No abstract available

[en] In this paper, we derive a nonlinear magnetic circuit model of an electromagnetic control-rod actuator in the SMART. The results of the nonlinear model are compared with those by linear circuit model and finite-element analyses. gnetic circuit modeling is a useful tool when designing an electromagnetic actuator, as it allows fast calculations and enables parametric studies. It is particularly essential when the actuator is to be used in a very complex system such as a nuclear reactor. Important design parameters must be identified at the early stage of the design process. Once the design space is narrowed down, more accurate methods such finite-element analyses (FEA) can be employed for detailed design. Magnetic circuit modeling is based on the assumption that a flux path consists of sections in each of which field quantities are constant with linear constitutive relations. This assumption fails to hold when portions of the flux path become saturated. The magnetic circuit must be modified in order to accurately describe the nonlinear behavior of saturation

[en] A severe accident is analyzed for Korea SMR reactor, SMART. Core melt down sequences are analyzed for SMART reactor core using MELCOR version 1.8.5. MELCOR is developed by Sandia National Laboratory for US NRC for the simulation of severe accidents in nuclear power plants. Two cases are simulated here and compared between them; one is the case for core having 3 concentric rings and the other is the case for core having 5 concentric rings. One inch break LOCA scenario is simulated and compared between these two core models. Time sequences for the thermal hydraulic behaviors of RPV and thermal heatup behaviors of reactor core are explained in graphically. Thermal hydraulic behavior such as the change of pressure, level, mass, and temperature of RPV is explained. Thermal heatup behavior of reactor core such as oxidation of cladding, hydrogen generation, core slumping down to lower plenum, and finally creep rupture of PRV lower head is explained. Engineered safety features such as safety injection systems (SIS), and Passive residual heat removal systems (PHRS), etc. are assumed to be not working. One inch break of severe accident is simulated on Korean SMR (SMART) Integral PWR with MELCOR code version 1.8.5. Core melt progression and lower head failure time is very slow compared to other commercial reactors. Simulation on 3 and 5 radial rings core models gives very similar pattern in core cell failure timings. Other various accident scenarios (for example, SBO in Fukushima) will be tried further. Containment behaviors and source term behaviors in severe accident conditions will be analyzed in future

[en] UCFR (Ultra-long-life Core Fast Reactor) is a 260MW_th /100MW_e sodium-cooled fast reactor which requires no on-site refueling and meets the need for future nuclear energy systems. UCFR is a pool type reactor with metallic fuels, four intermediate heat exchangers, two steam generators, and passive decay heat removal systems. Because gallium has the chemical reaction safety such as low oxygen reactivity compared to sodium, it can be used as a boundary material between sodium and atmosphere to enhance the nuclear safety of UCFR. In this research, design studied for neutronics and thermal-hydraulics are included. The safety performance of UCFR will be analyzed with MARS-LMR. Although MARS-LMR was originally intended for a safety analysis of liquid metal-cooled reactor, gallium properties were newly added to this code which is applicable for gallium-cooled systems. The properties of various liquid metals are indicated in table II. Considering needs to improve uranium utilization and solve the nuclear proliferation, ultra-long cycle fast reactor has been developed. UCFR is a 260MW_th/100MW_e sodium-cooled fast reactor which requires no on-site refueling during design period with metallic fuels (U-5Zr), HT-9 cladding, four intermediate heat exchangers, two steam generators, and Ga-based PDHRS. Through this paper, new PDHRS using gallium that can be remove decay heat passively for an infinite time is suggested. In Ga-based PDHRS, the both water and air as an ultimate heat sink will be can be considered because gallium has the chemical reaction safety. In this research, design study for neutronics and thermal-hydraulics were also included. For safety analysis of UCFR using MARS-LMR, detailed design of UCFR with Ga-based PDHRS will be required

[en] The material data of fatigue crack growth and creep crack growth for robust structural integrity evaluations lacks in Subsection NH while it provides material properties of G91 steel for design purposes at high temperature conditions. Creep-fatigue crack initiation and growth tests for a G91 tubular specimen, including a machined defect, have been performed by Kim and it attempted to assess a high temperature crack behavior of G91 side plate specimen by Lee. The fatigue crack growth tests of a G91 compact tension (CT) specimen were performed by Kim at three different temperatures (500 .deg. C, 550 .deg. C, and 600 .deg. C), three loading frequencies (0.1Hz, 1Hz, and 20Hz), and two loading ratio values of 0.1 and 0.3, respectively; thus total 18 test conditions were applied. In this study, complementary tests were performed for selected 8 test conditions among total 18 test conditions to assess the effects of temperature and loading frequency on the fatigue crack growth rate of G91 steel and the test results were discussed. It is known that the fatigue crack growth rate increases as loading frequency decreases, as temperature increases, and load ratio (R) increases, it depends on the test conditions and relative sensitivity. In this study, the fatigue crack growth tests for a G91 compact tension specimen were performed for a various loading frequencies, loading ratios, and temperatures. As shown in Fig. 5 ∼ Fig. 10, it was confirmed that the fatigue crack growth rate increases apparently as temperature increases. The effects of loading frequency and load ratio were assessed by comparing above results and it was found that the fatigue crack growth increases as loading frequency decreases from 20Hz to 0.1Hz and load ratio increases from 0.1 to 0.3. Collected data for high temperature fatigue crack growth of G91 steel would be utilized for the structural integrity assessment of SFR components

