Takeshi Yokoo

Development and validation of ALFUS: An irradiation behavior analysis code for metallic fast reactor fuels

An irradiation behavior analysis code for metallic fast reactor fuel, ALFUS, has been revised so that it can be applied to stress-strain analysis of U-Pu-Zr ternary fuel pins. The stress-strain cal...

Reactions of Uranium-Plutonium Alloys with Iron

In metallic U-Pu-Zr fuel for fast reactors, metallurgical reactions occur between the fuel alloy and the stainless steel cladding, and a liquid phase may be formed in the reaction zone at a higher temperature. In order to clarify the condition for liquefaction at the fuel-cladding interface, the reactions of U-Pu alloys with Fe have been examined at 923 and 943 K. The test results confirmed that the liquid phase is not formed at 923 K in any region of the reaction zone when the maximum Pu content in the (U,Pu)6Fe phase is less than the Pu solubility limit in this phase. Comparison of the present test results with the liquefaction data from the various tests on metallic fuel-cladding compatibility suggested that the liquefaction condition is independent of the Zr content in the fuel alloy and can be expressed as a function of the atom fraction ratio of Pu/(U+Pu) in the fuel alloy and the reaction temperature. At 923 K, liquefaction will occur when the Pu/(U+Pu) ratio is larger than 0.25.

Reactions between U-Pu-Zr Alloys and Fe at 923 K

In metallic U-Pu-Zr fuel, metallurgical reactions occur between the fuel slug and the cladding, and a phase of which the melting point is relatively low is formed in the reaction zone. If a liquid phase is formed, it can degrade cladding integrity. The potential for liquid phase formation near the cladding, therefore, should be excluded during normal reactor operation. In order to clarify the mechanism of liquefaction, the authors have conducted diffusion experiments at 923 K using two couples: U-13 at%Pu-22 at%Zr/Fe and U-22at%Pu- 22at%Zr/Fe, and examined the influence of the Pu content in the fuel alloy on the phases formed in the reaction zones. The liquid phase has been observed in the U-22 at%Pu-22 at%Zr/Fe couple. An assessment of the diffusion paths for these couples has indicated that the Pu content in the (U, Pu)6Fe-type phase in the reaction zone is a crucial factor in determining the conditions that lead to liquefaction. The Pu content in the (U, Pu)6Fe-type phase increases with that in the initial U-Pu-Zr alloy. The fuel design specifications to exclude the potential for liquefaction will be clarified by further examination of the Pu content in the (U, Pu) 6Fe-type phase for various U-Pu-Zr/Fe couples annealed at different temperatures.

Phase relations in the quaternary Fe–Pu–U–Zr system

Abstract The phase diagram in the quaternary Fe–Pu–U–Zr system was established at 923 K in the uranium-rich region to understand better the compatibility between the metal fuel and stainless steel cladding in a fast reactor. The experimental phase relation data obtained in this study was applied to the thermodynamic methodology for construction of the phase diagram. The calculated phase diagram was consistent and well within the experimental data. The applicability of the thermodynamic model to other temperatures was confirmed by comparing the present results of differential thermal analysis with the calculated phase diagrams. These consistencies mean that both the thermodynamic model and the assessed parameters in the binary and ternary subsystems developed so far are reasonable. The calculated phase diagram established in this study was also in good agreement with the analysis of the diffusion zone in tests on Pu–U–Zr/Fe couples. This suggests that the diffusion zone formed at the fuel–cladding interface in the reactor system can be assessed using the phase diagram in the quaternary Fe–Pu–U–Zr system.

Low-Burnup Irradiation Behavior of Fast Reactor Metal Fuels Containing Minor Actinides

Abstract Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and/or rare earths (REs) have been irradiated in the fast reactor PHÉNIX to examine the effects of adding those elements on metal fuel irradiation behavior. In this experiment, two MA-containing metal fuel pins, in which the test alloys U-19Pu-10Zr-2MA-2RE and U-19Pu-10Zr-5MA/U-19Pu-10Zr-5MA-5RE (wt%) were loaded into part of a standard U-19Pu-10Zr alloy fuel stack, and a reference fuel pin of U-19Pu-10Zr alloy without MAs or REs was set in an irradiation capsule. Two other capsules with this same configuration are also irradiated. Postirradiation examinations are conducted at ~2.5, ~7, and ~11 at.% burnup. For the low-burnup fuel pins, nondestructive tests after irradiation have been performed, and the integrity of the pins was confirmed. The irradiation behavior of MA-containing metal fuels up to 2.5 at.% burnup was analyzed using the ALFUS code. The calculation results, such as the axial swelling distribution of a fuel slug or the extrusion behavior of bond sodium to the gas plenum, are consistent with the measurement data regardless of the addition of MAs and REs to the U-Pu-Zr alloy fuels. This observation result indicates that the macroscopic irradiation behavior of U-Pu-Zr fuels containing MAs and REs of 5 wt% or less is similar to that of U-Pu-Zr fuels up to ~2.5 at.% burnup.

