• Nuclear Engineering and Technology
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• 제27권3호
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• pp.327-335
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• 1995

원자로의 가동중지 중이나. 재장전시 중기 발생기의 세관검사 및 보수작업을 병행하면 원전의 운전정지보수기 간을 현저하게 단축할 수 있다. 이때 원자로가 설치되어 있는 수조의 냉가수가 중기발생기내로 유입되는 것을 막는 장비로써 노즐댐이 있다. 노즐댐의 설치는 고방사선환경과 제한된 공간에서 작업을 해야 하는 특수성 때문에 작업자들이 기피하는 현상을 보인다. 현재 쓰이고 있는 무거운 노즐댐은 노즐댐설치 및 제거작업에 장애가 되는 가장 큰 요인이다. 본 논문에서는 노즐댐의 재질선정과 구조설계를 병행하여 현재 쓰이고 있는 노즐댐보다 가벼우면서도 굽힘강성 대 무게비와 비 강도가 증가된 노즐댐을 설계하였으며, 탄소섬유강화 복합재료로 경량노즐댐을 제작 완료하였다.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.257-264
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• 1994

원자력발전소 중대사고시 용융된 노심과 잔류냉각수가 증기폭발을 일으켜 원자로 격납용기의 건전성을 위협할 수 있다. 본 연구에서는 증기폭발을 모사할 수 있는 실험 장치를 제작하고, 물과 프레온을 사용하여 증기폭발실험을 수행하였다. 이때 고속카메라를 사용하여 폭발현상을 관측하였고, 동압측정기와 압력분출관을 이용하여 생성되는 폭발압력과 기계적인 에너지를 계측하였다. 이를 토대로 증기폭발의 중요인자들(물의 온도, 물의 주입속도, 물의 주입 시간, 그리고 냉매의 깊이)에 대한 민감도 분석을 수행하였다. 그리고, 압력용기 바닥의 구조물이 용융/냉각재의 반응에 미치는 영향을 살펴보기위하여 실험용기 내부에 그리드를 설치하여 폭발실험을 실시하였다. 물/프레온의 폭발실험에서 계측된 기계적에너지를 이용한 에너지효율은 0.5∼l.6%인 것으로 계산되었다.

• Nuclear Engineering and Technology
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• 제19권3호
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• pp.198-204
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• 1987

발전소 과도현상과 비냉각재 상실사고를 모의할 수 있는 가압경수로발전소 모의코드 MCSIM을 개발하였다. 원자로 냉각재계통은 에너지 방정식과 운동량 방정식을 분리 취급하면서 Drift Flux 2상 유동모델, 적분 운동량 방정식 등을 사용하여 모델링하였다. 증기발생기의 모사는 Pot Boiler 모델을 사용하였고, 2차계통을 위해서는 분리 취급된 정상상태 에너지 방정식과 운동량방정식을 핵출력 계산을 위해서는 점 동특성 방정식을 사용하였다. 현재의 코드성능을 시험하기 위해 완전 냉각재 유동상실사고와 제어봉 집합체 인출 사고를 계산하여 그 결과를 원자력 5/6호기 최종 안전 보고서의 결과와 비교하였다.

• 한국소음진동공학회논문집
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• 제20권3호
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• pp.215-222
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• 2010

Long and slender body, like a fuel rod, oscillating in axial flow can be unstabilized even by the small cross flow which can be activated by the flow mixer or turbulent generator. It is important to include these effects of flow disturbance in dynamic stability analysis of nuclear fuel rod. This work shows how eigen frequency of a multi-span fuel rod can be changed by the swirl flow, which is discretely generated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was calculated. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.382-395
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• 2002

An experimental study on transient critical heat flux (CHF) under flow coastdown has been performed for the water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady-state CHF The transient CHF experiments have been performed for three kinds of flow transient modes based on the coastdown data of a nuclear power plant reactor coolant pump. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to- CHF becomes large as the heat flux decreases. The critical mass flux has the largest value for slow flow reduction rate. There is a pressure effect on the ratio of the transient CHF data to steady-state CHF data. Except under low system pressure conditions, the flow transient CHF was revealed to be conservative compared with the steady-state CHF data. Bowling CHF correlation and thermal hydraulic system code MARS show promising results for the prediction of CHF occurrence .

