Liquid Breeder Research Articles

An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

Al-based anti-corrosion and T-permeation barrier development for future DEMO blankets

In the Helium-Cooled-Liquid-Lead (HCLL) design of Test-Blanket-Modules (TBM’s) for a future fusion power plant Pb–15.7Li is used as liquid breeder which is in direct contact with the structure material, e.g. EUROFER steel. Compatibility testing showed that high corrosion attack appears and that the dissolved steel components form precipitates with a high risk of system blockages. A reliable operation needs coatings as corrosion barriers. The earlier developed Hot-Dip Aluminisation (HDA) process has shown that Al-based scales can act as anti-corrosion as well as T-permeation barriers. Meanwhile two advanced electro-chemically based processes for deposition of Al-scales were successfully developed. The first (ECA=Electro-Chemical Al-deposition) is working with an organic electrolyte and the second one (ECX=Electro-Chemical-X-metal-deposition) is based on ionic liquids. Coatings in the μm-range were deposited homogeneously with exact controllable thicknesses. Metallurgical investigations showed the successful generation of protective scales and compatibility testing demonstrated the barrier function.

Nitriding Treatment of Reduced Activation Ferritic Steel as Functional Layer for Liquid Breeder Blanket

The development of functional layers such as a tritium permeation barrier and an anti-corrosion layer is the essential technology for the development of a molten salt type self cooled fusion blanket. In the present study, the characteristics of a nitriding treatment on a reduced activation ferritic steel, JLF-1 (Fe-9Cr-2W-0.1C) as the functional layer were investigated. The steel surface was nitrided by an ion nitriding treatment or a radical nitriding treatment. The nitridation characteristic of the steel surface was made clear based on the thermodynamic stability. The thermal diffusivity, the hydrogen permeability and the chemical stability in the molten salt Flinak were investigated. The results indicated that the nitriding treatment can improve the compatibility in the Flinak without the decrease of the thermal diffusivity, though there was little improvement as the hydrogen permeation barrier.

Nitrogen contamination effect on yttrium gettering of hydrogen in liquid lithium

Yttrium is a candidate for hydrogen hot trapping in a liquid tritium breeder of a fusion reactor blanket and in the flowing target of the International Fusion Materials Irradiation Facility, IFMIF. Nitrogen impurity in lithium is expected to decrease hydrogen trapping but this has not been confirmed. This study, directed at developing a hydrogen permeation device as a hydrogen concentration monitor in lithium, investigated hot trapping by yttrium in lithium. The hydrogen trapping rate of yttrium first aged in lithium containing nitrogen was smaller than the rate without nitrogen contamination. Nitrogen in contaminated yttrium did not form yttrium nitride, and the nitrogen diffused much deeper into the yttrium than did oxygen.

Numerical study of MHD pressure drop in rectangular ducts with insulating coatings

The MHD (magnetohydrodynamic) effect is a key issue, and quite often a constraint, in the development of liquid metal breeder blankets. The insulating coating is an effective approach to reduce the MHD effect. However, the insulating performance of coatings may be weakened by cracks caused by mechanical stresses, corrosion and so on. A three-dimensional code named MTC-H 1.0 was adopted to simulate the MHD flow in rectangular ducts with perfect and imperfect coatings. The results showed that perfect coatings could maintain the pressure drop to the level with fully electrically insulated wall when coating resistance was greater than 0.01, and longer crack distance ( L) would lead to higher MHD pressure drop in the duct with imperfect coatings.

