The accelerator driven system strategy in Japan

Recent status of study on accelerator driven system (ADS) in Japan is presented. The double-strata fuel cycle concept has been proposed by Japan Atomic Energy Research Institute (JAERI) under the `OMEGA' program as a system for partitioning and transmutation (P&T) of long-lived radioactive nuclides. The ADS dedicated to transmutation is the key technology of the double-strata fuel cycle concept. The proposed system is a lead-bismuth cooled, nitride fuel 800 MW ADS for transmutation of minor actinides and long-lived fission products. JAERI has carried out a conceptual design study under the joint project between JAERI and KEK (High Energy Accelerator Research Organization) on high-intensity proton accelerators. In addition to a high-intensity proton accelerator complex, two experimental facilities are planned for development and demonstration of accelerator driven transmutation technology: The major objective of an ADS physics experimental facility is to obtain reactor physics data of hybrid subcritical system. The major objective of an ADS engineering experimental facility is to accumulate material data for the design of a lead-bismuth target system.     

Thermal conductivities of irradiated UO2 and (U, Gd)O2 pellets

Thermal diffusivities of UO₂ and (U, Gd)O₂ pellets irradiated in a commercial reactor (maximum burnups: 60 GWd/t for UO₂ and 50 GWd/t for (U, Gd)O₂) were measured up to about 2000 K by using a laser flash method. The thermal diffusivities of irradiated UO₂ and (U, Gd)O₂ pellets showed hysteresis phenomena: the thermal diffusivities of irradiated pellets began to recover above 750 K and almost completely recovered after annealing above 1400 K. The thermal diffusivities after recovery were close to those of simulated soluble fission products (FPs)-doped UO₂ and (U, Gd)O₂ pellets, which corresponded with the recovery behaviors of irradiation defects for UO₂ and (U, Gd)O₂ pellets. The thermal conductivities for irradiated UO₂ and (U, Gd)O₂ pellets were evaluated from measured thermal diffusivities, specific heat capacities of unirradiated UO₂ pellets and measured sample densities. The difference in relative thermal conductivities between irradiated UO₂ and (U, Gd)O₂ pellets tended to become insignificant with increasing burnups of samples.     

Preliminary studies on PbO reduction in liquid Pb-Bi eutectic by flowing hydrogen

In the framework of the Italian research program TRASCO for ADS, the device ORE has been designed and operated at the ENEA C.R. Brasimone in order to study the kinetics of lead oxide reduction by flowing hydrogen, diluted at 3% by volume in argon. From the measurement of water generated by the reaction between hydrogen and lead oxide, the rate of PbO reduction was determined and a first estimation of the activation energy was found. The experimental results so far achieved in terms of reaction rate, activation energy and kinetic constant, as well as the experimental set-up and the methodology, are here summarised and discussed.     

Liquid metal corrosion/erosion investigations of structure materials in lead cooled systems: Part 1

In future lead cooled nuclear power systems the heavy liquid-metal pump will be placed in the hot temperature region of the reactor. This combines corrosion problematic at elevated temperatures with erosion at the impeller of the pump. Several steels designed for conventional mechanical high loaded pumps and the SiSiC have been tested in oxygen containing stagnant lead (10⁻⁶ and 10⁻⁸ wt%) at 550 and 600 °C for 2000 and 4000 h in the COSTA-facilities. Only SiSiC shows no influence by the liquid metal. No dissolution attack occurs at cast iron steels but inner oxidation takes place. NORILOY shows no dissolution attack. All other steels are attacked by liquid lead at one of the conditions. To evaluate the erosion-corrosion attack a new test facility allowing velocities of the lead of up to 20 m/s was designed and constructed. With a CFD-code the behaviour and flow velocity of the lead are simulated.     

Pyrochemical reduction of uranium dioxide and plutonium dioxide by lithium metal

The lithium reduction process has been developed to apply a pyrochemical recycle process for oxide fuels. This process uses lithium metal as a reductant to convert oxides of actinide elements to metal. Lithium oxide generated in the reduction would be dissolved in a molten lithium chloride bath to enhance reduction. In this work, the solubility of Li₂O in LiCl was measured to be 8.8 wt% at 650 °C. Uranium dioxide was reduced by Li with no intermediate products and formed porous metal. Plutonium dioxide including 3% of americium dioxide was also reduced and formed molten metal. Reduction of PuO₂ to metal also occurred even when the concentration of lithium oxide was just under saturation. This result indicates that the reduction proceeds more easily than the prediction based on the Gibbs free energy of formation. Americium dioxide was also reduced at 1.8 wt% lithium oxide, but was hardly reduced at 8.8 wt%.     

