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Metallic fuels, such as the U-Zr alloy system for sodium-cooled fast reactor (SFR), are melted in a Y2O3 slurry-spray-coated graphite crucible in order to prevent melt/material interactions. Reactive and porous coatings such as the Y2O3 slurry-spray-coating are a source of melt contamination and fuel loss, respectively. Therefore, a dense plasma-spray coating of non-reactive materials is desirable for a reusable melting crucible. In this study, ceramic materials such as Y2O3 and 8 mol% Y2O3-st. ZrO2, are selected as candidate protective coating materials for reusable crucibles for melting metallic fuel slugs. Studies are conducted to characterize the thermal cycling performance of the coatings, and the interactions between the U-10wt% Zr fuel and the coating layer on the substrates. The thermal cycling tests of the ceramic plasma-spray coatings exhibit good thermal cycling characteristics with few interconnected cracks. The Y2O3 plasma-spray coating does not form a significant reaction layer between the melt and the coating layer due to the thermodynamic stability of Y2O3. Hence, the Y2O3 plasma-spray coating exhibits the most promising performance among the ceramic coatings investigated, for use in reusable melting crucibles.

Microstructural development of irradiated U-10Zr fuel slug with T92 cladding specimen was examined after thermal exposure of 750 °C for 1 h. Optical microscopy, scanning electron microcopy and electron microprobe analysis were employed to examine the microstructure, constituent migration, and eutectic penetration. Migration phenomena of U, Zr, Fe, and Cr indicative of Soret effect was observed, and Nd lanthanide fission product was found at the eutectic melting region. Eutectic penetration was quantified and correlated to a thermal activation model with a good agreement. Compared to the previously reported eutectic penetration rates for the unirradiated U-10Zr fuel slug with FMS (Ferritic Martensitic Steel, HT9) cladding specimens, the eutectic penetration rate determined from this study for the irradiated fuel specimen was higher. This phenomenon can be caused by the effect of lanthanide fission product migration into fuel slug-cladding interface during irradiation, and lowering the eutectic threshold temperature for the irradiated fuel.

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.

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU)–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

As one of the candidate coating materials for a melting crucible, yttrium oxides were deposited on graphite and niobium substrates using slurry and plasma spraying methods. Thermal cycling tests and interaction studies between U–Zr/U–Zr–RE fuel melt and the Y2O3 coatings were carried out to evaluate the performance as reusable coatings for a melting crucible. A multi-layer coating method was also applied to overcome the issue of a thermal expansion mismatch between the coating and substrate. The results showed that the plasma-sprayed coatings showed a good consolidation after deposition compared to slurry coating. The plasma-sprayed Y2O3 coating on the niobium substrate showed better thermal cycling resistance than those coated on a graphite substrate. The proposed TaC/Y2O3 double-layer coating which was plasma-sprayed on the niobium substrate showed improved characteristics with no reaction layer formation and no separation from the substrate after the interaction with the U–Zr–RE melt. Copyright © 2014 John Wiley & Sons, Ltd.

Separation of rare-earth (RE) elements from minor actinides is difficult in pyroprocessing, especially for light REs (e.g., Nd, Ce, Pr, and La). Herein, U–10 wt%Zr fuel slugs containing 0, 3, and 7 wt%RE (RE: a rare-earth alloy comprising 53 % Nd, 25 % Ce, 16 % Pr, and 6 % La by weight) were prepared by gravity casting. Their casting soundness was characterized by gamma-ray radiography and immersion density measurement, while their microstructures and compositions were analyzed using scanning electron microscopy and energy-dispersive X-ray spectroscopy. Their actual compositions and impurities were confirmed using inductively coupled plasma atomic emission spectrometry. The compatibility of casting components and recovery of radioactive materials were evaluated based on the RE content.

Studies were carried out to evaluate the effect of lanthanide elements which are generated during the fission process on the interaction between metal fuel and cladding material. Quartenary lanthanide alloy as Nd-Ce-Pr-La and Sm-contained quintenary alloy were manufactured by vacuum arc remelting and a diffusion couple test with HT9 (12Cr1MoWV) alloy was carried out at 660 and 700 °C for up to 231 h followed by microstructural analysis. The results showed that interdiffusion between lanthanide element (Nd, Ce) and Fe took place as the diffusion couple test proceeds to form intermetallic compounds as an RE2(Fe,Cr)17 (RE=Nd, Ce, Pr, La) type. As the number of alloying elements increased, the reaction layer thickness increased, but a deviation of the theoretical non-linear relationship as well as an increase in the rate exponent occurred. Instantaneous diffusion of Fe into a lanthanide side and its associated Fe depletion beneath the interface acted as the major driving force in forming the reaction layer. The addition of Ce enhanced the Nd diffusion and increased the layer growth when compared to pure Nd alone.

