Helium Cooled Solid Breeder Blanket Research Articles

The effect of Li2TiO3 pebbles on the mechanical properties of ARAA under breeding blanket conditions

The breeding blanket serves as a critical element for achieving a self-sustaining tritium cycle and generating sufficient energy in fusion reactors. This study focuses on candidate materials of the helium-cooled solid breeder blanket designed for a demonstration fusion reactor breeding blanket in Korea. We conducted compatibility tests between advanced reduced-activation alloy (ARAA) and Li2TiO3 pebbles at 550 °C for 28 days in He-0.1 vol% H2 purge gas conditions. Surface oxidation on ARAA was observed regardless of pebble presence. Double oxide films comprised of Fe-rich and Cr-rich layers formed on the surface of ARAA with pebbles. Lithium deposition was verified in both contact and non-contact areas of ARAA. Despite surface oxidation, the impact on tensile strength and elongation was found to be negligible.

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER, considering the 400 s ON time and 1400 s dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.

Neutronics optimization study on the first wall design of CFETR for TBR enhancement

Tritium breeding capability is one of the most important objectives for fusion reactor design, and the tritium self-sufficiency has been set as a crucial target for the China Fusion Engineering Test Reactor (CFETR). The first wall (FW) is the first component of the blanket facing plasma, which bears high heat flux and high neutron current. Due to the neutrons entering into the blanket pass through the FW and then absorbed by the breeder materials, the FW design directly influences the blanket neutronics and tritium breeding performance. In this study, the helium-cooled solid breeder (HCSB) blanket of CFETR is selected as the reference design, and the impacts of the FW design parameters on tritium breeding performance have been studied comprehensively. A new conceptual mixed armor design and FW neutronics optimization guidelines have been proposed to enhance the tritium breeding ratio (TBR). Based on the mixed armor design and the optimization guidelines, the preliminarily optimized helium-cooled FW design shows about 9% TBR enhancement.

Development of a three-dimensional method for thermal-hydraulics/neutronics coupling analysis and its application on CFETR helium-cooled solid breeder blanket

Blankets are of practical importance in the fusion reactor due to their main functions for engineering application. As the fusion power designed for the Chinese Fusion Engineering Test Reactor (CFETR) is increased from 200 MW to 1.5 GW during its phase II operation, more challenges are issued for blankets from the points of view of thermal-hydraulics, neutronics, mechanics, and safety. In this paper, the Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC) was developed using the user-defined function (UDF) in the CFD software FLUENT. The explicit coupling scheme based on cell-to-cell mapping approach was adopted to transfer data between neutronics calculation and thermal-hydraulics simulation. Three-dimensional thermal-hydraulic model and neutronics model were established using Fluent and MCNP4C, respectively. The helium-cooled solid breeder (HCSB) blanket of CFETR was simplified as a slice model to reduce the computing resources. With the slice model of CFETR HCSB blanket, the coupling analysis was carried out using the developed CATNIC. The calculation results from coupling and non-coupling analyses were compared to verify the feasibility of the coupling method in this study. As the helium coolant was considered as the transparent medium for neutron, the thermal-hydraulic and neutronics analyses were both converged after two iterations. However, using the developed coupling method in this study, the detailed change of tritium production ratio (TPR) and nuclear heat deposition in different parts of the blanket was revealed, which could provide important guidance for the design of the distribution for both coolant and purge gas.

A Two-Way Neutronics/Thermal-Hydraulics Coupling Analysis Method for Fusion Blankets and Its Application to CFETR

