Tritium Breeding Blanket Research Articles

DEM-CFD simulation of helium flow characteristics in randomly packed bed for fusion reactors

The helium cooled pebble bed (HCPB) packed spherical pebbles of lithium orthosilicate (Li4SiO4) blanket has been applied widely in fusion reactors. Helium flow and tritium transport characteristics are important for the design of the tritium breeding blanket. In this paper, the geometric models of randomly packed bed with different tube-to-particle diameter ratios (D/Dp) were established by discrete element method (DEM) to investigate helium flow characteristics. The results show that the wall effect can't be neglected, and the particle friction factor (fp) decreases with the increase of D/Dp. In addition, the relationships among porosity, velocity, and pressure distributions were investigated. It is found that the dimensionless velocity (Uave/U) and the dimensionless pressure (P/ΔP) distributions are not related to the inlet velocity, but only depend on the structure of bed. Besides, the transient tritium transport simulation shows that the transport efficiency in the regions away from the wall mainly depends on the velocity distribution. The efficiency in the regions near the wall is lower than other positions due to the wall effect. The steady tritium transport simulation shows that the transport efficiency depends not only on the velocity distribution but also on the distribution of tritium production.

Study on the Chemical Compatibility Study Between Li2TiO3 Pebbles and 14Cr-ODS Steel

In order to investigate the chemical compatibility between tritium breeder Li2TiO3 pebbles and tritium breeder blanket material oxide dispersion strengthened (ODS) steel, the contact interface between Li2TiO3 pebbles and ODS steel heated in argon atmosphere at 500, 600 and 700 °C for 300 h was studied. It was found that the ions of pebbles could diffuse and corrode with the cladding material after a long-time reaction at high temperature. The corrosion area formed on the surface of Li2TiO3 pebbles was small. With the increase of temperature, a zone with enriched iron was found on the surface of the pebble. This part of the surface was the direct contact surface between the pebble and the steel. At the same time, the relative density of the pebbles increased and the crush load was decreased to 30 N. In addition, a slight corrosion phenomenon was found on the surface of ODS steel. It has been proved that the main components of the corrosion products were the complex oxide containing Fe and Cr and the complex oxide containing Li and Fe.

Ab initio study of Be and Be12Ti for fusion applications

The paper applies ab initio methods for investigation of possible origin of the easier tritium release from titanium beryllide Be12Ti as compared with that of beryllium. Both materials are considered as neutron multiplier materials in some design concepts of tritium breeding blanket of future fusion reactor. The only stable tetragonal crystal structure of Be12Ti was studied in this work. Hydrogen solution energy in all non-equivalent interstitial positions of Be12Ti was found to be lower than that for beryllium suggesting easier dissolution of hydrogen atoms in Be12Ti. Maximal binding energy of hydrogen atoms in vacancy linearly decreases with the number of hydrogen atoms both in beryllium and Be12Ti. The highest hydrogen binding in a vacancy is 0.4 eV lower in Be12Ti, predicting easier release of hydrogen from beryllide with respect to beryllium. No formation of chemical bonds between hydrogen atoms inside both beryllium and Be12Ti vacancies was observed.

Evaluation of Cross Section of 10Be Radioactive Nuclei Production at Deuteron Interaction with 9Be

The demand for more precise nuclear data on beryllium interactions is growing along with the development of controlled fusion facilities, where beryllium is already used or is planned to be used as a structural material in the coating of the first inner wall of the tokamak (the tritium breeding blanket). Evaluated cross sections of the 9Be(d, x)10Begs reaction at deuteron energies ranging from 0 to 22 MeV are presented. All available experimental data and the mathematical tools of the SaBa [1] electronic library developed at the Russian Federal Nuclear Center All-Russian Research Institute of Experimental Physics are used to obtain these cross sections.