Metal fuel slugs of U–Zr alloys for a sodium-cooled fast reactor (SFR) have been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents such as Am can cause problems in a conventional injection casting method. Therefore, in this study, several injection-casting methods were applied to evaluate the volatility of the metal-fuel elements and control the transport of volatile elements. Mn was selected as a volatile surrogate alloy since it possesses a total vapor pressure equivalent to that of minor actinide-bearing fuels for SFRs. U–10 wt% Zr and U–10 wt% Zr–5 wt% Mn metal fuels were prepared, and the casting processes were evaluated. The casting soundness of the fuel slugs was characterized by gamma-ray radiography and immersion density measurements. Inductively coupled plasma atomic emission spectroscopy was used to determine the chemical composition of fuel slugs. Fuel losses after casting were also evaluated according to the casting conditions.

Metal fuel slugs of U–Zr alloys for a sodium-cooled fast reactor (SFR) have conventionally been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents, such as Am, are problematic in a conventional injection casting method. As an alternative fabrication method, low pressure gravity casting has been developed. Casting soundness, microstructural characteristics, alloying composition, density, and fuel losses were evaluated for the following as-cast fuel slugs: U–10 wt% Zr, U–10 wt% Zr–5 wt% RE, and U–10 wt% Zr–5 wt% RE–5 wt% Mn. The U and Zr contents were uniform throughout the matrix, and impurities such as oxyen, carbon, and nitrogen satisfied the specification of total impurities less than 2,000 ppm. The appearance of the fuel slugs was generally sound, and the internal integrity was shown to be satisfactory based on gamma-ray radiography. In a volatile surrogate casting test, the U–Zr–RE–Mn fuel slug showed that nearly all of the manganese was retained when casting was done under an inert atmosphere.

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.

Candidate coating materials for re-usable metallic nuclear fuel crucibles, TaC, TiC, ZrC, ZrO2, and Y2O3, were plasma-sprayed onto a niobium substrate. The microstructure of the plasma-sprayed coatings and thermal cycling behavior were characterized, and U-Zr melt interaction studies were carried out. The TaC and Y2O3 coating layers had a uniform thickness, and high density with only a few small closed pores showing good consolidation, while the ZrC, TiC, and ZrO2 coatings were not well consolidated with a considerable amount of porosity. Thermal cycling tests showed that the adhesion of the TiC, ZrC, and ZrO2 coating layers with niobium was relatively weak compared to the TaC and Y2O3 coatings. The TaC and Y2O3 coatings had better cycling characteristics with no interconnected cracks. In the interaction studies, ZrC and ZrO2 coated rods showed significant degradations after exposure to U-10 wt.% Zr melt at 1600°C for 15 min., but TaC, TiC, and Y2O3 coatings showed good compatibility with U-Zr melt.

The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the development of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.

U-10wt%Zr spherical particles for use as particulate fuel were prepared by centrifugal atomization and subjected to pressureless sintering, which is one of the simplest powder processing techniques. At sintering temperature of 1100°C for 30 or 60 min, all samples ranging from +50 to −325 mesh showed no apparent bonding between the particles. However, at 1150°C (80 min), all samples formed a bulk body and the microstructures showed apparent sintering stages. Particularly, sample B (50–70 mesh) and sample C (70–100 mesh) showed pore characteristics suitable for a particulate fuel. The results suggest that pressureless sinterability for U-10Zr particulate fuel can be improved by adding small-size (–325 mesh) particles.

