Helium Cooled Ceramic Breeder Research Articles

Preliminary accident analyses of in-vessel LOCA for the chinese fusion test reactor HCCB blanket concept

For the helium-cooled ceramic breeder (HCCB) blanket concept improved in 2018 and the primary heat transfer system following China Fusion Engineering Test Reactor (CFETR) in 2019, an in-vessel loss of coolant accident (LOCA) has been investigated with the assumption of a break of the first wall (FW) coolant channels and a double-ended guillotine break of a helium cooling system (HCS) pipe in vacuum vessel (VV). The break leads to helium inject into the VV. After the accident occurs, the helium release leads a plasma disruption and pressure increase in the VV. This paper presents the analysis results of the in-vessel LOCA accidents in CFETR and gives the accident sequence compare with the break in the FW and header pipe. The accident analysis and influence compares can provide a reference for the iterative design.

Electromagnetic analysis of HCCB TBM-set during plasma major disruptions

Electromagnetic (EM) loads (i.e., Lorentz and Maxwell loads) on the Chinese (CN) Helium-Cooled Ceramic Breeder (HCCB) Test Blanket Module set (TBM-set) as well as on the installation Frame are numerically estimated using the ANSYS software. ITER Global Model (IGM) based on the edge-flux formulation is utilized in EM analysis. The current Preliminary Design (PD) model of HCCB TBM-set is taken into account. Four typical plasma Major Disruptions (MDs) are considered in accordance with the system load specifications. It is discovered that both kinds of EM loads are dominated by the radial component. The MD_DW_III is the worst case in terms of Lorentz loads, while the considered MDs have almost the same maximum Maxwell loads which appear under the condition of plasma start of burn. Regarding the EM loads on two TBMs, the Maxwell Force (MF) is much larger compared with the Lorentz Force (LF), however the Maxwell Torque (MT) is far smaller than the maximum Lorentz Torque (LT). The obtained results will be applied in corresponding structural analysis to ensure the integrity of HCCB TBM-set in the further work.

Three-dimensional thermal-hydraulics/neutronics coupling analysis on the full-scale module of helium-cooled tritium-breeding blanket

Blanket is of vital importance for engineering application of the fusion reactor. Nuclear heat deposition in materials is the main heat source in blanket structure. In this paper, the three-dimensional method for thermal-hydraulics/neutronics coupling analysis is developed and applied for the full-scale module of the helium-cooled ceramic breeder tritium breeding blanket (HCCB TBB) designed for China Fusion Engineering Test Reactor (CFETR). The explicit coupling scheme is used to support data transfer for coupling analysis based on cell-to-cell mapping method. The coupling algorithm is realized by the user-defined function compiled in Fluent. The three-dimensional model is established, and then the coupling analysis is performed using the paralleled Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC). The results reveal the relatively small influence of the coupling analysis compared to the traditional method using the radial fitting function of internal heat source. However, the coupling analysis method is quite important considering the nonuniform distribution of the neutron wall loading (NWL) along the poloidal direction. Finally, the structure optimization of the blanket is carried out using the coupling method to satisfy the thermal requirement of all materials. The nonlinear effect between thermal-hydraulics and neutronics is found during the blanket structure optimization, and the tritium production performance is slightly reduced after optimization. Such an adverse effect should be thoroughly evaluated in the future work.

Preliminary neutronics/thermal-hydraulics design and analyses of the helium-cooled mixed bed breeder blanket for CFETR under 1.5 GW fusion power

Chinese Fusion Engineering Test Reactor (CFETR) is a tokamak device designed with different power levels to validate fusion engineering technologies in China. Helium-cooled ceramic breeder (HCCB) blanket and Water-cooled ceramic breeder (WCCB) blanket are the main candidates. Compared with WCCB, HCCB generally has a higher tritium breeding ratio (TBR) because helium has smaller neutron absorption cross sections than water. However, the TBR of the HCCB decreases with the increase of the fusion power. In this paper, the preliminary neutronics and thermal-hydraulics design of the Helium-Cooled Mixed bed Breeder (HCMB) blanket is proposed. The blanket structure with a fusion power of 1.5 GW is designed based on the current HCCB design of the CFETR. The HCMB adopts helium as the coolant and the mixed Li4SiO4/Be12Ti as the tritium breeder. The neutronics and thermal-hydraulics performances of HCMB blanket have been analyzed in the paper. The results show that the HCMB blanket design has a TBR up to 1.25 with the thermal-hydraulics requirements satisfied.