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.908-917
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• 2020

Corrosion of structural materials presents a critical challenge in the use of lead-bismuth eutectic (LBE) as a nuclear coolant in an accelerator-driven system. By forming a protective layer on the steel surfaces, corrosion of steels in LBE cooled reactors can be mitigated. The amount of oxygen concentration required to create a continuous and stable oxide layer on steel surfaces is related to the oxidation process. So far, there is no oxidation experiment in fuel assemblies (FA), let alone specific oxidation detail information. This information can be, however, obtained by numerical simulation. In the present study, a new coupling method is developed to implement a coupling between the oxygen mass transfer model and the commercial computational fluid dynamics (CFD) software ANSYS-CFX. The coupling approach is verified. Using the coupling tool, we study the oxidation process of the FA and investigate the effects of different inlet parameters, such as temperature, flow rate on the mass transfer process.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1127-1133
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• 2021

A Spent Fuel Pool (SFP) is designed to store spent fuel assemblies in the pool. And, a SFP cooling and cleanup system cools the SFP coolant through a heat exchanger which exchanges heat with component cooling water. If the cooling system fails or interfacing pipe (e.g., suction or discharge pipe) breaks, the cooling function may be lost, probably leading to fuel damage. In order to prevent such an incident, it is required to properly cool the spent fuel assemblies in the SFP by either recovering the cooling system or injecting water into the SFP. Probabilistic safety assessment (PSA) is a good tool to assess the SFP risk when an initiating event for the SFP occurs. Since PSA has been focused on reactor-side so far, it is required to study on the framework of PSA approach for SFP and identify the key factors in terms of fuel damage frequency (FDF) through a case study. In this study, therefore, a case study of SFP-PSA on the basis of design information of APR-1400 has been conducted quantitatively, and several sensitivity analyses have been conducted to understand the impact of the key factors on FDF.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1125-1134
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• 2024

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19-and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2099-2112
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• 2024

This study examined the effects of CRUD on reflood heat transfer behaviors of nuclear fuel rods during a loss-of-coolant-accident (LOCA) in a pressurized water reactor using a best-estimate thermal-hydraulic analysis code. Changes in thermal properties and boiling heat transfer characteristics of the CRUD layer were extensively reviewed, and a set of correction factors to reflect the changes was implemented into the code. A heat structure layer reflecting the effects of CRUDs on the properties was added to the outer surface of the fuel cladding. Numerical simulations were conducted to examine the effects of CRUDs on reflood cooling of overheated fuel rods for representative separate and integral effect tests, FLECHT-SEASET and LOFT. In LOFT analysis, the average cladding temperature was increased due to the low thermal conductivity of CRUD during steady-state operation; however, in both analyses, the peak cladding temperature decreased, and the quenching time was reduced. Obtained results revealed that when the porous CRUD layer is deposited on the fuel cladding, two opposite effects appear. Low thermal conductivity of the CRUD layer always increases fuel temperature during normal operation; however, its hydrophilic porous structures may contribute to accelerated reflood cooling of fuel rods during a LOCA.

• 대한산업공학회지
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• 제26권1호
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• pp.54-65
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• 2000

The objective of this study is to evaluate the operator performance relating to hand switches with two or three buttons in the main control rooms of nuclear power plants. Based on the comparative analysis of the nuclear power plants, two different subjective workload-rating scales were used to evaluate the performance of 48 operators: the Overall Workload(OW) and National Aeronautics and Space Administration Task Load Index (NASA TLX). The survey questions consisting of the eight-items were asked to evaluate the operating experiences for the two different switch types. The OW scales ratings were applied to measure the workload of the switch-related tasks. The ratings revealed that signal detection tasks caused less workload in the three-buttoned-switch operators than the other switch group. However, in the switch operation tasks, the switch types did not show statistically significant effects on workload level. The NASA TLX scale ratings were performed based on detailed task scenarios that assumed the accident of small break loss of coolant, what we call, the small LOCH. The NASA TLX was administered to three different task groups: the reactor, the turbine, and the electric operator groups. Based on the NASA TLX, the two-buttoned switch groups showed higher workload than those with the three-buttoned switches. However, a statistically significant difference was found only in the reactor operator groups. When the current switch type was assumed to be changed for the other type, all of the three-buttoned switch groups were predicted to have higher workload than the other switch groups, respectively. The implications of these findings were discussed.