In order to investigate a fabrication process of metallic fuel for sodium-cooled fast reactor, U-Zr fuel slugs were fabricated with gravity casting and injection casting. In order to reduce the formation of radioactive wastes, and to have a fine microstructure, spherical powder of U-Zr alloy was also fabricated by centrifugal atomization. The microstructural characteristics were investigated by using electron microscopes and energy-dispersive electron X-ray spectroscope. U-10wt.%Zr fuel slugs had some dispersion phases of Zr precipitates and Zr compounds of about 5 µm in size. The matrix was composed of fine laminar structure of about 0.15 µm in thickness. U grains became finer from about 30 µm in conventionally cast fuel slug to about 2 µm in atomized powder. Laminar structure became so fine from 0.2 µm in conventionally cast fuel slug to 0.1 µm in atomized powder. It is possible for atomized metallic fuel to have a fine microstructure, resulting in a higher fission gas release rate during irradiation. Copyright © 2012 John Wiley & Sons, Ltd.

The intermetallic compounds formation at interface between rare earth elements and clad material were investigated to demonstrate the effects of rare earth elements on fuel-cladding chemical interaction (FCCI) behavior. Mischmetal (70Ce-30La) and Nd were prepared as rare earth elements. Diffusion couple testing was performed on the rare earth elements and cladding (9Cr2W steel) near the operation temperature of (sodium-cooled fast reactor) SFR fuel. The performance of a diffusion barrier consisting of Zr and V metallic foil against the rare earth elements was also evaluated. Our results showed that Ce and Nd in the rare earth elements and Fe in the clad material interdiffused and reacted to form intermetallic species according to the parabolic rate law, describing the migration of the rare earth element. The diffusion of Fe limited the reaction progress such that the entire process was governed by the cubic rate law. Rare earth materials could be used as a surrogate for high burnup metallic fuels, and the performance of the barrier material was demonstrated to be effective.

Microstructural and mechanical property were evaluated at the medium-sized HT9 (12Cr-1MoWV) forged steel which was considered as primary candidate for the fuel cladding in sodium-cooled fast reactor (SFR). Material was forged at after the induction melting to make round bar as 160mm diameter, 7000mm length then the radial distribution of microstructure as well as microhardness was evaluated. The results showed that overall microstructure exhibited as ferrite-martensite structure, where small amount (2~3%) of delta ferrite was formed throughout the specimen and maximum 15% of transformed ferrite was formed at the center, where it gradually decreased toward the radial direction. Sensitivity analysis of the cooling curve and Time-Temperature-Transformation (TTT) diagram revealed that formation of transformed ferrite could be avoided when the diameter was decreased down to 120mm.

The effect of hot rolling on the mechanical and microstructural property has been investigated to simulate the effect of hot extrusion during the manufacturing process of the fuel cladding for sodium cooled fast reactors (SFRs). Hot rolling of modified 9Cr–1Mo steel was carried out either at 1050 °C or 950 °C upon cooling after normalizing. Continuous annealing right after the hot rolling at 950 °C for 1 h has been carried out followed by the mechanical testing and microstructural analysis. The results showed that hot rolling without any annealing or tempering treatment leaves residual stress so that it leads to the abrupt increase of material strength that would affect cladding formability. Continuous annealing right after the hot rolling process can alleviate residual stress without decreasing too much of material strength. Hot rolling either at 1050 °C or 950 °C increases the number density of the remained precipitate which leads to the precipitation hardening. Introduction of continuous annealing results in an increase in the fraction of secondary V-rich MX precipitate that leads to an increase in the stability at high temperature mechanical property.

Candidate coating materials for re-usable metallic nuclear fuel crucibles, HfN, TiC, ZrC, and Y 2 O 3 , were plasma-sprayed onto niobium substrates. The coating microstructure and the thermal cycling behavior were characterized, and U–Zr melt interaction studies carried out. The Y 2 O 3 coating layer had a uniform thickness and was well consolidated with a few small pores scattered throughout. While the HfN coating was not well consolidated with a considerable amount of porosity, but showed somewhat uniform thickness. Thermal cycling tests on the HfN, TiC, ZrC, and Y 2 O 3 coatings showed good cycling characteristics with no interconnected cracks forming even after 20 cycles. Interaction studies done on the coated samples by dipping into a U–20wt.%Zr melt indicated that HfN and Y 2 O 3 did not form significant reaction layers between the melt and the coating while the TiC and the ZrC coatings were significantly degraded. Y 2 O 3 exhibited the most promising performance among HfN, TiC, ZrC, and Y 2 O 3 coatings.

