Neutron Multiplier Material Research Articles

Tritium breeding mock-up experiments containing lithium titanate ceramic pebbles and lead irradiated with DT neutrons

Experiments were conducted with breeding blanket mock-up consisting of two layers of breeder material lithium titanate pebbles and three layers of pure lead as neutron multiplier. The radial dimensions of breeder, neutron multiplier and structural material layers are similar to the current design of the Indian Lead–Lithium cooled Ceramic Breeder (LLCB) blanket. The mock-up assembly was irradiated with 14MeV neutrons from DT neutron generator. The local tritium production rates (TPR) from 6Li and 7Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% 6Li and 7.54% 6Li) in Li2CO3. Tritium production in the multiplication layers were also measured with above mentioned two types of pellets to compare the experimental tritium production with calculations. TPR from 6Li at one location in the breeder layer was also measured by direct online measurement of tritons from 6Li(n, t)4He reaction using silicon surface barrier detector and 6Li to triton converter. Additional verification of neutron spectra (En>0.35MeV) in the mock-up zones were obtained by measuring 115In(n, n′)115mIn reaction rate and comparing it with calculated values in all five layers of mock-up. All the measured nuclear responses were compared with transport calculations using code MCNP with FENDL2.1 and FENDL3.0 cross-section libraries. The average C/E ratio for tritium production in enriched Li2CO3 pellets was 1.11 in first breeder zone and 1.09 in second breeder zone with uncertainty 8.3% at 1σ level. The experimental details and results are discussed in this paper.

Tritium release from highly neutron irradiated constrained and unconstrained beryllium pebbles

Beryllium is the reference neutron multiplier material in the Helium Cooled Pebble Bed (HCPB) breeding blanket of fusion power plants. Significant tritium inventory accumulated in beryllium as a result of neutron-induced transmutations could become a safety issue for the operation of such blankets as well as for the nuclear waste utilization. To provide a related materials database, a neutron irradiation campaign of beryllium pebbles with diameters of 0.5 and 1mm at 686–1006K, the HIDOBE-01 experiment, has been performed in the HFR in Petten, the Netherlands, producing up to 3020appm helium and 298appm tritium. Thermal desorption tests of irradiated unconstrained and constrained beryllium pebbles were performed in a purge gas flow using a quadrupole mass-spectrometer (QMS) and an ionization chamber. Compared to unconstrained pebbles, constrained beryllium pebbles have an enhanced tritium release at all temperatures investigated. Small elongated sub-grains formed under irradiation in the constrained pebbles promote formation of numerous channels for facilitated tritium release.

Thermal Conductivity of Composites of Beryllia and Lithium Titanate

The International Thermonuclear Experimental Reactor (ITER) is designed to demonstrate the scientific and technological feasibility of fusion power for energy purposes. Wide varieties of solid breeders and multiplier materials have been proposed for fusion blankets. Beryllium and lithium titanate (Li2TiO3) have been accepted as neutron multiplier and breeder materials, respectively. However, swelling of beryllium due to helium and tritium permeation through metallic beryllium and low thermal conductivity of lithium titanate have caused serious limitations when ITER is in the demo version. It has been well established that BeO due its highest thermal conductivity among the known ceramics, low neutron absorption cross section, and high neutron reflection cross section is a good neutron multiplier. In the present investigation, a novel ceramic single compound of BeO-Li2TiO3 was synthesized, keeping the BeO content to Li2TiO3 in the volume ratio of 80:20, 75:25, 65:35, and 55:45 with the aim of maintaining the tritium breeding ratio as more than one, and characterized for phases present by x-ray diffraction and scanning electron microscopy.

Ab initio static and molecular dynamics studies of helium behavior in beryllium

Beryllium is an effective neutron multiplier material widely exploited in nuclear applications. It will be used in the helium-cooled beryllium pebble bed of fusion reactor blankets for increasing the efficiency of tritium production. Macroscopic effects of irradiation (e.g., swelling) on beryllium are greatly influenced by accumulated transmutation helium. Atomic scale simulations of beryllium behavior under irradiation are necessary for understanding the basic mechanisms and reliable prediction of microstructural changes.In this study, we investigate the behavior of interstitial and substitutional helium, its diffusion pathways and interaction with point defects present in irradiated beryllium by means of static and molecular dynamics ab initio simulations. It was shown that a mixed dumbbell consisting of self-interstitial and helium atoms represents the ground-state configuration for interstitial helium. At low temperatures, the mixed dumbbell migrates in the basal plane through a series of in-basal-plane rotations, while at higher temperatures it jumps also between adjacent basal planes.It was revealed that, as in many other metals, interstitial helium atoms are bound to each other (Eb∼1eV). In beryllium, two mixed dumbbells meet each other so that helium atoms are the nearest neighbors, whereas the helium pair can be oriented either in- or out-of-basal plane. Diffusion modes of the pair are discussed.In addition, it was found that helium is very strongly bound to vacancies: the binding energy of more than 3eV found by static ab initio calculations suggests that helium is unlikely to be released from a mono-vacancy at temperatures below the melting point. As has been shown previously, vacancy clusters are unstable in beryllium. This study shows that the addition of helium stabilizes di-vacancies. Thus, it is confirmed that the presence of transmutation gas is necessary for the development of a porous microstructure in beryllium.

