• Nuclear Engineering and Technology
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• 제32권1호
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• pp.57-66
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• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.517-524
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• 2017

Cyber security is an important issue in the field of nuclear engineering because nuclear facilities use digital equipment and digital systems that can lead to serious hazards in the event of an accident. Regulatory agencies worldwide have announced guidelines for cyber security related to nuclear issues, including U.S. NRC Regulatory Guide 5.71. It is important to evaluate cyber security risk in accordance with these regulatory guides. In this study, we propose a cyber security risk evaluation model for nuclear instrumentation and control systems using a Bayesian network and event trees. As it is difficult to perform penetration tests on the systems, the evaluation model can inform research on cyber threats to cyber security systems for nuclear facilities through the use of prior and posterior information and backpropagation calculations. Furthermore, we suggest a methodology for the application of analytical results from the Bayesian network model to an event tree model, which is a probabilistic safety assessment method. The proposed method will provide insight into safety and cyber security risks.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1765-1775
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• 2019

Multiple spurious operations (MSOs) mean multiple fire induced circuit faults causing an undesired operation of one or more systems or components. The Nuclear Energy Institute (NEI) of the United States published NEI 00-01 as guidelines for solving MSOs. And this guideline includes MSO scenarios of pressurized water reactor (PWR) and boiling water reactor (BWR). Nuclear power plant operators in U.S. analyzed MSOs under MSO scenarios included in NEI 00-01 and operators of PWRs in Korea also analyzed MSOs under the scenarios of NEI 00-01. As there are no pressurized heavy water reactors (PHWRs) in the United States, MSO scenarios of PHWRs are not included in the NEI 00-01 and any feasible scenarios have not been developed. This paper developed MSO scenarios which can be applied to PHWRs by reviewing the 63 MSO scenarios included in NEI 00-01. This study found that seven scenarios out of the 63 MSO scenarios can be applied and three more scenarios need to be developed.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.501-509
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• 2019

In this paper a body-centered cubic(BCC) crystal plasticity model based on microscopic dislocation mechanism is introduced and numerically implemented. The model is coupled with irradiation effect via tracking dislocation loop evolution on each slip system. On the basis of the model, uniaxial tensile tests of unirradiated and irradiated RPV steel(take Chinese A508-3 as an example) at different temperatures are simulated, and the simulation results agree well with the experimental results. Furthermore, crystal plasticity damage is introduced into the model. Then the damage behavior before and after irradiation is studied using the model. The results indicate that the model is an effective tool to study the effect of irradiation and temperature on the mechanical properties and damage behavior.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2475-2490
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• 2022

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO₂-Zr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1053-1064
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• 2009

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.47-58
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• 2015

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. Thefsensitivity of cooling time before prior to pyro-processing was studied. As the cooling time sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ${\leq}20%$ in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1993년도 하계학술대회 논문집 A
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• pp.140-143
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• 1993

Station Blackout(SBO) event means the complete loss of alternating current(AC) electrical power to the essential & nonessential switchgear buses in a nuclear power plants. Since many safety systems in nuclear power plants depend upon alternative current power, the SBO event could be an important contributor to damage of reactor core. Therefore, the SBO events have been considered as a very important safety issues in a nuclear power plants. In this paper, as evaluating the design characteristics of offsite & emergency power systems, an acceptable minimum SBO duration is calculated. And it is presented that the design method for alternative AC(AAC) sources to cope with the SBO events.

• 한국안전학회지
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• 제36권4호
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• pp.62-70
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• 2021

Representative systematic accident analysis methods proposed so far include AcciMap, STAMP, and FRAM. This study used these three techniques to analyze a fire accident case that occurred during routine manufacturing work in a domestic chemical plant and compared the results. The methods used different approaches to identify the cause of the accident, but they all highlighted similar causal factors. In addition to technical issues, the three accident analysis methods identified factors related to safety education, risk assessment, and the operation of the process safety management system, as well as management philosophy and company culture as problems. The AcciMap and STAMP models play complementary roles because they use hierarchical structures, while FRAM is more effective in analyses centered on human and organizational functions than in technical analyses.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.501-512
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• 2014

HANARO is a multipurpose research reactor located at the Korea Atomic Energy Research Institute (KAERI). Since the commencement of its operation in 1995, various neutron irradiation facilities, such as rabbit irradiation facilities, fuel test loop (FTL) facilities, capsule irradiation facilities, and neutron transmutation doping (NTD) facilities, have been developed and actively utilized for various nuclear material irradiation tests requested by users from research institutes, universities, and industries. Most irradiation tests have been related to national R&D relevant to present nuclear power reactors such as the ageing management and safety evaluation of the components. Based on the accumulated experience as well as the sophisticated requirements of users, HANARO has recently supported national R&D projects relevant to new nuclear systems including the System-integrated Modular Advanced Reactor (SMART), research reactors, and future nuclear systems. This paper documents the current state and utilization of irradiation facilities in HANARO, and summarizes ongoing research efforts to deploy advanced irradiation technology.