We analyzed the reaction between a rare earth element and ferritic-martensitic stainless steel in order to evaluate the role of the rare earth element in the fuel cladding chemical interaction (FCCI). A diffusion couple test between Misch metal (70Ce-30La) and 9Cr-2W (Gr.92) steel at both 660 °C and 800 °C was conducted, and a microstructural analysis was carried out. The result showed that Fe in the Gr.92 material and Ce in the Misch metal mainly diffused and reacted to form intermetallic compounds, namely Fe2Ce and Fe17Ce phases, above 660 °C, whereas La hardly diffused or reacted in the Gr.92 material. In the specimen tested at 660 °C, the Fe diffused outside of the clad interface so that Cr-rich precipitated beneath the clad interface. In the specimen tested at 800 °C, a local reaction caused by the eutectic transformation of the (Fe,Cr)17Ce phase resulted in a local penetration across the clad.

This study aimed to evaluate the performance of diffusion barriers in order to prevent fuel-cladding chemical interaction (FCCI) between the metallic fuels and the cladding materials, a potential hazard for nuclear fuel in sodium-cooled fast reactors. In order to prevent FCCI, Zr or V metal is deposited on the ferriticmartensitic stainless steel surface by physical vapor deposition with a thickness up to 5μm. The diffusion couple tests using uranium alloy (U-10Zr) and a rare earth metal such as Ce-La alloy and Nd were performed at temperatures between 660~800°C. Microstructural analysis using SEM was carried out over the coupled specimen. The results show that significant interdiffusion and an associated eutectic reaction ocurred in the specimen without a diffusion barrier. However, with the exception of the local dissolution of the Zr layer in the Ce-La alloy, the specimens deposited with Zr and V exhibited superior eutectic resistance to the uranium alloy and rare earth metal. (Received January 17, 2011)

This paper focuses on the long-term creep characterization of Gr. 91 steel using creep constitutive equations. The models of three such equations, a combination of power-law form and omega model (CPO), a combination of exponential form and omega model (CEO), and a combination of logarithmic form and omega model (CLO), which are described as sum decaying primary creep and accelerating tertiary creep, are proposed. A series of creep rupture data was obtained through creep tests with various applied loads at 600 °C. On the basis of the creep data, a nonlinear least-square fitting (NLSF) analysis was carried out to provide the best fit with the experimental data in optimizing the parameter constants of an individual equation. The results of the NLSF analysis showed that in the lower stress regions of 160 MPa (σ/σys 0.65), the CPO model showed better agreement than the other two models. It was found that the CEO model was superior to the CPO and CLO models in the modeling of long-term creep curves. Using the CEO model, the long-term creep curves of Gr. 91 steel were numerically characterized, and its creep life was predicted accurately.

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.

This study deals with the fabrication of high strength ferritic/martensitic steels by a control of both the carbon concentration and the fabrication process parameters. The 9Cr-2W steels containing a carbon concentration of 0.05, 0.07 and 0.11 wt% were normalized at 1050oC for 1 h, followed by a tempering at 550 and 750oC for 2 h, respectively. The results of the tensile tests at room temperature indicated that the tensile strengths were increased with an increase of the carbon concentration from 0.05 wt% to 0.07 wt%, but no more increase was observed when the carbon concentration was increased further up to 0.11%. After a cold rolling from a 4 mm to a 1 mm thickness without/with an intermediate heat treatment and a final heat treatment, the results of the tensile tests exhibited that superior tensile properties were obtained when the fabrication processes were composed of a tempering at 550oC, and a cold rolling with several intermediate heat treatments. These results could be attributed to the finely distributed precipitates in the partially recrystallized matrix. The optimized carbon concentration and the controlled fabrication process parameters are thus suggested to fabricate a high strength 9Cr-2W steel sheet.

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.

The effects of an intermediate heat treatment during a cold rolling on the tensile strength of a 9Cr–2W steel were evaluated. Before a cold rolling, the steel was normalized at 1050 °C and tempered at 550 °C in order to avoid the formation of M 23 C 6 and V-rich MX precipitates in the martensitic structure. A 75% cold rolling and a heat treatment at 750 °C for 30 min induced the formation of large M 23 C 6 carbides in a fully recrystallized structure. However, three cold rollings with an intermediate heat treatment at 750 °C for 10 min after each cold rolling led to the formation of fine and uniform M 23 C 6 carbides in a partially recrystallized structure, providing an enhanced tensile strength at 650 °C. It is thus concluded that an intermediate heat treatment during a cold rolling could be an effective procedure for fabricating a high strength 9Cr–2W steel at high temperatures.

