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.     

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.     

聚变堆交叉冷却固态包层中子学设计优化

针对聚变堆固态包层设计路线,提出了一个交叉排列氦冷固态包层概念。设计采用Be、Li₂TiO₃分层球床。两种尺寸的氦气冷却管道交叉排列,分两个回路同时冷却,以增加系统安全可靠性。分析比较了4种⁶Li富集度布置方案。结果表明：径向远离第一壁降低⁶Li富集度较为合理,靠近第一壁的增殖层⁶Li富集度不能过低,以减少长期运行中Li的消耗对氚增殖性能的影响。借助蒙特卡罗程序MCNP建立11.25°对称模型,全堆包层氚增殖率为1.176,包层寿期内产氚性能稳定,在包层寿命运行时间内的燃耗分布相对均匀。     

Preliminary design of a helium cooled molten lithium test blanket module for the ITER test in Korea

Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) blanket with ferritic steel (FS) as a structural material in the International Thermonuclear Experimental Reactor (ITER) program. The preliminary design and its performance of KO HCML test blanket module (TBM) are introduced in this paper. It uses He as a coolant at an inlet temperature of 300 °C and an outlet temperature up to 400 °C and Li is used as a tritium breeder by considering its potential advantages. Two layers of graphite are inserted as a reflector in the breeder zone to increase the tritium breeding ratio (TBR) and the shielding performances. A 3-D Monte Carlo analysis is performed with the MCCARD code for the neutronics and the total TBM power is designed to be 0.739 MW at a normal heat flux from the plasma side. From the analysis results with CFX-10 for the thermal-hydraulics, the He cooling path is determined and it shows that the maximum temperature of the first wall does not exceed 550 °C at the structural materials and the coolant velocities are 45 and 11.5 m/s in the first wall and breeding zone, respectively. The obtained temperature data is used in the thermal-mechanical analysis with ANSYS-10. The maximum von Mises equivalent stress of the first wall is 123 MPa and the maximum deformation of it is 3.73 mm, which is lower than the maximum allowable stress.     

Preliminary neutronic assessment of a helium-cooled Li8PbO6 breeding blanket design for DEMO

A preliminary neutronic assessment of the performances of a helium-cooled Li₈PbO₆ breeding blanket for the conceptual design of a DEMO fusion reactor is given. The study mainly focuses on TBR, power density responses and shielding factor optimization to estimate the feasibility of the design under the prescribed radiation deposition limits at TF-coils superconducting magnets. Computational analyses are based on three-dimensional 30° sector using the Monte Carlo code MCNPX 2.6. The scoping interest of helium-cooled Li₈PbO₆ blanket designs is based on a large potential minimization of the amount of Be required and the strong relaxation of ⁶Li enrichment requirements for this solution when compared to other solid breeder blanket options.     

Numerical Study on Effects of Fuel Mixture Fraction and Li-6 Enrichment on Neutronic Parameters of a Fusion–Fission Hybrid Reactor

This study presents the effects of mixture fractions of nuclear fuels (mixture of fissile–fertile fuels and mixture of two different fertile fuels) and ⁶Li enrichment on the neutronic parameters (the tritium breeding ratio, TBR, the fission rate, FR, the energy multiplication ratio, M, the fissile breeding rate, FBR, the neutron leakage out of blanket, L, and the peak-to-average fission power density ratio, Γ) of a deuterium–tritium (D–T) fusion neutron-driven hybrid blanket. Three different fertile fuels (²³²Th, ²³⁸U and ²⁴⁴Cm), and one fissile fuel (²³⁵U) were selected as the nuclear fuel. Two different coolants (pressurized helium and natural lithium) were used for the nuclear heat transfer out of the fuel zone (FZ). The Boltzmann transport equation was solved numerically for obtaining the neutronic parameters with the help of the neutron transport code XSDRNPM/SCALE4.4a. In addition, these calculations were performed by also using the MCNP4B code. The sub-limits of the mixture fractions and ⁶Li enrichment were determined for the tritium self-sufficiency. The considered hybrid reactor can be operated in a self-sufficiency mode in the cases with the fuel mixtures mixed with a fraction of equal to or greater than these sub-limits. Furthermore, the numerical results show that the fissile fuel breeding and fission potentials of the blankets with the helium coolant are higher than with the lithium coolant.     

