• Nuclear Engineering and Technology
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• 제37권6호
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• pp.575-586
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• 2005

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.419-433
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• 2016

One of the important severe accident management measures in the Light Water Reactors is water injection to the reactor core. The related phenomena are investigated by performing experiments and computer simulations. One of the most widely known is the QUENCH test-program. A number of analyses on QUENCH tests have also been performed by different computer codes for code validation and improvements. Unfortunately, any deterministic computer simulation is not free from the uncertainties. To receive the realistic calculation results, the best estimate computer codes should be used for the calculation with combination of uncertainty and sensitivity analysis of calculation results. In this article, the QUENCH-03 and QUENCH-06 experiments are modelled using ASTEC and RELAP/SCDAPSIM codes. For the uncertainty and sensitivity analysis, SUSA3.5 and SUNSET tools were used. The article demonstrates that applying the best estimate approach, it is possible to develop basic QUENCH input deck and to develop the two sets of input parameters, covering maximal and minimal ranges of uncertainties. These allow simulating different (but with the same nature) tests, receiving calculation results with the evaluated range of uncertainties.

• 시스템엔지니어링학술지
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• 제17권1호
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• pp.21-34
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• 2021

The use of Field-Programmable Gate Arrays (FPGAs) in the development of safety-related Human-Machine Interface (HMI) systems has gained much momentum in nuclear applications. Recently, one of the application areas for the Advanced Power Reactor 1400 (APR1400) is in the development of the advanced Component Interface Module (CIM) of the Engineered Safety Features Actuation System (ESFAS). Using systems engineering approach, we have developed a new FPGA-based advanced CIM software. The first step of our software development process involves the Preliminary Hazard Analysis (PHA) based on the previous CIM design. In this paper, we describe the qualitative approach used in performing the preliminary hazard analysis. The paper presents the methodology for applying a modified Hazard and Operability (HAZOP) procedure for the conduct of PHA which resulted in a qualitative risk-ranking scheme that informed the decisions for the safety criteria in the requirements specification phase. The qualitative approach provided the justification for design changes during the advanced CIM software development process.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.339-352
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• 2023

Prediction of the time-related traits of pebble flow inside pebble-bed HTGRs is of great significance for reactor operation and design. In this work, an image-driven approach with the aid of a convolutional neural network (CNN) is proposed to predict the remaining time of initially loaded pebbles and the time interval of paired flow images of the pebble bed. Two types of strategies are put forward: one is adding FC layers to the classic classification CNN models and using regression training, and the other is CNN-based deep expectation (DEX) by regarding the time prediction as a deep classification task followed by softmax expected value refinements. The current dataset is obtained from the discrete element method (DEM) simulations. Results show that the CNN-aided models generally make satisfactory predictions on the remaining time with the determination coefficient larger than 0.99. Among these models, the VGG19+DEX performs the best and its CumScore (proportion of test set with prediction error within 0.5s) can reach 0.939. Besides, the remaining time of additional test sets and new cases can also be well predicted, indicating good generalization ability of the model. In the task of predicting the time interval of image pairs, the VGG19+DEX model has also generated satisfactory results. Particularly, the trained model, with promising generalization ability, has demonstrated great potential in accurately and instantaneously predicting the traits of interest, without the need for additional computational intensive DEM simulations. Nevertheless, the issues of data diversity and model optimization need to be improved to achieve the full potential of the CNN-aided prediction tool.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 1997년도 춘계 학술발표회 논문집 Proceedings of EESK Conference-Fall 1997
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• pp.217-224
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• 1997

The safety related equipments for use in nuclear power plants should be subjected to the seismic qualification in order to insure the safety of the nuclear power plant. This paper summarizes the seismic qualification test on the Low Voltage Motor Control Centers(MCC's) for use in Wolsong Nuclear Power Plants, Units 2, 3 and 4. The seismic qualification test was performed on the two prototype MCC's(a two-bay wide unit for Phase #1 Test and a five-bay wide unit for Phase #2 Test). The specimens were electrically powered and monitored during the test process. It was demonstrated that the specimens possessed sufficient structural and electrical integrity to withstand the required seismic conditions.

