• Nuclear Engineering and Technology
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• 제7권1호
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• pp.53-68
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• 1975

The world's nuclear powered ships have been reviewed mainly based on development of the marine nuclear reactor and the present trends of nuclear warships and merchant ships in the world. In particular, emphasis was on the four nonmilitary nuclear ships, Russian Ice breaker Lenin, American Cargo-passenger ship Savannah, German Ore carrier Otto Hahn, and Japanese Cargo ship Mutsu. They are the only civilian nuclear ships which have entered service at the present time in the world. The nuclear fleets in United States, United Kingdom, Soviet, and France were described in view of historical development and the present stock of the nuclear ships. The present projects and the future trends for the nuclear merchant ships in the main shipbuilding countries have been also discussed. The nuclear fission and reactor were briefly discussed in the beginning of this article.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.275-282
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• 2024

Lead bismuth eutectic (LBE) is used as coolant for MicroURANUS, a small marine nuclear power plant, and this coolant is transported in the plant by an electromagnetic pump. Given the considerable heat generated by the electromagnetic pump, the cooling of the pump is essential. This study compared air cooling and water-cooling methods and found that the maximum temperatures during air and water cooling were 640 K and 372 K, respectively. These findings were utilized to design an electromagnetic pump with water-cooling. The maximum temperature of the pump was lower than the boiling point of water; thus, the pump did not require a separate pressurization. Consequently, the resistance problem of the coil and the deformation problem of the material caused by generated heat can be solved through water-cooling.

• Journal of Advanced Marine Engineering and Technology
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• 제28권5호
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• pp.827-836
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• 2004

The design parameter of the heat loss for the pressurized water reactor has been studied. The heat loss from the reactor vessel through the air gap. insulation are analysed by using the computational fluid dynamics code, FLUENT. Parametric study has been performed on the air gap width between the reactor vessel wall and the inner surface of the insulation, and on the insulation thickness. Also evaluated is the performance degradation due to the chimney effect due to gaps left between the panels during the installation of the insulation system. From the analysis results, the optimal with of air gap and insulation thickness and the value of heat loss are obtained The results show how the heat loss varies with the air gap width and insulation thickness. The temperature and the velocity distributions are also presented. From the results of the evaluation. the optimal air gap width and the optimal insulation thickness are obtained. As the difference between the predicted heat loss and measured heat loss from the reactor vessel is construed Primarily as losses due to chimney effect. the contribution of the chimney effect to the total heat loss is quantitatively indicated.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.514-519
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• 1997

The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control roe drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation.

• 한국표면공학회지
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• 제40권5호
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• pp.234-240
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• 2007

As a reactor coolant pump (RCP) is operated in the nuclear power system for a long time, so its surface is continuously contaminated by radioactive scales. In order to maintain for RCP internals, a special chemical decontamination process should be used to reduce the radiation from the RCP surface. In this study, applicable possibility in chemical decontamination for RCP was investigated for the various stainless steels. The stainless steel (STS) 304 showed the best electrochemical properties for corrosion resistance and the lowest weight loss ratio in chemical decontamination process model 3-1 than other materials. However, the pitting corrosion was generated in both STS 415 and STS 431 with the increasing numbers of cycle. The intergranular corrosion in STS 415 was sporadically observed. The sizes of their pitting corrosion were also increased with increasing cycle numbers.

• Journal of Advanced Marine Engineering and Technology
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• 제31권1호
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• pp.84-94
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• 2007

As a reactor coolant pump(RCP) is operated in the nuclear power system for a long time. so its surface is continuously contaminated by radioactive scales. In order to perform regular or emergency repair about RCP internals a special decontamination process should be used to reduce the radiation from the RCP surface by means of chemical cleaning. In this study, applicable possibility in chemical decontamination for RCP was investigated on the various materials. The STS 304 showed the best electrochemical properties for corrosion resistance than other materials. However, the pitting corrosion was slightly generated in both STS 415 and STS 431 with the increasing numbers of cycle and intergranular corrosion were sporadically observed. The size of their pitting corrosion and intergranular corrosion were also increased with increasing cycle numbers.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4335-4349
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• 2023

Nuclear marine propulsion has been emerging as a next generation carbon-free power source, for which proper passive residual heat removal systems (PRHRSs) are needed for long-term safety. In particular, the characteristics of unlimited operation time and compact design are crucial in maritime applications due to the difficulties of safety aids and limited space. Accordingly, a compact and long-term-operable PRHRS has been proposed with the key design concept of using both air cooling and seawater cooling in tandem. To confirm its feasibility, this study conducted system design and a transient analysis in an accident scenario. Design results indicate that seawater cooling can considerably reduce the overall system size, and thus the compact and long-term-operable PRHRS can be realized. Regarding the transient analysis, the Multi-dimensional Analysis of Reactor Safety (MARS-KS) code was used to analyze the system behavior under a station blackout condition. Results show that the proposed design can satisfy the design requirements with a sufficient margin: the coolant temperature reached the safe shutdown condition within 36 h, and the maximum cooling rate did not exceed 40 ℃/h. Lastly, it was assessed that both air cooling and seawater cooling are necessary for achieving long-term operation and compact design.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1429-1435
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• 2020

Applying an accurate parametric prediction model to identify abnormal or false pressurizer water levels (PWLs) is critical to the safe operation of marine pressurized water reactors (PWRs). Recently, deep-learning-based models have proved to be a powerful feature extractor to perform high-accuracy prediction. However, the effectiveness of models still suffers from two issues in PWL prediction: the correlations shifting over time between PWL and other feature parameters, and the example imbalance between fluctuation examples (minority) and stable examples (majority). To address these problems, we propose a cost-sensitive mechanism to facilitate the model to learn the feature representation of later examples and fluctuation examples. By weighting the standard mean square error loss with a cost-sensitive factor, we develop a Cost-Sensitive Long Short-Term Memory (CSLSTM) model to predict the PWL of PWRs. The overall performance of the CSLSTM is assessed by a variety of evaluation metrics with the experimental data collected from a marine PWR simulator. The comparisons with the Long Short-Term Memory (LSTM) model and the Support Vector Regression (SVR) model demonstrate the effectiveness of the CSLSTM.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1388-1395
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• 2017

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal-hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal-hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.

• Journal of Advanced Marine Engineering and Technology
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• 제40권2호
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• pp.112-119
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• 2016

Many studies have been conducted on the performance of a passive autocatalytic recombiner (PAR), but not many have focused on the locations where the PAR is installed. During a severe accident in a nuclear reactor containment, a large amount of hydrogen gas can be produced and released into the containment, leading to hydrogen deflagration or a detonation. A PAR is a hydrogen mitigation method that is widely implemented in current and advanced light water reactors. Therefore, for this study, a PAR was installed at different locations in order to investigate the difference in hydrogen reduction rate. The results indicate that the hydrogen reduction rate of a PAR is proportional to the distance between the hydrogen induction location and the bottom wall.