[en] The BDL-443 CANFLEX-RU bundle AKW was fabricated at Korea Atomic Energy Research Institute (KAERI) for power ramp irradiation testing in NRU reactor. The bundle was fabricated with IDR and ADU fuel pellets in adjacent elements and contains fuel pellets enriched to 1.65 wt% ²³⁵U in the outer and intermediate rings and also contains pellets enriched to 2.00 wt% ²³⁵U in the inner ring. This bundle does not have a center element to allow for insertion on a hanger bar. KAERI produced the IDR pellets with the IDR-source UO₂ powder supplied by BNFL. ADU pellets were fabricated and supplied by AECL. Bundle kits (Zircaloy-4 end plates, end plugs, and sheaths with brazed appendages) manufactured at KAERI earlier in 1996 were used for the fabrication of the bundle. The CANFLEX bundle was fabricated successfully at KAERI according to the QA provisions specified in references and as per relevant KAERI drawings and technical specification. This report covers the fabrication activities performed at KAERI. Fabrication processes performed at AECL will be documented in a separate report

[en] A static and finite-element (FE) analysis model was developed to simulate out-reactor fuel string strength tests with use of the structural analysis computer code ABAQUS. The FE model takes into account the deflection of fuel elements and stress and displacement in end-plates subjected to hydraulic drag loads. It was adapted to the strength tests performed for CANFLEX 43-element bundles and the existing 37-element bundles. The FE model was found to be in good agreement with the experiment results. With use of the FE model, the static behavior of the fuel bundle strings, such as load transfer between ring elements, end-plate rib effects, hydraulic drag load incurring plastic deformation in fuel string and hydraulic flow rate effects were investigated

[en] There are 380 fuel channels in a CANDU-6 reactor, and twelve fuel bundles are loaded into each fuel channel. High-pressure, heavy water coolant passes through the fuel bundle string to remove heat generated from the fuel. Fuel sheath collapses down around the uranium dioxide pellet due to the coolant pressure when the fuel is loaded into the reactor. Longitudinal ridges may form in CANDU fuel element sheaths as a result of sheath collapse onto the pellets. A static analysis, finite-element (FE) model was developed to simulate the longitudinal ridging of the fuel element with use of the structural analysis computer code ABAQUS. Collapse pressures were calculated for the fifty-one cases for which test results of WCL in 1973 and 1975 are available. Calculation results under-predicted the critical collapse pressure but it showed significant relationship against test results

[en] This report intends to select the test channel groups in the Wolsong-3 reactor for the evaluation of PHWR fuel integrity in the two phase flow condition. First, a detailed work schedule for fuel inspection, fuel loading, storage of the discharged fuel, and in-bay visual and dimensional examination was established on the CIGAR work plan of the Wolsong-3 reactor. According to the schedule, it is necessary to suggest the appropriate test channel groups with two phase flow before the new fuel loading. For the selection of the test channel groups, the distributions of the channel flow rate and the channel exit quality are calculated by the NUCIRC code with the monthly measured operation data of the Wolsong-3 reactor during May 2001 to December 2002. The monthly calculated data could sort out the channels with the channel exit quality greater than 0.2%, in the high power channels with the power higher than 6300 kW and the flow rate greater than 25.5 kg/s, and in the low power channels with the power higher than 4800 kW and less than 5300 kW. Based on the frequency rank of these sorted channels, O16 channel with thirty other channels in the high power region and D17 channel with ten other channels in the low power region are suggested as the test channel groups

[en] A structural analysis was performed to simulate the impact of the fuel bundle string on the inlet shield plug during a 100% Reactor Inlet Header (R.I.H) brake accident in a CANDU-6 reactor. Any significant damage to either the fuel or the fuel channel due to the collision could result in coolant flow blockage, and thus pose additional safety related concerns beyond those addressed for the initial loss-of-coolant accident. A Finite-Element (FE) model for simulating the collision was developed using the structural analysis computer code ABAQUS. The FE model was validated against the test results that have been obtained during the normal refueling impact test performed at KAERI in 1996. With use of the FE model, dynamic behavior of the fuel bundle string impacted on the shield plug was investigated and its effects on the fuel bundles and pressure tube were evaluated. The overall integrity of the fuel bundles as well as the possibility of bundle sticking or coolant flow blockage in the pressure tube was assessed

[en] The effect of the CANFLEX-NU fuel element bowing on the critical heat flux is reviewed and analyzed, which is requested by KINS as the Government design licensing condition for the use of the fuel bundles in CANDU power reactors. The effect of the gap between two adjacent fuel elements on the critical heat flux and onset-of-dryout power is studied. The reduction of the width of a single inter-rod gap from its nominal size to the minimum manufacture allowance of 1 mm has a negligible effects on the thermal-hydraulic performance of the bundle for the given set of boundary conditions applied to the CANFLEX-43 element bundle in an uncrept channel. As expected, the in-reactor irradiation test results show that there are no evidence of the element bow problems on the bundle performance

