Helium Cooled Ceramic Reflector Research Articles

Development of fabrication technologies for the first wall of the HCCR blanket

The helium-cooled ceramic reflector (HCCR) concept is a primary candidate among demonstration fusion power reactor (DEMO) blanket options in Korea. The HCCR blanket is composed of a first wall (FW), a breeding zone, side cover plates and a back manifold. This study focuses on fabrication technologies for the FW using laser-beam welding (LBW) and hot isostatic pressure (HIP) bonding. Permanent backing strips in the pressure boundaries of the HCCR blanket are not allowed according to regulations related to pressure vessels and nuclear pressure vessels because high-pressure helium gas at 8 MPa is used as the coolant. LBW conditions and procedures for the fabrication of a FW without backing strips were investigated using small mock-ups with an advanced reduced-activation alloy called ARAA, which was developed in Korea. The laser beam power, thickness of the cover plate, and width of the backing strips were considered as parameters. The LBW conditions and procedures were optimized to remove the backing strip and achieve uniform and continuous back beads. After LBW, the fabrication of the FW with small mock-ups was completed by joining the back plate with HIP bonding. The small mock-ups were subject to very high temperatures exceeding 1000°C during the LBW and HIP bonding processes, which led to changes in the microstructures and hardness levels. The microstructures and hardness levels were recovered through post-heat treatments performed after HIP bonding. The microstructures of the HIP bonding interface after the heat treatments were observed. A partial HIP bonding line was confirmed and transverse tensile tests were carried out to investigate the impact of the partial HIP bonding line on the mechanical properties of the HIP bonding interface. The tensile strength of the HIP bonding interface was approx. 2% lower than that of the base metal, indicating that the partial HIP bonding line caused a slight decrease in the tensile strength, but the impact on the HCCR blanket design would be negligible.

The effects of side interactions on the neutronic analysis of HCCR breeding blanket

Helium Cooled Ceramic Reflector (HCCR) is a candidate as Multi-Module Segmentation Breeding Blanket used the Korean DEMO Reactor (K-DEMO). Neutronic analyses of the breeding blanket are substantial due to the prominent role of neutrons in tritium breeding, energy deposition, and radiation damages. This work aims to consider the effects of side reactions with low cross-sections on the neutronic study to make the simulation results more accurate. The neutronic analysis of the HCCR blanket has been performed using the MCNPX.2.6.0 3D Monte Carlo code and ENDF/B-VII nuclear data library. The effects of side reactions have been considered on Tritium Breeding Ratio (TBR) and Neutron Multiplication Factor (NMF) computations. The 9Be(n,T)7Li is a side reaction to tritium breeding, and the 7Li(n,2n) and W(n,2n) reactions with low cross-section cause neutron multiplication. Furthermore, the deposited energy by alpha particles has been investigated. Finally, the reaction rate of the photonuclear response has been examined using MCNPX.

Status of HCCR TBM program for DEMO blanket

Development of the Helium Cooled Ceramic Reflector (HCCR) breeding blanket is underway according to the development strategy of core technology for K-DEMO under the fusion energy development roadmap in Korea. The main goals are to validate integrated design tools and to develop core technologies in the field of materials, manufacturing and joining technologies, helium cooling and tritium technologies, system integration, and safety technologies. A unique graphite reflector concept was adopted to save cost by reducing the amount of beryllium neutron multiplier. The main functions and design concepts of the HCCR breeding blanket are to be verified in ITER through the HCCR Test Blanket Module (TBM) program. Not only breeder modules are to be tested in ITER but also the technologies of cooling, coolant purification, tritium extraction, etc. are to be tested or proved in ITER before employed in DEMO. This paper introduces the current updates of the design and R&D activities of the HCCR TBM systems in the preliminary design stage.

