• Nuclear Engineering and Technology
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• 제56권6호
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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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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• pp.505-510
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• 2001

Advanced Pressurized Reactor 1400(APR1400), which is a standard evolutionary advanced light water reactor(ALWR), has been developed from 1992 as one of long-term Government Project(G-7). The APR-1400 is designed to operate at the rated output of 4000MWt to produce an electric power output of around 1450MWe. Due to the increased electric power, In Nuclear Power plant huge quantities of heat are generated in the thermo-dynamic process used for producing electrical energy. So, There is considerationly additional cooling, Heat transfer area and increased cooling water of Heat Exchanger which take care of the different smaller cooling duties within the nuclear power plant. We review applying to PRE instead of Shell-and-Tube Heat exchanger. In this paper, we describe the major design features of PRE, Comparison between a PHE and a Shell-and-Tube Heat Exchanger, and then Applicability of Plate Heat Exchanger in Nuclear Power Plant Component Cooling water systems.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.

• 한국의학물리학회지:의학물리
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• 제5권1호
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• pp.75-85
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• 1994

2000년대에 진입하게 되면 월성원자력발전소에서는 4기의 가압중수로형 원자로가 상업발전을 하게 되어서 많은 양의 삼중수소($^{3}$H)가 필연적으로 주변환경에 누출될 것이다. 이러한 방사성 핵종은 삼중수소의 형태로 편재되어 있으면서도 지속성을 갖고 있어서 우리의 환경에 쉽게 분포된다. 삼중수소는 베타방사선량 계측과 보건위해 평가를 위해 독특한 과제를 제시하는 특성을 갖고 있어서 본 논문에서는 삼중수소에 관한 여러가지 문제들을 보건물리와 관련하여 특성과 원천, 신진대사와 선량계측, 미세선량계측, 방사생물, 위해평가, 환경 경로 및 순환 등의 견지에서 정리하였다.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.668-674
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• 2017

Experimental verification of boron species in fuel CRUD (Chalk River Unidentified Deposit) would provide essential and important information about the root cause of CRUD-induced power shifts (CIPS). To date, only bonaccordite and elemental boron were reported to exist in fuel CRUD in CIPS-troubled pressurized water reactor (PWR) cores and lithium tetraborate to exist in simulated PWR fuel CRUD from some autoclave tests. We have reevaluated previous analysis of similar threadlike crystals along with examining some similar threadlike crystals from CRUD samples collected from a PWR cycle that had no indications of CIPS. These threadlike crystals have a typical [Ni]/[Fe] atomic ratio of ~2 and similar crystal morphology as the one (bonaccordite) reported previously. In addition to electron diffraction study, we have applied electron energy loss spectroscopy to determine boron content in such a crystal and found a good agreement with that of bonaccordite. Surprisingly, such crystals seem to appear also on corroded surfaces of Alloy 600 that was exposed to simulated PWR primary water with a dissolved hydrogen level of $5mL\;H_2/kg\;H_2O$, but absent when exposed under $75mL\;H_2/kg\;H_2O$ condition. It remains to be verified as to what extent and in which chemical environment this phase would be formed in PWR primary systems.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• 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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• 제55권9호
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• pp.3164-3182
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• 2023

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

• Corrosion Science and Technology
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• 제12권2호
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• pp.85-92
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• 2013

This work is concerned with an evaluation of dynamic boric acid corrosion (BAC) of low alloy steel for reactor pressure vessel of a pressurized water reactor (PWR). Mockup test method was newly established to investigate dynamic BAC of the low alloy steel under various conditions simulating a primary water leakage incident. The average corrosion rate was measured from the weight loss of the low alloy steel specimen, and the maximum corrosion rate was obtained by the surface profilometry after the mockup test. The corrosion rates increased with the rise of the leakage rate of the primary water containing boric acid, and the presence of oxygen dissolved in the primary water also accelerated the corrosion. From the specimen surface analysis, it was found that typical flow-accelerated corrosion and jet-impingement occurred under two-phase fluid of water droplet and steam environment. The maximum corrosion rate was determined as 5.97 mm/year at the leakage rate of 20 cc/min of the primary water with a saturated content of oxygen within the range of experimental condition of this work.

• Corrosion Science and Technology
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• 제18권5호
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• pp.190-195
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• 2019

SA508 low-alloy steel for a reactor vessel was exposed to primary water in a pressurized water reactor (PWR) plant because the cladding layer of type 309 stainless steel for the RPV was removed, due to an accident in which the detachment of the thermal sleeve occurred. The major advantage of the electrochemical deposition (ECD) Ni plating technique is that the reactor pressure vessel can be repaired without significant thermal effects, and Ni has solid corrosion resistance that can withstand boric acid. The corrosion rate assessment of the damaged part was performed, and its trend was analyzed. Essential variables of the Ni plating for repair of the damaged part were derived. These conditions are applicable variables for the repair plating device, and have been carefully adjusted using the repair plating device. The process for establishing ASME technical standards called Code Case N-840 is described. The process of developing Ni-plating devices, and the electroplating procedure specification (EPS) are described.