Jung Won Lee

Multi-layer coating of silicon carbide and pyrolytic carbon on UO2 pellets by a combustion reaction

Abstract The coating layers of silicon carbide and pyrolytic carbon on UO 2 pellets were prepared by using a combustion reaction between the carbon and silicon layers. The pyrolytic carbon and silicon were deposited by thermal decomposition of propane at 1250°C in a chemical vapor deposition unit and microwave pulsed electron cyclotron resonance plasma enhanced chemical vapor deposition (ECR PECVD) using silane at 500°C. Microstructural observation of the layers with scanning electron microscopy (SEM) showed that an inner layer existed following the surface contour of the pellet and the outer layer had a small number of fine pores inside. Chemical analyses with Auger electron spectroscopy (AES) and X-ray diffractometry (XRD) revealed that the inner and outer layers were pyrolytic carbon and silicon carbide, respectively. From the transmission electron microscopy (TEM) observation, the silicon carbide formed during the combustion reaction was identified as fine crystalline β-SiC. The temperature distribution of the specimen during the combustion reaction was estimated by a finite element method, which showed that preheating above 1300°C was required for the combustion reaction between silicon and carbon to propagate well through the specimen.

Fractional Release Behavior of Volatile and Semivolatile Fission Products During a Voloxidation and OREOX Treatment of Spent PWR Fuel

Abstract Quantitative analysis of the fission gas release characteristics during the voloxidation and oxidation and reduction of oxide fuel (OREOX) processes of spent pressurized water reactor (PWR) fuel was carried out by spent PWR fuel in a hot cell of the DUPIC Fuel Development Facility. The release characteristics of 85Kr and 14C fission gases during voloxidation process at 500°C are closely linked to the degree of conversion efficiency of UO2 to U3O8 powder, and it can be interpreted that the release from grain boundary would be dominated during this step. Volatile fission gases of 14C and 85Kr were released to near completion during the OREOX process. Both the 14C and 85Kr have similar release characteristics under the voloxidation and OREOX process conditions. A higher burnup spent fuel showed a higher release fraction than that of a low burnup fuel during the voloxidation step. It was also observed that the release fraction of semivolatile Cs was ~16% during a reduction at 1000°C of the oxidized powder, but over 90% during the voloxidation at 1250°C.

Hydride formation by high temperature cathodic hydrogen charging method and its effect on the corrosion behavior of Zircaloy-4 tubes in acid solution

Zircaloy-4 cladding tube was hydrided using the high temperature cathodic hydrogen charging method. The optimum conditions for charging more than 1000 ppm of hydrogen was 0.4 A/cm2 for 24 h at 150°C–0.5°C in an electrolyte containing hydrobisulphate ions. After vacuum annealing at 400°C for 3 h, thin platelet-shaped hydrides were formed within the tubing and were preferentially oriented along circumferential direction, which was related to the texture of the material. The hydride formed was identified as the δ-ZcH1.6 and γ-ZrH phases by X-ray diffraction. The corrosion potential of the hydrided alloy was +830 mVSCE in 90% HNO3 at 25°C and the material was rapidly corroded by anodic polarization. The corrosion potential was dramatically decreased in a 20% hydrochloric acid solution containing small amount of a strong oxidizer, such as ferric ion due to the instability of the passive film on zirconium in this environment. The corrosion potentials of the hydrided alloy were lower than those of the as-received alloy in the corrosive environments.

EFFECT OF IMPURITIES ON THE MICROSTRUCTURE OF DUPIC FUEL PELLETS USING THE SIMFUEL TECHNIQUE

The influence of fission products contents on the DUPIC fuel powder and pellet properties was experimentally evaluated using SIMFUEL as a surrogate for actual spent PWR fuel due to the high radioactivity of spent fuel. Pure and SIMFUEL pellets with fission products equivalent to a burn-up of 35,000 MWd/tU and 60,000 MWd/tU were used as impurities in this study. The specific surface area of the powder milled after the OREOX treatment increased and resulted in sintered pellets with a theoretical density (TD) higher than 95%, regardless of the impurity contents. However, the grain size of the sintered pellets decreased with the increasing impurity contents. As a result of the dissolved oxides in from the impurity groups, the specific surface area of the OREOX powder increased with an increase of the impurities. The grain size of the sintered pellets was significantly decreased by the metallic and oxide precipitates.

