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Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

Thermal cycling and melt reaction studies of ceramic coatings plasma-sprayed on Nb substrates were carried out to evaluate the performance of barrier coatings for metallic fuel casting applications. Thermal cycling tests of the ceramic plasma-sprayed coatings to 1450 degrees C showed that HfN, TiC, ZrC, and Y2O3 coating had good cycling characteristics with few interconnected cracks even after 20 cycles. Interaction studies by 1550 degrees C melt dipping tests of the plasma-sprayed coatings also indicated that HfN and Y2O3 do not form significant reaction layer between U–20 wt.% Zr melt and the coating layer. Plasma-sprayed Y2O3 coating exhibited the most promising characteristics among HfN, TiC, ZrC, and Y2O3 coating.

The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxide and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.

As the fuel-cladding chemical interaction (FCCI) between metallic fuel and cladding material can occur at a relatively low temperature, electroplating of Cr as an FCCI barrier on the inner wall of the cladding was proposed. U–10 wt.%Zr and HT9 disks were used as fuel and cladding material to conduct diffusion couple annealing tests at 740 and 800 °C. Two kinds of Cr layers dependent on the temperature of plating bath were electroplated on the surface of the HT9 disks. The Cr layer of 20 μm in thickness, after being electroplated at a bath temperature of 50 °C and a current density of 35 A/dm2 for 1 h, had some cracks and a small-grained structure. However, the Cr layer plated at a bath temperature of 80 °C and a current density of 56 A/dm2 for 2 h showed a dense structure. Both Cr layers inhibited the interaction between U–10 wt.%Zr and HT9 at a relatively high annealing temperature. U penetrated into the Cr layer plated at a bath temperature of 50 °C due to its penetrable structure. The Cr layer plated at a bath temperature of 80 °C showed a better performance than the Cr layer plated at a bath temperature of 50 °C as a diffusion barrier material with less U penetration.

The diffusion behavior in an interface between a U–10 wt.%Zr alloy and HT-9 steel at 700 °C and 730 °C was investigated. The Zr-rich layer near the interface between two alloys was formed mainly due to the decomposition of the δ-UZr 2 phase in U–Zr alloy, and acted to interrupt the interdiffusion of the alloying elements. Considerable interdiffusion was observed only when the Zr-rich layer was destroyed locally. As a result, the diffusion–reaction layer was mainly composed of a U-rich, (U,Zr)(Fe,Cr) 2 compound, Zr-rich and Zr-depleted layers. Fe elements which had diffused from HT-9 were mainly observed in the boundaries of the Zr-rich particles in the U–Zr alloy.

Diffusion couple tests of U–Zr or U–Zr–Ce alloys vs. ferritic martensitic steels such as HT9 or T91 were carried out in order to evaluate the performance of the diffusion barrier candidates. Elemental metal foils of Zr, Nb, Ti, Mo, Ta, V and Cr were very effective in inhibiting interdiffusion between these fuels and steels. Eutectic melting between the fuels and steels was not observed in any of the diffusion couples using these diffusion barrier foils at annealing temperatures up to 800 °C. Among the metallic foils evaluated in this study, V and Cr exhibited the most promising performances as a diffusion barrier material for eliminating the fuel cladding chemical interaction problem. However, Zr, Nb and Ti showed an active interaction with the fuel mainly due to the large U solubility.

In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.

The effects of the fabrication process parameters such as a tempering temperature, cold rolling and annealing condition on the precipitates and mechanical properties of a normalized 9Cr–2W–V–Nb steel were evaluated. Nb-rich MX precipitates were found in the specimen tempered at 550 °C while M 23 C 6 , Nb- and V-rich MX ones were observed in the specimen tempered at 750 °C. A cold rolling and an annealing at 750 °C of the specimen tempered at 550 °C induced the formation of large inhomogeneous M 23 C 6 carbides, causing a reduced tensile strength. However, the cold rolling of the specimen tempered at 750 °C provided fine precipitates due to a fragmentation of some of the M 23 C 6 carbides, and an annealing at 700 °C for 30 min was found to be suitable to recover the degraded mechanical properties from a cold working.

Thermal conductivity of RE-bearing U–Zr fuels was investigated. Under an assumption of a RE-rich phase forming a macroscopic mixture with the matrix, the thermal conductivity was estimated for U–Zr–RE alloy. It was evaluated that the thermal conductivity of the U–Zr fuels would be lowered by less than 5% due to the addition of RE. The measurement of the thermal properties of U–Zr–Ce supported the present estimation.

A two-step two-stage model is developed in this study on the basis of the recent theoretical model. This model incorporates a two-step burn-up factor in the two-stage diffusion processes in the grain lattice and at the grain boundary during the fission gas release. In-pile data sets available in FRAPCON-3 code are used to validate the model. Results show that the predictions are in better agreement with the experimental measurements than those of any other models built in the code over the entire burn-up range up to 75 000 MWd/MTU.