Development of an oxygen sensor for molter 44.5% lead-55.5% bismuth alloy

A potentiometric sensor for measuring oxygen activity in molter 44.5% lead-55.5% bismuth alloy is under development. Three parts form this sensor: an In/In₂O₃ reference electrode, a ZrO₂/Y₂O₃ solid electrolyte, and a molybdenum working electrode. The oxygen partial pressure in the melt is calculated applying the Nernst equation to the potential difference measured by the sensor. The minimum oxygen partial pressure detected by this sensor is 10⁻⁴⁰ bar. The sensor has been calibrated with the following metal/metal oxide mixtures: Pb-Bi/PbO, Pb/PbO, Sn/SnO₂, and Bi/Bi₂O₃. Reproducible measurements of the oxygen pressure of PbO formation have been obtained. The sensor performance has been evaluated in 99.999% N₂ and air (20% O₂) environments.     

Effects of nitrogen on low-cycle fatigue properties of type 304L austenitic stainless steels tested with and without tensile strain hold

A quantitative analysis of the effects of nitrogen on high temperature low-cycle fatigue without and with tensile strain hold at 600 °C has been conducted for type 304L stainless steels. For better understanding of the role of nitrogen on grain boundary precipitation, the grain boundary segregation of nitrogen was analyzed by Auger electron spectroscopy. The nitrogen addition is found to give relatively better resistance to creep-fatigue than continuous low-cycle fatigue. This in turn improves the fatigue life. This is due to the retardation of the precipitation of carbides at the grain boundary and reduction in the density of grain boundary cavitation sites which are the main factor of grain boundary damage under creep-fatigue test.     

Calculation of thermodynamic parameters of U–Pu–N system with carbon and oxygen impurities

Mixed nitride fuels are being considered for advanced FBR, but very little is known about the thermodynamic properties of these fuels. For an overall composition of the nitride fuel with small amounts of oxygen and carbon impurities, thermodynamic properties, e.g. carbon activity and partial pressures of nitrogen, carbon-monoxide, plutonium and uranium, were calculated in present work. These calculations were based on standard Gibbs free energies of the binary compounds, present in this multi-component system (U,Pu)–C–N–O. For an over all composition of the fuel, stable phase-field was determined by minimization of the Gibbs free energy of the system. The fabrication experiences of various workers, reported in literature, have shown that depending on the impurity content, nitride fuel can exist in two phase fields, mono-nitride phase in equilibrium with sesquinitride phase or mono-nitride phase in equilibrium with dioxide phase. Therefore, in present calculations special attention was given to the thermodynamic behavior of these two phase-fields. A comparison of calculated thermodynamic properties indicated that nitride fuel with dioxide as second phase will be superior to the one with sesquinitride.     

Analysis of the monoclinic-tetragonal phase transition of zirconia under irradiation

Zirconia produced by the oxidation of zirconium alloys in nuclear reactors exhibits a phase transition under ionic irradiation, simulating a neutron irradiation. To understand the mechanism responsible for this irradiation driven phase transition, different kinds of projectiles were used to irradiate pure monoclinic zirconia samples. The evolution of these irradiated samples as a function of dpa has been studied using grazing X-ray diffraction. The Rietveld method has been applied on collected X-ray diffraction diagrams to study the phase produced under irradiation and the kinetics of its formation. Even at high dpa values, only the monoclinic and tetragonal phases were used to simulate X-ray diffraction diagrams. No amorphisation of zirconia was observed. The evolution of unit cells and short range strains in both phases under irradiation leads us to think that the irradiation driven transition is martensitic. Supposing that the inelastic stopping power in sub-cascades is responsible for the irradiation driven phase transition, we propose a model based on the Landau-Ginzburg effective hamiltonian to explain both the m→t transition observed under irradiation and the t→m transition measured during isochronal annealing after irradiation.     

The effects of low dose rate irradiation and thermal aging on reactor structural alloys

As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10⁻⁸ dpa/s) irradiation at 380–410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.     

The experimental accelerator driven system (XADS) designs in the EURATOM 5th framework programme

The Accelerator-Driven System has the potential to safely fission Minor Actinides and transmute, on industrial scale, selected long-lived fission products into isotopes of lesser concern. European leading nuclear Industrial Companies and Research Centers,within the EURATOM 5th Framework Programme, have studied three solutions of eXperimental Accelerator Driven Systems (XADS), different for the power (80 and 50 MW) or the primary coolant (Lead Bismuth Eutectic and Gas). Two main concepts of Target Unit are envisaged: the Window Target Unit which features a thin metallic sheet as a barrier between the LBE target and the Proton Beam Pipe and the Windowless Target Unit in which the proton beam impinges directly on the free surface of the liquid LBE target. These designs are under study in order to assess and compare them on a common basis, and to outline the main R&D needs.     