An update on the latest and most relevant improvements achieved in the European helium-cooled ceramic breeder TBM design

The European TBM (Test Blanket Module) program under development will test two alternative concepts for tritium generation and extraction in ITER: a helium cooled / ceramic pebble breeder concept and a water cooled / lithium lead breeder concept.The former implements Advanced Ceramic Breeder pebbles that act as tritium breeding material, and Beryllium pebbles as neutron multiplier. A Reduced Activation Ferritic Martensitic steel, the EUROFER-97 (X10CrWVTa9-1), is used as structural material and pressurized Helium technology is implemented for efficient heat extraction (300-500°C, 8 MPa).The design of the TBM box involves challenges that must be tackled with a holistic approach to simultaneously consider a variety of aspects. Some of these challenges include: the thermal performance and the structural integrity of the TBM under ITER normal operations and accidental conditions, the implementation of pressurized equipment regulation(s), the convergence with manufacturing activities being developed in parallel and the high congestion in the reduced TBM volume –especially when considering instrumentation integration– without affecting too much the tritium breeding capability. It is also key to develop advanced methodologies based on design by analysis to virtually test alternative design solutions in an agile manner. These should be linked to the applicable nuclear design code (RCC-MRx) to consolidate design choices. In this way, a traceable evolution of the TBM design process can be built where alternatives are tested, compared and either discarded or selected in a consistent and comprehensive framework.This paper presents an overview of the very significant progress achieved in the design of the helium cooled concept TBM over the last two years, a period that corresponds to the first steps of the preliminary design phase towards the Preliminary Design Review (PDR) gate. This includes a significant change in the design of the back manifold area and the connection with the rest of the TBM-Set that solves long-standing historical problems inherited from pre-conceptual design stage. An update on the latest and most relevant methodological developments achieved is also provided in this paper, since these have been key to progress with the design evolutions in a sound and traceable manner. To end, the paper also presents some examples of the instrumentation definition and its complex integration within and outside the TBM box.

Application of perturbation theory and simulated annealing algorithm on neutronics optimization for CFETR HCCB TBB

For the neutronics design of a tritium breeding blanket (TBB) of fusion reactor, it is an essential object to produce enough tritium to achieve the tritium self-sufficiency with the tritium breeding ratio (TBR) >1. This paper presents an application study of perturbation theory and simulated annealing (SA) algorithm on neutronics optimization for the China Fusion Engineering Test Reactor (CFETR) helium-cooled ceramic breeder (HCCB) TBB. Firstly, an innovative method of neutronics optimization on the basis of the perturbation theory and the SA algorithm was studied. Next, the neutronics optimization code for the solid breeder TBB was developed, and both the convergence stability and the optimization performances were verified. Finally, the neutronics optimization application on a CFETR HCCB TBB typical module was performed. Compared with the optimized results by the hill-climbing algorithm based on neutronics transport calculations, the SA algorithm based on the perturbation calculations shows a stronger ability for TBR enhancement and demonstrates lower computational costs and higher optimization efficiency.

Mixed hydrogen isotopes plasma-driven permeation through CLF-1 RAFM steel for ITER HCCB TBM

The first wall of the helium-cooled ceramic breeder test blanket module (TBM) will be subjected to charge exchange neutral from the mixed deuterium (D) and tritium (T) deuterium–tritium (D–T) mixture plasma irradiation. To understand the isotope effects on mixed plasma-driven permeation (PDP) through the TBM wall, which is typically made by reduced activation martensitic/ferritic steel (RAFM), mixed D and protium (H) PDP experiments for a Chinese RAFM CLF-1 have been performed. Permeation fluxes of H and D through ∼mm thick CLF-1 have been recorded at various sample temperatures and particle incident energies. The co-permeation of H and D is found to be diffusion-limited in the bulk and obeys classical mass effects of diffusion species. The H/D ratio of the permeation fluxes is found to be stable when changing sample temperature and is insensitive to ion incident energy. In addition, the mixed H can provide more channels for D recombination to reduce the D permeation.

