Currently, the Korean nuclear industry uses ZIRLO as material for nuclear fuel cladding(zirconium alloy). KEPCO Nuclear Fuel is in the process of developing a HANA alloy to enable domestic production of cladding. Cladding manufacture involves multistage heat treatments and pickling processes, the latter of which is vital for the removal of defects and impurities on the cladding surface. SMUT that forms on the cladding surface during such pickling process is a source of surface defects during heat treatment and post-treatment processes if not removed. This study analyzes ZIRLO, HANA-4, and HANA-6 alloy claddings to extensively study the SEM/EDS, XRD, and particle size characteristics of SMUT, which are second phase particles that are formed on the cladding surface during pickling processes. Using the analysis results, this study observes SMUT formation characteristics according to Nb concentration in Zr alloys during the washing process following the pickling process. In addition, this study observes SMUT removal characteristics on cladding surfaces according to concentrations of nitric acid and hydrofluoric acid in the acid solution.

According to the nuclear safety act, the enforcement regulations and the notification of the atomic energy commission, a preliminary decommissioning plan must be submitted at all domestic nuclear facilities. In accordance with this preliminary decommissioning plan, it is required to prepare eleven items from the outline of the decommissioning plan of the nuclear facility to the fire protection. Currently, the nuclear fuel cycle facility operated by the Korea Atomic Energy Research Institute (KAERI) consists of a radioactive waste form test facility (RWFTF), a post irradiation examination facility (PIEF), a radioactive waste treatment facility (RWTF), and a radioactive waste storage facility (RWSF). The decommissioning strategies, decommissioning methods and dismantling activities of these nuclear facilities are described in this paper. The scope of decommissioning, the dismantling method, the final conditions of the site, the management of radioactive waste, and the cost of decommissioning are established in the decommissioning strategy. The decommissioning schedule, work order, basic principle and technical feasibility are determined at the method of decommissioning. The disinfection techniques and activity plans for facilities and sites contaminated with radioactive materials are described at the dismantling activity. Therefore, this paper describes the concept of decommissioning of the nuclear fuel cycle facilities and prepares a preliminary decommissioning plan to be prepared afterwards.

In this paper, a structural integrity on the test rig with assembly plug to perform intermediate examination is evaluated. Structural analysis results between the test rig with non assembly plug and assembly plug are compared, because the assembly plug has an effect on the flow of the coolant in the test rig. A equivalent stress value on the test rig with assembly plug is increased more than the stress on the test rig with non-assembly plug. A shape optimization of the assembly plug is performed to decrease the stress. Considering a connection with the transport tool, a optimized shape of the assembly plug is presented to minimize the stress on the test rig. Using the optimized assembly plug, the equivalent stress on the test rig with the optimized plug is less than the stress on the test rig with the non-optimized plug.

Tri-isotropic (TRISO) coatings on zirconia surrogate beads are deposited using a fluidized-bed vapor deposition (FB-CVD) method. The silicon carbide layer is particularly important among the coated layers because it acts as a miniature pressure vessel and a diffusion barrier to gaseous and metallic fission products in the TRISO-coated particles. In this study, we obtain a nearly stoichiometric composition in the SiC layer coated at 1400oC, 1500oC, and 1400oC with 20 vol.% methyltrichlorosilane (MTS), However, the composition of the SiC layer coated at 1300-1350oC shows a difference from the stoichiometric ratio (1:1). The density decreases remarkably with decreasing SiC deposition temperature because of the nanosized pores. The high density of the SiC layer (≥ 3.19 g/cm2) easily obtained at 1500oC and 1400oC with 20 vol.% MTS did not change at an annealing temperature of 1900°C, simulating the reactor operating temperature. The evaluation of the mechanical properties is limited because of the inaccurate values of hardness and Young’s modulus measured by the nano-indentation method.

초고온가스로는 고온의 원자로 열을 이용하여 대량의 청정 수소와 고효율의 전기를 생산할 수 있는 제 4세대 원자로이다. 초고온가스로는 0.5mm 직경의 우라늄을 세라믹으로 3중 코딩해 직경 약 0.9mm의 TRISO라고 불리 는 피복입자를 사용한다. TRISO는 크기가 작을 뿐 아니라 특수 코팅 처리가 되어 있어 우라늄이 직접 공기 중 에 노출될 일이 없다. 핵연료 품질 유지 측면에서 TRISO 제작시 핵연료의 동일한 구형성, 밀도 및 피복층 두께 를 유지하는 것이 중요하다. 본 논문에서는 X-선 래디오그래피 기술을 이용한 비파괴 방법을 적용하여 측정한 TRISO 피복입자의 피복층 두께 자료를 바탕으로 피복입자핵연료의 대량 제조 관리를 가정하였고, 이 경우 X-R 관리도를 이용하여 생산 공정의 공정 이상 여부를 시뮬레이션하였다.(한글초록 300자 내외)