The MEGAPIE 1 MW target in support to ADS development: status of R&D and design

The MEGAPIE project is aimed at designing, building and operating a liquid metal spallation neutron target as a key experiment on the road to an experimental accelerator driven system and to improve the neutron flux at the PSI spallation source. The design of the target system has been completed. The target configuration and the operating conditions have been defined and the expected performance assessed. A preliminary safety analysis has been performed considering normal, off-normal and accident conditions and a corresponding report has been submitted to the authorities for licensing. The experience gained up to now shows that MEGAPIE may well be the first liquid metal target to be irradiated under high power beam conditions.     

The development of radiation embrittlement models for US power reactor pressure vessel steels

A new approach of utilizing information fusion technique is developed to predict the radiation embrittlement of reactor pressure vessel steels. The Charpy transition temperature shift data contained in the Power Reactor Embrittlement Database is used in this study. Six parameters-Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature - are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.     

Liquid metal embrittlement of the martensitic steel 91: influence of the chemical composition of the liquid metal.: Experiments and electronic structure calculations

In previous works [Scripta Mater. 43 (2000) 997; J. Nucl. Mater. 296 (2001) 256], we showed that the martensitic steel 91 is prone to liquid metal embrittlement (LME) by liquid lead provided that some metallurgical conditions are fulfilled. In this work, we report results of LME of the steel 91 in contact with Pb-Bi and other low melting temperature metals such as Sn and Hg. Our experimental results can be interpreted within the framework of the surface energy reduction models for LME. To account for the experimental observations, we performed electronic structure calculations to assess the chemical interaction between low melting temperature metal atoms and iron surfaces. Our results allow to establish a simple criterion that can give trends on the embrittlement power of a liquid metal in contact with iron.     

Corrosion behavior of Al-surface-treated steels in liquid Pb-Bi in a pot

Corrosion tests were performed in oxygen-saturated liquid Pb-Bi at 450 °C and 550 °C in a pot for 3000 h for Al-surface-treated steels containing various levels of Cr contents. The Al surface treatments were achieved using a gas diffusion method and a melt dipping method. Al₂O₃, FeAl₂ and AlCr₂ produced by the gas diffusion method exhibited corrosion resistance to liquid Pb-Bi, while the surface layer produced by the melt dipping method suffered a severe corrosion attack. Fe₄Al₁₃ and Fe₂Al₅ produced by the melt dipping method disappeared during the corrosion test at 550 °C and only FeAl remained.     

Radiation effects in structural materials of spallation targets

Effects of radiation damage by protons and neutrons in structural materials of spallation neutron sources are reviewed. Effects of atomic displacements, defect mobility and transmutation products, especially hydrogen and helium, on physical and mechanical properties are discussed. The most promising candidate materials (austenitic stainless steels, ferritic/martensitic steels and refractory alloys) are compared, and needed investigations are identified.     

Corrosion behaviour of aluminized martensitic and austenitic steels in liquid Pb-Bi

The Pb-Bi liquid alloy is under consideration as a spallation target material in the hybrid systems due to its suitable nuclear and physical properties. In order to limit the risks of corrosion of the structural elements in contact with the liquid Pb-Bi, protection by means of aluminized coatings was investigated for 316L austenitic steel and T91 martensitic steel. For both steels, no damages were observed after immersions in static Pb-Bi up to 500 °C for low oxygen concentrations and long durations. However, at 600 °C in the same conditions, a non-uniform degradation of the coatings was observed. Only coated 316L was tested in dynamic conditions. The results were generally satisfying for temperatures from 350 to 600 °C and for fluid velocities up to 2.3 m s⁻¹. However, in both the IPPE loops and the CICLAD device, some localized damage of the coatings, attributed to erosion, was observed.     

Kinetics and mechanisms of the reaction of air with nuclear grade graphites: IG-110

The work presented in this report is part of an ongoing effort in the microgravimetric evaluation of the intrinsic reaction parameters for air reactions with graphite over the temperature range of 450 to 750°C. Earlier work in this laboratory addressed the oxidation/etching of H-451 graphite by oxygen and steam. This report addresses the air oxidation of the Japanese formulated material, IG-110. Fractal analysis showed that each cylinder was remarkably smooth, with an average value, D, the fractal dimension of 0.895. The activation energy, E_a, was determined to be 187.89 kJ/mol indicative of reactions occurring in the zone II kinetic regime and as a result of the porous nature of the cylinders. IG-110 is a microporous solid. The low initial reaction rate of 9.8×10⁻⁵ at 0% burn-off and the high value (764.9) of Φ, the structural parameter confirm this. The maximum rate, 1.35×10⁻³ g/m²s, was measured at 34% burn-off. Reactions appeared to proceed in three stages and transition between them was smooth over the temperature range investigated. Both E_a and ln A did not vary with burn-off. The value of ΔS, the entropy of activation, was −41.4 eu, suggesting oxygen adsorption through an immobile transition state complex. Additional work is recommended to validate the predictions that will be made in relation to accident scenarios for reactors such as the modular high temperature gas-cooled reactor where fine grained graphites such as IG-110 could be used in structural applications.