Preliminary Design of the CFETR HCCB Blanket With S-Shaped Cooling Plate in BU and Thermal–Mechanical Analysis of the First Wall

Three concepts of tritium breeding blanket have been proposed for China Fusion Engineering Test Reactor. One of the concepts is helium-cooled ceramic breeder (HCCB) blanket. The HCCB blanket has the S-type cooling pipes in breeding unit (BU), and the BU consists of lithium ceramic pebble beds and beryllium pebble beds. The breeding material and the multiplier material are separated by the cooling plates. Theoretical calculations have been done to obtain temperature distribution and pressure drop for the preliminary design of the breeding blanket. Then, ANSYS CFX is employed to verify the thermal performance of the first wall (FW) in radial–toroidal and poloidal–toroidal directions. The temperature distribution in FW is obtained, and the optimization of cooling channels is proposed according to the simulation results. Finally, according to the temperature distribution, structural stress analysis has been done to verify the feasibility of the design.

Conceptual studies of gamma ray diagnostics for DEMO control

The future tokamak demonstration fusion reactor (DEMO) will operate at unprecedented physical and technological conditions where high reliability of the system components is required. The conceptual study of a suite of DEMO diagnostics is on-going. Among these, a Gamma-Ray Spectrometric Instrument (GRSI) is being investigated to assess its performance and information quality in view of DEMO control. The GRSI foresees radial-orthogonal multi-line of sight viewing DEMO plasma across its poloidal section as a further development of the Gamma-Ray Camera of JET and of the Radial Gamma-Ray Spectrometers proposed for ITER but with stricter technological constraints. These include surface availability in the Tritium Breeding Blankets of DEMO vessel inner wall for diagnostics collimators openings, diagnostics distance from the plasma, neutron irradiation and activation of the reactor structures. On DEMO the gamma-ray (γ) emission from DT plasmas consists of T(d,γ)5He (Eγ = 16.63 MeV) and T(p,γ)4He (Eγ = 19.81 MeV) reactions which for their high Eγ would allow in principle for background-free measurements. This work reports the assessment on the GRSI diagnostic capability. Reactions cross sections are assessed and used for the calculations of the reactions γ emission energy spectrum under DEMO DT plasma conditions and compared with 14 MeV neutron emissions before and after the GRSI collimator. Investigation of the GRSI γ spectrometers performance is also presented. Measurement of the γ emission intensity of T(d,γ)5He can be in principle used as an independent assessment of DEMO DT plasmas fusion power.

Numerical investigation of the pebble bed structures for HCCB TBM

The packing structures of tritium breeder and neutron multiplier pebble beds are important to estimate the thermal mechanical properties and the flow characteristics of purge gas in solid tritium breeding blanket. In this study, the pebble packing structures were numerical studied based on DEM simulation. The packing structures of cylinder pebble beds were further analyzed. The results show that the aspect ratio has a great influence on packing structure of pebble bed. Increasing the aspect ratio, a higher average packing factor can be obtained. Further, the U-shaped Li4SiO4 pebble beds were investigated. The results show that when the width of the U-shaped channel, the height of the bed and the diameter of the pebbles are given, the dimensions of the U-shaped container have little effect on the pebble bed packing structures. In addition, the influences of pebble size on packing structures of U-shaped pebble bed were analyzed and discussed. By decreasing the pebble size, the average packing factor can be increased, the wall effect region of the U-shaped pebble bed can be also reduced relatively and a more uniform pebble bed can be obtained.

Maintainability of the helical reactor FFHR-c1 equipped with the liquid metal divertor and cartridge-type blankets

The maintenance scheme in the compact helical fusion reactor, FFHR-c1, equipped with the liquid metal divertor, REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) and the cartridge-type molten salt blankets, CARDISTRY-B (CARtridges Divided and InSerTed RadiallY − Blanket), is investigated. The magnetic configuration of the compact variant FFHR-c1 is similar to that of the Large Helical Device (LHD), while the device size is 2.8 times enlarged from LHD and a strong magnetic field strength of ∼8 T at the plasma center is adopted. The maintenance of the REVOLVER-D is simpler than that of the helical divertor with a complicated structure as seen in the LHD. In the REVOLVER-D, showers of molten tin are injected into the ergodic layer at 10 inner ports. To circulate the molten tin, 10 sets of the shower system including a liquid metal pump, ducts, a showerhead, a pool, and a heat exchanger, are installed. These can be replaced with simple up/down motions. The CARDISTRY-B consists of 320 tritium breeding blanket cartridges, which are toroidally segmented every two degrees. These cartridges are maintained by using heavy manipulators with a simple 4DOF motion at a constant toroidal angle.

