• Nuclear Engineering and Technology
• /
• v.28 no.6
• /
• pp.576-582
• /
• 1996

The integrity of reactor vessel support system of the Korean Standard Nuclear Power Plant (KSNPP) is investigated for a postulated reactor vessel closure head drop event. The closure head is disassembled from the reactor vessel during refueling process or general inspection of reactor vessel and internal structures, and carried to proposed location by the head lift rig. A postulated closure head drop event could be anticipated during closure head handling process. The drop event may cause an impact load on the reactor vessel and supporting system. The integrity of the supporting system is directly relevant to that of reactor vessel and reactor internals including fuels. Results derived by elastic impact analysis, linear and non-linear buckling analysis and elasto-plastic stress analysis of the supporting system implied that the integrity of the reactor vessel supporting system is intact for a postulated reactor vessel closure head drop event.

• Progress in Superconductivity and Cryogenics
• /
• v.15 no.4
• /
• pp.53-58
• /
• 2013

KSTAR in-vessel cryo-pump has been installed in the vacuum vessel top and bottom side with up-down symmetry for the better plasma density control in the D-shape H-mode. The cryogenic helium lines of the in-vessel cryo-pump are located at the vertical positions from the vacuum vessel torus center 2,000 mm. The inductive electrical potential has been optimized to reduce risk of electrical breakdown during plasma disruption. In-vessel cryo-pump consists of three parts of coaxial circular shape components; cryo-panel, thermal shield and particle shield. The cryo-panel is cooled down to below 4.5 K. The cryo-panel and thermal shields were made by Inconel 625 tube for higher mechanical strength. The thermal shields and their cooling tubes were annealed in air environment to improve the thermal radiation emissivity on the surface. Surface of cryo-panel was electro-polished to minimize the thermal radiation heat load. The in-vessel cryo-pump was pre-assembled on a test bed in 180 degree segment base. The leak test was carried out after the thermal shock between room temperature to $LN_2$ one before installing them into vacuum vessel. Two segments were welded together in the vacuum vessel and final leak test was performed after the thermal shock. Commissioning of the in-vessel cryo-pump was carried out using a temporary liquid helium supply system.

• Nuclear Engineering and Technology
• /
• v.47 no.1
• /
• pp.66-73
• /
• 2015

Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the element-birth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.

• Nuclear Engineering and Technology
• /
• v.44 no.3
• /
• pp.237-248
• /
• 2012

Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.

• Nuclear Engineering and Technology
• /
• v.32 no.2
• /
• pp.99-107
• /
• 2000

A probabilistic integrity analysis method is presented for a reactor vessel under pressurized thermal shock(PTS) based on Monte Carlo simulation. This method can be applied to the structural integrity assessment of a reactor vessel subjected to pressurized thermal shock where the coolant temperature transient cannot be expressed explicitly as a time function. An axially or circumferentially oriented infinite length surface crack is assumed to be in the beltline weld region of the rector vessel's inside surface. The random variables are the initial crack depth, neutron fluence on the vessel's inside surface, the copper and nickel content of the vessel materials, R $T_{NDT}$ , $K_{IC}$ , and K/aub la/. The reliability of a sample reactor vessel under PTS is assessed quantitatively and the influence of the amount of neutron fluence is also examined by applying the present method.sent method.

• Ocean and Polar Research
• /
• v.37 no.3
• /
• pp.235-247
• /
• 2015

The main purpose of procuring the oceanographic research vessel with state-of-the-art technology is to provide a floating laboratory to conduct field work on the global oceans. The vessel should be properly utilized to locate and evaluate unexplored natural resources as well as to contribute international efforts to better understand and manage global environmental issues. Top priorities in the vessel design are high safety standards, noise and vibration control efficiency, and effective application of research equipment. For the accomplishment of all activities, the vessel length over all should be extended ~100 m with a gross tonnage of ~5,900 ton. In particular, the dynamic positioning system II will essentially operate at sea state 6. The high efficiency emissions reduction system will also be adopted in preparation for entry into force of 3rd exhaust emission control (Tier III). About 130 navigational and scientific instruments will be installed. The final design and model test of the new research vessel were reviewed and completed, respectively, in 2014. Currently, the ship is being built on schedule and expected to be delivered in December 2015. Within the near future, the new vessel will assume the role of carrying out multidisciplinary oceanographic researches of the highest standards in a technologically advanced and environment friendly manner.