To precisely assemble the fuel test rod, an orbital TIG welding system was designed and developed to accurately conduct orbital TIG welding for the nuclear fuel test rod. Using this system, a welding process needs to confirm the welding properties for orbital TIG welding. Therefore, preliminary weld tests were performed on the cladding tubes under various conditions, and the results show that the width and depth of HAZ of the cladding specimen welded using identical power during an orbital TIG welding cycle was continuously increased from a welded start-point to a welded end-point because of heat accumulation. The performance tests were conducted under the welding conditions considered through preliminary welding tests, and the properties of the specimens were conformed through surface and microstructure analyses.

To conduct a nuclear fuel irradiation test, the inside of the nuclear fuel rod must be assembled along with the test fuel, several different parts, and sensors, and then filled with high-pressure and high-purityhelium gas. Therefore, it is necessary to develop helium gas filling techniques that can achieve exact TIG (Tungsten Inert Gas) spot welding at a pin-hole of the nuclear fuel rod to fill helium gas into the nuclear fuel test rod. However, previous apparatuses do not have repeatability for TIG spot welding as they lack an electrode position control jig to exactly fix a TIG electrode in a high-pressure chamber, and they consume a large amount of helium gas. Therefore, a TIG spot welding apparatus was developed to easily and accurately conduct TIG spot welding and significantly reduce the gas consumption. In addition, the optimum welding conditions of this welding apparatus were established through various weld tests.

This paper describes the spherical ammonium diuranate gel particles which are the intermediated material of the microsphere for an VHTR(very high temperature reactor) nuclear fuel. The characteristics of the intermediate-ADU gel particles prepared by AWD(ageing, washing, and drying) and FB(fluidized-bed) apparatus were examined and compared in a sol-gel fabrication process. The electrical conductivity of washing filtrate from the FB treating and the surface area of dried-ADU gel particles were higher than those of AWD treating. Also, an internal pore volume in dried-ADU gel particles showed a more decrease in AWD treatment than FB treatment because of decomposition of PVA affected by the washing time. However, the internal microstructures of ADU gel particles were similar regardless of the process variation.

When a new nuclear fuel is developed, irradiation test needs to be carried out in the research reactor to analyze the performance of the new nuclear fuel. It is necessary to attach sensors in the fuel rod and connect them with instrumentation cables to check the performance of the nuclear fuel during the burn up test in the test loop. A thermocouple is installed at the center of the fuel rod to check the centerline temperature of a fuel rod during the irradiation test. Therefore, A hole needs to be made at the center of a fuel pellet to put the thermocouple. However, it is difficult to make a small fine hole on the sintered UO2 pellet with a simple drilling machine, because the hardness and density of a sintered UO2 pellet are very high. In this study, an instrumented fuel rod mock-up was fabricated using an automated precise drilling machine. Four sintered alumina were drilled off and assembled into the zircaloy tube and a thermocouple was instrumented in the fuel rod mock-up. Sealing of an instrumented fuel rod mock-up was performed in the following two methods. It is sealing of similar metals perform by welding method, and sealing of dissimilar metals perform by swagelok method.

The Heat Transport system loop stability of CANDU-6 reactor as Wolsong-1 with the CANFLEX fuel bundle has been studied. The SOPHT modelling of the CANFLEX fuel bundle and the ROH interconnection line was made and the stability analysis response of Wolsong-1 reactor with CANFLEX fuel bundle was obtained. The mechanics of the flow instability caused by two phase flow was reviewed. Without the ROH interconnection line the Heat Transport system loop is unstable while the Heat Transport system is stable within ±1 % of nominal flow with the ROH interconnection line

The effects of thermal treatment conditions on ADU (ammonium diuranate) prepared by SOL-GEL method, so-called GSP (Gel supported precipitation) process, were investigated for kernel preparation. In this study, ADU compound particles were calcined to particles in air and Ar atmospheres, and these particles were reduced and sintered in 4%-/Ar. During the thermal calcining treatment in air, ADU compound was slightly decomposed, and then converted to phases at . At , the phase appeared together with . After sintering of theses particles, the uranium oxide phases were reduced to a stoichiometric . As a result of the calcining treatment in Ar, more reduced-form of uranium oxide was observed than that treated in air atmosphere by XRD analysis. The final phases of these particles were estimated as a mixture of and .

The small break loss-of-coolant accidents for the HANARO fuel test loop have been predicted by MARS code. Conservative method was used for the prediction of the loss-of-coolant accidents. The maximum peak cladding temperature was calculated as 1286K, which was lower than the design limit temperature (1477K) of nuclear fuels for the HANARO fuel test loop. The maximum peak cladding temperature occurred for the cold leg break in the HANARO pool. The hydrogen generation and oxidation of the fuel cladding were also negligible. Consequently, it is ensured that the emergency cooling water system for the HANARO fuel test loop is appropriate for the small break loss-of-coolant accidents.