Maxwell force analysis of CN HCCB TBM based on preliminary design

The Chinese (CN) Helium Cooled Ceramic Breeder (HCCB) Test Blanket Module (TBM) will be tested in ITER according to the ITER TBM program. In order to well achieve the test goals and to meet the ITER technical requirements, the design of CN HCCB TBM is continuously updating since 2004 and it is currently in preliminary design (PD) phase. As an in-vessel component of ITER, TBM will be exposed to strong magnetic field. Consequently, large electromagnetic (EM) loads may generate on TBM structures and should be quantitatively evaluated to provide input for the structural integrity assessment of TBM. In this paper, the Maxwell force (MF), one of the EM loads, is numerically calculated and discussed. To improve the calculation efficiency, the finite element (FE) model is simplified by ignoring the cooling channels. Hence, the corrected B-H curve based on effective volume ratio of ferrite material is adopted to exclude the influence caused by absence of cooling channels. Current obtained results are comparable with those in previous work and agree well with analytical ones also, which indicates that the global FE model developed here is acceptable and the analytical equation derived in this paper can be used to approximately predict the MF on TBM.

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.

Study on Multiphysics Coupling and Automatic Neutronic Optimization for Solid Tritium Breeding Blanket of Fusion Reactor

A tritium breeding blanket (TBB) is an essential component in a fusion reactor, which has functions of tritium breeding, energy generation and neutron shielding. Tritium breeding ratio (TBR) is a key parameter to evaluate whether the TBB could produce enough tritium to achieve tritium self-sufficiency (TBR > 1) for a fusion reactor. Current codes or software struggle to meet the requirements of high efficiency and high automation for neutronic optimization of the TBB. In this paper, the multiphysics coupling and automatic neutronic optimization method study for a solid breeder TBB is performed, and a corresponding code is developed. A typical module of China fusion engineering test reactor (CFETR) helium cooled ceramic breeder (HCCB) TBB was selected, and a 3D neutronics model of an initial scheme is developed. The automatic neutronic optimization was performed by using the developed code for verification. Results indicate that the TBR could increase from 1.219 to 1.282 (~5.17% improvement), and that the maximum temperature of each type of material in the optimized scheme is below the allowable temperature. It is of great scientific significance and engineering value to explore and study the algorithm for automatic neutronic optimization and the code development of the TBB.

Neurtonics/Thermal-hydraulic analyses of the CFETR HCCB blanket for multiple operation modes under the poloidal nonuniform neutron wall loading condition

Helium Cooled Ceramic Breeder (HCCB) blanket is one significant component of China Fusion Engineering Test Reactor (CFETR). The neutronics and thermal-hydraulic performances of the blanket are the bases of the safe operation for CFETR, and both of them are influenced by the neutron wall loading (NWL) distribution on the blanket. In our previous work, the influences of the poloidal nonuniform NWL on the neutronics performances of the CFETR HCCB blanket have been preliminarily investigated. However, as the nuclear heating rate in each blanket component is adopted as the internal heat source for the thermal-hydraulic analyses, it's necessary to further investigate the influences of the poloidal nonuniform NWL on the thermal-hydraulic performances of the blanket. Besides, CFETR aims to cover multiple fusion powers (from 200 MW to 1.5 GW) at the present stage. Therefore, it's also essential to compare the different influences of the poloidal nonuniform NWL on blanket performances for multiple operation modes.

Annihilation kinetics of irradiation defects in promising tritium breeding pebbles

