• Nuclear Engineering and Technology
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• 제46권3호
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• pp.343-352
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• 2014

Uncertainty and sensitivity analyses with respect to nuclear data are performed with depletion calculations for BWR and PWR fuel assemblies specified in the framework of the UAM-LWR Benchmark Phase II. For this, the GRS sampling based tool XSUSA is employed together with the TRITON depletion sequences from the SCALE 6.1 code system. Uncertainties for multiplication factors and nuclide inventories are determined, as well as the main contributors to these result uncertainties by calculating importance indicators. The corresponding neutron transport calculations are performed with the deterministic discrete-ordinates code NEWT. In addition, the Monte Carlo code KENO in multi-group mode is used to demonstrate a method with which the number of neutron histories per calculation run can be substantially reduced as compared to that in a calculation for the nominal case without uncertainties, while uncertainties and sensitivities are obtained with almost the same accuracy.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3665-3674
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• 2021

One of the important parts of the in-core fuel management is loading pattern optimization (LPO). The loading pattern optimization as a reasonable design of the in-core fuel management can improve both economic and safe aspects of the nuclear reactor. This work proposes the hybrid of fuzzy logic controller with harmony search algorithm (HS) for loading pattern optimization in a pressurized water reactor. The music improvisation process to find a pleasing harmony is inspiring the harmony search algorithm. In this work, the adjustment of the harmony search algorithm parameters such as the bandwidth and the pitch adjustment rate are increasing performance of the proposed algorithm which is done through a fuzzy logic controller. Hence, membership functions and fuzzy rules are designed to improve the performance of the HS algorithm and achieve optimal results. The objective of the method is finding an optimum core arrangement according to safety and economic aspects such as reduction of power peaking factor (PPF) and increase of effective multiplication factor (Keff). The proposed approach effectiveness has been tried in two cases, Michalewicz's bivariate function problem and NEACRP LWR core. The results show that by using fuzzy harmony search algorithm the value of the fitness function is improved by 15.35%. Finally, with regard to the new solutions proposed in this research it could be used as a trustworthy method for other optimization issues of engineering field.

• 대한기계학회논문집A
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• 제34권1호
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• pp.49-54
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• 2010

탄성변형 접촉 조건을 가지는 이중냉각 핵연료봉과 이를 지지하는 지지구조체 사이에서 발생하는 프레팅 마멸거동을 모사 시험편을 이용하여 실험적으로 분석하였다. 이중냉각 핵연료봉은 기존의 핵연료에 비해 외경이 증가하므로 새로운 형상을 가지는 지지구조체의 적용이 필수적이며 이에 대한 진동 특성 및 마멸 저항성에 대한 평가가 필수적이다. 본 연구에서는 현재까지 제안된 다양한 형상 중에서 대표적으로 엠보싱 형상을 가지는 지지구조체의 모사 시험편을 이용하여 이중냉각 핵연료봉의 내마멸 특성을 분석하였다. 개발된 지지구조체 특성 시험장비를 이용하여 모사된 지지구조체 시험편의 특성시험을 수행하였으며 이를 해석에 의한 결과와 비교하였다. 또한 기존의 핵연료 내마멸시험과 동일한 조건 및 장비를 이용하여 프레팅 마멸시험을 수행하여 이중냉각 핵연료봉의 프레팅 마멸거동을 관찰하였다. 본 논문에서는 실험결과로부터 지지구조체 특성과 프레팅 마멸거동 사이의 상관관계에 대하여 자세히 논하였다.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1602-1611
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• 2021

A varying degree of eccentricity always exists in the initial configuration of a nuclear fuel rod. Its impact on traditional LWR fuel is limited as the radial gap closes relatively early during irradiation. However, the effect of misalignment is expected to be more relevant in rods with highly conductive fuels, large initial gaps and low conductivity filling gases. In this paper, we study similar characteristics in the experimental setup of two fuel disc irradiation campaigns carried out in the OECD Halden Boiling Water Reactor. Using the multi-dimensional fuel performance code OFFBEAT, we combine 2-D axisymmetric and 3-D simulations to investigate the effect of eccentricity on the fuel temperature distribution. At the same time, we illustrate how the advent of modern tools with multi-dimensional capabilities might further improve the design and interpretation of in-pile separate-effect tests and we outline the potential of such an analysis for upcoming experiments.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2415-2421
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• 2020