In order to investigate a nitriding process of spent oxide fuel and the subsequent change in thermal properties after nitriding, simulated spent fuel powder was converted into a nitride pellet with simulated fission product elements through a carbothermic reduction process. Nitriding rate of simulated spent fuel was decreased with increasing of the amount of fission products. Contents of Ba and Sr in simulated spent fuel were decreased after the carbothermic reduction process. The thermal conductivity of the nitride pellet was decreased by an addition of fission product element but was higher than that of the oxide fuel containing fission product elements.

FTL(Fuel Test Loop) is a facility that confirms performance of nuclear fuel at a similar irradiation condition with that of nuclear power plant. FTL construction work began on August, 2006 and ended on March, 2007. During Construction, ensuring the worker's safety was the top priority and installation of the FTL without hampering the integrity of the HANARO was the next one. The installation works were done successfully overcoming the difficulties such as on the limited space, on the radiation hazard inside the reactor pool, and finally on the shortening of the shut down period of the HANARO. The Commissioning of the FTL is on due to check the function and the performance of the equipment and the overall system as well. The FTL shall start operation with high burn up test fuels in early 2008 if the commissioning and licensing progress on schedule.

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It is consisted of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS which is localed inside the pool is divided into 3-parts; they are in-pool pipes, IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor collant pressure boundary. The localization of the IVA is achieved by manufacturing through local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique of the instrument lines has been checked for its functionality and yield. A IVA has been manufactured by local technique and will be finally tested under out of the high temperature and high pressure test.

The nuclear fuel cladding temperatures of the HANARO fuel test loop have been calculated by MARS code for the large break loss-of-coolant accidents. Conservative method was used for the analysis of the loss-of-coolant accidents. Consequently, the maximum peak cladding temperature was predicted as 1235K, which was lower than the design limit temperature (1477K) of nuclear fuels for the HANARO fuel test loop. This means that the cooling capability of the emergency cooling water system for the HANARO fuel test loop is sufficient for the large break loss-of-coolant accidents.

The conservative method on the analysis of loss-of-coolant accidents for the HANARO fuel test loop was established based on the guide of evaluation method for the emergency core cooling systems of pressurized light water reactors. The evaluation models, the Moody model for discharge rate calculation and the Baker-Just model for water-metal reaction calculation, were used. In order to calculate conservative peak cladding temperatures for accidents the multipliers to the correlations of heat transfer coefficients in the MARS were also introduced. Consequently it is found that the maximum peak cladding temperature predicted by using the conservative method is sufficiently greater than that calculated by using the best-estimated models.

본 연구에서는 경수로용 핵연료집합체의 전체지지격자(Full Size Grid)와 부분지지격자(Small Size Grid)에 대한 정적 좌굴강도 실험과 전체 지지격자와 부분지지격자를 구성하는 지지격자판(Grid Strap)에 대한 정적 좌굴해석을 수행하여 지지격자의 좌굴특성을 분석하였으며, 분석결과를 이용하여 전체지지격자와 부분지지격자에 대한 좌굴하중값의 예측 가능성을 평가하였다. 좌굴강도 실험은 웨스팅하우스형 연료의 셀을 갖는 전체지지격자와 등의 셀을 갖는 부분지지격자에 대하여 수행하였으며, 실험결과를 이용하여 지지격자의 좌굴강도와 지지격자의 행(rows)과 열(columns) 사이의 관계식을 제시하였다. 좌굴강도 해석은 범용 유한요소해석코드인 ANSYS를 이용하여 수행하였으며, 해석결과를 이용하여 지지격자의 좌굴특성을 분석하고 실험결과와 비교평가 하였다.

경수로 원자로 하부구조물에서 발생되는 유포의 불균일성에 기인하는 교차류와 핵연료집합체의 수력저항의 차이에 의해 발생하는 교차류, 그리고 축류 등에 의해 유발되는 연료봉의 불안정성은 핵연료손상의 원인이 될 수 있으므로, 새로운 연료 개발 시 연료봉에 대한 진동 및 안정성 해석을 수행하여 연료봉 진동과 불안정성 발생 여부를 확인하고 있다. 본 연구에서는 새로 개발된 고리 2호기용 형 개량핵연료 집합체에 대한 연료봉의 진동 및 안정성 해석을 수행하여 지지격자 높이와 위치, 그리고 지지조건 등이 연료봉의 진동특성 및 안정성에 미치는 영향을 평가하였다 그리고 해석결과에 근거하여 개량연료 집합체에서 중간지지격자 높이와 각 지지격자의 위치를 제안하였다.