HCCR breeding blankets optimization by changing neutronic constrictions

The neutronic analysis of Helium Cooled Ceramic Reflector (HCCR) breeding blankets has been performed using the 3D Monte Carlo code MCNPX and ENDF nuclear data library. This study aims to reduce 6Li percentage in the breeder zones as much as possible ensuring tritium self-sufficiency. This work is devoted to investigating the effect of 6Li percentage on the HCCR breeding blanket's neutronic parameters, such as neutron flux and spectrum, Tritium Breeding Ratio (TBR), nuclear power density, and energy multiplication factor. In the ceramic breeders at the saturated thickness, increasing the enrichment of 6Li reduces its share in the tritium production. Therefore, ceramic breeders typically use lower enriched Li from 30% to 60%. The investigation of neutronic analysis in the suggested geometry shows that using 60% 6Li in Li2TiO3 can yield acceptable TBR and energy deposition results, which would be economically feasible.

Effects of Temperature and Pressure on Purge Gas Chemical Compositions in Tritium Breeding Blanket

The effects of temperature and pressure conditions on the equilibrium chemical compositions of purge gas at the outlet of the test blanket module (TBM) in the helium-cooled ceramic reflector (HCCR) are studied. As the chemical species in the equilibrium states, nine chemical species are considered: H, T, O, H2, HT, T2, H2O, HTO, and T2O. The mole fractions of these chemical species are calculated using a Gibbs free energy minimization method starting from the initial state of a H2-HTO mixture. The standard Gibbs free energies for the tritium species used in the study are calculated from the molecular constants obtained by a coupled-cluster calculation. The effects of pressure variations on the equilibrium compositions are shown to be negligible. The effects of temperature variations are also insignificant when the temperature exceeds 800 K. When the initial H2/HTO ratio is more than 10, more than 90% of tritium is expected to be recovered in the form of HT.

Conceptual design and analysis of the HCCR breeder blanket for the K-DEMO

A helium-cooled ceramic reflector (HCCR) blanket was studied as a candidate breeder blanket for the Korean DEMO reactor. A detailed three-dimensional neutronics model based on a K-DEMO model and a simplified thermal-hydraulic model were established for the HCCR-DEMO blanket conceptual design analysis. The insertion of a graphite reflector resulted in more than 33 % reduction of the required total beryllium multiplier volume, while its influence on the tritium breeding ratio (TBR) was evaluated to be less than 2 %. The final overall TBR of the HCCR-DEMO blanket was estimated as being more than 1.15 even after operation of 20 effective full power years (EFPYs) without replacement.

Hydrogen adsorption performance for large-scale cryogenic molecular sieve bed

Cryogenic hydrogen adsorption using molecular sieve beds is considered to be one of the main candidate processes for recovery of produced tritium from purge gas in breeding blankets and it has been chosen for separating hydrogen isotopes in Tritium Extraction System (TES) of Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS). Various adsorbents and their performance have been studied for the cryogenic adsorption using small-scale experiments. However, large-scale experiments comparable to TBS-relevant scale are required to have sufficient confidence for component design and performance prediction of Cryogenic Molecular Sieve Bed (CMSB) for the TBS and beyond. To properly evaluate hydrogen adsorption performance of a large size CMSB, a series of experiments have been performed using PGLoop facility which is constructed and operated in the National Fusion Research Institute. The experimental conditions were set to include breeding-blankets-relevant parameters. As such, the effects of swamping ratio, total pressure, and flow rates on the performances of CMSB were studied in the range of hydrogen partial pressures from 100 to 700 Pa. While a slight reduction in hydrogen adsorption performance is observed in comparison to the small isotherm experiments, which can be attributed to scale-up effects, it shows that the experimental results agree reasonably well with existing literature data.

Neutronic assessment of HCCR breeding blanket for DEMO

Helium-cooled ceramic reflector (HCCR) blanket concept was proposed as a DEMO breeding blanket in Korea. It adopted lithium ceramic breeder, graphite reflector to reduce the beryllium multiplier, reduced activation ferritic-martensitic steel as a structural material, and high pressure/temperature helium as a coolant. As a pre-conceptual design study, neutronic assessment of the HCCR blanket was carried out by inserting the newly developed model into the K-DEMO three-dimensional neutronic calculation model. As results, the peak neutron wall load of 4.06 MW/m2 was obtained at the outboard midplane blanket first wall for the fusion power of 3 GW. The overall tritium breeding ratio (TBR) was estimated as 1.10. And the poloidal mean global TBR distribution at the blankets were calculated.