Oxidation behavior of plasma sintered beryllium–titanium intermetallic compounds as an advanced neutron multiplier

Beryllium intermetallic compounds (beryllides) such as Be12Ti are very promising candidates for advanced neutron multiplier materials in a demonstration fusion power reactor (DEMO). However, beryllides are too brittle to be fabricated either into pebble-type or rod-type shapes via conventional methods (i.e. arc melting and hot isostatic pressing). We have proposed a plasma sintering technique as a new method for beryllide fabrication, and our studies on the properties of plasma sintered beryllides are ongoing.In the present work, the oxidation properties of plasma sintered beryllides were investigated at 1273K for 24h in a dry air atmosphere to evaluate the high temperature properties of this material. Thermal gravimetry measurements indicate that specimens with larger fractions of Be12Ti phase corresponding to samples that have been sintered for longer time periods, exhibit superior oxidation properties.Our evaluation of the oxidation behavior of each phase in our beryllide samples is as follows: Be12Ti and Be17Ti2 both have good oxidation resistance, owing to the formation of dense and protective scales, while the Be and Be2Ti phases are mainly responsible for thermal-gravimetry (TG) weight gains, which is indicative of severe oxidation. We attribute the degradation in oxidation resistance specifically to Be2Ti that transforms into TiO2, and also find this phase to be the cause of deterioration in the mechanical properties of samples, owing to cracks near Be2Ti phase conglomerates.

A Study on 8–18Be Isotopes Used on Neutron Multiplier in Reactor Design

Neutronic characterization and development of structural materials, neutron multiplier materials, tritium breeders are primarily important for fusion and hybrid reactors. In order to improve neutron economy, beryllium, lead, bismuth, zirconium are considered and used as neutron multiplier materials in fusion and hybrid reactor design. In this study, rms charge radii, neutron radii, mass radii and neutron skin thickness were calculated for 8–18Be isotopes nuclei. The neutron and proton density are calculated for Be isotopes. The results obtained were compared with the experimental and theoretical results of other researchers by using Hartree–Fock method with an effective interaction with Skyrme forces.

Radiation growth of beryllium

Beryllium will be used as a neutron multiplier material in the present European HCPB (Helium Cooled Pebble Bed) blanket concept of the DEMO fusion reactor. Therefore, investigation of neutron irradiation influence on dimension stability of beryllium is very important. In this paper, the radiation damage of the TE-56 beryllium grade manufactured by hot extrusion was investigated. The beryllium samples were irradiated in the SM reactor at the temperatures of 70 °С and 200 °С up to neutron fluence of (1.3–14.2) 10 22 cm −2 (Е > 0.1 MeV). The measurements of the sample dimensions and density as well as microstructure examinations by X-rays and TEM were carried out. It was shown that superposition of two processes – radiation growth and anisotropic swelling occurs in beryllium under neutron irradiation.

Recent results on beryllium and beryllides in Japan

Recently, several research and development (R&D) programs of beryllium and its alloys for fusion reactor application have been conducted in Japan and the community of beryllium research has grown. In the R&D area of beryllium technology for a solid breeding blanket, a major subject is applications of beryllides, for which lifetime evaluation as neutron multiplier, impurity effect on irradiation behavior and recycling of irradiated materials are under investigation. As for neutron multiplier materials, recent R&D has focused on Be–Ti alloys based on Be 12Ti as an advanced option, because of their superior properties compared with beryllium metal in such respects as compatibility with structural materials and oxidation resistance at higher temperatures. In this paper, recent results on beryllide R&D in Japan are described.

ITER-Test blanket module functional materials

Solid breeder and liquid breeder blanket concepts are being developed for testing in ITER. In addition to fundamental investigations on the use of reduced activation ferritic/ferritic–martensitic steels (RAFMS) or V-alloys as the structural material, tritium breeder and neutron multiplier materials, and corresponding corrosion issues, there are three classes of functional materials being developed. Performance of these functional materials will significantly impact the performance of respective blanket concepts. Two classes of materials are related to the self-cooled liquid breeder design concepts. The first class is MHD coatings for the self-cooled Li breeder concepts. The second class is for flow channel inserts (FCI) for the dual coolant PbLi concepts. The third class of functional material is tritium permeation barrier material commonly needed for solid and liquid breeder blanket concepts. A description and the development status of these functional materials are described in this paper.