The China Fusion Engineering Test Reactor (CFETR) is a tokamak device to validate and demonstrate fusion engineering technology. In CFETR, the breeding blanket is a vital important component that is closely related to the performance and safety of the fusion reactor. Neutronics/thermal-hydraulics (N/TH) coupling effect is significant in the numerical analysis of the fission reactor. However, in the numerical analysis of the fusion reactor, the existing coupling code system mostly adopts the one-way coupling method. The ignorance of the two-way N/TH coupling effect would lead to inaccurate results. In this paper, the MCNP/FLUENT code system is developed based on the 3D-1D-2D hybrid coupling method. The one-way and two-way N/TH coupling calculations for two typical blanket concepts, the helium-cooled solid breeder (HCSB) blanket and the water-cooled ceramic breeder (WCCB) blanket, are carried out to study the two-way N/TH coupling effect in CFETR. The numerical results show that, compared with the results from the one-way N/TH coupling calculation, the tritium breeding ration (TBR) calculated with the two-way N/TH calculation decreases by −0.11% and increases by 4.45% for the HCSB and WCCB blankets, respectively. The maximum temperature increases by 1 °C and 29 °C for the HCSB and WCCB blankets, respectively.

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.

Tritium transportation modeling and analysis for tritium extraction system of HCSB blanket

Tritium is essential to achieve self-sufficiency as the indispensable fuel of DT reactions for future fusion power plant. Tritium plant is intended for recycling the tritium bred by breeder blankets and the unburned fuel from core plasma, and maintaining tritium cycle transportation. As a critical system in tritium plant, tritium extraction system (TES) is used for extracting tritium from blanket breeders and purifying hydrogen isotopes in the purge gas from blankets. In order to obtain the detailed behavior for tritium self-sufficiency analysis and safety performance assessment of TES during its operation, tritium transportation analysis in TES needs to be carried out. In this paper, a detailed TES model is developed referring to the tritium extraction processing for helium cooled solid breeder (HCSB) blanket. Several behaviors of tritium including tritium permeation, tritium leakage and chemical equilibrium movement including HT and HTO, are taken into consideration. The results and analyses of tritium inventories and tritium losses in TES are obtained. PRF (Permeation Reduction Factor) sensitivity analysis of components in TES is also carried out. The requirements of flow rate in glove box (GB) for atmosphere detritiation and makeup flow for He-H2 makeup unit are given.

Improving the optimization algorithm of NTCOC for application in the HCSB blanket for CFETR Phase II

Blanket is the key component of Chinese Fusion Engineering Test Reactor (CFETR). In our previous work, considering that the radial arrangement of the internal functional zones has the greatest influence on the blanket performances and is also the most important basis of the 3D design, NTCOC, a Neutronics/Thermal-hydraulic Coupling Optimization Code which adopts 1D neutronics and 2D thermal-hydraulic simplified calculation models, has been independently developed for facilitating the radial build design and optimization of the CFETR helium cooled solid breeder (HCSB) blanket. Through setting three leading design requirements as the optimization criteria, this code has been applied on optimizing the radial build arrangement of a conceptual design of HCSB blanket for CFETR Phase I, which preliminarily verifies the reasonability of the code. However, recently, the design phase of CFETR has gradually changed from I to II, the fusion power of which will be quintuple to the previous I. This means that the internal nuclear power deposition in each blanket component under II operation condition will be much larger than the previous I. Therefore, much larger coolant inlet mass flowrate will be needed and more breeder units should to be inserted, both of which will make the total pressure drop of blanket for Phase II much larger than the previous I and even exceed the limit of CFETR HCSB blanket greatly. This can’t be sustained in CFETR, a power plant that should produce electricity. To solve this problem, in this paper, the previous integrated neutronics/thermal-hydraulic optimization algorithm is improved by adding the new pressure drop limit into the optimization criteria. Then the improved code has been applied on optimizing the radial build arrangement of a conceptual design of HCSB blanket for CFETR Phase II through constant and variable step length two methods. The calculation results verifies the feasibility and reasonability of the improved optimization algorithm preliminarily.

