Solid Breeder Concept Research Articles

Development of water-cooled cylindrical blanket in JA DEMO

The concept of the tritium breeding blanket for Japan’s DEMOnstration fusion reactor (JA DEMO) has been developed with pressure tightness against in-box loss-of-coolant accidents based on a water-cooled solid breeder concept. The cooling conditions are designed on the pressurized-water reactor water conditions which are the coolant temperature of 290 °C–325 °C and the operating pressure of 15.5 MPa, respectively. The point of the blanket design is to reduce the amount of structural material in casing as well as to ensure its pressure tightness. This is because a decrease in the amount of structural material improves tritium breeding ratio (TBR). A cylindrical structure, a thin wall casing structure which ensures pressure tightness and could increase TBR, is feasible. However, a relatively larger useless space is expected between modules when the cylindrical blanket modules are arranged in a vacuum vessel, which could decrease TBR. Therefore, the cylindrical blanket modules are to be in a close-packed arrangement to reduce useless space. The Be12Ti block (which shows a minor swelling compared to Be) is selected to achieve the target TBR as the net Be density is equivalent to the case of the Be pebble. The use of Be12Ti blocks can reduce or remove the cooling piping inside the module as the Be12Ti block has a higher thermal conductivity than pebbles. As a result of the neutronics, finite element method, and computational fluid dynamics analyses, it was found that the target TBR value can be achieved in cylindrical structure blanket that ensure pressure tightness.

Tritium Breeding Ratio Evaluation of Solid Breeder Concepts for the FESS-FNSF

This paper presents a parametric study of the Fusion Energy System Studies-Fusion Nuclear Science Facility’s (FNSF’s) tritium breeding performance for several solid breeder concepts, neutron multiplying materials, and blanket materials, assuming volume fractions based on the most recent FNSF design as a realistically representative fusion facility. In this study, we initially surveyed the tritium breeding ratio (TBR) of several solid breeder concepts by employing a simplified but efficient one-dimensional (1-D) infinite cylinder reduced-order model (ROM). Parametric studies were performed with the ROMs for the full range of breeder-to-multiplier ratios to identify the optimum mixture compositions for each breeder type that would lead to a maximum TBR. These optimized breeder-multiplier combinations were then homogenized with FNSF blanket component materials to estimate their impacts on the TBR. Subsequently, as a validation step for the optimal designs, TBR calculations were performed using a more realistic modified 1-D ROM with inner and outer breeding regions, as well as with a fully detailed 22.5-deg three-dimensional (3-D) sector of the FNSF to assess the impact of geometry details on the TBR. The differences between the two 1-D models were negligible, while the ROMs were able to correctly predict trends and identify the maximum and minimum TBR cases, as well as show consistent biases relative to the results produced by the full 3-D, 22.5-deg sector for specific breeder/multiplier combinations. Solid breeder concepts such as , , and outperformed all others in this study in terms of TBR performance when combined with all the neutron multiplier materials selected. An underlying goal of this study was to develop and improve rapid and reliable ROMs to aid designers during parametric optimizations of highly complex and computationally expensive fusion models.

High temperature oedometric compression of Li2TiO3 pebble beds for Indian TBM

Lithium Metatitanate (Li2TiO3 or LMT) and Lithium Orthiosilicate (Li4SiO4 or LOS) are the prominent among the suitable candidate materials for breeders. Robust design of the breeder blankets requires a thorough understanding of the thermo-mechanical response of the breeder materials at different loading conditions. In this context, the material characterization plays a vital role. Solid breeder concept of various member nations comprises of pebbles in steel container. Determination of effective properties of the pebble beds helps in understanding and formulation of constitutive relations for such granular systems. In order to develop simulation tools to understand the complex thermo-mechanical response of the breeder units, the properties of the material should be known. Hence, the objective of this work is to characterize crush load of Li2TiO3 pebbles and investigate the thermo-mechanical response of the pebble beds. The study involves determining the crush load of individual pebbles and pebble-bed elastic modulus. Uniaxial oedometric tests are performed to investigate the influence of temperature on the effective elastic modulus of the pebble beds.