Utilization of thorium in a high power density hybrid reactor with innovative coolants

This study presents the neutronic analysis of a high power density (DT) fusion–fission (hybrid) reactor fuelled with thorium fuel for an operation period of 48 months. The effect of the innovative coolants, Li20Sn80 and Flinabe, used to breed tritium as well as transfer nuclear heat out of the blanket on the neutronic performance of the hybrid reactor is investigated. For comparison to these two coolants, natural lithium is chosen to be used in the same blanket. Calculations are performed on a simple experimental hybrid blanket with the aid of SCALE4.3 by solving the Boltzmann transport equation with the XSDRNPM code in 238 groups and a S8–P3 approximation. Numerical results bring out that the blanket using Li20Sn80 or Flinabe shows higher neutronic performance with respect to energy multiplication and fuel enrichment than that with natural lithium. However, when Li20Sn80 or Flinabe is used as coolant, neutron multiplier material will be required to maintain tritium self-sufficiency for the fusion driver at startup condition of the reactor.

Design and technology development of solid breeder blanket cooled by supercritical water in Japan

This paper presents results of conceptual design activities and associated R&D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed.By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study.In parallel with the design activities, engineering R&D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was successfully fabricated. It withstood the high heat flux test at 2.7 MW m−2. Also, a correlation parameter of the Li2TiO3 pebble bed made by the sol–gel method was verified by measurement of the thermal conductivity of the breeder pebble bed, which is one of the most important design data.

Preliminary neutronic estimation for demo blanket with beryllide

In preliminary neutronic assessment was conducted for the design of the DEMO blanket with neutron multiplier and tritium breeder materials packed in a multi-layered area of a blanket container. Beryllium metal and beryllides (Be 12W, Be 12Ti, Be 12V, etc.) are suggested as candidate neutron multiplier materials and material property evaluations of each candidate neutron multiplier materials under neutron irradiation have been carried out. This study estimates the effect of neutron multiplier materials and the effect of the packing configuration on the TBR of the SSTR-based blanket. This work shows that the tritium-breeding ratio (TBR) in a mixed type blanket is about 1.2 times greater than that of separated type blanket. This study shows that TBR of mixed type blanket using Be 12Ti or Be 12V will be about 1.2 (required TBR is about 1.3), which is 90% of the TBR for a mixed type blanket using Be. It is concluded that Be 12Ti and Be 12V are the most promising neutron multipliers from the view of the tritium breeding.

Preliminary analysis of waste recycling scenarios for future fusion power plants

As in the case of fission power plants, fusion power plants will have a need to transport radioactive material from the plant site, to the ultimate disposal facility. The objective of this study was to identify and assess means for minimizing the off-site transport of radioactive materials, and ultimately the waste quantity to be disposed of. The results indicate that there are practical options for achieving the objective, such as, on-site refurbishing of in-vessel components (IVCs). Refurbishing can also reduce the overall cost associated with the regular replacement of IVCs (about nine times in the reactor lifetime). On-site IVC refurbishing, however, requires that fusion power plants be constructed as multi-unit plants, preferably four or more units per site. The weakest element of the present study is the quantification of the unit cost for reprocessing neutron multiplier and breeder materials. To reduce the study uncertainties, better unit costs need to be developed for future studies.

Progress in blanket designs using SiC f/SiC composites

This paper summarizes the most recent design activities concerning the use of SiC f/SiC composite as structural material for fusion power reactor breeding blanket. Several studies have been performed in the past. The most recent proposals are the TAURO blanket concept in the European Union, the ARIES-AT concept in the US, and DREAM concept in Japan. The first two concepts are self-cooled lithium–lead blankets, while DREAM is an helium-cooled beryllium/ceramic blanket. Both TAURO and ARIES-AT blankets are essentially formed by a SiC f/SiC box acting as a container for the lithium–lead which has the simultaneous functions of coolant, tritium breeder, neutron multiplier and, finally, tritium carrier. The DREAM blanket is characterized by small modules using pebble beds of Be as neutron multiplier material, of Li 2O (or other lithium ceramics) as breeder material and of SiC as shielding material. The He coolant path includes a flow through the pebble beds and a porous partition wall. For each blanket, this paper describes the main design features and performances, the most recent design improvements, and the proposed manufacturing routes in order to identify specific issues and requirements for the future R&D on SiC f/SiC.