This paper provides an introduction to and a summary of the remaining papers in this issue of Nuclear Technology. The papers in this issue present the important results from a U.S. Department of Energy-sponsored Nuclear Engineering Research Initiative (NERI) project to study the efficacy of the thorium-uranium dioxide (ThO2-UO2) once-through fuel cycle in current light water reactors. The project addressed fuel cycle neutronics and economics; ThO2-UO2 fuel manufacturing; the in-pile thermal/mechanical behavior of ThO2-UO2 fuel during normal, off-normal, and accident conditions; and the long-term stability of ThO2-UO2 waste. Results from this work show that a small-scale separation of the uranium and thorium will enhance the fuel reactivity and achievable burnup from uranium-thorium dioxide fuels. Under conditions that meet the thermal requirements in present pressurized water reactors (PWRs), a properly designed microheterogeneous fuel will have more reactivity than all-uranium fuel, and the overall production of plutonium is significantly reduced. The use of thorium as a host for actinide fuels when PWRs are used for actinide transmutation was also explored. It was also determined that there were no fundamental obstacles to converting the current plants that manufacture uranium oxide-only fuel to a mixed ThO2-UO2 fuel. Also, the in-service and transient thermal andmore » mechanical performance of homogeneous ThO2-UO2-based fuels with respect to safety is generally equal to or better than that of all-uranium fuel. Furthermore, a mixed thorium-uranium dioxide spent fuel appears to be a much more stable waste form than uranium oxide spent fuel.« less

Techniques to fabricate thorium-uranium dioxide fuel [(Th,U)O2] have been developed, and the thermal conductivity of (Th,U)O2 pellets has been measured. Mixtures of thorium dioxide (ThO2) and urani...

Co-based hardfacing alloys exposed to PWR primary coolant may be replaced with Co-free alloys to lower occupational radiation exposure. To evaluate the viability of Co-free hardfacing alloys, we conducted hot-water tests for gate valves hardfaced with NOREM™ 02 (Fe-base), Deloro™ 50 (Ni-base), and Stellite™ 6 (Co-base). Using a high flow test loop, on–off cycling tests were conducted in 280 °C water. It was observed that NOREM 02 exhibited galling and excessive leak after 1000 cycle test whereas no leakage was developed with Deloro 50 after 2000 cycles. To estimate the activity reduction effect of Co-free hardfacing alloys, an existing activity transport model was modified. It is found that the main contributor of Co activity buildup is the corrosion of steam generator (SG) tubing. The Korean Next Generation Reactor (APR-1400) tubed with alloy 690 having a reduced cobalt impurity allowance is expected to have 73% lower Co activity on SG surface compared with the case of alloy 600 tubing. The complete replacement of Stellite 6 with Co-free hardfacing alloys is expected to cut additional 5% of activity which may be too small to justify the risk of galling and leakage development as revealed by the hot-water test.

A mechanistic model of fission gas release and swelling for the U–Pu–10Zr metallic fuel in a fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix), was developed. Fission gas bubbles are assumed to nucleate isotropically from gas atoms in the metallic fuel matrix since they could nucleate at both the grain boundaries and the phase boundaries, which are randomly distributed inside the grain. Bubbles can grow to larger sizes by gas atom diffusion and coalescence with other bubbles. When bubble density or swelling due to bubbles reaches a threshold value, bubbles are assumed to be interconnected to each other to form an open channel to the external free space and then the fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are successively released through the open bubbles. The GRSIS model was validated by comparison with the irradiation test results of U–Pu–10Zr fuels as well as parametric studies of the key variables in the model.

The RAPID ( RA dial power and burnup P rediction by following fissile I sotope D istribution in the pellet) model was developed to predict the radial distribution of power, burnup and fissionable nuclide densities in the LWR UO 2 fuel pellets to be used in the fuel performance analysis code. It considers the specific radial variations of the neutron reactions of all the fissionable nuclides as functions of burnup and 235 U enrichment in the pellet. Comparison of the RAPID prediction with the measured data of the irradiated fuels and the predicted results by other codes showed good agreement, and therefore, the RAPID model can be used for UO 2 fuel of up to 10 w/o 235 U enrichment and 150 MWD/(kg U) pellet average burnup under the LWR environment.

A formation mechanism of high burnup structure (HBS) or rim effect in UO2 fuel was proposed. The fission gas bubbles in the HBS region may be nucleated and stabilized at the threshold atomic concentration of gas atoms inside the grain of UO2 fuel. HBS formation in HBS region may be driven not by the conventional thermal diffusion process, which should increase with the temperature, but by the cascade of atomic movements by the fission fragments. Based upon the above mechanism, the threshold local burnup for HBS initiation was derived as a function of temperature, fission rate and grain size. The threshold burnup for HBS initiation was predicted to increase with temperature, and at high temperature, HBS would never form due to the fission gas release. Variations of HBS widths among the different irradiation tests could be explained by the proposed mechanism.

The effects of irradiation on fracture toughness of modified 9Cr–1Mo steel in the transition region were investigated. Half size precracked Charpy specimens were irradiated up to 1.2 × 10 21 n/cm 2 ( E > 0.1 MeV) at 340 °C and 400 °C in the Korean research reactor. The irradiation induced transition temperature shift for a modified 9Cr–1Mo was evaluated by using the Master Curve methodology. The T 0 temperature for the unirradiated specimens were measured as −67.7 °C and −72.4 °C from the tests with standard PCVN (precracked charpy V-notch) and half sized PCVN specimens, respectively. The T 0 shifts of specimens after irradiation at 340 °C and 400 °C were 70.7 °C and 66.1 °C, respectively. The Weibull slopes for the fracture toughness data obtained from the unirradiated and irradiated modified 9Cr–1Mo steels were determined to confirm the applicability of master curve methodology to modified 9Cr–1Mo steel.