• Journal of Ocean Engineering and Technology
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• v.37 no.5
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• pp.205-214
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• 2023

As the regulations on greenhouse gas emissions at sea become strict, efforts are being made to minimize environmental pollutants emitted from fossil fuels used by ships. Considering the large sizes of ships in conjunction with securing stable supplies of environment-friendly energy, interest in nuclear energy to power ships has been increasing. In this study, the neutron irradiation that occurs during the nuclear reactor operation and its effect on the structural responses of the stiffened-plate structures are investigated. This is done by changing the material properties of DH36 steel according to the research findings on the neutron-irradiated steels and then performing the structural response analyses of the structures using analytical and finite-element numerical solutions. Results reveal the influence of neutron irradiation on the structural responses of the structures. It is shown that both the strength and stiffness of the structures are affected by the neutron-irradiation phenomenon as their maximum flexural stress and deflection are increased with the increase in the amount of neutron irradiation. This implies that strength and stiffness need to be considered in the design of ships equipped with marine nuclear reactors.

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.542-550
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• 1996

Nuclear ship reactors have several features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been peformed under rolling, heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removal to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1738-1748
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• 2019

Previous research has shown that the inclined condition has an impact on the natural circulation (natural circulation) mode operation of Floating Nuclear Power Plant (FNPP) mounted on the movable marine platform. Due to its compact structure, small volume, strong maneuverability, the Integral Pressurized Water Reactor (IPWR) is adopted as marine reactor in general. The OTSGs of IPWR are symmetrically arranged in the annular region between the reactor vessel and core support barrel in this paper. Therefore, many parallel natural circulation loops are built between the core and the OTSGs primary side when the main pump is stopped. and the inclined condition would lead to discrepancies of the natural circulation drive head among the OTSGs in different locations. In addition, the flow rate and temperature nonuniform distribution of the core caused by inclined condition are coupled with the thermal hydraulics parameters maldistribution caused by OTSG group operating mode on low power operation. By means of the RELAP5 codes were modified by adding module calculating the effect of inclined, heaving and rolling condition, the simulation model of IPWR in inclined condition was built. Using the models developed, the influences on natural circulation operation by inclined angle and OTSG position, the transitions between forced circulation (forced circulation) and natural circulation and the effect on natural circulation operation by different OTSG grouping situations in inclined condition were analyzed. It was observed that a larger inclined angle results the temperature of the core outlet is too high and the OTSG superheat steam is insufficient in natural circulation mode operation. In general, the inclined angle is smaller unless the hull is destroyed seriously or the platform overturn in the ocean. In consequence, the results indicated that the IPWR in the movable marine platform in natural circulation mode operation is safety. Selecting an appropriate average temperature setting value or operating the uplifted OTSG group individually is able to reduce the influence on natural circulation flow of IPWR by inclined condition.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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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.

• Nuclear Engineering and Technology
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• v.55 no.12
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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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• v.52 no.7
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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.

• Journal of Energy Engineering
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• v.5 no.2
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• pp.146-152
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• 1996

A small power reactor core of 108 MW$\_$th/ was designed with some design constraints: 2 year refueling cycle length, soluble boron free operation, low power density, and proven fuel assembly design - Uljin 3'||'&'||'4 design specifications. CASMO-3 and KINS-3 was used to evaluate operational capability for power level control via control rods. Cycle length, power peaking factor, M.T.C., and power coefficients were also checked. Designed core loaded with KOFAs satisfied all design goals. We found that much more burnable poisons are to be loaded with axial enrichment zoning. Control rod assemblies should be located at every other assemblies with more than 3 banks. Additional shutdown banks are proposed for the safe plant cooldown, which could be located at core periphery.

• Transactions of the Korean Society of Mechanical Engineers
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• v.12 no.3
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• pp.520-527
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• 1988

Liquid ejectors are widely used as marine pumps, inducer stage for the boiler feed water pump, boiler recirculating pump, cooling water recirculating pump in boiling water type nuclear reactor and a deep well pump, because of their high working confidence and simplicity. Furthermore, it requires only a modest net positive suction head for cavitation-free operation and it can be installed in remote location from mechanical power source. It is not easy to presume the friction losses, because it is complicately affected by area ratio, flowrate ratio, nozzle spacing, throat length, shape of liquid ejector and so on. Therefore, the optimization of liquid ejector design is still dependent, to a large extent, on the experimental results and empirical procedures. On the design of the liquid ejector, the area ratio and the nondimensional throat length are the most important design factors among the mentioned above. In this experiment, the effects of the area ratio and the nondimensional throat length to ejector efficiency are carried out systematically by the combination of 4 kinds of motive nozzle and 2 kinds of throat length. In this paper, the present experimental results are compared with the calculated ones by the previous computer aided design program based on one dimensional flow equation. And also, an empirical equation for the working limit of liquid ejector is reported.