Helium Cooled Ceramic Breeder Research Articles

Preliminary verification of a transient analysis code TSACO for helium cooled system

This work mainly focuses on the verification process of the thermal-hydraulic and safety analysis code (TSACO) applied to the Chinese helium-cooled ceramic breeder test blanket module cooling system. The HE-FUS3 facility, which is an integral test facility for helium cooling systems, has been chosen to assess the capability of the TSACO code. The steady state and transient behavior of two loss-of-flow accidents (LOFAs) were simulated by TSACO. The steady state temperature distributions of the two LOFA tests were well reproduced by TSACO. The transient results of mass flowrates and outlet temperatures predicted by TSACO at the compressor and test section compared well with the experimental data. The good agreement between the calculated results and experimental data verifies the capability of the TSACO code to simulate helium cooling systems. In addition, an uncertainty analysis was conducted for the test section outlet temperatures of LOFA30 and LOFA50. The power, inlet temperature, mass flowrate, physical material properties, and heat transfer models were chosen as the input uncertainty parameters. The results show that the experimental data are easily covered by the uncertainty bands calculated by TSACO during transient analysis. This demonstrates the accuracy and fidelity of the TSACO code.

Microstructure evolution of beryllium with argon ion irradiation

Beryllium is planned to be used as a neutron multiplier in the Helium-cooled ceramic breeder (HCCB) test blanket module (TBM), which is the primary option of the Chinese TBM program. The characteristic of the microstructural change of beryllium under irradiation is an important factor contributing to the understanding of degradation of physical-mechanical properties of the material. Beryllium metal was irradiated with 100keV Ar+ ions to a dose of 1 ×1021/m2 at room temperature. The microstructural change of Beryllium has been investigated using transmission electron microscopy (TEM). After the irradiation, a high density of black-dot-like defects were induced in the sample, some of which were resolved as small dislocation loops. Gas-filled bubbles with an average size of about 33.7nm were observed. Ion irradiation induced swelling at room temperature was also discussed.

24 - Impact of Gene Mutations on Response to Lenalidomide and OS in Lower-Risk Non-Del(5Q) MDS Patients Ineligible/Refractory to Erythropoiesis-Stimulating Agents (ESAS)

As one candidate tritium breeding blanket concept for CFETR, a conceptual structure design of helium cooled ceramic breeder (HCCB) blanket was proposed. After the engineering design of CFETR HCCB blanket as well as its associated ancillary systems, a safety assessment on two postulated accident scenarios were carried out for the prediction of the potential consequences. Each scenario was modeled and computed by RELAP5 code. The transient accidents analysis was performed based on the steady state results. The comparison of thermal-hydraulic parameters computed by RELAP5 and ANSYS CFX code was to verify the accuracy of the RELAP5 modeling. Results involving a number of parameters investigated indicate good agreement for better integrity of FW and BU design under the two assumed accident scenarios and infer sufficient heat removal capability of the analyzed blanket module under postulated accidents.

Neutronics study on HCCB blanket for CFETR

Chinese Fusion Engineering Testing Reactor (CFETR) is aimed to obtain the technologies to fill the gaps between ITER and DEMO. The helium cooled ceramic breeder (HCCB) blanket is one of the candidates for CFETR. Ceramics Li4SiO4, beryllium and helium of 8MPa were selected as tritium breeding material, neutron multiplier and coolant, respectively. CLF steel, one of reduced activation ferritic/martensitic (RAFM) steels, was applied as structural materials of blanket. Neutron and gamma transport was simulated with MCNP code and FENDL/MC-2.1 neutron cross section data. Activation calculation was performed with FISPACT-2007 code and EAF-2007 activation file. In this study, neutronics characteristic of HCCB blanket such as tritium breeding ratio (TBR) and shielding were assessed. The arrangement of Be, Li4SiO4 and cooling plate in blanket module was optimized to improve TBR. The sensitive study for First Wall on TBR was also performed. The radioactivity inventories of HCCB blanket were estimated. The dominant radionuclides in structural materials, Li4SiO4 and Beryllium were discussed, respectively. The damage of plasma facing components such as First Wall and divertor due to DT neutron irradiation was analyzed in terms of DPA value and gaseous product.