Overview of the TBM R&D activities in Japan

In Japan, development of Water Cooled Ceramic Breeder (WCCB) blanket is being performed as the primary candidate of Test Blanket Module (TBM) toward DEMO. Also development of high temperature Li–Pb blanket, Li/V blanket and Flibe blanket are being performed toward the advanced DEMO blanket. Regarding the WCCB TBM development, real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS), F82H and successfully tested by heat flux equivalent to ITER TBM condition, 0.5 MW/m 2, 80 cycles with the same coolant condition as DEMO. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure are successfully fabricated. Furthermore, asemmbling test of the real scale FW plate mockup and SW plate mockup was performed to clarify the welding condition to form TBM box structure. All key technologies of blanket module fabrication has been addressed and almost achieved toward TBM. Regarding the development of liquid breeder blankets, all key technologies, such as, material compatibility and mass transfer, tritium recovery performance, insulation coating development, were covered, by using elementary experiments to convection loops of Li–Pb, Li and Flibe. High temperature heat exchange test up to 900 °C has been initiated by using Li–Pb test loop.

Recent activities on tritium technologies of BA DEMO-R&D program in JAEA

The R&D for tritium technologies to a demonstration reactor (DEMO) are planned to be carried out in the Broader Approach (BA) program in Japan by JAEA with Japanese universities: (1) tritium analysis technology; (2) basic tritium safety research; and (3) tritium durability test. The EU joins the discussions and assessment of the R&D results. As the recent activities on the tritium technologies of the BA program, a multi-purpose RI facility, where the above R&D subjects will be carried out, has been designed in detail. A preliminary safety study has also been carried out for the amount of tritium released to the environment and for the radiation dose of workers. The facility is now under construction at Rokkasho in Aomori. The main subjects of the R&D of tritium analysis are the technologies for real-time analysis for hydrogen isotopes, gas, liquid and solid (such as microGC and calorimeter). The materials of interest include F82H, SiC, ZrCo, solid and liquid advanced breeder and multipliers. In the tritium durability tests, organic materials and metals planned to be used in a DEMO plant are studied for the radiation and the corrosion damage. A series of preliminary studies for the above subjects has been started.

Development of a potential based code for MHD analysis of LLCB TBM

A two dimensional solver is developed for MHD flows with low magnetic Reynolds’ number based on the electrostatic potential formulation for the Lorentz forces and current densities which will be used to calculate the MHD pressure drop inside the channels of liquid breeder based Test Blanket Modules (TBMs). The flow geometry is assumed to be rectangular and perpendicular to the flow direction, with flow and electrostatic potential variations along the flow direction neglected. A structured, non-uniform, collocated grid is used in the calculation, where the velocity ( u), pressure ( p) and electrostatic potential ( ϕ) are calculated at the cell centers, whereas the current densities are calculated at the cell faces. Special relaxation techniques are employed in calculating the electrostatic potential for ensuring the divergence-free condition for current density. The code is benchmarked over a square channel for various Hartmann numbers up to 25,000 with and without insulation coatings by (i) comparing the pressure drop with the approximate analytical results found in literature and (ii) comparing the pressure drop with the ones obtained in our previous calculations based on the induction formulation for the electromagnetic part. Finally the code is used to determine the MHD pressure drop in case of LLCB TBM. The code gives similar results as obtained by us in our previous calculations based on the induction formulation. However, the convergence is much faster in case of potential based code.

Comparison of Two Types of Blanket Concepts for a Fast Ignition ICF Reactor FALCON-D

A Fast ignition Advanced Laser fusion reactor CONcept with a Dry first-wall and a high repetition laser (FALCON-D) has been proposed to investigate the potential of the fast ignitionin the reactor concepts. For the blanket system, two types of blanket concepts, i.e.asolid and a liquid metal breeder types using the reduced activation ferritic steel (F82H) were proposed. In this study, two types of blankets were designed, where the thickness of the blankets was minimized while keeping the net TBR larger than 1.07. One of the blanket concepts for FALCON-D is based on the solid breeder (Li2TiO3) with beryllium (Be) neutron multiplier and water cooling. The second blanket concept is based on liquid metal breeder (Li17Pb83) with water cooling. The maintenance method for FALCON-D is applicable to both blanket types. The net electric power of the solid breeder blanket is 110 MW larger than that of the liquid metal breeder blanket. This is mainly caused by the differences in the neutron energy multiplication. In the case of the liquid metal breeder blanket with water cooling, the net TBR 1.09 is achieved without Be as the neutron multiplier. Such design without Be can remove a risk of accident due to the chemical reaction between beryllium and water. From the economical point of view, the solid breeder blanket with water cooling, which generates a larger electric power, is desirable. On the other hand, if the combination of beryllium and water cooling was not acceptable from a viewpoint of safety, the blanket system with the liquid metal would be another possible option.