• Nuclear Engineering and Technology
• /
• v.41 no.7
• /
• pp.921-928
• /
• 2009

As part of an evaluation for an in-vessel severe accident management strategy, a coolant injection into the reactor vessel under depressurization of the reactor coolant system (RCS) has been evaluated in detail using the SCDAP/RELAP5 computer code. A high-pressure sequence of a small break loss of coolant accident (SBLOCA) has been analyzed in the Optimized Power Reactor (OPR) 1000. The SCDAP/RELAP5 results have shown that safety injection timing and capacity with RCS depressurization timing and capacity are very effective on the reactor vessel failure during a severe accident. Only one train operation of the high pressure safety injection (HPSI) for 30,000 seconds with RCS depressurization prevents failure of the reactor vessel. In this case, the operation of only the low pressure safety injection (LPSI) without a HPSI does not prevent failure of the reactor vessel.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05d
• /
• pp.438-442
• /
• 1996

The design of the vacuum vessel of KT-2(a large-aspect-ratio, mid-size tokamak) is presented. The KT-2 vacuum vessel provides necessary environments to contain a plasma of double-null configuration with elongation of up to 1.8. The vacuum vessel is designed as an all-metal welded structure. Eddy currents are induced on the vessel during all stages of the plasma operation. Influences of the continuous vessel on the plasma were investigated. No significant effect of the vessel on the plasma in every aspect of null formation, plasma initiation, plasma control was found. Stresses and deformations in the vessel by atmospheric pressure and electromagnetic forces due to the eddy currents were calculated using 3D FEM code.

• Journal of the Korean Society of Fisheries and Ocean Technology
• /
• v.53 no.2
• /
• pp.169-176
• /
• 2017

Tuna purse seine fishery in the western and central Pacific Ocean (WCPO) has been rapidly developed since early 1980s due to massive investment of major distant water fishing nations, and catch by purse seine fishery operating in the WCPO accounts for nearly half of the world's tuna total catch. As fishing efficiency is reflected by not only improving of individual vessel's capacity but also increasing number of active vessel, it is essential to understand vessel capacity for reliable assessment result on how fishery affects stock status of target species. In this study, fishing efficiency was analyzed by main factors which are representative of vessel capacity using fishing data and vessel information related to Korean tuna purse seine fishery operating in the western and central Pacific Ocean from 1992 to 2014. It showed that fishing efficiency of vessel tends to be higher when having larger vessel tonnage, higher engine power, lower vessel age and larger length of vessel. As for fishing efficiency by set type, CPUE of associated set with floating objects was generally higher than that of free school set, and CPUE of free school set seemed to have a greater effect on engine power and vessel age compared to other factors.

• Nuclear Engineering and Technology
• /
• v.53 no.2
• /
• pp.532-548
• /
• 2021

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies are needed that can directly monitor severe accident phenomena. This paper presents an interacting multiple model (IMM) based fault detection and diagnosis (FDD) approach for the identification of in-vessel phenomena to provide the accident propagation information using reactor vessel (RV) out-wall temperature distribution during severe accidents in a nuclear power plant. The estimation of wall temperature is treated as a state estimation problem where the time-varying wall temperature is estimated using IMM employing three multiple models for temperature evolution. From the estimated RV out-wall temperature and rate of temperature, the in-vessel phenomena are identified such as core meltdown, corium relocation, reactor vessel damage, reflooding, etc. We tested the proposed method with five different types of SA scenarios and the results show that the proposed method has estimated the outer wall temperature with good accuracy.