Effect of weld spacing on microstructure and mechanical properties of CLAM electron beam welding joints

China low activation martensitic (CLAM) steel has been chosen as the primary structural material in the designs of dual function lithium-lead (DFLL) blanket for fusion reactors, China helium cooled ceramic breeder (HCCB) test blanket module (TBM) for ITER and China fusion engineering test reactor (CFETR) blanket. The cooling components of the blankets are designed with high density cooling channels (HDCCs) to remove the high nuclear thermal effectively. Hence, the welding spacing among the channels are small. In this paper, the welded joints of CLAM steel with different weld spacings have been fabricated with electron beam welding (EBW). The weld spacing was designed to be 2mm, 3mm, 4mm, 6mm and 8mm. The microstructure and mechanical properties such as microhardness, impact and tensile were investigated at different welding spacing for both conditions of as-welded and post weld heat treatment (PWHT). The PWHT is tempering at 740°C for 120min. The results showed that the grain size in the heat affected zone (HAZ) increased with the increasing weld spacing, and the joint with small weld spacing had a better performance after PWHT. This work would give useful guidance to improve the preparation of the cooling components of blanket.

Preliminary accident analyses of in-vessel LOCA and In-box LOCA for China helium-cooled ceramic breeder test blanket system

The preliminary transient accident analysis has been carried out since 2005 for China Helium Cooled Ceramic Breeder Test Blanket Module (CN HCCB TBM) on ITER equatorial port #2. In this paper, the RELAP5 model is recreated according to the design of the CN HCCB 1×4 TBM and its cooling system, and two design basis accidents (In-Vessel Loss of Coolant Accident (LOCA) and In-Box LOCA) are investigated. The influences of different break sizes during these accidents are comprehensively analyzed. The analysis results can provide a reference for the following parameter optimization design of CN HCCB Test Blanket System (TBS).

Recent progress of China HCCB TBM tritium system

China tritium system including Tritium Extraction System (TES) with Tritium Accountancy System (TAS) integrated in and Coolant Purification System (CPS), which is subordinate to Helium Coolant System (HCS), is of great importance for China Helium Cooled Ceramic Breeder Test Blanket Module (CN HCCB TBM). The purge gas (99.9% He+0.1% H2) carrying Q2O (Q=H, D, T) and Q2 from Li4SiO4 ceramic breeder flows through the reduction bed where Q2O is reduced into Q2 and then absorbed by the getter bed. The HT/HTO ratio and the total tritium are determined by TAS. Catalytic oxidation combines with molecular sieve absorption and hot metal purification are applied to remove tritium and other impurities in helium coolant. A loop including depressurization, helium-sweeping assisted thermal desorption, and cold trapping for the regeneration of saturated molecular sieve bed until the concentration of the desorbed Q2O is reduced to an acceptable level.This paper introduces the recent progress of China tritium system including updated conceptual designs of TES and CPS, as well as the developments of the functional materials and the devices.

Current progress of Chinese HCCB TBM program

ITER is an unique opportunity to test tritium breeding blanket mock-ups in integrated Tokamak operating conditions. Helium-cooled ceramic breeder (HCCB) test blanket module will be the primary option of the Chinese ITER TBM program. China plans to test its own test blanket modules (TBM) during ITER different operation phase (H-H,D-D, D-T). A preliminary design of HCCB TBM have been completed in 2013. In order to reduce the effects of magnetic field ripple, the HCCB TBM design has been updated with reduced RAFM mass recently. Related R&D on key components fabrication, welding technology, materials development and mock-up test of TBM modules and components have being implemented. In this paper, the current status on design and related R&D, as well as the development of auxiliary system (TES, CPS,HCS, etc.,) of Chinese HCCB TBS program were introduced.

Analysis of the impacts of the J-TEXT TBM mock-up on the equilibrium magnetic field

The Test Blanket Module (TBM) will be used in the test port of ITER to demonstrate tritium self-sufficiency and the extraction of high grade heat for electricity production. J-TEXT TBM mock-up using reduced activation ferritic/martensitic (RAFM) steel as structural material was designed and fabricated to perform and validate relevant electromagnetic and thermal technologies of the China Helium-Cooled Ceramic Breeder Test Blanket Module (CN HCCB-TBM) on the J-TEXT. Its size is one third of the CN HCCB-TBM. By using the finite element analysis technology, this paper analyzed the impacts on the equilibrium magnetic field over the plasma region after introducing the structure material RAFM steel. The distribution of toroidal field (TF) ripple and the magnitude of the error field with the mock-up at different positions were given. Simulation shows the distribution of the null field region formed by poloidal field (PF). The influence to tokamak discharge has been evaluated by drawing the magnetic field lines. Based on the results above, we have optimized and finished the installation of the mock-up to J-TEXT which meets the needs of the experiments and to ensure the normal discharge.

