三维蒙特卡罗程序MCNP.计算结果表明,不同的冷却剂对混合堆系统中子能谱影响很大:氦冷系统的能谱最硬,主要发生快中子裂变,氚增殖效果最好;水冷系统的能谱最软,产能最多,但需提高TBR;Flibe冷系统的能谱较硬,产能最少.     

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

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

Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

China Fusion Engineering Test Reactor(CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO.One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2to ensure tritium self-sufficiency.A concept design for a water cooled ceramics breeding blanket(WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR.Based on this concept,a one-dimensional(1D) radial built breeding blanket was first designed,and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build.A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models,addressing neutron wall loading(NWL),tritium breeding ratio(TBR),fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components.The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design.     

Neutronic analysis of an inertial fusion energy breeder with SiCf/SiC

Neutronic performance of a blanket driven ICF (Inertial confinement fusion) neutron based on SiC_f/SiC composite material is investigated for fissile fuel breeding. The investigated blanket is fueled with ThO₂ and cooled with natural lithium or (LiF)₂BeF₂ or Li₁₇Pb₈₃ or ⁴He coolant. MCNP4B Code is used for calculations of neutronic data per DT neutron. Calculations have show that values of TBR (tritium breeding ratio) being one of the main neutronic paremeters of fusion reactors are greater than 1.05 in all type of coolant, and the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Calculations show that natural lithium coolant blanket has the highest TBR (1.298) and M (fusion energy multiplication) (2.235), Li₁₇Pb₈₃ coolant blanket has the highest FFBR (fissile fuel breeding ratio) (0.3489) and NNM (net neutron multiplication) (1.6337). ⁴He coolant blanket has also the best Γ (peek-to-average fission power density ratio) (1.711). Values of neutron leakage out of the blanket in all type of coolants are quite low due to SiC reflector and B₄C shielding.     

A neutron point kinetic model for fusion relevant calculations

In the framework of research activities on fusion reactors a great effort is dedicated by the scientific community to the development of tritium breeding blankets. One of the main goals is to assess the neutronic behaviour of such devices to analyse their tritium breeding performance and to evaluate the required data for their thermal–mechanic and thermal–hydraulic design. Many papers have been published on this topic considering some stationary condition to calculate such important quantities as heating power, gas production and dpa rates, tritium breeding ratio, etc., but not much attention has been focussed to neutronic transport analyses in transient conditions. The present paper proposes a simple model based on the point kinetics approximation, which has been set up deriving an alternative formulation of the time-dependent neutron transport equation. This approach allows to define some physical characteristics that can be interpreted in a statistical way, making possible to calculate these quantities numerically by the Monte Carlo method. The adoption of the aforementioned numerical method has the great advantage that complex geometries (as the fusion reactor's blankets are) can be analysed with acceptable computational times. Some simple cases have been investigated to implement the theoretical model proposed with MCNP5 code and to show its potentiality. Then, applications to fusion reactor ITER blanket module and to the Helium Cooled Test Blanket Module, to be tested in ITER, have been taken into account in order to assess their neutronic time-dependent behaviour and the results obtained have been critically discussed.     

Neutronic analysis of PROMETHEUS reactor fueled with various compounds of thorium and uranium

In this study, neutronic performance of the DT driven blanket in the PROMETHEUS-H (heavy ion) fueled with different fuels, namely, ThO₂, ThC, UO₂, UC, U₃Si₂ and UN is investigated. Helium is used as coolant, and SiC is used as cladding material to prevent fission products from contaminating coolant and direct contact fuel with coolant in the blanket. Calculations of neutronic data per DT fusion neutron are performed by using SCALE 4.3 Code. M (energy multiplication factor) changes from 1.480 to 2.097 depending on the fuel types in the blanket under resonance-effect. M reaches the highest value in the blanket fueled with UN. Therefore, the investigated reactor can produce substantial electricity in situ. UN has the highest value of ²³⁹Pu breeding capability among the uranium fuels whereas UO₂ has the lowest one. ²³⁹Pu production ratio changes from 0.119 to 0.169 according to the uranium fuel types, and ²³³U production values are 0.125 and 0.140 in the blanket fueled with ThO₂ and ThC under resonance-effect, respectively. Heat production per MW (D,T) fusion neutron load varies from 1.30 to 7.89 W/cm³ in the first row of fissile fuel breeding zone depending on the fuel types. Heat production attains the maximum value in the blanket fueled with UN. Values of TBR (tritium breeding ratio) being one of the most important parameters in a fusion reactor are greater than 1.05 for all type of fuels so that tritium self-sufficiency is maintained for DT fusion driver. Values of peak-to-average fission power density ratio, Γ, are in the range of 1.390 and ∼1.476 depending on the fuel types in the blanket. Values of neutron leakage out of the blanket for all fuels are quite low due to SiC reflector. The maximum neutron leakage is only ∼0.025. Consequently, for all cases, the investigated reactor has high neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium required for (D,T) fusion reaction.     