Steady State and Transient Thermal Analysis of the Updated Helium-Cooled Solid Breeder Blanket for CFETR

This paper presents the results of the steady state and transient thermal analysis of the updated helium-cooled solid breeder blanket for the Chinese fusion engineering test reactor (CFETR). The updated design of the helium-cooled solid breeder blanket for the CFETR has been described. The commercial finite element method code ANSYS 15.0 is used for the thermal analysis in this paper. The steady state thermal analysis of the updated blanket has been performed, showing that the temperatures of different materials of the blanket module are below the corresponding temperature limits. The 3-D transient thermal analysis of the updated helium-cooled solid breeder blanket for the CFETR under normal pulsed operation has been conducted. The temperature of different components during flattop burning time calculated by transient thermal analysis matches well with that of steady state thermal analysis. Furthermore, the transient thermal behavior of the blanket under the in-box loss of coolant accidents condition are also described and critically discussed.

The assessment of shutdown dose rate and radioactive waste of HCSB during its replacement in CFETR

Chinese Fusion Engineering Testing Reactor (CFETR) is a new test superconducting tokamak in China aimed to obtain the state-of-the-art technologies and bridge the gaps between ITER and DEMO. The helium cooled solid breeder blanket (HCSB) is one of three main blanket candidates for CFETR. During reactor operations, the highly radioactive nuclides will be produced by the blanket components which are located in the high-energy neutron radiation field. In order to protect the workers from too high radiation and ensure the normal operation of machinery, the Shutdown Dose Rate (SDR) of HCSB during its replacement is evaluated based on the rigorous two-step (R2S) method. Besides, according to the recommended standards provided by Safety and Environmental Assessment of Fusion Power (SEAFP) of the EU fusion program, the initial evaluation of the high level radioactive waste of the blanket is carried out as well. The results show that, based on current design of HCSB, the maximum SDR of the blanket is decreased to 4728 Sv/h after the reactor shutdown one day. And after 50 years of cooling, most of the structural materials can be recycled by simple remote operation.

Numerical research on the coupling optimization design rule of the CFETR HCSB blanket using the NTCOC code

In this study, a three-step approach of ordinal 3D-1D&2D-3D is independently proposed to derive a coupling optimization design rule of the Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder (HCSB) blanket. Firstly, the three dimensional (3D) neutronics model of CFETR is built with all of the 15 blanket modules adopting the original radial build arrangement parameters. Based on this model, both the neutron wall loading (NWL), which will serve as the normalization factors for the following local neutronics calculations, and the tritium breeding ratio (TBR) distributions on each blanket module are obtained. Secondly, the radial build arrangements of all the 7 inboard and the 8 outboard blanket modules in an 11.25° toroidal CFETR sector are optimized by taking both the tritium breeding and thermal performances into account as the optimization objectives. NTCOC, a Neutronics/Thermal-hydraulic Coupling Optimization Code, which has been independently developed by Xi’an Jiaotong University for accelerating the radial build design and optimization of the fusion reactor HCSB blanket concept with adopting the simplified 1D neutronics and 2D thermal-hydraulic models, is used for these work. Thirdly, based on the final optimized schemes of all the blanket modules, a comprehensive research on the coupling optimization design rule of the CFETR HCSB blanket is performed in-depth and a coupling optimization design integration approach has been finally summarized. Finally, the 3D CFETR neutronics model is updated by replacing all of the original blanket radial build arrangement parameters with the final optimized schemes. The calculation results show that after the radial optimizations, the total TBR of all the 15 blanket modules has increased from the previous 1.092–1.274, which can well meet the self-sufficiency tritium breeding requirement of CFETR.

Preliminary safety assessment on two LOCAs under ITER-like condition for the optimized CFETR helium cooled solid breeder blanket

A conceptual design of helium cooled solid breeder (HCSB) blanket, one of three blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR), has been performed recently. The transient analysis for different possible accidents should be carried out to assess its safety performance. In this paper, the ITER-like conditions are adopted since the associated system for CFETR is missing, such as helium cooling system, plasma shutdown condition, pump behavior, etc. The complete model is recreated including the optimized typical outboard HCSB blanket (NO.12) and its ancillary helium cooling loop, and accident analyses of two loss of coolant accident (LOCA) cases are investigated using RELAP5. The influences of different break areas under in-vessel LOCA are compared, and the accident consequence after small area break is chosen to be investigated. Regarding the ex-vessel LOCA, the influences of different break locations are thoroughly analyzed. Since the plasma cannot terminate passively after ex-vessel pipe break, the plasma termination behaviors are investigated with different shutdown time. The computational results show that with the safety criteria for ITER the HCSB blanket can be cooled down effectively by the helium cooling system (HCS) and the integrity of pressure barriers can be guaranteed for both accidents.