Parametric survey for benefit of partitioning and transmutation technology in terms of high-level radioactive waste disposal

Benefit of implementing Partitioning and Transmutation (P&T) technology was parametrically surveyed in terms of high-level radioactive waste (HLW) disposal by discussing possible reduction of the geological repository area. First, the amount and characteristics of HLWs caused from UO2 and MOX spent fuels of light-water reactors (LWR) were evaluated for various reprocessing schemes and cooling periods. The emplacement area in the repository site required for the disposal of these HLWs was then estimated with considering the temperature constrain in the repository. The results showed that, by recycling minor actinides (MA), the emplacement area could be reduced by 17–29% in the case of UO2-LWR and by 63–85% in the case of MOX-LWR in comparison with the conventional PUREX reprocessing. This significant impact in MOX fuel was caused by the recycle of 241Am which was a long-term heat source. Further 70–80% reduction of the emplacement area in comparison with the MA-recovery case could be expected by partitioning the fission products (FP) into several groups for both fuel types. To achieve this benefit of P&T, however, it is necessary to confirm the engineering feasibility of these unconventional disposal concepts.

Analytical study on deformation and fission gas behavior of metallic fast reactor fuel

Abstract In order to analytically investigate irradiation behavior of metallic fast reactor fuels, the authors have developed the ALFUS (ALoyed Fuel Unified Simulator) code. The ALFUS can mechanistically simulate gas release and deformation behavior of the uranium-zirconium alloy fuel. The stress-strain analysis model into which anisotropic strain due to cavitation at the grain and/or phase boundary in the a-uranium phase has been introduced simulates anisotropic deformation of the uranium-zirconium alloy fuel. The models included in the ALFUS are thought reasonable and consistent with knowledge obtained from the irradiation test results. When the fuel slug swells out and comes into contact with the cladding, compressive stress is produced in the slug and decreases volume of the open pore if it has been formed. This is an essential process to keep fuel-cladding mechanical interaction (FCMI) small in case of lower smear density fuel. Analyses with the ALFUS indicate a significant level of FCMI in case of higher than 80% smear density fuel.

A Design Study on the FBR Metal Fuel and Core for Commercial Applications

A design methodology for the FBR metal fuel and core is developed. It consists of the fuel integrity prediction by a mechanistic analysis code, and the core neutronic/thermal-hydraulic design in which the effect of the characteristic fuel behavior and fuel cycle technologies are properly considered. Based upon this methodology, a fuel and core design study for reactors of various output scale (150–1500 MWe) is conducted, and the feasibility of metal fuel FBRs for the future commercial applications is reviewed. The results show that the averaged burnup of as high as 90–150 MWd/t is achievable at the refueling interval of 1–2 years with 3–4 batches, regardless of the output scale. The maximum allowable cladding temperature is assumed to be 650°C to avoid the liquid phase formation, which leads to the achievable core outlet temperature of ∼510°C based on the current hotspot factors estimation. It is found that high breeding ratio of ≥1.2 can be enabled with relatively small blanket amount, owing to the very high internal conversion capability. Another advantage is that it is possible to significantly reduce the burnup reactivity loss at the slight expense of the burnup, in which case the core excess reactivity becomes as small as a few dollars.

Core performance of fast reactors for actinide recycling using metal, nitride, and oxide fuels

Core performance analyses are conducted for fast reactors that accept and recycle the plutonium and minor actinides (MAs) recovered from light water reactor (LWR) spent fuel, together with the plutonium and MAs from the fast reactors` own production. Metal, nitride, and oxide are the fuel materials used to compare the neutronic and safety parameters and to discuss acceptable minor actinide content. Based on the material balance of the analyzed cores, an LWR-fast reactor fuel cycle model is used to calculate the mass flow of the plutonium and MAs and to estimate their total amount in the waste stream.

Phase relations in the Fe-Pu-U ternary system

An isothermal section of the Fe-Pu-U ternary system at 650 °C was assessed in a previous study. In the present study, the predictions of the phase relations in the Fe-Pu-U system to higher and lower temperatures were performed by applying the interaction parameters determined at 650 °C. Differential thermal analysis (DTA) for the Fe-Pu-U alloys was also carried out to confirm the phase relations in the temperature region of 500 to 800 °C. Both results agreed well. On the basis of the predicted ternary phase diagram, the phase relations for a region surrounded by Fe2Pu, Fe2U, U, and Pu were described by a reaction scheme and a projection of the liquidus surface.