• Nuclear Engineering and Technology
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• 제25권2호
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• pp.248-254
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• 1993

본 연구에서는 법규에서 규정하고 있는 주변온도 38$^{\circ}C$의 정상수송조건하에서 수송용기의 건식수송조건에 대한 열해석을 평가하였다. 수송용기는 1회에 PWR 핵연료집합체 4개를 운반할 수 있는 용량을 가지며, 설계기준 핵연료는 연소도 38,000 MWD/MTU, 냉각기간 3년을 기준으로 하였다. 건식수송조건에 대한 열해석을 평가하기 위하여 COBRA-SFS 전산코드를 이용하였다. 수송용기 내부 cavity에 공기, 질소 및 헬륨가스를 채우는 세가지 조건에 대한 해석을 수행하였으며, 최대 핵연료봉의 온도는 수송용기 내부 cavity가 공기인 경우에는 277$^{\circ}C$, 헬륨인 경우에는 226$^{\circ}C$로 계산되었다. 이 값은 건식수송조건에서 수송용기 내부에 장전된 PWR 핵연료집합체가 열적으로 건전성을 유지하기 위한 규정온도보다 낮은 것으로 나타났다.

• 한국방사선학회논문지
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• 제11권7호
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• pp.571-578
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• 2017

원자로 내 사고발생 시 냉각수의 비등으로 기포가 발생하고, 기포율을 측정하기 위하여 열수력 안전 분야에서는 주로 Optical Fiber Probe(OFP)나 광학 카메라를 이용하여 측정하지만 기하학적 구조의 한계로 인해 $17{\times}17$ 배열의 봉 다발 내에 장비를 설치하는 것에는 어려움이 있다. 본 연구는 예비 연구로서 봉 다발에 적용하기 전 X선 시스템과 다양한 모사 팬텀을 이용하여 연구 가능성 평가를 수행하였다. 라디오그라피 및 토모그라피 실험을 통해 X선 발생 장치의 관전압 130 kVp, 관전류 1 mA가 적합하였다. 또한, 기포 해상도 팬텀을 통해 가시적으로 1 mm 크기의 구멍에 대해 측정이 가능하였으며 막대 팬텀을 이용한 대조도 평가의 경우 프레온 내부에서 대조도가 상대적으로 떨어짐을 확인할 수 있었다. 그러나 영상 재구성 시 일그러짐이 없는 좋은 영상을 획득할 수 있었다. 기포 발생 팬텀 실험을 통해 기포의 유동 방향 확인 및 단층 영상을 획득할 수 있었고, Image J 툴을 이용하여 하나의 단층영상에 대해 18 %의 기포율을 측정할 수 있었다. 본 연구는 핵연료 주변 기포율 측정을 위한 선행 연구를 수행하였고 지속적인 연구를 위한 기초 연구로서 활용할 수 있을 것이다.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

• Nuclear Engineering and Technology
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• 제16권2호
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• pp.106-115
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• 1984

원자로 안전한계 설정의 근본목적은 핵연료 및 원자료 계통의 건전성을 보장할 수 있도록 원자로 운전조건을 제한하자는 데 있다. 원자로 보호계통은 원자로 운전변수들이 트립설정치에 도달하게 되면 원자로를 긴급정지시켜 운전조건이 안전한계를 초과하지 못하도록 한다. 따라서 이들 트립설정치의 생산을 위해서는 계산과 측정오차를 충분히 고려해 주어야 한다. 본 기술보고서에서는 웨스팅하우스 원자로 보호계통 트립설정치의 생산에 따른 기본원리와 노심 안전한계의 개발방법 및 트립설정치의 생산절차를 검토하였다. 웨스팅하우스 보호계통 트립설정치의 생산원리는 노심의 안전한계를 계산하고 측정 및 계산에 따른 불확실성을 충분히 고려하여 보수적인 트립설정치를 생산함으로써 핵연료의 용융과 DNB가 발생하지 않도록 하자는 데 있다.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2004년도 학술대회지
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• pp.37-40
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• 2004