• γ-Ray irradiation mainly introduces E′-center and O-related (O - -center, O 2 - center et.al) defects in Li 2 TiO 3 , Li 4 SiO 4 and Li 2 TiO 3 -Li 4 SiO 4 pebbles. • Isochronal and isothermal annealing experiment were carried out to illustrate the annihilation kinetics of different breeding pebbles. • Li 2 TiO 3 pebbles have a better irradiation stability than Li 4 SiO 4 pebbles. Lithium-based tritium breeding materials will be adopted to produce tritium in the D-T fusion reactor. Li 2 TiO 3 and Li 4 SiO 4 pebbles have been proposed as candidates in water-cooled ceramic breeder blanket (WCCB) and helium-cooled ceramic breeder blanket (HCCB) respectively. Biphasic ceramics of core–shell Li 2 TiO 3 -Li 4 SiO 4 have been considered as advanced breeding materials due to combining the superiority of Li 2 TiO 3 and Li 4 SiO 4 . The defects will be introduced into tritium breeders in the operation of fusion reactor. They have important effects on tritium release. It is necessary to carry out the experiment on irradiation defects evolution. The annihilation kinetics of defects induced by γ-ray irradiation were investigated. Fluka and Flair were used to calculate the DPA (displacement per atom) of these irradiated pebbles. EPR (Electron Paramagnetic Resonance) characterization and EasySpin simulation were adopted to analyze the evolution and annihilation kinetics processes of irradiation defects. The concentration of defects decreased as the annealing temperature increased. There were still a certain amount of defects in Li 4 SiO 4 and Li 2 TiO 3 -Li 4 SiO 4 when the defects in Li 2 TiO 3 disappeared by annealing. Li 2 TiO 3 pebbles have a better irradiation stability than that of Li 4 SiO 4 pebbles. According to the results of EasySpin simulation, the defect concentration of E’-center and O-related was obtained respectively. The kinetics parameters for the defects of E’-center and O-related center in Li 2 TiO 3 and Li 4 SiO 4 were acquired. The evolution behavior of Ti 3+ in core–shell Li 2 TiO 3 -Li 4 SiO 4 contributes to the recovery of defects. The correlation between annihilation of irradiation defects and tritium release was discussed. The anti-irradiation damage stability of three kinds of ceramic breeders were evaluated. This work carries out the research on the defect kinetics of the new core–shell structure tritium breeder. Meanwhile, a comprehensive comparison of the annihilation properties of three promising breeders has been made. Li 2 TiO 3 pebbles present excellent tritium release performance due to higher annihilation rate constant. Compared the performances of these breeder pebbles, Li 2 TiO 3 plays a positive role in irradiation tolerance and tritium release performance, and Li 4 SiO 4 has higher lithium density which benefit for tritium production.

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

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

Preliminary thermal-hydraulic influence of the poloidal nonuniform neutron wall loading on the CFETR HCCB blanket

Abstract Helium Cooled Ceramic Breeder (HCCB) blanket, which has been determined as one of the blanket candidates for China Fusion Engineering Test Reactor (CFETR), is the critical component of the fusion reactor. The neutronics and thermal-hydraulic performances of the blanket are the bases of the safe operation, and they are also the initial boundary conditions for the other performance analyses. Both performances are influenced by the neutron wall loading (NWL) distribution on the blanket and the radial structural arrangement of the internal functional zones. In our previous work, the influences of the poloidal nonuniform NWL on the neutronics performances of the blanket have been investigated preliminarily. However, as the nuclear power deposition in each blanket component is the internal heat source for the thermal-hydraulic analyses, it’s essential to further investigate the influence of the poloidal nonuniform NWL on the thermal-hydraulic performances of the blanket. In this paper, taking the typical CFETR HCCB blanket module as the research object, firstly, both the neutronics and the thermal-hydraulic models are established. For investigating the influence of the poloidal nonuniform NWL on blanket performances, both models are divided into five identical parts along the poloidal direction. Afterwards, by referencing the actual poloidal NWL distribution in CFETR, we have aggregately proposed three kinds of poloidal nonuniform NWL distribution forms, including two symmetrical and one stepped NWL. On these bases, both the neutronics and the thermal-hydraulic analyses under the poloidal nonuniform NWL distributions are conducted, and the calculation results are compared with those corresponding to the uniform NWL distribution. The comparison results show that the maximum temperature of every functional zone calculated under the three nonuniform NWL distributions is generally higher than that corresponding to the uniform NWL distribution, which means that the thermal-hydraulic results calculated under the simplified uniform NWL distribution are not conservative. Under the condition of nonuniform NWL distributions, there is obvious poloidal temperature nonuniformity in every functional zone. Besides, the temperature deviations between the stepped and the uniform NWL distributions are generally larger than those between the symmetrical and the uniform ones. This paper can provide some valuable references for the detailed engineering design, analyses, and optimizations of the CFETR HCCB blanket.

Mechanical assessment of first wall / blanket for DEMO

As a critical part of a breeding blanket of fusion reactor or DEMO, it is a common method to design and optimize the first wall separately before the mechanical evaluation of the fully integrated blanket. To understand the difference between the simplified first wall models and the fully integrated first wall/blanket model, the differences of the mechanical behavior between the simplified and integrated models are studied in this paper. The results provide a reference for the conceptual design and optimization of the first wall. Based on the HCCB (helium-cooled ceramic breeder) DEMO blanket model, the mechanical analysis of the integrated model was carried out, and the relatively accurate mechanical characteristics of the first wall for M-type and C-type damage were obtained. Considering the results of the integrated model as a basis, the mechanical analysis of the M-type and C-type for the 2 typical simplified model was carried out. For the separate first wall model, the maximum differences of M-type and C-type are -19% and -17%, respectively. For the most simplified model, the maximum differences of M-type are -17%, the differences of C-type are seriously deviated from the results of the integrated model. The results indicate that one quarter first wall model with reasonable constraints is closer to the results of the fully integrated FW/blanket model. The analysis results of one quarter first wall model can be used as a reference for the conceptual design of the first wall and can be utilized to optimize the first wall design based on the stress limits for prevention of the M-type and C-type damage modes.