Numerical and experimental characterization of ceramic pebble beds under cycling mechanical loading

All solid breeder concepts considered to be tested in ITER (International Thermonuclear Experimental Reactor), make use of lithium-based ceramics in the form of pebble-packed beds as tritium breeder. A thorough understanding of the thermal and mechanical properties of the ceramic pebble beds under fusion relevant conditions is essential for the design of the breeder blanket modules of future fusion reactors. In this study, the effect of cyclic loading on the mechanical behaviour of pebble bed assemblies was investigated using a Discrete Element Method (DEM) code. The numerical simulations were compared with the experimental outcomes. The results of numerical simulations show that the pebble size distribution affects noticeably the stress-strain behaviour of the assemblies. A good qualitative agreement between experimental and simulation results was found in terms of difference between residual strains of consecutive cycles. An increase of the oedometric modulus with the compressive load was observed for all investigated compositions in both experimental and DEM simulations. The numerical results show an increase of the oedometric modulus (E) with progressive compaction of the assemblies due to the cycling loading, while no significant influence of the pebbles size distribution was observed.

ChemInform Abstract: Recent Developments of Solid Breeder Fabrication

AbstractReview: 25 refs.

Size-Dependent Crush Analysis of Lithium Orthosilicate Pebbles

The crushing strength of the breeder material [lithium orthosilicate (Li4SiO4 or OSi)] in the form of pebbles to be used for EU solid breeder concept is investigated. The pebbles are fabricated using a melt-spray method, and hence, a size variation in the pebbles produced is expected. Knowledge of the mechanical integrity (crush strength) of the pebbles is important for a successful design of a breeder blanket. In this paper, we present the experimental results of the crush (failure) loads for spherical OSi pebbles of different diameters ranging from 250um to 800um. The ultimate failure load for each size shows a Weibull distribution. Furthermore, the mean crush load increases with increase in pebble diameter. It is also observed that the level of opacity of the pebble influences the crush load significantly. The experimental data presented in this paper and the associated analysis could possibly help us to develop a framework for simulating a crushable polydisperse pebble assembly using discrete element method.

Recent developments of solid breeder fabrication

Worldwide research and development has been performed on solid breeder materials as a tritium source for future fusion power reactors. These ceramic materials will be used in the International Thermonuclear Experimental Reactor (ITER) to check their performance and screen their applicability for use in fusion blankets. While all members of the ITER project are engaged in the development of solid breeder blanket concepts, five are actively involved in the fabrication of tritium breeding ceramics for the different test blanket modules to be tested in ITER.All solid breeder concepts proposed for ITER are based on the use of ceramics in the form of pebble beds with pebble diameters of about 1mm or smaller. Although alternative materials are still under investigation, the main candidates for tritium breeding ceramics are lithium metatitanate and lithium orthosilicate. Recent developments and preparatory work for ITER are presented.

Safety assessment of Chinese ITER TBM design with helium-cooled solid breeder concept

Safety analyses have been performed for the Chinese ITER test blanket module (TBM) design with helium-cooled solid breeder (HCSB) concept for testing in a ITER device based on the design developed with the China blanket programme. Some safety issues associated with this TBM design are explored. In particular, radiological inventories, afterheat, waste disposal ratings (WDRs), electromagnetic analysis, LOCA accident analysis and tritium safety management for the CH HCSB TBM design are given. The analysis results show that the total radiological inventory in the TBM at shutdown is very low and drops rapidly within a minute due to the decay of the Mn-56 isotope, the total afterheat is as low as 8.77 × 10−3 MW, which is attributed mainly to the structure and the WDRs for Ni-59, Zr-93 and Nb-94 are 1.48 × 10−2, 4.72 × 10−9 and 1.034 × 10−3, respectively, and according to the US 10CFR61 regulation, these materials are qualified for shallow land burial. Results show that this test blanket concept is feasible within the existing domestic technologies and has good safety characteristics. In addition, future directions in the safety and environmental area are proposed.

An overview of US ITER test blanket module program

A testing strategy and corresponding test plan have been presented for the two proposed US candidate breeder blankets: (1) a helium-cooled solid breeder concept with ferritic steel structure and Be neutron multiplier, but without a fully independent TBM and (2) a dual-coolant helium-cooled ferritic steel structure with self-cooled LiPb breeding zone that uses a flow channel insert as MHD and thermal insulator. Example test module designs and configuration choices for each line of ITER TBM are shown and discussed in the paper. In addition, near-term R&D items for decision-making on testing of both solid breeder and dual-coolant PbLi liquid breeder blanket concepts in ITER are identified.