• 수산해양교육연구
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• 제29권3호
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• pp.920-931
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• 2017

Since the export of Korean-type APR 1400 in 2009 to the UAE, Korea has been achieved management performance, quality inspections, training, nuclear fuel exports for the nuclear power plant. Despite this apparent growth, there are lacking of the research on the marine transportation of radioactive waste. And the terrible accident at the Japan nuclear power plant in 2011 has caused another reconsideration such as emergency response training and plan, reinforcement of safety regulation. According to the Korean government aims to rebuild the appropriate regulation, training, education that is necessary in order to ensure the safety of marine transportation of radioactive waste. Therefore, this study analyzed the various problems identified by the team of experts for the radioactive waste and marine field, the investigation of relevant legal basis, the need for emergency response training for the person in charge of radioactive waste and suggested the simulation-based interactive curriculum during the process of safety verification related to the marine transport of mid- and low-level radioactive waste generated at the Yeon-ggwang nuclear power(Hanbit) plant in 2015.

• 한국안전학회지
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• 제17권2호
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• pp.92-99
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• 2002

Safety-related motor operated valve(MOV) safety significance for Ulchin Unit 3 was categorized. The safety evaluation of MOV of domestic nuclear power plants affects the generic data used for the quantification of MOV common cause failure(CCF) events in Ulchin Units 3&4 PSA. Therefore, in this paper, MGL(multiple greek letter)parameter ${\beta}$, used for the evaluation of MOV CCF probabilities in Ulchin Units 3&4 probabilistic safety assessment(PSA), was re-estimated and the MOV safety significance was categorized. The re-estimation results of MGL parameter show that the value of(is decreased by 30% compared with the current value used in Ulchin Unit 3&4 PSA. The categorization results of MOV safety significance using the changed value of MGL parameter(show that the number of HSSCs(high safety significant components) is decreased by 54.5% compared with those using the current value of it used in Ulchin Units 3&4 PSA.

• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1175-1184
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• 2022

The pore structure of uranium-bearing sandstone is one of the critical factors that affect the uranium leaching performance. In this article, uranium-bearing sandstone from the Yili Basin, Xinjiang, China, was taken as the research object. The fractal characteristics of the pore structure of the uranium-bearing sandstone were studied using mercury intrusion experiments and fractal theory, and the fractal dimension of the uranium-bearing sandstone was calculated. In addition, the effect of the fractal characteristics of the pore structure of the uranium-bearing sandstone on the uranium leaching kinetics was studied. Then, the kinetics was analyzed using a shrinking nuclear model, and it was determined that the rate of uranium leaching is mainly controlled by the diffusion reaction, and the dissolution rate constant (K) is linearly related to the pore specific surface fractal dimension (D_S) and the pore volume fractal dimension (D_V). Eventually, fractal kinetic models for predicting the in-situ leaching kinetics were established using the unreacted shrinking core model, and the linear relationship between the fractal dimension of the sample's pore structure and the dissolution rate during the leaching was fitted.

• 한국방사선학회논문지
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• 제16권3호
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• pp.327-334
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• 2022

본 연구에서는 방사선(학)과 재학 중 교내 실습의 만족도에 대한 설문과 방사선관계종사자 및 방사선작업종사자의 현황 등을 조사해 방사선(학)과에 대한 원자력안전법의 규제에 대한 실효성 및 타당성에 대한 기초 연구에 목적을 두었다. 방사선(학)과 재학 중 수시출입자로 지정되어 교내 실습 중 방사선 발생장치를 취급 및 조작하지 못한 종사자의 실습 만족도는 만족하지 못한다가 34.62%로 나타났다. 반면 재학 중 방사선작업종사자로 지정되거나 원자력안전법의 규제 이전에 재학하여 방사선 발생장치를 취급 및 조작한 종사자의 실습 만족도는 만족한다가 50%로 나타났다. 또한 교육기관의 방사선작업종사자의 연간 피폭선량은 0.05 mSv 이하로 나타났다. 방사선작업종사자와 방사선관계종사자의 추이를 보면 방사선(학)과를 졸업한 학생들은 의료기관 중에서도 방사선관계종사자로 등록된 진단용 방사선 발생장치를 취급하는 분야로 가장 많은 취업을 한다는 것을 알 수 있다. 따라서 현재의 원자력안전법의 규제를 완화하거나 의료법 및 진단용 방사선 발생장치의 안전관리에 관한 규칙 등을 개정하여 의료기사 등에 관한 법률에서 정한 방사선(학)과 학생들의 학습권을 보장하고 실무 중심의 교육이 이루어질 수 있도록 하는 것이 필요하다고 사료된다.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.293-305
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• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.