Conceptual design of HCCR-TBM diagnostics

The design of diagnostics system for the Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) has now progressed from generic approaches to the conceptual design integration phases. It is proposed to monitor the temperature, strain, magnetic field strength and neutron flux in selected positions of the corresponding Test Blanket Module (TBM). The temperature can be measured by thermocouple sensors to be integrated for the first wall and breeding zones of TBM. The strain will be monitored by appropriate gauges on the first wall of TBM. Several magnetic pick-up coils could be installed to measure magnetic field strength inside the TBM structure. For the neutron flux monitoring, pairs of Neutron Activation System (NAS) probes and Micro-Fission Chambers (MFCs) would be installed in the TBM structure. In this paper, a brief description on the proposed diagnostics system as well as the key components is presented, and the integration challenges of HCCR TBS diagnostics system are outlined.

Investigation of technical gaps between DEMO blanket and HCCR TBM

Korea has own programs toward DEMO and fusion reactors that will require further improved DEMO blanket and energy utilization systems. Primary goals of DEMO blanket development in Korea are to develop and verify the integrated blanket design tools; to develop blanket materials, cooling and tritium fuel cycle technologies; and to develop and evaluate fabrication and joining technologies. The concept of helium-cooled ceramic reflector (HCCR) blanket is adopted to be tested in ITER as Test Blanket Module (TBM). Currently, the design and R&D activities are mainly performed through the ITER TBM program in Korea. It is expected to demonstrate the major objectives of the breeding blanket: extraction of heat from burning plasma, tritium self-sufficiency and its integrity within the whole systems considering safety features. Although TBM program will provide very essential data and experience, there will be a considerable technical gap from DEMO blanket. This paper presents the technical gap in the main technical categories of the breeding blanket based on the experience of the development of HCCR TBM and breeding blanket. It is found that still about 60–70% of the DEMO blanket technology can be achieved through the HCCR TBM depending on the technology maturity level, and some ways to DEMO blanket from HCCR TBM are proposed.

Components Design and Analysis for the HCCR TBM Preliminary Design

After Conceptual Design (CD) Approval, Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) design is being updated for the preparation of the preliminary design phase, in which the manufacturability is the main concern based on the TBM-set model of CD phase. The major design parameters at the preliminary design phase were introduced. According to these parameters, the design of the main components such as first wall (FW), side wall (SW), and the breeding zone (BZ) was updated through the thermal-hydraulic analysis with ANSYS CFX v14.5. Also, the detailed design was performed on the back manifold (BM), TBM-shield, and the connecting supports. The internal channels for the He coolant and purge gas are designed in BM, in which the characteristics of the He channel were analyzed to meet the requirements of the temperature. The design concept of the connecting supports was described. The design of the TBM-shield is updated in order to simplify production and assembly.

Numerical Investigation on Cooling Performance of Breeding Zone in HCCR TBM

The helium cooled ceramic reflector (HCCR) test blanket module (TBM) has been designed to be installed in ITER and to verify the tritium production and the heat extraction in Korea. Lithium, beryllium, and graphite are used as a breeder, a neutron multiplier, and a reflector, respectively, which called as breeding zone (BZ) including cooling plate. The BZ was operated with the highest temperature in the TBM due to the nucler heating not only in breeding material but also structure. The margin to the allowable temperature for the breeder is very small in the current conceptual design of HCCR TBM. In the present study, feasible methods were investigated to lower the maximum temperature of the BZ. The thermal resistance and the effect of each factor were studied with a conventional CFD code, ANSYS-CFX v14.5. It is found that the thermal resistance related to the pebble beds layer was main factor to determine the breeder temperature, and the installation of the cooling fins could reduce the heat transfer resistance and lower the maximum temperature of breeder about 80°C.

