• 방사성폐기물학회지
• /
• 제19권2호
• /
• pp.205-214
• /
• 2021

Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.

• 한국안전학회지
• /
• 제20권2호
• /
• pp.32-37
• /
• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.

• 소음진동
• /
• 제5권4호
• /
• pp.515-524
• /
• 1995

Tubes in heat exchanger of fuel rods in reactor core are supported at intemediate point by support p0lates or springs. Current practice is, in case of heat exchanger, to allow clearance between tube and support plate for design and manufacturing consideration. And in case of fuel rod the clearance in support point can be generated due to the support spring force relaxation. Flow-induced vibration of a tube can cause it to impact or rub against support plate or against adjacent tubes and can result in fretting-wear. The tube-to- support dynamic interaction is used to relate experimental wear data from single-span test rigs to real multi-span heat exchanger configurations. The dynamic interaction cna be measured during experimental wear tests. However, the dynamic interaction is difficult to measure in real heat exchangers and, therefore, analytical techniques are required to estimate this interaction. This paper describels the nonlinear impact model of DAGS(Dynamic Analysis of Gapped Structure) code which simulates the tube response to external sinusodial or step excitation and predicts tube motion and tube-to-support dynamic interaction. Three experimental measurements-two single span rods excited by sinusodial force and a two span rod impacted by a steel ball are compared from the simulation nonlinear model of DAGS code. The simulation results from DAGS code are in good agreement with measurements. Therefore, the developed model of DAGS code is good analytical tool for estimating tube-to-support dynamic interaction in real heat exchangers.

• Nuclear Engineering and Technology
• /
• 제54권3호
• /
• pp.842-848
• /
• 2022

Parametric studies of heat transfer and fluid flow are very important research of interest because the design and operation of fluid flow and heat transfer systems are guided by these parametric studies. The safety of the system operation and system optimization can be determined by decreasing or increasing particular fluid flow and heat transfer parameter while keeping other parameters constant. The parameters that can be varied in order to determine safe and optimized system include system pressure, mass flow rate, heat flux and coolant inlet temperature among other parameters. The fluid flow and heat transfer systems can also be enhanced by the presence of or without the presence of particular effects including gravity effect among others. The advanced Generation IV reactors to be deployed for large electricity production, have proven to be more thermally efficient (approximately 45% thermal efficiency) than the current light water reactors with a thermal efficiency of approximately 33 ℃. SCWR is one of the Generation IV reactors intended for electricity generation. High Performance Light Water Reactor (HPLWR) is a SCWR type which is under consideration in this study. One-eighth of a proposed fuel assembly design for HPLWR consisting of 7 fuel/rod bundles with 9 coolant sub-channels was the geometry considered in this study to examine the effects of system pressure and mass flow rate on wall and fluid temperatures. Gravity effect on wall and fluid temperatures were also examined on this one-eighth fuel assembly geometry. Computational Fluid Dynamics (CFD) code, STAR-CCM+, was used to obtain the results of the numerical simulations. Based on the parametric analysis carried out, sub-channel 4 performed better in terms of heat transfer because temperatures predicted in sub-channel 9 (corner subchannel) were higher than the ones obtained in sub-channel 4 (central sub-channel). The influence of system mass flow rate, pressure and gravity seem similar in both sub-channels 4 and 9 with temperature distributions higher in sub-channel 9 than in sub-channel 4. In most of the cases considered, temperature distributions (for both fluid and wall) obtained at 25 MPa are higher than those obtained at 23 MPa, temperature distributions obtained at 601.2 kg/h are higher than those obtained at 561.2 kg/h, and temperature distributions obtained without gravity effect are higher than those obtained with gravity effect. The results show that effects of system pressure, mass flowrate and gravity on fluid flow and heat transfer are significant and therefore parametric studies need to be performed to determine safe and optimum operating conditions of fluid flow and heat transfer systems.

