• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1955-1962
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• 2020

In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.

• Journal of Advanced Marine Engineering and Technology
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• 제10권2호
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• pp.156-164
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• 1986

In the case of boiling on high temperature wall, vapor film covers fully or parcially the surface. This phenomenon, film boiling or transition boiling, is very important in the surface heat treatment of metal, design of cryogenic heat exchanger and emergency cooling of nuclear reactor. Mainly supposed hydraulic-thermal accidents in nuclear reactor are LCCA (Loss of Coolant Accident) and PCM (Power-Cooling Mismatch). Recently, world-wide studies on reflooding of high temperature rod bundles after the occurrence of the above accidents focus attention on wall temperature history and required time in transient cooling process, wall superheat at rewet point, heat flux-wall superheat relationship beyond the transition boiling region, and two-phase flow state near the surface. It is considered that the further systematical study in this field will be in need in spite of the previous results in ref. (2), (3), (4). The paper is the study about the fast transient cooling process following the wall temperature excursion under the CHF (Critical Heat Flux) condition in a forced convective subcooled boiling system. The test section is a vertically arranged concentric annulus of 800 mm long and 10 mm hydraulic diameter. The inner tube, SUS 304 of 400 mm long, 8 mm I.D, and 7 mm O.D., is heated uniformly by the low voltage AC power. The wall temperature measurements were performed at the axial distance from the inlet of the heating tube, z=390 mm. 6 chromel- alumel thermocouples of 76 .mu.m were press fitted to the inner surface of the heating tube periphery. To investigate the heat transfer characteristics during the fast transient cooling process, the outer surface (fluid side) temperature and the surface heat flux are computed from the measured inner surface temperature history by means of a numerical method for inverse problems of transient heat conduction. Present cooling (boiling) curve is sufficiently compared with the previous results.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1514-1524
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• 2019

Hydrogen-steam gas mixture may be injected into containment with flow regime varying both spatially and transiently due to wall effect and pressure difference between primary loop and containment in severe accidents induced by loss of coolant accident. Preliminary CFD analysis is conducted to gain information about the helium flow regime transition process from jet to buoyancy plume for forthcoming experimental study. Physical models of impinging jet and wall condensation are validated using separated effect experimental data, firstly. Then helium transportation is analyzed with the effect of jet momentum, buoyancy and wall cooling discussed. Result shows that helium distribution is totally dominated by impinging jet in the beginning, high concentration appears near gas source and wall where jet momentum is strong. With the jet weakening, stable light gas layer without recirculating eddy is established by buoyancy. Transient reversed helium distribution appears due to natural convection resulted from wall cooling, which delays the stratification. It is necessary to concern about hydrogen accumulation in lower space under the containment external cooling strategy. From the perspective of experiment design, measurement point should be set at the height of connecting pipe and near the wall for stratification stability criterion and impinging jet modelling validation.

• 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.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.727-734
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• 2012

A new advanced safety feature of DVI+ (Direct Vessel Injection Plus) for the APR+ (Advanced Power Reactor Plus), to mitigate the ECC (Emergency Core Cooling) bypass fraction and to prevent switching an ECC outlet to a break flow inlet during a DVI line break, is presented for an advanced DVI system. In the current DVI system, the ECC water injected into the downcomer is easily shifted to the broken cold leg by a high steam cross flow which comes from the intact cold legs during the late reflood phase of a LBLOCA (Large Break Loss Of Coolant Accident)For the new DVI+ system, an ECBD (Emergency Core Barrel Duct) is installed on the outside of a core barrel cylinder. The ECBD has a gap (From the core barrel wall to the ECBD inner wall to the radial direction) of 3/25~7/25 of the downcomer annulus gap. The DVI nozzle and the ECBD are only connected by the ECC water jet, which is called a hydrodynamic water bridge, during the ECC injection period. Otherwise these two components are disconnected from each other without any pipes inside the downcomer. The ECBD is an ECC downward isolation flow sub-channel which protects the ECC water from the high speed steam crossflow in the downcomer annulus during a LOCA event. The injected ECC water flows downward into the lower downcomer through the ECBD without a strong entrainment to a steam cross flow. The outer downcomer annulus of the ECBD is the major steam flow zone coming from the intact cold leg during a LBLOCA. During a DVI line break, the separated DVI nozzle and ECBD have the effect of preventing the level of the cooling water from being lowered in the downcomer due to an inlet-outlet reverse phenomenon at the lowest position of the outlet of the ECBD.

• Nuclear Engineering and Technology
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• 제24권1호
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• pp.86-97
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• 1992

이 논문의 목적은 C-E System 80+ 원자로에서의 직접용기주입 설계를 가압 열 충격의 견지에서 평가하는 것이다. 영의 출력에서의 주증기관 파단과 0.05 ft$^2$면적의 소형파단 냉각재상실사고가 가능성있는 가압열충격 사고로 선정되었다. 원자로 다운카머 영역에서의 유체 성층효과를 예측하기 위하여 주증기관 파단사고에 대하여는 COMMIX-IB 전산코드를, 그리고 0.05 ft$^2$소형파단 냉각재상실사고에 대하여는 REMIX 전산코드를 사용하여 유체혼합해석이 수행되었다. 압력과 온도의 과도변화를 받는 원자로용기 벽내의 응력분포는 두 사고에 대하여 OCA-P전산코드를 사용하여 계산되었다. 해석결과, 붕괴열의 고려가 없는 소형파단 냉각재 상실사고의 경우 용기내 균열발생의 가능성이 있으나 붕괴열을 고려하면 용기의 수명기간중 균열발생의 가능성은 없다.