Reducing the Peak-to-Average-Power-Ratio in Fusion Blankets

Breeding blankets with integrated first wall are one of the most critical components of nuclear fusion reactors. Blankets breeding zones are characterized by steep nuclear heating gradients due to the exothermic nuclear reaction 6Li(n, )T and the high intensity neutron flux in the proximity of the first wall. Non-uniformity in nuclear heating can generate sharp temperature gradients that deeply affect material properties. This conceptual study explores an original way to flatten nuclear heating profiles by proposing a blanket characterized by layers of different 6Li enrichment in the breeder region while maximizing Tritium Breeding Ratio (TBR) and power generation. Two types of fusion blanket are studied: (1) Helium Cooled Ceramic Reflector (HCCR) and (2) Dual Coolant Lead Lithium (DCLL). For HCCR, it is found in the optimal design case, that the power peak-to-average can be reduced by 47.85%, 42.45% and 54.13% in the front, middle and back channel respectively when compared to the reference design. On the other side, we found that this method of profile flattening is not appealing for DCLL, under the geometrical configuration and material selection in this particular blanket design, since most of nuclear heating is caused by photon heat deposition.

Accident analysis on LOCA in HCCR-TBS towards CCWS-1

Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be operated at elevated temperature with high pressure helium coolant during normal operation in ITER. One of the main ancillary systems of HCCR-TBS is Helium Cooling System (HCS) which play an important role to extract heat from HCCR Test Blanket Module (TBM) by the helium coolant to keep the operational temperature and the extracted heat is finally transferred to ITER CCWS-1 (Component Cooling Water System) by a Printed Circuit Heat Exchanger (PCHE) in the HCS. In such circumstances if Loss Of Coolant Accident (LOCA) occurs in the PCHE, the high pressure helium coolant in the primary side goes into the lower pressure water in the secondary side thus pressurizing CCWS-1. In addition, since the helium coolant contains tritium due to permeation from the TBM, tritium migrates into CCWS-1, a non-nuclear system. In this paper, accident analysis for LOCA in the heat exchanger is presented. For the analysis, GAMMA-FR code which has been developed for fusion applications was used. Main components in the HCS and CCWS-1 were modelled as volume and junctions. The accident analysis was performed for the reference case with ten channels rupture and sensitivity study was also performed by changing the crack size. The results show that pressure and tritium requirement of CCWS-1 can be met in spite of LOCA in the heat exchanger of the HCCR-TBS HCS.

Study on Equilibrium Compositions of HTO/H2 Mixture in Purge Gas Flow in Tritium Breeding Blanket

The equilibrium chemical compositions of purge gas at the outlet of the Test Blanket Module in Helium Cooled Ceramic Reflector (HCCR) Test Blanket System are studied. Mole fractions of H, T, O, H2, HT, T2, H2O, HTO, and T2O in the equilibrium state are calculated by a Gibbs free energy minimization method starting from the initial state of H2/HTO mixture. The standard Gibbs free energy for tritium species obtained by the density functional theory is used in the calculations. The tritium recovery rates in the form of HT, T2, HTO and T2O are estimated from the equilibrium chemical compositions obtained in the calculations. The effects of H2 concentration in the purge gas on the tritium recovery process are also investigated by parametric study with variations of purge gas flow rate and volume ratio of hydrogen to helium in the purge gas.

Mechanical properties of ARAA steel after electron beam welding

Korea has designed a helium cooled ceramic reflector (HCCR) test blanket module (TBM) that includes a TBM shield, called a TBM set, that will be tested in ITER. The HCCR TBM is composed of four sub-modules and a back manifold. In addition, each sub-module is composed of a first wall (FW), a breeding box with a seven-layer breeding zone (BZ), and side walls with a cooling path. Korean RAFM steel was developed as a structural material for the HCCR TBM, and advanced reduced activation alloy (ARAA) was selected as the primary candidate from various program alloys. Welding technologies for fabrication of the HCCR TBM were developed using ARAA. Tensile, impact, bend tests were performed after post-weld heat treatment, and hardness, microstructure characteristics were determined before and after post-weld heat treatment to evaluate the welded specimen under the chosen welding conditions.

