• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3910-3917
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• 2021

Reactor fuel's power distribution plays a vital role in designing the new generation thermionic Space Reactor Power Systems (SRPS). In this paper, the 1/12th SPACE-R's full reactor core was numerically analyzed with two kinds of different axial power distribution, to identify their impacts on thermal-hydraulic and thermoelectric characteristics. In the benchmark study, the maximum error between numerical results and existing data or design values ranged from 0.2 to 2.2%. Four main conclusions were obtained in the numerical analysis: a) The axial power distribution has less impact on coolant temperature. b) Axial power distribution influenced the emitter temperature distribution a lot, when the core power was cosine distributed, the maximum temperature of the emitter was 194 K higher than that of the uniform power distribution. c) Comparing to the cosine axial power distribution, the uniform axial power distribution would make the maximum temperature in each component of the reactor core much lower, reducing the requirements for core fuel material. d) Voltage and current distribution were similar to the axial electrode temperature distribution, and the axial power distribution has little effect on the output power.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1791-1801
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• 2023

The multi-physics coupled methodologies that have been widely used to analyze the complex process occurring in nuclear reactors have also been used to the R&D of numerical reactors. The advancement in the field of computer technology has helped in the development of these methodologies. Herein, we report the integration of ADPRES code and RELAP5 code into the SALOME-ICoCo framework to form a multi-physics coupling platform. The platform exploits the supervisor architecture, serial mode, mesh one-to-one correspondence and explicit coupling methods during analysis, and the uncertainty analysis tool URANIE was used. The correctness of the platform was verified through the NEACRP-L-335 benchmark. The results obtained were in accordance with the reference values. The platform could be used to accurately determine the power peak. In addition, design margins could be gained post uncertainty analysis. The initial power, inlet coolant temperature and the mass flow of assembly property significantly influence reactor safety during the rod ejections accident (REA).

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.280-285
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• 2001

This paper presents a modelling of thermal discharge performance for a static ice-on-coil ice-storage tank. Through the present study, discharging characteristics were examined with the existing results of theoretical and numerical heat transfer analyses. Also, an experiment was conducted to obtain a real set of discharge performance. The thermal effectiveness, the ratio of the actual heat transfer rate to the maximum possible heat transfer rate, decreased when the stored energy decreased during discharging period. And the effectiveness increased as the coolant flow rate through the storage increased, of which increasing rate decreased abruptly near the maximum and the minimum stored energy. An empirical correlation was obtained from the experimental and the numerical analysis data.

• Nuclear Engineering and Technology
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• 제29권5호
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• pp.406-416
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• 1997

A computational method is developed for predicting the steady-state temperature field in an LMR core. Detailed core-wide coolant temperature profiles are efficiently calculated using the simplified energy equation mixing model[1] and the subchannel analysis method. The $\theta$-method is employed for discretizing the energy equations in the axial direction. The interassembly coupling is achieved by interassembly gap flow. Cladding and fuel temperatures are calculated with the one-dimensional conduction model and temperature integrals of conductivities. The accuracy of the method is tested by performing several benchmark calculations for too LMR problems. The results indicate that the accuracy is comparable to the other methods based on ENERGY model. It is also shown that the implicit scheme for the axial discretization is more efficient than the explicit scheme.

• 대한전기학회논문지
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• 제31권10호
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• pp.109-115
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• 1982

Reactor dynamics were studied by reactor noise at TRIGA MARK Il reactor whose rated power is 250KW thermal. The power spectral densities(PSD) of the noise were measured by stochastic method with high resolution digital filters and Fast Fourier Transformers. The transfer function of the reactor at zero power was identical to the theoretical characteristics. When the power was increasec above 1KW, reactor showed its poswer resonances at 3Hz and 10 Hz. It was analyzed that 3Hz peak was generated by heat transfer and coolant flow effects and 10Hz peak by nuclear reaction effects.

