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

• 제어로봇시스템학회:학술대회논문집
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• 2001.10a
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• pp.133.6-133
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• 2001

Reactor vessel is one of the most important equipment of Nuclear Power Plant (NPP) with regard to the nuclear safety. Thus reactor vessel must be examined periodically by certified experts. Currently, ultra-sonic(UT) non-destructive inspection is executed on reactor vessel. Two different techniques are used in this inspection. One is using the movable manipulator fixed with the support-guide placed on the vessel, and the other is using mobile robot moving in the vessel. Movable manipulator machine is very heavy, hard to handle, and very expensive. Mobile robot equipment is small and convenient but has a weak point on positional precision. To solve these problems we developed a reactor inspection system based on laser-driven mobile robot. This paper describes the main concept and structure of integrated inspection units and the feature of implemented units.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2375-2387
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• 2024

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.16-23
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• 2018

In order to investigate the ways to enhance the seismic performance of APR1400 seismic fragile equipment by direct design changes, four equipment such as Reactor Vessel Support, Integrated Head Assembly, Remote Shutdown Console, and Pressurizer are reviewed using information of the main dimensions, seismic stress evaluation results, design FRS, etc. in this paper. In addition to the direct reinforcement of equipments, the feasibility of seismic isolation for the safety related cabinet is also investigated and the actual adaption plan of a commercial spring-damper system is briefly reviewed.

• Structural Engineering and Mechanics
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• v.43 no.4
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• pp.475-502
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• 2012

In this paper, the seismic analysis modeling technologies for sodium-cooled fast reactor (SFR) are presented with detailed descriptions for each structure, system and component (SSC) model. The complicated reactor system of pool type SFR, which is composed of the reactor vessel, internal structures, intermediate heat exchangers, primary pumps, core assemblies, and core support structures, is mathematically described with simple stick models which can represent fundamental frequencies of SSC. To do this, detailed finite element analyses were carried out to identify fundamental beam frequencies with consideration of fluid added mass effects caused by primary sodium coolant contained in the reactor vessel. The calculation of fluid added masses is performed by detailed finite element analyses using FAMD computer program and the results are discussed in terms of the ways to be considered in a seismic modeling. Based on the results of seismic time history analyses for both seismic isolation and non-isolation design, the functional requirements for relative deflections are discussed, and the design floor response spectra are proposed that can be used for subsystem seismic design.

• Journal of Radiation Protection and Research
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• v.32 no.4
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• pp.158-167
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• 2007

It is required that the neutron dosimetry be present to monitor the reactor vessel throughout its plant life. The Ex-vessel Neutron Dosimetry Systems which consist of sensor sets, radiometric monitors, gradient chains, and support hardware have been installed for 3-Loop plants after a complete withdrawal of all six in-vessel surveillance capsules. The systems have been installed in the reactor cavity annulus in order to characterize the neutron energy spectrum over the beltline region of the reactor vessel. The installed dosimetry were withdrawn and evaluated after a irradiation during one cycle and then compared to the cycle specific neutron transport calculations. The reaction rates from the measurement and calculation were compared and the results show good agreements each other.

• Structural Engineering and Mechanics
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• v.51 no.6
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• pp.1005-1016
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• 2014

Roles of reactor internals are to support nuclear fuel, provide insertion and withdrawal channels of nuclear fuel control rods, and carry out core cooling. In case of functional loss of the reactor internals, it may lead to severe accidents caused by damage of nuclear fuel assembly and deterioration of reactor vessel due to attack of fallen out parts. The present study is to examine fluid flows in reactor internals subjected to hydrodynamic loads. In this context, an integrated model was developed and applied to two kinds of numerical analyses; one is to analyze periodic loading effect caused by pump pulsation and the other is to analyze random loading effect employing different turbulent models. Acoustic pressure distributions and flow velocity as well as pressure and temperature fields were calculated and compared to establish appropriate analysis techniques.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4014-4021
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• 2021

If safety injection systems (SISs) do not work in the event of a loss-of-coolant accident (LOCA), the accident can progress to a severe accident in which the reactor core is exposed and the reactor vessel fails. Therefore, it is considered that a technology that provides recoverable maximum time for SIS actuation is necessary to prevent this progression. In this study, the corresponding time was defined as the golden time. To achieve the objective of accurately predicting the golden time, the prediction was performed using the deep fuzzy neural network (DFNN) with rule-dropout. The DFNN with rule-dropout has an architecture in which many of the fuzzy neural networks (FNNs) are connected and is a method in which the fuzzy rule numbers, which are directly related to the number of nodes in the FNN that affect inference performance, are properly adjusted by a genetic algorithm. The golden time prediction performance of the DFNN model with rule-dropout was better than that of the support vector regression model. By using the prediction result through the proposed DFNN with rule-dropout, it is expected to prevent the aggravation of the accidents by providing the maximum remaining time for SIS recovery, which failed in the LOCA situation.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.2
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• pp.179-184
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• 2009

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.