Validation of Modeling Tools to Describe the Corrosion/Precipitation Behavior of Eurofer Steel in Flowing Pb-17Li

RAFM steels (e.g. Eurofer) are considered as struc-tural material for blanket components of future fusion power plants. One of the envisaged blanket concepts to be tested in ITER foresees the application of a liquid breed-er, the eutectic lead alloy Pb-17Li. Various corrosion experiments have been made in the past, mostly conducted up to temperatures of ca. 480°C, with respect to deter-mine corrosion rates and mechanisms and comparison of the results with earlier tested RAFM-steels of type F82H-mod., Optifer and Manet. In the mean time the envisaged operational temperature increased to around 550°C and flow rates may also have changed. Thus extrapolations of the RAFM-steel corrosion behavior determined in the past to the higher working conditions may be problematic due to large uncertainties in reliability and, additionally, only low knowledge on transport of dissolved components in the Pb-17Li flow is present.Therefore, the development of modeling tools for de-scribing Pb-17Li corrosion was of absolute necessity. The modular structured code MATLIM is based on physical, chemical and thermo-hydraulic parameters and, in the first stage, the development was focused on the dissolution of Eurofer steel and validation with test results ob-tained at 480 and 550°C in the lead-lithium loop PICOLO of FZK.

Toward the ultimate goal of tritium self-sufficiency: Technical issues and requirements imposed on ARIES advanced power plants

Due to the lack of external tritium sources, all fusion power plants must demonstrate a closed tritium fuel cycle. The tritium breeding ratio (TBR) must exceed unity by a certain margin. The key question is: how large is this margin and how high should the calculated TBR be? The TBR requirement is design and breeder-dependent and evolves with time. At present, the ARIES requirement is 1.1 for the calculated overall TBR of LiPb systems. The Net TBR during plant operation could be around 1.01. The difference accounts for deficiencies in the design elements (nuclear data evaluation, neutronics code validation, and 3D modeling tools). Such a low Net TBR of 1.01 is potentially achievable in advanced designs employing advanced physics and technology. A dedicated R&D effort will reduce the difference between the calculated TBR and Net TBR. A generic breeding issue encountered in all fusion designs is whether any fusion design will over-breed or under-breed during plant operation. To achieve the required Net TBR with sufficient precision, an online control of tritium breeding is highly recommended for all fusion designs. This can easily be achieved for liquid breeders through online adjustment of Li enrichment.

Tritium permeation experiment at IFMIF Medium Flux Test Module

Tritium release experiments using different breeding material candidates are planned for the medium flux region of the IFMIF Test Cell. Nowadays, only ceramic breeder materials have been suggested to be tested in the Tritium Release Module located in the Medium Flux Test Module of IFMIF. Liquid breeder blankets are very promising and for that reason, several concepts will be tested in ITER. One of the main problems concerning the liquid blankets is the permeation of the generated tritium in the breeder throughout the walls. Since tritium permeation is highly influenced by irradiation conditions, IFMIF is a suitable scenario to perform tritium permeation related experiments. In this paper, a preliminary design of a tritium permeation experiment for the Medium Flux Test Module of IFMIF is proposed, in order to contribute to the progress of the liquid breeder blanket concept validation. The conceptual design of the capsule in which the experiment will be performed is carried out, taking into consideration the experiment necessities and its implementation in the Tritium Release Module. In addition to this, some thermal hydraulic calculations have been performed to evaluate the thermal behaviour of the irradiation capsule.