Three-dimensional neutronic calculations for a fusion breeder APEX reactor using some libraries

The effects of evaluated nuclear data files on neutronics characteristics of a fusion–fission hybrid reactor have been analyzed; three-dimensional calculations have been made using the MCNP4C Monte Carlo Code for ENDF/B-VII T = 300 K, JEFF-3.0 T = 300 K, and CENDL-2 T = 300 K evaluated nuclear data files. The nuclear parameters of a fusion–fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, fissile fuel breeding and nuclear heating in a first wall, blanket and shield have been investigated for the mixture components of 90% Flibe (Li₂BeF₄) and 10% UF₄ for a blanket layer thickness of 50 cm. The contributions of each isotope of Flibe (⁶Li, ⁷Li, ¹⁹F, ⁹Be) and UF₄ (²³⁵U, ²³⁸U) to the integrated parameter values were calculated. The neutron wall load is assumed to be 10 MW/m².     

Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR).Some updating of neutronics analyses was needed,because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket,including the optimization of radial build-up and customized structure for each blanket module.A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses.The tritium breeding capability,nuclear heating power,radiation damage,and decay heat were calculated by the MCNP and FISPACT code.The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency.The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW.The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60,respectively.The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module # 3.The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time.The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years.     

Neutronic optimization for a blanket of the laser fusion reactor,SENRI-I

Neutronic calculations were performed to optimize the SENRI blanket in terms of energy multiplication as well as tritium breeding ratio. The blanket employs a thick ( 64-cm) Li layer as breeder/coolant. Three approaches were taken here to achieve the goal: (1) reduction of⁶Li in the lithium, (ii) replacement of the Li layer by a molten-salt (flibe) layer, and (iii) shipment of excess tritium to a nonbreeding blanket. It was found that the excess tritium produced in the SENRI blanket could be used effectively to obtain additional power by fueling a nonbreeding D-T reactor.     

Radiation Damage and Tritium Breeding Study in a Fusion Reactor Using a Liquid Wall of Various Thorium Molten Salts

A new magnetic fusion reactor design, called APEX uses a liquid wall between fusion plasma and solid first wall to reach high neutron wall loads and eliminate the replacement of the first wall structure during the reactor’s operation due to the radiation damage. In this paper, radiation damage behavior of the inboard and outboard first walls made of a ferritic steel, 9Cr-2WVTa, in the APEX blanket using various thorium molten salts, 75% LiF-25% ThF₄, 75% LiF-24% ThF₄-1% ²³³UF₄ and 75% LiF-23% ThF₄-2% ²³³UF₄ was investigated. Furthermore, tritium breeding potential of these salts in such a blanket was also examined. Computations were carried out using the code Scale 4.3 by solving Boltzmann neutron transport equation. Numerical results brought out that only the liquid wall containing the molten salt, 75% LiF-23% ThF₄-2% ²³³UF₄ and having a thickness of ≥38 cm would be suitable to be used in the APEX reactor with respect to radiation damage criteria for the first wall structures and tritium self-sufficiency for the (DT) fusion driver.     

A tritium permeation model for conceptual fusion reactor designs

A transient tritium permeation model is developed based on a simplified conceptual DT-fueled fusion reactor design. The major design features described in the model are a solid breeder blanket, a low pressure purge gas in the blanket, and a high pressure helium primary coolant. Tritium inventory in the breeder is considered to be due to diffusive hold-up and solubility effects. It is assumed that diffusive hold-up is the dominant factor in order to separate the solution for the breeder tritium concentration. The model was applied to the STARFIRE-Interim Reference Design, whose system parameters yielded a breeder tritium inventory on the order of grams, based on an average pellet radius of 10^?3 cm. The breeder pellets reach their steady-state tritium content in approximately 1.4×10⁴ s from system start-up, assuming continuous full power operation. Both the steady-state breeder tritium concentration and the time to reach that steady-state are proportional to the pellet radius squared. Other candidate solid breeders were considered, and their effect on the blanket tritium inventory was noted. The addition of oxygen to the primary coolant loop was required in order to keep the tritium losses through the heat exchanger to within the design goal of 0.1 Ci/day.     

聚变堆氦冷固态包层结构和6Li富集度对产氚率的影响

在聚变堆固态包层基本参数基础上，建立简化20°模型，包层分第1壁装甲、第1壁冷却板、氚增殖区和支撑结构。分别选择Li₄SiO₄和Li₂O做增殖材料，应用MCNP程序，研究第1壁结构布置和6Li富集度对产氚率的影响。结果表明：₆Li富集度适宜选择在30％~80％之间；第1壁选择Be装甲可提高产氚率；冷却管板的厚度应取3cm以下，以避免对产氚造成不利的影响。     