Concept for a vertical maintenance remote handling system for multi module blanket segments in DEMO

The anticipated high neutron flux, and the consequent damage to plasma-facing components in DEMO, results in the need to regularly replace the tritium breeding and radiation shielding blanket. The current European multi module segment (MMS) blanket concept favours a less invasive small port entry maintenance system over large sector transport concepts, because of the reduced impact on other tokamak systems – particularly the magnetic coils. This paper presents a novel conceptual remote maintenance strategy for a Vertical Maintenance Scheme DEMO, incorporating substantiated designs for an in-vessel mover, to detach and attach the blanket segments, and cask-housed vertical maintenance devices to open and close access ports, cut and join service connections, and extract blanket segments from the vessel. In addition, a conceptual architectural model for DEMO was generated to capture functional and spatial interfaces between the remote maintenance equipment and other systems. Areas of further study are identified in order to comprehensively establish the feasibility of the proposed maintenance system.     

Preliminary performance analysis and optimization based on 1D neutronics model for Indian DEMO HCCB blanket

India, under its breeding blanket R&D program for DEMO, is focusing on the development of two tritium breeding blanket concepts; namely the lead-lithium-cooled ceramic breeder and the helium-cooled ceramic breeder (HCCB). The study presented in this paper focuses on the neutronic design analysis and optimization from the tritium breeding perspective of the HCCB blanket. The Indian concept has an edge-on configuration and is one of the variants of the helium-cooled solid breeder blanket concepts proposed by several partner countries in ITER. The Indian HCCB blanket having lithium titanate (Li₂TiO₃) as the tritium breeder and beryllium (Be) as the neutron multiplier with reduced-activation ferritic/martensitic steel structure aims at utilizing the low-energy neutrons at the rear part of the blanket. The aim of the optimization study is to minimize the radial blanket thickness while ensuring tritium self-sufficiency and provide data for further neutronic design and thermal-hydraulic layout of the HCCB blanket. It is found that inboard and outboard blanket thicknesses of 40 cm and 60 cm, respectively, can give a tritium breeding ratio (TBR) >1.3, with 60% ⁶Li enrichment, which is assumed to be sufficient to cover potential tritium losses and associated uncertainties. The results also demonstrated that the Be packing fraction (PF) has a more profound impact on the TBR as compared to 6Li enrichment and the PF of Li₂TiO₃.     

Neutronics study for the installation of the LBVM in the medium flux area of IFMIF

IFMIF (International Fusion Materials Irradiation Facility) will be a fusion dedicated facility producing a large amount of neutrons with the appropriate energy spectrum to test materials and subcomponents for DEMO and future Fusion Power Plants.While the high flux area of IFMIF will be devoted to reduced activation structural materials for first wall and blanket, the medium flux area will be dedicated to functional materials for breeder blankets. In particular, the Liquid Breeder Validation Module (LBVM), will host experiments related with functional materials for liquid breeder blankets. Since IFMIF neutron spectra have been intended to fit the most irradiated areas of a fusion reactor in the high flux area, the irradiation conditions in the LBVM placed in the medium flux area of IFMIF have been assessed. The effect of some neutron shifter/reflector components to optimize the neutron spectra have been evaluated in order to find out the proper irradiation conditions for functional materials for liquid breeder blankets.Therefore, the objective of this report is to summarize the neutronic calculations developed to evaluate the viability of IFMIF neutron source to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for future nuclear fusion power reactors (ITER, DEMO). The irradiation parameters evaluated for this purpose are: the tritium production for liquid breeder material (Pb–17Li) and the damage dose (dpa) and gas production to damage dose ratios for Al₂O₃ and SiC functional materials.The main conclusion is that, it is possible to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for the future nuclear fusion reactor DEMO. Nevertless, the use of some shifter components will be needed to optimize some irradiation parameters.     

Physical investigation for neutron consumption and multiplication in fusion–fission hybrid test blanket module

Inelastic scattering of high energy fusion neutrons does affect the performance of fusion blanket based on the choice of different materials. It will also affect the behavior of source neutrons in a subcritical fusion fission hybrid blanket and consequently the transmutation and tritium breeding performance. A fusion fission hybrid test blanket module (HTBM) is designed which is presumed to be tested in a large sized tokamak and plasma neutron source is similar to ITER. In this preliminary design of HTBM the neutron source and loss factors are computed for the detailed neutronic performance analysis. The neutronic analysis of hybrid blanket module is performed for five different TRU fuel types: TRU-Zr, TRU-Mo, TRU-Oxide, TRU-Carbide and TRU-Nitride. In this module design, it is aimed to burn and transmute the TRU nuclides from high-level radioactive waste of PWR spent fuel. The effect of TiC reflector on transmutation and tritium breeding performance of HTBM is also quantified. MCNPX is used for neutronic computations. Neutron spectrum, capture to fission ratio and waste transmutation ratio of each fuel type are compared to evaluate their waste transmutation performance. Tritium breeding ratio is also compared for two coolant options: Li and LiPb eutectic.