Measurements of time-dependent liquid-metal magnetohydrodynamic flows in a flat rectangular duct

In the helium-cooled lead lithium (HCLL) blanket, which has been chosen as a reference concept for a liquid-metal breeding blanket to be tested in ITER, the heat is removed by helium cooled plates aligned with the strong toroidal magnetic field that confines the fusion plasma. The liquid breeder lead lithium circulates through gaps of rectangular cross-section between the cooling plates to transport the generated tritium towards external extraction facilities. Under the action of the strong magnetic field, liquid metal flows in conducting rectangular ducts exhibit jet-like velocity profiles in the thin boundary layers near the side walls, which are parallel to the magnetic field like the cooling plates in HCLL blankets. The velocity in these side layers may exceed several times the mean velocity in the duct and it is known that these layers become unstable for sufficiently high Reynolds numbers. The present paper summarizes experimental results for such unstable time-dependent flows in strong magnetic fields, which have been obtained in the MEKKA liquid metal laboratory of the Forschungszentrum Karlsruhe. In particular, spatial and temporal scales of perturbation patterns are identified. The results suggest that the flow between cooling plates in a HCLL blanket is laminar and stable. The observed time-dependent flow behavior appears at larger velocities so that the present results are more relevant for applications in dual coolant concepts where high-velocity jets have been predicted along side walls.

Recent advances and issues in development of silicon carbide composites for fusion applications

Radiation-resistant advanced silicon carbide (SiC/SiC) composites have been developed as a promising candidate of the high-temperature operating advanced fusion reactor. With the completion of the ‘proof-of-principle’ phase in development of ‘nuclear-grade’ SiC/SiC composites, the R&D on SiC/SiC composites is shifting toward the more pragmatic phase, i.e., industrialization of component manufactures and data-basing. In this paper, recent advances and issues in (1) development of component fabrication technology including joining and functional coating, e.g., a tungsten overcoat as a plasma facing barrier, (2) recent updates in characterization of non-irradiated properties, e.g., strength anisotropy and chemical compatibility with solid lithium-based ceramics and lead-lithium liquid metal breeders, and (3) irradiation effects are specifically reviewed. Importantly high-temperature neutron irradiation effects on microstructural evolution, thermal and electrical conductivities and mechanical properties including the fiber/matrix interfacial strength are specified under various irradiation conditions, indicating seemingly very minor influence on the composite performance in the design temperature range.

Status and strategy of fusion materials development in China

The liquid metal and solid ceramic pebble breeder blankets have become the most promising blankets for ITER-TBMs or DEMO reactors in China and the world due to their potential advantages. In recent years the corresponding research work on fusion reactor materials mainly focuses on structural materials, plasma facing materials and the functional materials for the blanket such as breeder, coating and flow channel insert etc. for the successful application of fusion energy in the near future. The R&D on those materials in the two kinds of blankets is being carried out widely in China, including fabrication and manufacturing techniques, physical/mechanical properties assessment before and after irradiation, joining techniques for structural materials, compatibility evaluation, and the development and verification of the criteria for fusion material designs. The progress on main R&D activities of fusion reactor materials in China is introduced and prospected in the paper.

Preliminary experiments on the corrosion of CLAM steel in flowing eutectic LiPb

China low activation martensitic steel (CLAM), which is one of the RAFMs and under development in ASIPP, is considered as the primary candidate structural material for blankets in FDS series fusion reactors and will be in contact with the liquid breeder LiPb, so its corrosion behavior is of significance for its successful application. In the thermal convection LiPb loop DRAGON-I, the corrosion experiment of CLAM in flowing LiPb at 480 °C was performed up to about 2500 h to evaluate the corrosion behavior of CLAM in LiPb. The exposed samples were observed and analyzed, respectively, by metallography and scanning electron microscopy (SEM) with energy dispersive X-ray (EDX) spectroscopy after 500, 1000, 1500, 2000 and 2500 h corrosion experiments, respectively. The results show that the corrosive attack started with the dissolution of the passivated oxides formed during the heat treatment on the surface of CLAM steel. Then the corrosive attack continued further into the steel matrix. The weight loss was about 0.38 mg/cm 2 after 2500 h exposure.