• Journal of Radiation Protection and Research
• /
• 제14권2호
• /
• pp.30-36
• /
• 1989

고리 1호기의 사용후 핵연료에 함유된 핵종중 핵임계도에 미치는 영향이 큰 22개 핵종에 대한 핵단면적 자료를 XLACS-II에 의해서 ENDF/B-IV로 부터 취해서 FISSLIB(51군 핵단면적 자료)를 만들었다. 그리고 AMPX 조직에 의해서 DLC-43/CSRL 자료로 부터 생산된 51군 핵단면적 자료와 함께 사용할 수 있도록 하였다. 한편 BNL-325와 FISSLIB을 비교하여 후자의 유용성을 확인하였다. 그리고 FISSLIB에 수록된 핵분열 생성물 핵종에 대한 핵군단면적을 기본 입력자료로 하여 무한 배열된 고리 1호기 사용후 핵연료봉에 대한 핵임계도 계산을 수행하였다. 그 결과 냉각기간에 따라 핵분열 생성물 핵종을 고려할 때의 핵임계도는 그것을 무시할 때보다 9-14%의 감소효과를 나타냈다.

• Nuclear Engineering and Technology
• /
• 제48권1호
• /
• pp.16-25
• /
• 2016

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

• 분석과학
• /
• 제16권3호
• /
• pp.191-197
• /
• 2003

플루토늄 표준물 (NBL, CRM No.122)을 이용하여 정확한 농도의 플루토늄 용액을 만들고 이 용액에 대하여 동위원소희석 질량분석법과 조절전위 전기량법으로 플루토늄을 정량하였다. 동위원소희석 질량분석법에 의한 결과는 약 $0.9{\mu}g$-Pu 시료량의 크기에서 모액 플루토늄의 농도에 대하여 얻어진 측정값의 비는 $1.002176{\pm}0.000452$이었으며, 상대 표준편차 0.045%(신뢰도 95%)의 좋은 정밀도를 보였다. 조절전위 전기량법에 의한 측정값은 사용한 시료량에 따라 0.9923~0.9960의 비를 보였다. 두 방법 간의 차이는 0.6~1% 범위 내에서 잘 일치하는 결과를 보였다. 따라서 조절전위 전기량법보다 좋은 결과를 보인 동위원소희석 질량분석법을 이용하여, 조사된 하나로 연소 핵연료봉을 상, 중 및 하단부분으로 구분하여 시료를 취하여 이를 용해하고 음이온교환 칼람을 사용 분리한 플루토늄을 각각 정량 하였다. 이 결과 각각 그람(fuel+clad) 시료당 1.155 mg, 2.483 mg 및 1.920 mg의 플루토늄 값을 얻었다.

• Nuclear Engineering and Technology
• /
• 제52권5호
• /
• pp.897-907
• /
• 2020

The flow characteristics of reactor fuel assembly always intrigue the designers and the experimentalists among the myriad phenomena that occur simultaneously in a nuclear core. In this work, the visual experimental method has been developed on the basis of refraction index matching (RIM) and particle image velocimetry (PIV) techniques to investigate the detailed flow characteristics in China fast reactor fuel subassembly. A 7-rod bundle of simulated fuel subassembly was fabricated for fine examination of flow characteristics in different subchannels. The experiments were performed at condition of Re=6500 (axial bulk velocity 1.6 m/s) and the fluid medium was maintained at 30℃ and 1.0 bar during operation. As for results, axial and lateral flow features were observed. It is shown that the spiral wire has an inhibitory effect on axial flow and significant intensity of lateral flow mixing effect is induced by the wire. The root mean square (RMS) of lateral velocity fluctuation was acquired after data processing, which indicates the strong turbulence characteristics in different flow subchannels.

• Nuclear Engineering and Technology
• /
• 제26권3호
• /
• pp.345-354
• /
• 1994

수치 해석을 통하여 이차유동을 포함한 핵연료봉 주위의 난류 유동장을 예측하였다. 등방성 난류와 점성계수 모델과 이차 유동을 모사하기 위해 단순화된 대수응력모델을 사용하여 난류 운동 에너지 (k)와 난류 에너지 소멸률($\varepsilon$)의 이 방정식 모델과 운동량 방정식을 유한 차분법으로 풀어 유동장내의 평균속도, 이차유동, 난류 운동 에너지, 난류 응력 분포 등을 구하였다. 수치해석 예측치를 실험데이타와 비교하여 만족할만한 결과를 얻었고 유동장내에서 이차유동의 영향을 확인하였다. 즉 이차유동이 절대 크기는 작더라도 대류 효과에 의해 큰 영향을 미치는 것을 본 연구를 통해 알 수 있었다.

• Nuclear Engineering and Technology
• /
• 제50권1호
• /
• pp.35-42
• /
• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.