This study dealt with how to mathematically obtain creep constants of C, Q/2.3R, T a and t a used in Time-Temperature Parametric (TTP) methods. A number of creep rupture data from 550 °C to 800 °C for Type 316LN stainless steel were collected through a worldwide literature survey and experimental data produced at KAERI. It was found that the creep rupture data revealed a linear relationship when the data were plotted by the semilogarithmic coordinates of log (t r ) versus stress. On the basis of this linear relationship, the creep constants of C, Q/2.3R, T a and t a used in the Larson-Miller (LM), the Orr-Sherby-Dorn (OSD), and the Manson-Haferd (MH) methods were obtained by a new numerical method. This approach can be utilized as a useful method to predict the long-term creep life of Type 316 stainless steel.

The mechanical properties and microstructural evolution of modified 9Cr-1Mo steel have been studied to investigate steel property changes after long-term isothermal aging at 600 °C for 50,000 h. The microhardness and strength were maintained constantly after aging but the impact energy was dramatically reduced by 62 % during the aging period. From the viewpoint of microstructural evolution after the aging process, Cr-enrichment and Fe-depletion took place within the M23C6-type precipitates in the as-aged steel and V-depletion also happened within the VX-type precipitates after aging. In addition, the precipitates of the M2Mo-type Laves phase and the segregation of the impurity atoms would be formed during the long-term aging period. It was considered that the sharp reduction of the impact energy could be related to the formation of the Laves phases and the impurity segregation after aging at 600 °C. The phase stability was also verified by the specific heat results up to 950 °C from a DSC test. It was concluded from this study that the modified 9Cr-1Mo steel would keep its microstructural stability at 600 °C during the long-term aging period of 50,000 h, which was equivalent to the in-service life of the SFR fuel cladding.

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.

In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical CO 2 Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.

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.

Based on a composition of 99.4 wt% AISI 316L stainless steel, 0.3 wt% Ti and 0.3 wt% Y2O3, an austenitic ODS steel was fabricated by a process of mechanical alloying, hot isostatic pressing and rolling. Fine oxide particles were observed in the matrix, and their chemical formulations were determined to be Y2Si2O7 and TiO. Heat treatment of the cold-rolled sample at 1200°C induced an isotropic tensile behavior at room temperature and at 700°C This result would be mainly attributed to the equiaxed grains that form as a result of the heat treatment for recrystallization.

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.

The Mg-6wt%Al-1wt%Zn alloy powders mixture produced by gas atomization were spark plasma sintered (SPSed). The SPSed sample was coated by plasma electrolytic oxidation (PEO) method, and the microstructure, mechanical and corrosion properties were investigated and compared to the normally sintered sample. The SPSed sample exhibited a microstructure of well-bonded grains with approximately 48 ,m in grain size, which was bigger than that of the normally sintered sample. The coatings on the SPSed sample were characterized by dense and porous layers, while, in the case of the normally sintered sample, the coatings had grown through the grain boundary of substrate with increasing the coating time. The surface roughness, thickness and micro-hardness of coatings on the SPSed sample increased gradually with increasing the coating time. The corrosion property of coatings on the SPSed sample had a tendency to improve with increasing the coating time in 3.5%NaCl solution. However, it was worse than that of coatings on the normally sintered sample.

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

Metallic fuel slugs of U–10Zr–5Mn (wt%), a surrogate alloy for the U–TRU–Zr (TRU: a transuranic element) alloys proposed for sodium-cooled fast reactors, were prepared by injection casting in a laboratory-scale furnace, and their characteristics were evaluated. As-cast U–Zr–Mn fuel rods were generally sound, without cracks or thin sections. Approximately 68% of the original Mn content was lost under dynamic vacuum and the resulting slug was denser than those prepared under Ar pressure. The concentration of volatile Mn was as per the target composition along the entire length of the rods prepared under 400 and 600 Torr. Impurities, namely, oxygen, carbon, silicon, and nitrogen, totaled less than 2,000 ppm, satisfying fuel criteria.

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