Thermochemical degradation of limestone aggregate concrete on exposure to sodium fire

Limestone aggregate concrete blocks were subjected to sodium fire conforming to a realistic scenario in order to qualify them as protective sacrificial layers over structural concrete flooring in liquid metal-cooled fast breeder reactors. Mid infrared absorption measurements were carried out on these sodium fire-exposed samples as a function of depth from the affected surface. Definite signatures of thermochemical degradation indicating dehydration and structural modification of the limestone concrete have been obtained. Control runs were carried out to delineate the thermal effects of sodium fires from that of the chemical interaction effects. Measurements on limestone aggregate samples treated with fused NaOH provided direct evidence of the exact mechanism of the sodium attack on concrete. The observed degradation effects were correlated to the mechanical strength of the concrete blocks and to the intensity of the sodium fire experienced.

Application of Proton Conductors to Hydrogen Monitoring for Liquid Metal and Molten Salt Systems

The chemical control of impurity such as hydrogen and oxygen in coolants is one of the critical issues for the development of liquid metal cooled fast reactors and self-cooled liquid breeder blankets for fusion reactors. Especially, hydrogen (isotopes) level is the key parameter for corrosion and mechanical properties of the in-reactor components. For fission reactors, the monitor of hydrogen level in the melt is important for safety operation. The control of tritium is essential for the tritium breeding performance of the fusion reactors. Therefore, on-line hydrogen sensing is a key technology for these systems. In the present study, conceptual design for the on-line hydrogen sensor to be used in liquid sodium (Na), lead (Pb), lead-bismuth (Pb-Bi), lithium (Li), lead-lithium (Pb-17Li) and molten salt LiF-BeF2 (Flibe) was performed. The cell of hydrogen sensor is made of a solid electrolyte. The solid electrolyte proposed in this study is the CaZrO3-based ceramics, which is well-known as proton conducting ceramics. In this concept, the cell is immersed into the melt which is containing the hydrogen at the activity of PH1 of ambient atmosphere. Then, the cell is filled with Ar-H2 mixture gas at regulated hydrogen activity of PH2. The electromotive force (EMF) is obtained by the proton conduction in the electro chemical system expressed as Pt, Melt(PH1) | Proton conductor | PH2, Pt. The Nernst equation is used for the evaluation of the hydrogen activity from the obtained EMF. The evaluations of expected performance of the sensor in liquid Na, Pb, Pb-Bi, Pb-17Li, Li and Flibe were carried out by means of the measurement test in gas atmosphere at hydrogen activities equivalent to those for the melts in the reactor conditions. In the test, the hydrogen activity in the gas varied from 2.2x10-14 to 1. The sensor exhibited good response, stability and reproducibility.

Corrosion of Structural Materials in Liquid Lead Alloys for Nuclear Applications

Abstract The corrosion behavior of structural materials in liquid lead alloys like Pb-17Li as a liquid tritium breeder for future fusion reactors and Pb-55Bi (lead-bismuth eutectic [LBE]) as an env...

Activation Analysis for Sequential Reactions of a Fusion Demo-Reactor

Low activation material is one of the important factors for constructing high power fusion reactors in future. Unexpected activation, however, may be produced through sequential reactions due to charged particles created by primary neutron reactions. In the present work, the effect of the sequential activation reaction was studied for candidate low activation materials of a fusion demo-reactor. The calculations were conducted by the ACT4 code developed in JAEA for the activation analysis of fusion reactor designs and revised for dealing with the sequential activation reactions. The results say that the real dose rate around vanadium alloy becomes larger after the cooling for 3 years by considering the reaction. Although metal hydrate is regarded as an excellent low activation shield material, the reactions due to recoil protons are influential and the dose rate around vanadium hydrate is several orders of magnitude larger than the value calculated without the sequential process after 2 weeks cooling. In case of liquid breeders, the effect of sequential reactions is popularly observed and it affects the breeder reprocessing and the shield design of circulation loop.