Preliminary assessment on safety performance of updated HCCB blanket module for CFETR

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.

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

The Indian test blanket module (TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R&D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices (ITER relevant and DEMO). The Indian Lead–Lithium Cooled Ceramic Breeder (LLCB) blanket concept is one of the Indian DEMO relevant TBM, to be tested in ITER as a part of the TBM program. Helium-Cooled Ceramic Breeder (HCCB) is an alternative blanket concept that consists of lithium titanate (Li2TiO3) as ceramic breeder (CB) material in the form of packed pebble beds and beryllium as the neutron multiplier. Specifically, attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions. These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.

New high performance computing software for multiphysics simulations of fusion reactors

Integrated modelling is an important element of tokamak fusion research. It contributes to the interpretation and planning of experiments, validation of theory against experimental results, development of plasma control techniques as well as the design of next step devices. Large efforts continue to be devoted to the integrated modelling of plasma physics effects in a fusion reactor core and have led to significant progress in physics understanding, while tools to model multiphysics phenomena taking place in fusion reactor components such as tritium breeding blankets have not yet reached a similar level of detail and completeness. We are developing a new high performance computing tool for simulating the latter effects. It is based a parallel computational mechanics code Alya which is specially designed for running with high efficiency standards in large scale supercomputers and is capable of solving different physics problems in a coupled way. As first step towards multiphysics simulations of fusion reactor components, we are developing a new neutron transport module based on the deterministic approach for coupling with the existing, implemented physics modules in Alya. This module, together with existing physics implemented in Alya will allow to solve future high fidelity multiphysics and multiscale problems. We present the capabilities of the Alya system and the latest version of the new neutron transport module currently under development.

Evaluation of Li mass loss from Li2TiO3 with excess Li pebbles in water vapor atmosphere

Understanding of Li mass transfer behavior in a tritium breeding blanket is an important issue from viewpoints of establishment of tritium cycle and tritium safety. In this work, weight reduction property of Li2TiO3 with excess Li pebbles, which were fabricated by National Institutes for Quantum and Radiological Science and Technology, was investigated and the amount of Li mass loss and the rate of Li mass loss in water vapor atmosphere at elevated temperatures were evaluated. The Li mass loss proceeding at 900 °C with a relatively high rate was limited. The Li mass loss of Pebble210 (Li/Ti = 2.10) was 1.2 wt% and that of Pebbe211 (Li/Ti = 2.11) was 1.4 wt%, eventually. The rate of the Li mass loss increased with increasing temperature and it seemed to increase proportionally to a square root of water vapor pressure. An empirical formula for the Li mass loss was proposed as a function of temperature, heating time and water vapor pressure in a purge gas.

Design and analysis of manifolds for Indian HCCB blanket module

The Indian LLCB blanket concept is one of the Indian DEMO relevant test blanket module, to be tested in ITER as a part of the TBM program. Helium-Cooled Ceramic Breeder (HCCB) is an alternative blanket concept for its DEMO, which consists of lithium titanate (Li2TiO3) as ceramic breeder (CB) material in the form of packed pebble beds and beryllium as the neutron multiplier. Indian TBM program in ITER is one of the major steps in Indian fusion reactor program towards DEMO. It establishes the synergy among research institutes and industries for carrying out the R&D activities in the critical areas like design of tritium breeding blankets relevant to DEMO. In this paper, the HCCB blanket based on ITER like fusion reactor has been proposed and analysed. The design of manifolds is an important part in a breeding blanket design since it determines how helium is getting distributed in to different circuits for cooling different zones and components in blanket module. The flow scheme for HCCB blanket has been optimized based on its flow parameters. The inlet helium coolant of 300C comes out at 360C from FW and is then distributed into breeding unit zones, grid plates, top plate and bottom plate in parallel configuration. Hydraulic analysis of different circuits using ANSYS CFX has been performed to estimate the flow distribution from manifolds, velocities and pressure drop in different circuits. Heat transfer coefficients have also been evaluated using the obtained velocities in different cooling channels. The detailed design and analyses of the flow schemes in different circuits of blanket module and the purge helium flow have been discussed in this report. The thermo-mechanical design and analysis at using ANSYS mechanical has also been performed at design and operating conditions for p type and s type damages including fatigue and discussed in this report. This study will further help in the design of DEMO blanket module.

Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor

A configuration of a fusion-driven transmutation reactor with a low aspect ratio tokamak-type neutron source was determined in a self-consistent manner by using coupled analysis of tokamak systems and neutron transport. We investigated the impact of blanket configuration on the characteristics of a fusion-driven transmutation reactor. It was shown that by merging the TRU burning blanket and tritium breeding blanket, which uses PbLi as the tritium breeding material and as coolant, effective transmutation is possible. The TRU transmutation capability can be improved with a reduced blanket thickness, and fast fluence at the first wall can be reduced. Article History: Received: July 10th 2017; Received: Dec 17th 2017; Accepted: February 2nd 2018; Available onlineHow to Cite This Article: Hong, B.G. (2018) Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor. International Journal of Renewable Energy Development, 7(1), 65-70.https://doi.org/10.14710/ijred.7.1.65-70

Testing of tritium breeder blanket activation foil spectrometer during JET operations

Accurate measurement of the nuclear environment within a test tritium breeding-blanket module of a fusion reactor is crucial to determine tritium production rates which are relevant to self-sufficiency of tritium fuel supply, tritium accountancy and also to the evaluation of localised power levels produced in blankets. This requires evaluation of the time-dependent spectral neutron flux within the test tritium breeding-blanket module under harsh radiation and temperature environments. The application of an activation foil-based spectrometer system to determine neutron flux density using a pneumatic transfer system in ITER has been studied, deployed and tested on the Joint European Torus (JET) machine in a recent deuterium - deuterium campaign for a selection of high purity activation foils. Deployment of the spectrometer system has provided important functional and practical testing of the detector measurement system, associated hardware and post processing techniques for the analysis of large data sets produced through the use of list mode data collection. The testing is invaluable for the optimisation of systems for future planned testing in tritium - tritium and deuterium - tritium conditions. Analysis of the time and energy spectra collected to date and the status of the development of methods for post-processing are presented in this paper.

Waste assessment of European DEMO fusion reactor designs

Predicting the amount of waste produced from a fusion power plant is vital to assess the likely environmental impact, disposal costs, and also to satisfy nuclear regulators. Inventory simulations are ideally suited to this task because they can be used to evolve in time, via numerically solving the ordinary differential rate equations, the chemical composition of reactor materials, both during operation, when exposed to neutron irradiation, and afterwards, during post-operation decay. The radiological response of the time-evolving material composition can then be used to assign a waste class to reactor components. Waste assessment has been performed for designs of the European demonstration power plant (DEMO) concept using the FISPACT-II inventory simulator, and using extensive Monte Carlo simulations of the neutron irradiation fields as input. The masses in each waste class (defined using the IAEA infrastructure) have been charted in time for in-vessel components (including the divertor and blanket), ex-vessel regions such as the coils, and for the reactor vacuum vessel (VV). Comparisons are made between the waste-class masses generated for different tritium breeding blanket concepts. Typical predictions include the observation that the majority of the VV will become low-level waste within 100 years, while plasma-facing components and tritium breeding units will remain classified as intermediate-level waste for longer. The waste classification implications of sub-dividing the large VV is considered, highlighting the potential benefits, for waste disposal and recycling prospects, of being able to separate low activity regions of a component from more active regions.