Neutron transmission measurement for natural W at nELBE

Korea has developed a Helium Cooled Ceramic Reflector Test Blanket Module (Ko HCCR TBM) related to the ITER project. Tungsten is considered as a prime candidate for the plasma facing materials in fusion reactors, and for the structure material of Ko HCCR TBM. KAERI (Korea Atomic Energy Research Institute) has been evaluating neutron cross sections of tungsten isotopes for neutron energy of up to 150 MeV based on nuclear reaction codes and available measurement data. New experimental data were measured at nELBE of HZDR (Helmholtz-Zentrum Dresden-Rossendorf) for a comparison with the evaluated and existing measurement data. The neutron source nELBE adopts a 40 MeV superconducting electron linac and a liquid Pb target for time-of-flight measurements. The nELBE neutron source uses no moderator and provides fast neutrons. An electron bunch length of 5 ps and a compact target provide a good neutron energy resolution with a relatively short flight length compared to other time-of-flight neutron sources. Transmission data of a natural tungsten sample were measured with a flight path length of 852.1 cm and a repetition rate of 101.56 kHz. The neutron total cross section of natural tungsten was obtained for an energy range of 100 keV to 10 MeV.

Thermal-Hydraulic Analysis for Conceptual Design of Korean HCCR TBM Set

With a conceptual design of Korean helium-cooled ceramic reflector test blanket module (TBM), including TBM shield for testing in ITER, thermal-hydraulic analyses are performed with a conventional CFD code, ANSYS-CFX v14.5, for the electromagnetic module and the integral module, including the TBM shield. With the same model and meshes, according to the ITER operation conditions of the H/He and D-T phases, the temperature distribution, flow rate, and pressure drop are investigated to meet the design requirements, and the temperature data are directly provided to a mechanical analysis.

Evaluation of ARAA steel E-beam welding characteristics for the fabrication of KO HCCR TBM

Korea has designed a helium cooled ceramic reflector (HCCR) test blanket module (TBM), including a TBM shield, called a TBM set, that will be tested in ITER. Korean RAFM steel was developed as a structural material for the HCCR TBM, and advanced reduced activation alloy (ARAA) was selected as the primary candidate from various program alloys. Fabrication technologies for the HCCR TBM were developed using ARAA to provide the method and procedure for fabricating the TBM for testing in ITER based on RCC-MRx, which was selected as the design and fabrication code and standard for the HCCR TBM. To establish and optimize welding procedures for electron beam welding of an ARAA material, variations in welding current and speed were investigated. A series of performance tests was performed before and after post-weld heat treatment to evaluate the welded specimen under the determined welding conditions.

Seismic analysis for conceptual design of HCCR TBM-set

Using the conceptual design of the Korean helium cooled ceramic reflector (HCCR) test blanket module (TBM) including the TBM-shield for testing in ITER, a seismic analysis is performed. According to the ITER TBM port plug (TBM PP) system load specifications, seismic events are selected as SL-1 (seismic level-1), SL-2 (seismic level-2), and SMHV (seismes maximaux historiquement vraisemblables, Maximum Histroically Probable Earthquakes). In a modal analysis a total of 50 modes are obtained. Then, a spectra response analysis for each seismic event is carried out using ANSYS based on the modal analysis results. For each event, the obtained Tresca stress is evaluated to confirm the design integrity, by comparing the resulting stress to the design criteria. The Tresca strain and displacement are also estimated for the HCCR TBM-set. From the analysis, it was concluded that the maximum stresses by the seismic events meet the design criteria, and the displacements are lower than the designed gap from the TBM PP frame. The results are provided to a load combination analysis.