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

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

Three-dimensional validation and analyses of the optimized CFETR HCCB blanket

The Helium Cooled Ceramic Breeder (HCCB) blanket is the critical component of the China Fusion Engineering Test Reactor (CFETR). The radial structural layout of the internal functional zones has the largest impact on the neutronics and thermal-hydraulic performance of the blanket, and it’s also the major determinant for the detailed 3D design. In the previous work, NTCOC, a Neutronics/Thermal-hydraulic Coupling Optimization Code, has been developed for application in the radial structural layout optimization of the CFETR HCCB blanket. However, for increasing the optimization efficiency, the NTCOC employs the simplified 1D neutronics and 2D thermal-hydraulic calculation models, the results of which may differ from the calculation results of the 3D actual models. Therefore, it’s essential to verify the reliability of the simplified models by comparing their calculation results with the corresponding 3D models. In this work, both the 3D neutronics and thermal-hydraulic analyses corresponding to the radial structural optimization processes of the CFETR HCCB blanket are performed and the corresponding calculation results are compared with NTCOC. The results show that adopting the 1D neutronics model in NTCOC can reflect the change tendency of the tritium breeding performance during the radial optimization processes accurately. However, the radial temperature distribution which is calculated by the 3D thermal-hydraulic model is generally higher than the NTCOC results, which means that the direct application of the 2D thermal-hydraulic model in NTCOC is not conservative. For solving this problem, a 3D thermal revision is added to the previous optimization method in the form of feedback. After revision, the 3D calculated temperature of the newly obtained blanket scheme is generally lower than the NTCOC’s, which means that the previous problem has been well solved. This work comprehensively verifies the reliability of the NTCOC and the improved optimization method after the 3D thermal revision for application in the radial structural optimization of the CFETR HCCB blanket.

Neutronics influence of poloidal nonuniform neutron wall loading on helium-cooled ceramic breeder blanket for CFETR

Abstract After its conceptual design, the Chinese Fusion Engineering Test Reactor (CFETR) is now at the stage of detailed engineering design. During this stage, many detailed factors should be considered for the components in the CFETR. In this paper, the neutronics influence of poloidal nonuniform neutron wall loading (NWL) is studied for helium-cooled ceramic breeder (HCCB) blanket of the CFETR. The blanket is divided into five parts along the poloidal direction. Each part is composed of the armor, the first wall, tritium breeders, neutron multipliers, and the manifold in the flow distribution zone. The nuclear heat deposition and the tritium breeding performance of the CFETR HCCB blanket are studied with symmetrical and stepped NWL distributions. The results obtained from the uniform NWL distribution are also presented as the reference. For each NWL distribution, the nuclear heat deposition and the tritium breeding capability in different blanket parts are compared. The results show that the nonuniform NWL distribution has almost no influence on the total nuclear heat deposition or the total tritium production. The overall variation is less than 1 ‰ compared to the results under the uniform NWL distribution. However, the local performances of different blanket parts are affected obviously. The maximum variation is over 20 %, which should be focused on during the design for the flow distribution of helium coolant and purge gas. The present work aims to provide an important guidance for the detailed design and structural optimization of the blanket for the CFETR, especially for consideration of the high-level fusion power during phase II (over 1 GW).

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.

A new integrated SSCs safety classification method and a case study on CN HCCB TES

Systems, Structures and Components (SSCs) have different safety significance for reactor operation, and it is necessary to classify SSCs into different safety classes for design, manufacture and maintenance of a reactor. In order to further identify the safety classes of SSCs in reactors, a new integrated safety classification method was proposed in this study, based on the 50-SG-D1 safety classification method, risk-informed safety classification method and integrating relative importance and functions importance classification. To demonstrate the advantages of the proposed method, the China Helium Cooled Ceramic Breeder Tritium Extraction System (CN HCCB TES) in the International Thermonuclear Experimental Reactor (ITER) was analyzed. New safety classes with more comprehensive information were obtained by applying the proposed method, indicating that the new safety classes of the items obtained by the proposed safety classification method could be of more significances than those derived from the present safety classification method.