• 한국안전학회지
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• 제11권2호
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• pp.52-59
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• 1996

This paper presents a new approach for assessing the dynamic reliability in a complex system such as a nuclear power plant. The method is applied to a dynamic analysis of the potential accident sequences which may occur during mid-loop operation. Mid-loop operation is defined as an operation to make RCS water level below the top of the flow area of the hot legs at the junction with the reactor vessel for repairs and maintenance of steam generators and reactor coolant pumps for a specific time. The Idea behind this approach consists of both the use of the concept of the performance achievement/requirement correlation and of a dynamic event tree generation method. The assessment of the system reliability depends on the determination of both the required performance distribution and the achieved performance distribution. The quantified correlation between requirement and achievement represents a comparison between two competing variables. It is demonstrated that this method is easily applicable and flexible in that it can be applied to any kind of dynamic reliability problem.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.111-120
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• 2021

During the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel and steel are solidified into debris particles, which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes the debris particles relocated and the bed flattened, which called debris bed relocation. Because the thickness of debris bed has great influence on the cooling ability of fuel debris in low plenum, it's very necessary to evaluate the transient changes of the shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a large number of debris bed relocation experiments were carried out by improved bottom gas-injection experimental method in this paper. The effects of different experimental factors on the relocation process were studied from the experiments. The experimental data were also used to further evaluate a semi-empirical onset model for predicting relocation.

• Nuclear Engineering and Technology
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• 제20권1호
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• pp.35-46
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• 1988

원자로의 설계나 안전성 분석을 위해서는 핵연료 집합체 내의 유동 구조와 열전달에 대한 지식이 매우 중요하다. 따라서 핵연료 집합체 내의 유체 온도 분포를 정확히 계산하기 위해서는 냉각재 유로 내에서의 속도분포를 정확히 알아야 하는데 이것은 복잡한 난류 현상 때문에 예측하기가 매우 어렵다. 본 연구는 비등방성을 고려한 2-방정식 모형을 사용하여 속도분포를 구하고 핵연료 표면에서의 균일열속을 가정하므로써 유로내에서의 속도 분포를 예측하였다. 수치해는 Galerkin유한 요소법에 의해 핵연료봉 표면까지 구하여졌다. 수치 결과는 알려진 실험치 및 계산치와 비교되어 잘 일치하고 있고, 또한 난류 비등방성이 유로 내의 평균속도와 온도분포에 영향을 미치고 있음을 보았다. 그리고 조밀한 삼각 배열 핵연료 집합체(P/D=1.05-1.3) 내에서 나트륨 냉각재를 사용한 경우의 Nu-P/D관계식을 수립하였다.

• 한국소음진동공학회논문집
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• 제26권1호
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• pp.55-62
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• 2016

Nuclear fuels are always exposed to hot temperature and high speed coolant flow during the reactor operation. Thus the fuel rod accompanies small amplitude vibration due to the turbulent flow. The random vibration causes friction between the fuel rod and the grid structure which provides the lateral supports. The friction is critical to the fuel rod fretting wear, and it degrades fuel performance when a severe wear is developed. LuGre friction model is introduced in the paper, and the performance was evaluated comparing to the classical Coulomb model. It is shown that the developed friction force considering the Coulomb friction is not enough to stop or delay the motion while the stick-slip can be simulated using LuGre friction model. Numerical solutions of the two dimensional spacer grid cell model with the modern friction are also reviewed, and it is discussed that the new friction model simulates well the nonlinear mechanism.

• Nuclear Engineering and Technology
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• 제35권3호
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• pp.179-190
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• 2003

The multiple steam generator tube rupture (SGTR) event scenario with available safety systems was experimentally and analytically evaluated. The experiment was conducted on the large scaled test facility to simulate the multiple SGTR event and investigate the effectiveness of operator actions. As a result, it indicated that the opening of pressurizer power operated relief valve was significantly effective in quickly terminating the primary-to-secondary break flow even for the 6.5 tubes rupture. In the analysis, the recent version of RELAP5 code was assessed with the test data. It indicated that the calculations agreed well with the measured data and that the plant responses such as the water level and relief valve cycling in the damaged steam generator were reasonably predicted. Finally, sensitivity study on the number of ruptured tubes up to 10 tubes was performed to investigate the coolant release into atmosphere. It indicated that the integrated steam mass released was not significantly varied with the number of ruptured tubes although the damaged steam generator was overfilled for more than 3 tubes rupture. These findings are expected to provide useful information in understanding and evaluating the plant ability to mitigate the consequence of multiple SGTR event.