[en] Each sub-module has a seven-layer breeding zone: three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebble packed tritium breeder layers, and a reflector layer packed with 1 mm diameter graphite pebbles to reduce the volume of beryllium. The abrasion of graphite structures due to the relative motion or thermal cycle during operation may produce graphite dust. It is thought that the graphite dust is more oxidative than bulk graphite, and thus oxidation behavior of graphite dust must be examined to analyze the safety of the reactors during an air ingress accident. In this study, the oxidation and explosion behaviors of ball-milled nuclear graphite powder were investigated. An examination was made to characterize the oxidation behavior of ball-milled nuclear graphite powder through a TG-DSC analysis. With the ball milling time, the BET surface area increased with a reduction of the particle size, but decreased with the chemisorption of O₂ on the activated surface. The enhancement of the oxidation after the ball milling is attributed to both increases in the specific surface area and atomic scale defects in the graphite structure

[en] This study was conducted to explore the potential of Supercritical Carbon Dioxide (S-CO₂) Brayton cycle for the HTGR application. The S-CO₂ cycle is being considered as a PCS due to its high thermal efficiency, simplicity, compactness and so on. Generally, the S-CO₂ Brayton cycle is characterized as a highly recuperated cycle which means that to achieve high thermal efficiency, the cycle requires a highly effective recuperator. Argonne National Laboratory (ANL) showed that direct application of the standard S-CO₂ recompressing Brayton cycle to the HTGR or the Very High Temperature Reactor (VHTR) is difficult to achieve high thermal efficiency due to the mismatch of the temperature difference between the temperature drop of helium as the primary reactor coolant and the temperature rise of CO₂ as the PCS coolant through an Intermediate Heat Exchanger (IHX). Therefore, our research team suggests a novel S-CO₂ cycle configuration, the S-CO₂ Brayton and Rankine hybrid cycle, to solve this limitation. This S-CO₂ hybrid concept is utilizing the waste heat of the S-CO₂ Brayton cycle as heat input to the S-CO₂ Rankine cycle. Dividing the thermal capacity of the heat source in to the Brayton cycle part and Rankine cycle part of the S-CO₂ hybrid cycle appropriately, the temperature difference at the IHX could be reduced, therefore the net system performance and operating range can be improved. In this study, the ANL research is reviewed by the in-house cycle analysis codes developed by the Korea Advanced Institute of Science and Technology (KAIST) research team. And the S-CO₂ Brayton and Rankine hybrid cycle is studied as a PCS for the VHTR condition which was utilized by ANL research team; it was assumed that the core outlet temperature to be 850 .deg. C and the core inlet temperature of 400 .deg. C

[en] TRISO fuel particles are designed to not crack due to the stresses from processes (such as differential thermal expansion or fission gas pressure) at temperatures up to and beyond 1600 .deg. C, and therefore can contain the fuel in the worst of accident scenarios in a properly designed reactor. The TRISO fuel particles are fabricated into compacts and placed in a graphite block matrix in a prismatic block gas cooled reactor. TRISO fuel particles have been developed in the KAERI as part of the NHDD project. We expect the TRISO fuel compacts to be used in the Korean type VHTR reactor in the near future. Two of the major quality components are the sphericity and exact kernel size. Thus, we present a TRISO quality control method to improve the quality of TRISO and cope with continuous mass manufacturing production

[en] Polycrystalline nuclear graphite has been proposed as a fuel element, moderator and reflector blocks, and core support structures in a very high temperature gas-cooled reactor. During reactor operation, graphite core components and core support structures are subjected to various stresses. It is therefore important to understand the mechanism of deformation and fracture of nuclear graphites, and their significance to structural integrity assessment methods. Digital image correlation (DIC) is a powerful tool to measure the full field displacement distribution on the surface of the specimens. In this study, to gain an understanding of compressive deformation characteristic, the formation of strain field during a compression test was examined using a commercial DIC system. An examination was made to characterize the compressive deformation behavior of nuclear graphite by a digital image correlation. The non-linear load-displacement characteristic prior to the peak load was shown to be mainly dominated by the presence of localized strains, which resulted in a permanent displacement. Young's modulus was properly calculated from the measured strain

[en] Diversity is the fundamental principle in safety system design of new nuclear power plants, which uses different mitigation measures to provide diverse ways of responding to a significant event. Regarding the diversity principle, EU-APR1400 (European APR1400) safety system should be in accordance with European design requirements. This paper provides how the core residual heat can be removed with open RCS (Reactor Coolant System) closure head condition using diverse cooling systems assuming a loss of SCS (Shutdown Cooling System). The loss of SCS results from a common cause failure of SCS or a loss of UHS (Ultimate Heat Sink). One of the postulated common cause failures which lead to the loss of SCS can be a manufacturing deficiency of shutdown cooling pumps, component cooling water pumps or essential service water pumps. The limiting event of the loss of SCS is the loss of UHS since the loss of UHS causes losses of safety systems including SCS, CCWS and ESWS. NO operator action inside the MCR (Main Control Room) during the first 30 minutes and outside the MCR during the first 60 minutes can be credited to recover the failed SCS. The diverse means of the EU-APR1400 for the residual heat removal function with RCS open condition have been developed to comply with the diversity principle of the European design requirements of new nuclear power plants. The results of the preliminary assessment show that the diverse design features are expected to satisfy the design criteria. In the future, more detailed design and assessment for this means will be performed considering interface requirements with completing the design of relevant systems and components