Helium-cooled Ceramic Breeder Test Research Articles

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.

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.

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.

Measurement of the impacts of the J-TEXT TBM mock-up on the toroidal and vertical magnetic field

A 1/3 scaled mock-up of the China Helium-Cooled Ceramic Breeder Test Blanket Module (CN HCCB-TBM) was integrated into the J-TEXT tokamak to study the eddy current distribution, electromagnetic load and thermal load modes on the TBM during plasma disruption. The J-TEXT TBM mock-up is made of reduced activation ferritic/martensitic (RAFM) steel and supported by a cantilever structure, with the ability to move along the radial direction. To investigate the effects on the equilibrium magnetic field in the presence of the mock-up containing ferromagnetic material, two poloidal arrays of printed circuit board (PCB) coil-type magnetic probes are developed and installed inside the J-TEXT vacuum vessel. The toroidal field (TF) ripple along the toroidal direction and the difference of vertical field (VF) over the plasma region are measured. Corresponding simulations are also performed, and the results fit well with the measured ones. It is shown that the TBM mock-up in J-TEXT will change the shape of equilibrium magnetic field in the vacuum vessel significantly, introduce large TF and VF ripple, especially in low field side in front of the mock-up. On the other hand, the effects will produce only a few potentially troublesome problems on the normal discharge when the recession of the mock-up is up to 15 cm away from the port.

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.

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).

Numerical Investigation of Channel for Helium Cooled China HCCB-TBM First Wall by Application Enhance Heat Transfer Method

To enhance heat transfer efficiency on first wall (FW) of ITER China Helium-Cooled Ceramic Breeder-Test Blanket Module (HCCB-TBM), CFD numerical simulation method is adopted. On the basis of calculating helium gas cooling scheme of FW smooth channel, FW structural temperature gradient, maximum wall temperature, average heat transfer coefficient, and pressure drop of channel are selected as evaluation indexes. Numerical simulation comparison are performed on heat transfer schemes like placing transversal ribs and V-shaped ribs in the flow channel of front wall and the helium gas turbulence intensity and the heat transfer area are improved through optimizing the distance and angle between V-shaped ribs and other parameters to enhance heat transfer. The optimization scheme of helium-cooled FW for HCCB-TBM through the three dimensional numerical simulation is: V-shaped ribs are placed on the inner surface of front wall, the rib cross section is 1 mm × 1 mm, the distance between rib pitches is 10 mm and the rib angle is 60°. Under the same helium cooling condition, compared with the FW smooth channel, the optimized V-shaped rib scheme enhances the average heat transfer efficiency by about 70 % and the FW maximum temperature drops by 349.3 K. The result provides support for further research on FW helium cooling heat transfer enhancement experiment and engineering design optimization for China HCCB-TBM.

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.

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.

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.

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.

Assessment of the activation, decay heat, and waste disposal of the US helium-cooled ceramic breeder test blanket module in ITER

The radioactivity inventory and decay heat in the US helium-cooled ceramic breeder (HCCB) test blanket module (TBM) have been assessed at shutdown and for several times thereafter. The sub-module will have its own FW and structural container box that houses the breeder and beryllium pebble bed units, arranged in an edge-on-configuration. Low activation ferritic steel (F82H) is used as the structure and helium is used as a coolant. The breeder beds are made of Li 2TiO 3 pebbles in which lithium has been enriched up to 75% in Li-6. Pulsed operation mode is assumed. During operation in the D-T phase, the total heating rate in the TBM is ∼263 kW. The total amount of tritium generated in the breeder and the beryllium multiplier is ∼9 g and 0.07 g, respectively, after reaching the 0.3 MWa/m 2 fluence limit. At shutdown, the total radioactivity and decay heat levels are ∼0.89 MCi and ∼0.002 MW, respectively. These values drop sharply after 1 min to ∼0.098 MCi and ∼0.0006 MW. The contribution from the F82H structure is the dominant one up to ∼10 years following shutdown. After ∼10 years, the contribution to the total activation and decay heat from the breeder material is the dominant one due to the generated tritium. The WDR of various components are far below unity and thus are well within ITER regulatory guidelines.