Solid breeder thermal transient under purge-line-break accidents

The breeder thermal performances under a purge line break have been analyzed for two blanket design options: a blanket design using a packed breeder bed and a blanket design using a sintered breeder product. Under a purge line break open to a vacuum environment, the packed bed breeder temperature exceeds its operating temperature limit at a faster rate than that of the sintered breeder blanket design for the same breeder temperature gradient. Depending on the breeder material and nominal operating conditions, the breeder reaches its maximum operating temperature in time ranging from 32 seconds to 125 seconds for a break area of 10 cm² in packed bed designs. However for the sintered product design, the consequence of this transient might not result in the breeder exceeding its maximum operating temperature if a reasonable contact pressure could be established at the interface. To reduce the safety hazards, the tritium concentration build up in the vacuum vessel in conjunction with the purge gas pressure inside the blanket module should be used as a measure for initiating the reactor shutdown for this type of accident. The consequence of the purge line break outside the vacuum vessel on the breeder transient thermal performance is less significant because of a longer transient time involved.     

On the proliferation issues of a fusion fission fuel factory using a molten salt

The fusion fission fuel factory (FFFF) is a hybrid fusion fission reactor using a neutron source, which is in this case taken similar to the source of the Power Plant Conceptual Study - Water Cooled Lithium Lead (PPCS-A) design, for fissile material production instead of tritium self-sufficiency. As breeding blanket the first wall of the ITER design is attached to a molten salt zone, in which ThF₄ and UF₄ solute salts are transported by a LiF-BeF₂ solvent salt. For this blanket design, the fissile material is assessed in quantity and quality for both the Th-U and the U-Pu fuel cycle.The transport of the initial D-T fusion neutrons and the reaction rates in this breeding blanket are simulated with the Monte Carlo code MCNP4c2. The isotopic evolution of the actinides is calculated with the burn-up code ORIGEN-S.For the Th-U cycle the bred material output remains below 10 g/h with a ²³²U impurity level of 30 ppm, while for the U-Pu cycle supergrade material is produced at a rate up to 100 g/h.     

Conceptual design of a water cooled breeder blanket for CFETR

China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by ⁶Li(n,α)T reaction. Li₂TiO₃ pebbles and Be₁₂Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li₂TiO₃ and Be₁₂Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be₁₂Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into account to replace Be plate in viewpoint of safety. In this contribution, study on neutronics and thermal design for a water cooled breeder blanket with superheated steam is reported.     

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.     

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.     

Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor(CFETR) operating on a Deuterium-Tritium(D-T) fuel cycle. It is necessary to study the tritium breeding ratio(TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder(WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket,the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code(MCNP) and the fusion activation code FISPACT-2007.The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation.In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW.     

CFETR氦冷固态增殖剂包层产氚性能研究及优化

氦冷固态增殖剂包层是中国聚变工程实验堆（CFETR）的3种候选包层概念之一。本文基于中国核工业西南物理研究院提出的一种氦冷固态增殖剂包层概念，通过蒙特卡罗输运程序MCNP5建立了包层三维中子学模型，探究了不同几何布置方案及结构设计参数对包层产氚性能的影响，得到了全堆氚增殖比（TBR）及极向各包层模块产氚分布，并由优化后的模型得到了包层模块核热分布。结果表明，优化后的TBR达到1.177，满足氚自持的最低要求。     

Optimized fissile and fusile breeding in a laser-fusion fissile-enrichment flux trap blanket

Optimization of fissile and fusile production in the SOLASE-H laser-fusion fissile-enrichment fuel-factory blanket is carried out. The objective is maximizing fissile breeding with the constraints of maintaining self-sufficiency in tritium production, and realistically accounting in the modeling for structural and coolant compositions and configurations imposed by the thermal-hydraulic and mechanical designs. The effect of radial and axial blanket zone thicknesses on fusile and fissile breeding is studied using a procedure which modifies the zones' effective optical thicknesses, rather than the actual three-dimensional geometrical configurations. A tritium yield per source neutron of 1.08 and a Th (n, ) reaction yield per source neutron of 0.43 can be obtained in such a concept, where ThO₂ Zircaloy-clad fuel assemblies for light water reactors (LWRs) are enriched in the²³³U isotope by irradiating them in a lead flux trap. This corresponds to 0.77 kg/[MW(th)-year] of fissile fuel production, and 1.94 years of irradiation in the fusion reactor to attain an average 3 w/o fissile enrichment in the fuel assemblies. For a once-through LWR cycle, a support ratio of 2–3 is estimated. However, with fuel recycling, more attractive support ratios of 4–6 may be attainable for a conversion ratio of 0.55, and of 5–8 for a conversion ratio of 0.70. These estimates are lower than those reported, around 20, for related designs.