Earticle

The Heat Transport system loop stability of CANDU-6 reactor as Wolsong-1 with the CANFLEX fuel bundle has been studied. The SOPHT modelling of the CANFLEX fuel bundle and the ROH interconnection line was made and the stability analysis response of Wolsong-1 reactor with CANFLEX fuel bundle was obtained. The mechanics of the flow instability caused by two phase flow was reviewed. Without the ROH interconnection line the Heat Transport system loop is unstable while the Heat Transport system is stable within ±1 % of nominal flow with the ROH interconnection line

A computer code COSMOS has been developed based on the CARO-D5 for the thermal analysis of LWR UO$_2$ and MOX fuel rod under steady-state and transient operating conditions. The main purpose of the COSMOS, which considers high turnup characteristics such as thermal conductivity degradation with turnup and rim formation at the outer part of fuel pellet, is to calculate temperature profile across fuel pellet and fission gas release up to high burnup. A new mechanistic fission gas release model developed based on physical processes has been incorporated into the code. In addition, the features of MOX fuel such as change in themo-mechanical properties and the effect of microscopic heterogeneity on fission gas release have been also taken into account so that it can be applied to MOX fuel. Another important feature of the COSMOS is that it can analyze fuel segment refabricated from base irradiated fuel rods in commercial reactors. This feature makes it possible to analyze database obtained from international projects such as the MALDEN and RISO, many of which were collected from refabricated fuel segments. The capacity of the COSMOS has been tested with some number of experimental results obtained from the HALDEN, RISO and FIGARO programs. Comparison with the measured data indicates that, although the COSMOS gives reasonable agreement, the current models need to be improved. This work is being performed using database available from the OECD/NEA.

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal-hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal-hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.

중대사고시 LMFBR의 에어로졸(aerosol) 동특성을 살피기 위해 전산코드인 MCAD (Multicomponent Aerosol Dynamics)가 개발되었다. 사고경과에 따른 두 방사능원의 상대적인 충돌확률을 적용하여 에어로졸계를 모사할 수 있다. Brownian 확산과 중력작용에 의한 결합 및 제거과정을 고려했으며, 입자형태를 묘사하기 위해 밀도보정과 형태요소(shape factor)를 동시에 고려하였다. ORNL의 NSPP-300 계열 실험자료와 기존의 코드를MCAD의 입증에 이용하였다. 그 결과 MCAD의 계산치와 실험치 및 기존의 코드 계산값이 일치함을 보여준다. 여러 입력자료의 불화실한 값들을 정의하고, 그들값의 한계로 설정하기 위하여 불확실성 및 민감도해석을 수행하였다. 14개의 입력자료를 선택하여 실험계획법과 Latin hypercube sampling에 의한 입력자료를 조합하여 그 회귀 (regression) 정도를 반응표면 계획법(Response surface method)에 의해 구하였다. 각 변수들의 중요성 및 시간경과에 따른 그들의 상대적인 등위를 결정하기 위하여 단계식 회귀방법 (Stepwise regression method)을 고려했다. LHS에 의한 회귀모형에 Monte Carlo Method를 적용하여 계산값 및 변수들에의 신뢰도를 향상시켰다.

To analyze the aerosol dynamics in severe accidents of LMFBR, a new computer code entitled MCAD (Multicomponent Aerosol Dynamics) has been developed. The code can treat two component aerosol system using relative collision probability of each particles as sequences of accident scenarios. Coagulation and removal mechanisms incorporating Brownian diffusion and gravitational sedimentation are included in this model. In order to see the effect of particle geometry, the code makes use of the concept of density correction factor and shape factors. The code is verified using the experimental result of NSPP-300 series and compared to other code. At present, it fits the result of experiment well and agrees to the existing code. The input variables included are very uncertain. Hence, it requires uncertainty and sensitivity analysis as a supplement to code development. In this analysis, 14 variables are selected to analyze. The input variables are compounded by experimental design method and Latin hypercube sampling. Their results are applied to Response surface method to see the degree of regression. The stepwise regression method gives an insight to which variables are significant as time elapse and their reasonable ranges. Using Monte Carlo Method to the regression model of LHS, the confidence level of the results of MCAD and their variables is improved.

The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction(MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass, The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment.