Impact analysis of the time trend of TBR and irradiation damage assessment of HCSB blanket for CFETR

Chinese Fusion Engineering Testing Reactor (CFETR) is a test tokamak reactor to bridge the gap between ITER and future fusion power plants and to demonstrate generation of fusion power in China. In fusion power plants, tritium is generated from the reaction of neutron and Lithium. One of the missions of CFETR is the full cycle of tritium self-sufficiency. For the mission, a Helium Cooled Solid Breeder blanket (HCSB) was proposed for CFETR and its conceptual design has been carried out. In order to assess the capacity of the tritium breeding and irradiation damage of first wall of the HCSB blanket during the 8 years’ engineering test stage, this paper presents the time trend of TBR analysis and irradiation damage assessment of HCSB blanket based on the three-dimensional (3D) neutronics model which is created by McCad. In the 3D neutronics model, the outboard blanket on equatorial plane is described based on the detailed 3D engineering model. The calculations were performed by MCNP and FISPACT with FENDL/2.1 data library. The impact analysis of the thickness of coolant plates (CP) and the structural material content in CPs to the TBR is assessment.

Preliminary Design of the Helium Cooling System for the Helium Cooled Solid Breeder Blanket of CFETR

Chinese Fusion Engineering Test Reactor (CFETR) is an ITER-like fusion device that was proposed to achieve 50–200 MW fusion power, 30–50 % duty time factor, and tritium self-sufficiency. As a candidate blanket concept for CFETR, a helium cooled solid breeder blanket was designed following the specific requirements. The design of a helium cooling system (HCS) was conducted. The HCS is an important ancillary system of helium cooled solid breeder blanket for CFETR. The general description of HCS design was presented in this paper. Specifically, the design principles and thermal design requirements were defined, the process flow diagram of the system was given and the main components were introduced. Accordingly, the dimensions of the components were preliminary estimated to provide a space allocation reference for a CFETR blanket system.

Thermal–mechanical analysis of the first wall for CFETR helium cooled solid breeder blanket

Recently a conceptual structure design of helium-cooled solid breeder blanket was proposed as one of the candidates for the Chinese Fusion Engineering Test Reactor (CFETR). In the blanket, a trapezoidal U-shaped FW was adopted to decrease gaps between blanket modules. Due to high heat flux from plasma and intense irradiation by high-energy neutrons, thermal–mechanical analysis of the FW was carried out to ensure reasonable FW temperature and stress under normal load condition. A fluid–solid coupled code ANSYS was used to verify that the mechanical behavior conforms to the criteria of design codes and standards. The calculations showed that the FW material temperature limit was satisfied when the inner surface of the coolant channel was artificially roughened to 0.5mm in height. Furthermore, the influence of different FW geometrical configurations on the temperature field and stress distribution was investigated. The results led to improvements to the FW design that can reduce the maximum temperature and stress of FW structural material.

Thermal Analysis of Breeder Unit for Helium Cooled Solid Breeder Blanket of Chinese Fusion Engineering Test Reactor

As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor, a kind of helium cooled solid tritium breeder blanket was proposed. The blanket uses the pebbles of lithium ceramics (Li4SiO4) and beryllium as tritium breeder and neutron multiplier, respectively. The thermal conditions of breeder unit directly affect the performance of tritium breeding and the safety of blanket. Therefore, thermal analysis of the pebble beds is of vital importance for a reliable blanket design. State steady thermal hydraulic analysis of the breeder unit has been performed, showing that the temperature satisfied the corresponding material temperature limits. State steady thermo-mechanical analysis has also been carried out. The maximum von Mises stress was within the allowable stress. Parametric sensitivity studies have been conducted to investigate the influence of main parameters (e.g. coolant mass flow rate, inlet temperature and pebble bed thermal conductivity) on the temperature distribution of the pebble beds and cooling plates.