Conceptual Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR

The China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak reactor proposed by the China National Integration Design Group. The aim is to develop China’s next-step fusion device. A helium-cooled ceramic breeder (HCCB) blanket concept has been put forward by the blanket integration design team of the Institute of Plasma Physics of the Chinese Academy of Sciences. The present blanket configuration is based on the upper port dimension and maintenance scheme. The HCCB blanket comprises a U-shaped first wall, cap, breeder unit, middle plate, and back plate. The breeder unit includes a cooling plate, beryllium pebble bed, and lithium silicate pebble bed. The structure of the HCCB blanket concept is obviously different from the ITER HCCB test blanket module of China. The CFETR HCCB blanket will meet the requirements of CFETR long-pulse or steady-state operation with a duty cycle time greater than 0.3. The maintenance scheme of the blanket is introduced in this paper. Three-dimensional neutronic analysis results show that the tritium breeding ratio can satisfy the design requirement. Besides, the thermal-hydraulic behavior of the first wall has been studied using ANSYS CFX code.

Fault Diagnosis of HCCB TES for ITER Based on Functional Modeling

China has designed a test blanket system (TBS) based on helium cooled ceramic breeder (HCCB) test blanket module (TBM) for international thermonuclear experimental reactor (ITER). Since the system is complicated, it is so difficult to achieve the safety and stability requirement of TBS under operating condition that a real time fault diagnosis system need to be developed. This survey proposes a novel fault diagnosis method based on functional modeling, which can satisfy the safety and stability requirement of TBS. After giving a brief description of HCCB tritium extraction system (TES) that is one of the TBM’s ancillary systems, the paper presents the fault diagnosis method based on functional modeling by using multilevel flow modeling in detail. Then the TES is modeled and analyzed using this novel method. At last, a conclusion is drawn illustrating that this novel method is suitable for HCCB TBS and ITER.

Preliminary Quantitative Evaluation of Potential Risk for HCCB TBS Tritium Extraction System Based on Fault Tree Analysis

International Thermonuclear Experimental Reactor (ITER) is the first multinational fusion research facility, and test blanket module (TBM) is one of the most important components of ITER. The helium cooled ceramic breeder (HCCB) TBM and its auxiliary system are called HCCB test blanket system (TBS). Tritium extraction system (TES) plays an essential role in HCCB TBS. Potential risk of TES determines whether HCCB TBS can achieve tritium self-consistency and energy extraction. In this paper, potential risk analysis was performed on TES. By analyzing the potential failure modes of components, more than 60 basic failure events were obtained. The fault tree of TES failure is built. The qualitative analysis was performed on the tree model. Minimum cutsets (MCS) were obtained to identify the main impact factors of TES failure. Then quantitative evaluation was made based on MCSs, the mean failure rate of TES is 2.50e−3. The results showed the risk level of TES basically satisfied the operation requirements of ITER without regard of maintenance. The impact of basic events on TES failure probability is analyzed and ranked, the 10 high-impact basic events and corresponding components are obtained. The above mentioned analysis and results provide effective references for TES design optimization.

RAMI Analysis of HCCB TBS for ITER

ITER is the first worldwide international experimental nuclear fusion facility, which aims to prove the physics and technological basis for future fusion power plants. As main stages of ITER technical risk control, the reliability, availability, maintainability and inspectability (RAMI) approach should be applied to all ITER components during their design phase to reduce potential technical risks. Test blanket modules play a key role in ITER. Helium cooled ceramic breeder (HCCB) TBM is one of TBM concepts which were proposed by China. HCCB TBM and its ancillary system are called HCCB test blanket system (TBS). The RAMI analysis was performed on the conceptual design of the ITER HCCB TBS in this paper. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 3 main functions, 1 support function, 14 sub-functions and 50 basic functions. These functions were described using the IDEF0 method. Reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. The inherent availability of the HCCB TBS expected after implementation of mitigation actions was calculated to be 94.69 % over 2 years. A failure modes, effects and criticality analysis was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability.

Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

New progress on design and R&D for solid breeder test blanket module in China

ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R&D activities for each TBM module with the auxiliary system are introduced.The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li4SiO4 pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R&D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.

Overview of design and thermal–hydraulic analysis of Indian solid breeder blanket concept

India has developed two concepts of breeding blanket for the DEMO reactor: one is Lead Lithium Ceramic Breeder (LLCB), and the other one is Helium-cooled Ceramic Breeder (HCCB) concept. Indian HCCB concept is having edge on configuration of helium-cooled solid breeder with RAFMS structure. Li2TiO3/Li4SiO4 and beryllium are used as the tritium breeder and neutron multiplier, respectively. 2D thermal–hydraulic simulation studies using ANSYS have been performed based on the heat load obtained from neutronics calculations to confirm heat removal under ITER pulsed operation. Transient thermal analysis has been simulated in ANSYS for the ITER relevant operational conditions. Thermal analysis provides important information about the temperature distribution in different materials used and their temperature–time histories. Result of thermal–hydraulic simulations shows that in each cycle, the maximum temperature of all materials remains same. The peak temperatures of all materials are well within their limiting value. Concept designs of HCCB blanket and its thermal hydraulic analysis will be presented in this paper.

Progress on design and R&D for helium-cooled ceramic breeder TBM in China

The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. Current progress on the design and R&D for Chinese helium-cooled ceramic breeder TBM (CN HCCB TBM) in China is presented. The main updated design and related R&D of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being carried out. Recent status of the components and fabrication technology development is also reported. The neutron multiplier Be pebbles, tritium breeder Li4SiO4 pebbles, and structure material CFL-1 are being prepared in the laboratory scale. The fabrication of 1/3 sized mock-up and construction of a He test loop are being carried out. The key technology development is proceeding to the large scale mock-up fabrication and demonstration tests toward on ITER testing.

Neutronics Analysis in Support of the Fusion Development Facility Design Evolution

3-D neutronics analysis was performed for the baseline design of FDF. Two blanket concepts were considered; Dual Coolant Lead Lithium (DCLL), and Helium Cooled Ceramic Breeder (HCCB). A peak outboard neutron wall loading of 2 MW/m2 and a fluence of 6 MW-yr/m2 can be achieved with 240 MW fusion power. The tritium breeding ratio is adequate for both blankets. Modest magnet damage parameters were obtained. However, it is recommended that the PF coil in the divertor region be moved vertically farther from the mid-plane to allow adding ~15 cm of shield to reduce the peak organic insulator dose to an acceptable level.

Evaluations on reduction of the ITER TFC ripple generated by CN HCCB-TBM

Reduced Activation Ferritic/Martensitic (RAFM) steel has been chosen as structural material for China Helium-Cooled Ceramic Breeder Test Blanket Module (CN HCCB-TBM). The magnetization of RAFM definitely increases the toroidal field perturbation (called TF ripple) in international thermonuclear experimental reactor (ITER). The TF ripple could cause ripple loss of high-energy particles and result in a large localized heat load on the first wall of TBMs. Thus some positive measures to reduce the TF ripple generated by TBMs have been evaluated by finite element models (FEM) in this paper. It has been shown that under the intervention of ferromagnetic inserts (FIs) the TF ripple could be reduced to the acceptable level of ITER (namely, TF ripple ∼0.7% at R = 8.2 m of plasma edge near the equatorial plane) while fully considering several actual combinations (mass-reduction and recess) of HCCB-TBM and introduction of correction coils.

Conceptual and preliminary engineering design of experimental helium loop for China HCCB TBM components test

Chinese Helium Cooled Ceramic Breeder Test Blanket Module (CN HCCB TBM) had been confirmed to be one of the leader concepts, which will be tested in half of ITER equatorial port. It is responsible for each TBM does not jeopardize ITER operation. In order to ensure the future CN HCCB TBM safely and effectively operating in ITER, components and even full scale of TBM should be tested before delivery to ITER site. Therefore, a new high temperature and high pressure experimental helium loop will be constructed to validate the thermal-hydraulic and thermal-mechanical capability of those components. In this article, a rational conceptual design of the experimental helium loop was presented. At the same time, preliminary engineering design includes thermal-hydraulics assessment of equipments, thermal stress of pipes in the loop were performed. The assessment results show that technical issues of helium loop construction are expected to be solved successfully. Finally, possible test strategies have been presented to fulfill the TBM thermo-hydraulics campaigns.