Remote fabrication of DUPIC fuel pellets in a hot cell under quality assurance program

The Korea Atomic Energy Research Institute (KAERI) has been developing the Direct Use of Spent Pressurized Water Reactor (PWR) Fuel in the CANada Deuterium Uranium (CANDU) Reactors (DUPIC) fuel fabrication technology since 1992, and the basic DUPIC fuel fabrication process was established in 2002. In order to demonstrate the robustness of the DUPIC fuel fabrication process through the irradiation test, it is important that a Quality Assurance (QA) program should be in place before a fabrication of the DUPIC fuel. Therefore, the Quality Assurance Manual (QM) for the DUPIC fuel was developed on the basis of the Canadian standard, CAN3-Z299.2-85. This manual describes the quality management system applicable to the activities performed for the DUPIC fuel fabrication at KAERI. In order to demonstrate the DUPIC fuel fabrication technology and produce qualified DUPIC fuel pellets, the process qualification tests were performed, which include three pre-qualification tests and three qualification tests. The characteristics of the DUPIC fuel pellets such as the sintered density, grain size, and surface roughness were measured and evaluated in accordance with the QA procedures. The optimum fabrication process of the DUPIC fuel pellet was also established based on the qualification results. Finally a production campaign was carried out to fabricate the DUPIC fuel pellets at a batch size of 1 kg following the QA program. As a result of the production campaign, qualified DUPIC fuel pellets were successfully produced and, therefore, the remote fuel fabrication technology of the DUPIC fuel pellet was demonstrated.

Progress of the dupic fuel compatibility analysis-IV: Fuel performance

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and fuel handling system in the reactor core by both experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design, which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high-power and high-burnup conditions even though some material properties, such as the thermal conductivity, are a little lower compared to the uranium fuel. However, it is required that the current DUPIC fuel design be changed slightly to accommodate the high internal pressure of the fuel element. It is also strongly recommended that more irradiation tests of the DUPIC fuel be performed to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

Improvement of the sinterability of thermally-treated UO2 powder by horizontal rotary ball milling

Horizontal rotary ball milling has been demonstrated to be a useful method for reducing the particle size of ceramic powder in remote operation in shielded hot cells. Techniques, equipment and operating parameters, such as milling media, media wear and rotor speed were investigated with Al2O3 powder to evaluate its performance prior to contamination with nuclear fuel material. The established operating parameters were then verified with UO2 powder, which had been produced by a thermal process to make fuel pellets. The sintering of the milled UO2 powder showed the higher sintered densities obtainable by the milling, and the milling process seemed to be an important factor in improving the powder characteristics.

INPRO Studies of Proliferation Resistance for DUPIC Fuel Cycle

The evaluation of proliferation resistance on the DUPIC fuel cycle was performed using the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) user manual. The user manual was finally published from INPRO as a tool for proliferation resistance evaluation. Five user requirements were well organized under one basic principle, and each evaluation parameter for each user requirement has criteria for qualitative and quantitative evaluation for an innovative nuclear system. The DUPIC fuel cycle is to fabricate CANDU fuel from spent pressurized water reactor fuel by use of a dry thermal process without separating the stable fission products. The DUPIC process and fabricated fuel have very intense radiation background and a low amount of fissile plutonium and uranium. The DUPIC fuel cycle has a number of intrinsic features that enhance proliferation resistance. The number of assemblies in the DUPIC process to get 1 significant quantity is very large ([approximately]48 assemblies). The assessment results using the user manual show that the DUPIC fuel cycle is very strong against nuclear proliferation by the material property itself and the facility condition. Additionally, several suggestions and conditions were made to increase the proliferation resistance for innovative future nuclear energy system.

Neutron-Induced Data Evaluation for Selected Fission Products

The neutron-induced cross section data for 19 high-priority fission products were evaluated in the fast energy region using the following models: spherical and deformed optical model, multi-step compound and direct (MSC and MSD), pre-equilibrium exiton and Hauser-Feshbach models with a width fluctuation. The calculations were compared to recently measured data and to currently evaluated files. The results in the fast energy region were converted into the ENDF-6 format and merged with the evaluated resonance part. For consistency, the background was included with the merging energy region. The nuclear data set involves the (n, tot), (n, n), (n, n′), (n, 2n), (n, 3n), (n, nα), (n, np), (n,γ), (n, p), and (n, α) cross sections from the thermal to 20 MeV energy range. The results showed good agreement with the measured data. A considerable improvement was achieved for most reactions and nuclei. Format and physics checking codes were applied to each full data set and the NJOY code was used to perform a processing check on the individual cross sections.