[en] This extreme reactor condition makes serious material limitation and emphasizes the importance of safety analysis. To get permission of construction license, previous researches like preliminary safety research have been analyzed risk assessments of fusion reactors. To simulate the severe accidents in fusion reactor, a number of thermal hydraulic simulation codes were used(ECART, INTRA, ATHENA/RELAP and so on). Before construction, to obtain ITER license about safety issue, MELCOR is chosen as the thermal hydraulic code to be used to simulate radioactive material release from severe accidents. Capability of the simulation code in severe accident analysis is to simulate the cooling system in ITER, the transport of radionuclides during design basis accidents (DBAs) including beyond design basis accidents (BDBAs). MELCOR is fully integrated code that models the accidents in Light Water Reactor (LWR). To analyze the accidents in ITER, MELCOR 1.8.2 version is modified. The amount of release radioactive material is safety acceptance criteria in the nuclear fusion system. There are three kinds of radioactive materials in fusion reactor; tritium (or Tiritiated water: HTO), activation products from divertor or first-wall(AP) and activated corrosion products(ACP). In generic Site Safety Report (GSSR), table I lists the release guidelines for tritium and activation products for normal operation, incidents, and accidents. This small scale facility makes the experimental flexibility to develop fusion technology. Fusion source difference between KSTAR and ITER is D-D(Deuterium- Deuterium reaction) fusion and D-T(Deuterium- Tritium reaction) fusion. This D-D fusion makes Tritium in the 50 percent chance. The radioactivity of tritium is small to consider, but, the accident analysis is indispensable. In the present work, the conservatively estimated tritium inventory in KSTAR is used with one of the most severe accident in ITER; Fusion Power Termination System(FPTS) failure with multiple first wall pipe break. The MELCOR modified input deck is used to study and radioactive material leakage is simulated with aerosol release package to follow up the ITER safety analysis. In this research, follow-up study of safety analysis and simple safety analysis application in KSTAR was conducted with MELCOR

[en] A structural analysis was performed to simulate the impact of the fuel bundle string on the inlet shield plug during a 100% Reactor Inlet Header (R.I.H) brake accident in a CANDU-6 Reactor. Any significant damage to either the fuel or the fuel channel due to the collision could result in coolant flow blockage, and thus pose additional safety related concerns beyond those addressed for the initial loss-of-coolant accident. A three-dimensional Finite-Element (FE) model for simulating the collision was developed using the structural analysis computer code ABAQUS. The FE model was validated with the test results that have been obtained during the normal refueling impact test performed at KAERI in 1996. The analysis results agree well with the test results. With use of the FE model, dynamic behavior of the fuel bundle string impacted on the shield plug was investigated and its effects on the fuel bundles and pressure tube were evaluated. The overall integrity of the fuel bundles as well as the possibility of bundle sticking or coolant flow blockage in the pressure tube was assessed

[en] To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

[en] KAERI/AECL/BNFL Joint Program of CANFLEX-SEU/RU(Slightly Enriched Uranium/ Recovered Uranium) fuel for CANDU reactors has been conducted since 1996, where the RU is a product in the reprocessing of the irradiated uranium fuel. The total amount of RU produced from reprocessing operations in Europe and Japan was expected to be around 25,000 tons by the year 2000 with additional quantities also having arisen from reprocessing operation in the former Soviet Union. The RU fuel be burned up to the twice of the natural uranium fuel and so the spent fuel volume production rate is a half of the rate in the spent fuel production from the natural uranium fuel. The RU price is relatively lower to the SEU fuel with the same enrichment. CANFLEX-RU fuel bundle will increase 5% more the CCP of CANDU-6 reactor compared to the 37-element fuel bundle. As the economic and industrial point of view, the use of CANFLEX-RU fuel in a CANDU will be resulted in 1)reduction of reactor operating cost, 2)improvement of uranium utilization, 3)reduction of fuel cycle costs, 4)reduction of spent fuel volume production rate, and 5)reduction of an amount invested for the spent fuel storage. This document is the final phase report of the R and D on the development of recovered uranium fuel technology for PHWR. As one of the mid- and long-term project of national nuclear energy, the R and D project has been performed for 3 years since March 2000. It describes the R and D contents and results in the fields of the CANFLEX-RU fuel design and fabrication, the reactor physics, the thermalhydraulics, safety and so on. Also, it is included the fuel technical and test specifications of the fuel. The thermo-mechanical design criteria and analysis methodology of the CANFLEX-RU fuel element and bundle were established, and then the in-reactor performances and designs of CANFLEX-RU fuel element and bundle were analyzed preliminarily. In the reactor physics analysis of CANDU-6 reactor with CANFLEX-RU fuel bundle, Establishing the reactor physics analysis methodology and code system for the analysis, the CANFLEX-RU equilibrium and transition cores of a CANDU 6 reactor were analysis to find the optimized refuelling scheme. The ThermalHydraulic(T/H) analyses of CANDU-6 fuel channel with CANFLEX-RU fuel bundles and CANDU-6 fuel channel mixed with CANFLEX-RU fuel bundles and 37-element bundles were performed. Also, the subchannel analysis and the data analysis of high flow visualization tests were carried out. 6 scenarios of postulated accidents that caused fuel failures were selected and analyzed In the safety assessment of the reactor with CANFLEX-RU fuel. Finally, the characterizations of CANFLEX-RU design of materials/fabrications, physical and chemical and radiological properties of the CANFLEX-RU powder/pellet were evaluated.