The impact/sliding wear tests have been performed in high temperature high pressure water in order to evaluate the effect of spring shape on the wear behavior of a spring supported tube for nuclear fuel fretting study. The results indicate that the tube wear volume and the size of the wear scar are closely related to each spring shape. From the analysis of the wear scar, it is possible to extract the real worn area (Aw) from the size of the wear scar (At). In addition, we found that the wear volume has a linear relation with the real worm area rather than the size of wear scar and this was only determined by each spring shape in the high temperature and pressure water condition. From the above results, it is possible to evaluate the wear resistant spring using the correlation between the variation of the real worn area and the wear behavior at each spring.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.1373-1376
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• 2008

A supporting structure for a long tube bundle of a large diameter is considered in this paper. The primary purpose of the present study is to develop a spacer grid structure for a so-called "dual cooled nuclear fuel", which has been being studied for a nuclear power uprate. The outer diameter of the fuel rod increases considerably from the conventional one. So a completely new shape of the supporting structure (spacer grid) needs to be developed. One of the challenges is to insert a supporting tube into the cross points of the grid straps. To meet a supporting performance, the load vs. displacement characteristics should be obtained. So the present study focuses on the finite element analysis technology to evaluate the characteristics through a parametric study. As a result, major influencing parameters are investigated for an optimized spacer grid design.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.534-539
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• 1997

A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.14-18
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• 2020

Korea Atomic Energy Research Institute (KAERI) has carried out various neutronics experiments in the commissioning stage of the Jordan Research and Training Reactor (JRTR), and this paper introduces the results of first criticality prediction and experiment for the JRTR. The Monte Carlo Code for Advanced Reactor Design and analysis (McCARD) with the ENDF/B-VII.0 nuclear library was used for prediction calculations in the process of the first criticality approach, which was performed to provide reference for the first criticality experiment. In the experiment, fuel loading was carried out by measuring the inverse multiplication factor (1/M) to predict the number of fuel assemblies at the first criticality, and the first critical was reached on April 25, 2016. Comparing the first criticality prediction and experiment, the calculated and measured CAR (Control Absorber Rod) heights for the first criticality were 575 mm and 570.5 mm, respectively, that is, the difference between the two results was approximately 5 mm. From this result, it was confirmed that JRTR manufacturing and various experiments had successfully progressed as designed.

• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1181-1190
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• 2009

Knowing more about the Local Power Density (LPD) at the hottest part of a nuclear reactor core can provide more important information than knowledge of the LPD at any other position. The LPD at the hottest part needs to be estimated accurately in order to prevent the fuel rod from melting in a nuclear reactor. Support Vector Machines (SVMs) have successfully been applied in classification and regression problems. Therefore, in this paper, the power peaking factor, which is defined as the highest LPD to the average power density in a reactor core, was estimated by SVMs which use numerous measured signals of the reactor coolant system. The SVM models were developed by using a training data set and validated by an independent test data set. The SVM models' uncertainty was analyzed by using 100 sampled training data sets and verification data sets. The prediction intervals were very small, which means that the predicted values were very accurate. The predicted values were then applied to the first fuel cycle of the Yonggwang Nuclear Power Plant Unit 3. The root mean squared error was approximately 0.15%, which is accurate enough for use in LPD monitoring and for core protection that uses LPD estimation.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1525-1531
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• 2019

The indigenous nuclear power program of India is based mainly on a series of Pressurised Heavy Water Reactors (PHWRs). A burst correlation for Indian PHWR fuel claddings has been developed and empirical burst parameters are determined. The burst correlation is developed from data available in literature for single-rod transient burst tests performed on Indian PHWR claddings in inert environment. The heating rate and internal overpressure were in the range of 7 K/s-73 K/s and 3 bar-80 bar, respectively, during the burst tests. A burst criterion for inert environment, which assumes that deformation is controlled by steady state creep, has been developed using the empirical burst parameters. The burst criterion has been validated with experimental data reported in literature and the prediction of burst parameters is in a fairly good agreement with the experimental data. The burst criterion model reveals that increasing the heating rate increases the burst temperature. However, at higher heating rates, burst strain is decreased considerably and an early rupture of the claddings without undergoing considerable ballooning is observed. It is also found that the degree of anisotropy has significant influence on the burst temperature and burst strain. With increasing degree of anisotropy, the burst temperature for claddings increases but there is a decrease in the burst strain. The effect of anisotropy in the ${\alpha}$-phase is carried over to ${\alpha}+{\beta}$-phase and its effect on the burst strain in the ${\alpha}+{\beta}$-phase too can be observed.