A Conceptual Study on Helium-Cooled Solid Breeder Blanket for Prototype CH DEMO

A conceptual design of tritium-breeder blanket for prototype DEMO is introduced. In this concept, Li4SiO4 pebbles are chosen as tritium-breeder material and Be for neutron multiplication. To confirm the thermal effectiveness,a CFD code FLUENT is used for the calculation. The result indicated that, the thermal effectiveness of the concept is acceptable.

Current R&D Activities on Korean Helium Cooled Solid Breeder Test Blanket Module

Helium Cooled Solid Breeder (HCSB) blanket is one of two blanket concepts being considered as Korean Test Blanket Module (TBM) for ITER with the aim for testing and verifying the capability of the breeding blanket concepts. R&D activities being carrying out for HCSB TBM include the development of materials, fabrication technologies, and TBM associated system design. A small sample of ferritic/martensitic (FM) steel was fabricated. It was found that the tensile strength was close to previous value. A fabrication technique of sphere pebbles of lithium titanate breeder and graphite reflector was developed. A FM/FM TIG welding was performed and the results showed that tensile strength of the welded zone was decreased about 10 %. A small punch test method of mechanical property evaluation was introduced to verify the suitability of small specimen for irradiation test by examining the relationship of the conventional uniaxial tensile test, and the tensile strengths were compared. As TBM design has complicated square channel configuration, short square channel was fabricated successfully. Finally, the components and specifications of the TBM associated systems are described in this paper.

Current status of design and analysis of Korean Helium-Cooled Solid Breeder Test Blanket Module

It is known that the test of the capability of tritium self-sufficiency and the extraction of high-grade heat using tritium breeding blanket concepts is one of the major ITER missions. This requires the development of test blanket modules based on a corresponding DEMO blanket design, despite the differences in operating conditions between DEMO and ITER. Two blanket concepts such as a Helium-Cooled Solid Breeder (HCSB) blanket and a Helium-Cooled Molten Lithium (HCML) blanket are being considered as Korean Test Blanket Module (TBM) for ITER with the aim for verifying the capability of the breeding blanket design and manufacturability. In this paper, the current status of the design and analysis for the HCSB TBM is described. The key features of the design include the use of graphite neutron reflectors to reduce the amount of beryllium multiplier. Major analyses have been executed in order to optimize the design. The results of nuclear and thermo-hydraulic analysis show that the design parameters and requirements are satisfied. Also, the safety analysis results show that the design of the TBM system is sufficient to endure several accident events.

U.S. Plans and Strategy for ITER Blanket Testing

Testing blanket concepts in the integrated fusion environment is one of the principal objectives of ITER. Blanket test modules will be inserted in ITER from Day 1 of its operation and will provide the first experimental data on the feasibility of the D-T cycle for fusion. With the US rejoining ITER, the US community has decided to have strong participation in the ITER Test Blanket Module (TBM) Program. A US strategy for ITER-TBM has evolved that emphasizes international collaboration. A study was initiated to select the two blanket options for the US ITER-TBM in light of new R&D results from the US and world programs over the past decade. The study is led by the Plasma Chamber community in partnership with the Materials, PFC, Safety, and physics communities. The study focuses on assessment of the critical feasibility issues for candidate blanket concepts and it is strongly coupled to R&D of modeling and experiments. Examples of issues are MHD insulators, SiC insert viability and compatibility with PbLi, tritium permeation, MHD effects on heat transfer, solid breeder “temperature window” and thermomechanics, and chemistry control of molten salts. A dual coolant liquid breeder and a helium-cooled solid breeder blanket concept have been selected for the US ITER-TBM.