• 에너지공학
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• 제14권1호
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• pp.54-61
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• 2005

한국형 신형원자로1400(APR1400)은 3983MWt급의 2×4 루프 개량형 가압경수로(PWR)로서 대형 냉각재상실사고 발생시 안전주입수의 원자로용기 직접주입(DVI) 방식을 채택하고 있으며, 안전주입수탱크(SIT) 내부에 유량조절기(Fluidic Device, FD)를 장착하고 있다. 본 연구에서는 신형원자로 1400의 안전주입계통에 새로이 적용된 주요 특징 중 하나인 유량조절기에 대하여 최적안전해석코드인 TRAC-M/F90, 3.782버전을 이용한 성능평가 및 민감도 분석을 수행하였다. 연구결과 유량조절기가 안전주입수의 원자로 유입을 적절하게 조절하고 있음을 확인하였으며, 안전주입수탱크 내부의 압축질소체적 감소가 안전 주입수체적 감소에 비하여 노심의 급냉 완료 시점을 빠르게 하였다. 또한 안전주입계통의 전체 저항계수(K factor)가 최소 또는 최대일 때 노심의 급냉 완료 시점은 평균값인 경우보다 다소 늦어졌으나, 피복재 최고온도(PCT)는 상대적으로 큰 차이가 발생하지 않았다.

• Nuclear Engineering and Technology
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• 제14권4호
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• pp.196-203
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• 1982

본 논문에서는 고리 1호기의 기사용 핵연료 집합체를 수송하기 위한 Cask를 설계하였다. 이를 위하여 고리 1호기의 기사용 핵연료 집합체로부터 방출되는 감마선과 중성자를 계산하여 MORSE 및 ANISN전산 코드로써 차폐 계산을 수행하였다. 그 결과, 9개의 집합체를 동시에 수송할 수 있는 Steel Cask가 가장 적합하다는 것을 밝혔다. 이 Steel Cask에 대한 안전성을 평가하기 위하여 연료봉의 중심 온도와 복재온도를 계산하여 핵연료의 용융점보다 훨씬 낮음을 증명하였다. 또한 KENO와 MORSE전산 코드를 사용하여 임계도 계산을 수행하여 미임계 상태임을 증명하였다. 이로써 9개의 기사용 핵연료 집합체를 동시에 수송할 수 있는 Steel Cask를 간단히 설계하였다.

• Nuclear Engineering and Technology
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• 제13권3호
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• pp.121-129
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• 1981

많은 기존 공식중 세계의 가장 우수한 최대 유량공식을 찾아, 그 세개의 한계유량공식의 개요와, 냉각재상실사고시 격납용기격리에 실패하는 경우를 특정지을 수 있는 구멍의 크기와 격납용기의 압력 및 온도 등이 주어진 상태하에서 격납용기로부터 외부대기로 방출되는 개략적인 핵분열생성물의 양을 추정하기 위한 계산절차를 제시하였다. 이상기체의 임계유량공식과 이상유(two-phase flow)의 최대유량을 산출하기 위한 무디(Moody)의 도표를 이용하여 계산실예를 제시하였으며, 그 결과를 콘뎀프트-앨티(CONTEMPT-LT) 전산코드의 질량유출공식을 콘버징 노즐(converging nozzle)을 통과하는 음속류(sonic flow)의 경우에 적용하여 산출한 값과 비교하여 보았다. 이리하여 이상 기체의 임계유량공식은 무디(Moody)의 공식이 주는 값과 거의 비슷한 결과를 줌을 입증하였다. 또한 냉각재상실 사고시 격납용기로부터의 유출율을 추정하기 위해서는 콘템프트-앨티(CONTEMPT-LT)의 질량유출공식을 사용하는 것보다 이상 기체의 최대유량공식을 사용하는 것이 더 보수적인 방법임을 보여 주었다.

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.701-709
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• 1995

RELAP5 /MOD3/KAERl의 임계유동모델을 위한 실제적인 배출계수들을 9개의 MARVIKEN 임계유동실험 의 평가계산을 통하여 과냉각과 이상임계유동에 대하여 구하였다. 선택된 실험에는 높은 초기 과냉각도와 큰 노즐 세 장비(L/D)인 것들이 포함되었다. 코드의 평가결과는 RELAP5/MOD3/KAERI은 과냉각임계유동을 크게 예측하고 이 상임계유동은 작게 예측함을 보이고 있다. 이러한 결과들을 이용하여 임계유동모델의 실제적인 배출계수들을 반복법으로 정량화 하였다. 실제적인 배출계 수는 과냉각임계유동이 0.89 그리고 이상임계유동이 1.07로 결정되었으며 관련 표준편차는 각 각 0.0349과 0.1189이다. 본 연구로부터 얻어진 결과는 대형냉각재 상실사고의 실제적인 계통반응 계산과 비상노심냉각계통 성능평가에 적용할 수 있다.