Development of materials and fabrication of porous and pebble bed beryllium multipliers

Beryllium is considered to be a neutron multiplier material for the reference ITER breeding blanket. The main requirements for the porous beryllium multiplier for the breeding blanket are: (1) inherently open porosity within 20 ± 2% for easy removal of radioactive gases; (2) high thermal conductivity; (3) close contact with a stainless steel (SS) shell to provide high heat transfer. A beryllium multiplier can be fabricated by two different techniques: by manufacturing porous or pebble bed beryllium. The method designed (patent 2106931 RU) in SSC RF-VNIINM (Russia) provides for the production of porous beryllium conforming to the requirements mentioned above. For comparative fission tests and the optimization of breeding zone functional capabilities, porous (21.9%) and binary pebble bed (density=78%) beryllium multipliers were fabricated. DEMO breeding blanket models and a mock-up of fission (IVV-2M reactor) tests have been manufactured at SSC RF-VNIINM.

Beryllium R&D for fusion applications

Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma–structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal.

Design development and manufacturing sequence of the European water-cooled Pb-17Li test blanket module

In 1996, the European Community started the development of a water-cooled Pb-17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the Basic Performance Phase prior to d–t operation. The test module is designed to be representative for a DEMO breeding blanket and relies on the liquid alloy Pb-17Li as both tritium breeder and neutron multiplier material, and water at PWR pressure and temperature as coolant. The structural material is martensitic steel. The straight, box-like structure of this blanket confines a pool of liquid Pb-17Li which is slowly circulated for ex-situ tritium extraction and lithium adjustment. The box and the Pb-17Li pool are separately cooled, the former with toroido-radial tubes, the latter with a bundle of double-walled U-tubes, equally made of martensitic steel and equipped with a permeation barrier. This paper presents the latest design and three manufacturing schemes with different degrees of technology. Advanced techniques such as solid or powder HIP are proposed to provide design flexibility. With a 3D neutronics analysis, the power and tritium generation were determined.

Development of Tritium Release Apparatus Using Pulse Mode Heating

Tritium release apparatus with the function of pulse mode heating were developed by using the infrared ray furnace to demonstrate the pulse mode heating of tritium breeder blanket for the fusion reactor. This apparatus was installed in the glove box of the beryllium PIE facility that constructed the hot-laboratory of Japan Materials Testing Reactor. The performance of this apparatus is that the minimum time of rapid heating up to 1015 °C is about 119 s and maximum heating rate reached at 1000 °C/min. The maximum temperature is depended on the crucible materials because of the differences for infrared ray absorption. The conversion efficiency of the gaseous water by ceramic electrolysis cell is above 99.99 %. The pulse mode heating of the tritium breeder and neutron multiplier materials of the blanket could be demonstrate by using this apparatus.

ITER R & D on breeder blanket materials

The International Thermonuclear Experimental Reactor (ITER) activity includes specific R&D in critical areas to support the ITER design. Several of the critical tasks defined as part of the ITER validating R&D are materials related. This paper presents a summary of the breeder/blanket materials related R&D conducted by the four participating parties; viz., European Community, Japan, Soviet Union and the United States. The current effort includes several subtasks in each of the following four task areas: 1. (1) ceramic breeder materials, 2. (2) lead-lithium breeder, 3. (3) aqueous salt solutions, 4. (4) neutron multiplier materials. The objective and scope of each subtask is stated. Although this work was initiated in late 1988, important results have been obtained in all four areas. The work on ceramic breeders includes properties measurements, compatibility studies, and tritium recovery experiments with Li 2O and selected ternary ceramics. Investigations on the lead-lithium breeder have focused primarily on properties, compatibility, and tritium extraction from the 17Li-83Pb eutectic alloy. The work on the aqueous salt breeders includes corrosion, solution chemistry, and radiolysis effects in a radiation environment. Investigations defined under the neutron multiplier task are focused on beryllium; however, lead in the lithium lead alloy also serves as a neutron multiplier. The schedule and planned future work in each area are summarized.

Chemical interactions of beryllium with lithium-based oxides and stainless steel

The chemical compatibility of Be with Li 2SiO 3, Li 4SiO 4 and stainless steel (AISI 316) was investigated in the temperature range between 600 and 900° C with maximum annealing times of 1000 h. Beryllium is of interest as a neutron multiplier material in a fusion reactor. First chemical interactions in the Be Li 2SiO 3 , and Be Li 4SiO 4 systems occur at 650 °C. The compatibility of Be with Li 2SiO 3 seems to be sufficient up to 650 °C, whereas that with Li 4SiO 4 is sufficient even up to 700 °C. At higher temperatures the silicide reaction products Li x Si y , become liquid which results in a strong local attack and penetration into the lithium silicates. Be interacts with stainless steel locally already at 600 ° C. The compatibility behavior in the Be/Li-silicate/stainless steel system under isothermal conditions is therefore determined by the Be/steel interactions.