원자로 냉각 계통의 배관 파열에 근거한 냉각재 상실 사고를 방출계수 0.4에 대하여 분석하였다. 분석은 원자로 냉각계통의 배관 파열에 의하여 발생된 감압부터 노심 복구까지의 전 과도 상태를 포함한다. 계통 열수력과 핵연료 성능 평가를 위하여 BLOWDOWN 단계에서는 RELAP4/MOD6-EM 코드와 RELAP4/MOD6-HOT CHANNEL 코드를 사용하였으며 REFLOOD 단계에서는 RELAP4/ MOD6-FLOOD 코드와 TOODEE2 코드를 각각 사용하였다. LOWER PLENUM 충전을 고려하기 위하여 DOWNCOMER에서 증기-물역방향 유동과 과열벽효과를 근사하여 간단한 해석적 모델이 개발되었다. EOB 발생시의 정보를 근거로 하여 재충전지속 시간과 초기 복구 온도가 계산되었으며 RELAP4/MOD6에 의한 분석결과와 비교하여 상당한 일치를 보였다. 또한, 조기 EOB 발생에 영향을 미치는 계통변수의 연구가 수행되어졌다. DOWNCOMER와 UPPER HEAD사이의 마찰손실이 조기 EOB 발생에 지대한 영향을 미쳤으며 적당한 마찰손실계수의 선택을 통하여 조기 EOB 발생을 방지할 수 있었다. 노심 nodalization이 여섯 개인 경우와 세 개인 경우의 분석 결과가 계통열수력학적 면에서 유사한 결과를 나타내지만, 좋은 결과를 얻기 위하여 전자의 경우가 요구된다.

The loss of coolant accident based on a double-ended cold leg break is analyzed with the discharge coefficient (Ca) of 0.4. This analysis covers the whole transient period from the start of depressurization to the complete refilling of the core by using RELAP4/MOD6-EM and RELAP4/ MOD6-HOT CHANNEL for the system thermal-hydraulics and the fuel performance during the blowdown phase respectively, and RELAP4/MOD6-FLOOD and TOODEE2 during the reflood phase. A simple analytical method has been developed to account for the lower plenum filling by approximating steam-water countercurrent flows and superheated wall effects at the downcomer during the refill period. Based on the informations. at the time of EOB (end-of-bypass), the refill duration time and the initial reflooding temperature were estimated and compared with the results from the RELAP4/MOD6, resulting in a good agreement. In addition, some parametric studies on the EOB were performed. The form loss coefficient between upper head and upper downcomer was found to be sensitive to the occurrence of the spurious EOB. Appropriate form loss coefficients should be taken into account to avoid the flow oscillations at the downcomer. The analyses with the six and three volume core nodalizations, respectively, show much similar trends in the system thermal-hydraulic performance, but the former case is recommended to obtain good results.

원자력 발전소에서 발생하는 공통원인 고장을 분석하기 위한 컴퓨터 코드 COMCAF를 개발하였다. 공통원인 고장을 다룰 때, 먼저 계통의 최소 단절집합들을 공통원인 기본사상들이 고려되지 않은 고장수목으로부터 구한다. 그리고, 공통원인 고장들이 같은 최소 단절집합내의 부품들간에 있는지 또는 서로 다른 최소 단절집합들의 부품들간에 있는지를 고려하여 이들 최소 단절집합들의 발생 확률값을 계산한다. 유사하거나 동일한 부품들간에 공통원인 고장이 있을때는 Basic Para-meter 모델을 사용한다. 그러나, 서로 다른 부품들간에 공통원인 고장이 있을때는 Basic Para-meter모델에 쓰인 Symmetry Assumption을 두개 이상의 부품에 영향을 주는 기본사상들에만 적용한다. Inclusion-Exclusion방법을 사용하여 정점사상확률간을 구한다. 이 경우 같은 최소 단절 집합들에 있는 부품들의 공통원인 고장뿐만아니라 서로 다른 최소 단절집합들에 있는 부품들의 공통원인 고장도 쉽게 고려될 수 있다. 본 연구에서는 이러한 공통원인 고장분석을 가압경수로의 보조 급수계통에 적용하였다. 이들 정점사상의 확률값들을 공통원인 고장이 없는 경우와 비교하였다.

COMCAF, a computer code for the common-cause failure analysis, is developed to treat the common-cause failures in nuclear power plants. In the treatment of common-cause failures, the minimal cut sets of the system are obtained first without changing the fault-tree structure. The occurrence probabilities of the minimal cut sets are then calculated accounting for the common-cause failures among components in the same minimal cut set or in different minimal cut sets. The basic parameter model is used to model the common-cause failures between similar or identical components. For dissimilar components, the assumption of symmetry used in the basic parameter model is applied to the basic events affecting two or more components. The top event probability is evaluated using the inclusion-exclusion method. In addition to the common-cause failures of components in the same minimal cut sets, failures of components in the different minimal cut sets are also easily accounted for by this method. This study applied this common-cause failure analysis to the PWR auxiliary feedwater system. The results in the top event probability for the system are compared with those of no common-cause failures.