Study on the temperature control mechanism of the tritium breeding blanket for CFETR

The Chinese fusion engineering testing reactor (CFETR) will demonstrate tritium self- sufficiency using a tritium breeding blanket for the tritium fuel cycle. The temperature control mechanism (TCM) involves the tritium production of the breeding blanket and has an impact on tritium self-sufficiency. In this letter, the CFETR tritium target is addressed according to its missions. TCM research on the neutronics and thermal hydraulics issues for the CFETR blanket is presented. The key concerns regarding the blanket design for tritium production under temperature field control are depicted. A systematic theory on the TCM is established based on a multiplier blanket model. In particular, a closed-loop method is developed for the mechanism with universal function solutions, which is employed in the CFETR blanket design activity for tritium production. A tritium accumulation phenomenon is found close to the coolant in the blanket interior, which has a very important impact on current blanket concepts using water coolant inside the blanket. In addition, an optimal tritium breeding ratio (TBR) method based on the TCM is proposed, combined with thermal hydraulics and finite element technology. Meanwhile, the energy gain factor is adopted to estimate neutron heat deposition, which is a key parameter relating to the blanket TBR calculations, considering the structural factors. This work will benefit breeding blanket engineering for the CFETR reactor in the future.

Chemical characterization of lithium titanate and lithium aluminate as tritium breeders of fusion reactor by PIGE and INAA methods

Lithium titanate and lithium aluminate, proposed as tritium breeding blanket materials in D–T fusion reactor, were analyzed by nuclear analytical techniques as a part of chemical quality control exercise. Concentrations of Li, Ti and Al were quantified by particle induced gamma-ray emission using 4 MeV proton beam from folded tandem ion accelerator facility. Trace elemental concentrations belonging to long-lived nuclides were determined by instrumental neutron activation analysis using neutron flux from Dhruva research reactor at BARC. The analysis results are useful in designing a shielded facility for tritium release studies of these irradiated breeder materials.

Evaluation and optimization of tritium breeding, shielding and nuclear heating performances of the helium cooled solid breeder blanket for CFETR

Chinese Fusion Engineering Test Reactor (CFETR) is a new test tokamak device being designed in China aimed to bridge gaps between ITER and DEMO. As one of the candidate tritium breeding blankets, a conceptual design scheme of the helium cooled solid breeder blanket has been proposed and a series of preliminary analyses have been carried out to access the performances. However, both the required global tritium breeding ratio for CFETR not less than 1.2 and its poor working conditions under intense radiation need further thorough neutronics analyses and optimizations during the design phase. In this work, first, three-dimensional neutronics model of CFETR was built, and the neutron wall loading and global TBR were obtained. The nuclear and thermal calculation results were automatically coupled, which could make the neutronics calculation results more accurate and guaranteed they could always satisfy the corresponding thermal limits during the whole process. Then, the tritium breeding and shielding performances of both the outboard and inboard equatorial blanket modules were optimized for the comprehensive optimal schemes. The influences of Be/W armors on the shielding performance and TBR were also investigated. Finally, the nuclear heating rates and the neutron flux densities in different components were calculated based on the obtained comprehensive optimal scheme. In this paper, the neutronics analyses and optimizations verified that the optimized conceptual design could well meet the tritium self-sufficiency and neutron shielding requirements, and this could provide a valuable reference for the further thermal-hydraulic analysis and structural optimization of the CFETR helium cooled solid breeder blanket.

Discrete element modeling of pebble bed packing structures for HCCB TBM

In solid tritium breeder blanket, the tritium breeder and neutron multiplier are used in formed of pebbles. To reveal the inner packing structures of the pebble bed, in this paper, the discrete element method (DEM) was applied to simulate the pebble bed packing structures by pouring pebbles from the top of container. The current numerical results obviously show that with the increase of aspect ratios the packing factor can be significantly increased in both cylinder pebble bed and cubic pebble bed. Namely, the larger aspect ratios in cylinder and cubic pebble bed, the smaller the proportion of the wall affected region, the greater the average packing factor of the cylinder and cubic pebble bed. Furthermore, the pebble packing structures in a U-shaped container were also simulated and analyzed. In the bend pebble bed, the average packing factor of 0.6278 was obtained. Close to the convex and concave container wall, the wall effects of pebble distribution and packing factor were observed. With the increase of the distance to the container wall, the layered arrangements disappear gradually and the oscillating characteristics of local packing factor are also damped little by little. The oscillations of local packing factors are limited in about 5 diameters. A transition region is formed between the layered packing in the near wall region and the uniformly random packing in the inner bulk region. In addition, the contact force distribution and coordination number distribution were also detailed analyzed.