Vanadium-base alloys for fusion reactor applications — a review

Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys with emphasis on the V-15Cr-5Ti alloy, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R&D) needs are defined. The relatively high thermal conductivity and low thermal expansion coefficient of vanadium-base alloys, which result in lower thermal stresses for a given heat flux compared to most other candidate alloys, should enhance the reactor wall-load and lifetime capability. Since the mechanical strength of vanadium-base alloys is retained at relatively high temperatures, higher operating temperatures are projected for these alloys than for austenitic or ferritic steels. The refractory metals, including vanadium, characteristically exhibit good corrosion resistance in purified liquid metals. The vanadium alloys also exhibit favorable neutronic properties which include lower parasitic neutron absorption leading to better tritium breeding performance, lower bulk nuclear heating rates, and lower helium generation rates compared to the steels. Limited data from moderate fluence neutron irradiations (fast breeder reactor) and high-damage level ion irradiations indicate good radiation damage resistance. Major concerns regarding the use of vanadium-base alloys for fusion reactor applications relate to cost, reactivity with oxygen/air, and the limited data base. The impact of these concerns and potential design solutions are discussed. Because of the corrosion/compatibility considerations, the vanadium alloys offer greater promise for liquid metal breeder concepts than for the solid breeder concepts.

Water-Cooled Blanket Concepts for the Blanket Comparison and Selection Study

The primary goal of the Blanket Comparison and Selection Study (BCSS) was to select a limited number of blanket concepts for fusion power reactors, to serve as the focus for the U.S. Department of Energy blanket research and development program. The concepts considered most seriously by the BCSS can be grouped for discussion purposes by coolant: liquid metals and alloys, pressurized water, helium, and nitrate salts. Concepts using pressurized water as the coolant are discussed. Water-cooled concepts using liquid breeders-lithium and 17Li-83Pb (LiPb)-have severe fundamental safety problems. The use of lithium and water in the blanket was considered unacceptable. Initial results of tests at Hanford Engineering Development Laboratory using steam injected into molten LiPb indicate that use of LiPb and water together in a blanket is a very serious concern from the safety standpoint. Key issues for water-cooled blankets with solid tritium breeders (Li/sub 2/O, or a ternary oxide such as LiAlO/sub 2/) were identified and examined: reliability against leaks, control of tritium permeation into the coolant, retention of breeder physical integrity, breeder temperature predictability, determination of allowable temperature limits for breeders, and /sup 6/Li burnup effects (for LiAlO/sub 2/). The BCSS's final rankings and associated rationale for all water-cooledmore » concepts are examined. Key issues and factors for tokamak and tandem mirror reactor versions of water-cooled solid breeder concepts are discussed. The reference design for the top-ranked concept-LiAlO/sub 2/ breeder, ferritic steel structure, and beryllium neutron multiplier-is presented. Finally, some general conclusions for water-cooled blanket concepts are drawn based on the study's results.« less

Ceramic Lithium Compounds for Solid Breeder Concepts for Fusion Reactors

Thermophysical and thermochemical data on ceramic lithium compounds proposed as tritium breeding material for the solid breeder concept of fusion reactors using a purge gas to continuously extract the tritium bred are discussed. Diffusion coefficients for tritium in these materials from the data of tritium extraction experiments reported in the literature are critically examined and reevaluated. The effect of the diffusion rate and of the solubility of tritium on the tritium inventory in the breeder material is discussed. Only extended “in-pile” experiments in fission reactors can answer the question whether or not the application of a solid breeder concept is really feasible in a future fusion reactor.

Design and Analysis of the Solid Breeder Integral Simulation Test for PE-IIa

The PE-II Solid Breeder Concept Integral Simulation Test is designed to simulate (by way of nonnuclear heating) the thermal-hydraulic/ thermomechanical operating characteristics of a water-cooled, Li/sub 2/O fusion reactor blanket. The test piece consists of a central coolant tube surrounded by a cylinder of solid breeder material. This test piece, which simulates a unit cell of a blanket, is then subjected to an external heat flux and a purge flow of helium at carefully controlled conditions. Instrumentation provides information on temperature profiles, moisture levels, and purge flow conditions. Pretest thermal/stress analysis of this experiment indicates that it can provide an accurate simulation of actual conditions in a prototypical fusion reactor blanket.