The axial heterogeneous void distribution in a fuel channel is a relevant and important issue during nuclear reactor analysis for LWR, especially for boiling water channel-type reactors. Variation of the coolant density in fuel channel has an effect on the neutron spectrum that will in turn have an impact on the values of absolute reactivity, the void reactivity coefficient, and the fuel isotopic compositions during irradiation. This effect is referring to as the history effect in light water reactor calculations. As the void reactivity effect is positive in RBMK type reactors, the underestimation of water density heterogeneity in 3D reactor core numerical calculations could cause an uncertainty during assessment of safe operation of nuclear reactor. Thus, this issue is analysed with different cross-section libraries which were generated with WIMS8 code at different reference water densities. The libraries were applied in single fuel model of the nodal code of QUABOX-CUBBOX/HYCA. The thermohydraulic part of HYCA allowed to simulate axial water distribution along fuel assembly model and to estimate water density history effect for RBMK type fuel.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1190-1197
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• 2018

Analysis of the GRSW-A model coupled into the FALCON code is extended by simulation of central void formation in fuel pellets due to high-temperature fuel restructuring. The extended calculation is verified against published, well-known experimental data. Good agreement with the data for a central void diameter in pellets of the rod irradiated in an Experimental Breeder Reactor is shown. The new calculation methodology is employed in comparative analysis of modern BWR fuel behavior under assumed high-power operation. The initial fuel porosity is shown to have a major effect on the predicted central void diameter during the operation in question. Discernible effects of a central void on peak fuel temperature and Pellet-Cladding Mechanical Interaction (PCMI) during a simulated power ramp are shown. A mitigating effect on PCMI is largely attributed to the additional free volume in the pellets into which the fuel can creep due to internal compressive stresses during a power ramp.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.435-435
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• 2002

• 대한기계학회논문집A
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• 제39권3호
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• pp.249-257
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• 2015

원자력 발전소의 반응로에는 핵분열 에너지를 생성하고 방사성 물질의 유출을 막는 핵연료 집합체가 있으며, 이러한 집합체는 핵연료와 피복관으로 구성되어 있는 핵 연료봉으로 구성되어 있다. 원자로에서 핵연료봉 거동의 안전성을 평가하기 위해 해석적인 방법을 적용하며 이러한 평가 코드를 핵 연료 성능 코드라 한다. 경수로 핵연료 해석에서는 간극의 두께에 따라 열전도도가 크게 영향을 받는 간극 열전도도가 주요 거동해석에 영향을 미친다. 본 연구에서는 간극 두께에 따라 열전도도가 변화하는 3 차원 간극 요소(Gap element)를 제안하였으며, 이를 적용하기 위해 3 차원 열탄성 모듈을 FORTRAN90을 이용하여 개발하였다. 제안된 3 차원 간극 요소를 이용하여 핵 연료봉에서 발생할 수 있는 비대칭적인 형상인 핵 연료 표면에 결함이 생긴 경우 MPS(Missing Pellet Surface)와 핵연료봉의 편심(Eccentricity of the nuclear fuel rod) 형상에 대하여 3 차원 해석을 진행하였다.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.11-16
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• 2016

Fuel assembly's mechanical characterization test facility (FAMeCT) in KAERI was constructed with upgraded functional features such as increased loading capacity, underwater vibration testing and severe earthquake simulation for extended fuel design guideline. This facility is designed and developed to provide out-pile fuel data for accident analysis model and fuel licensing. Functional tests of FAMeCT were performed to confirm functionality, structural integrity, and validity of newly-built fuel assembly mechanical test facility. Test program includes signal check of data acquisition system, load delivering capacity using real-sized fuel assemblies and a standard loading cylindrical rigid specimen. Fuel assembly's lateral bending test was carried out up to 30 mm of pull-out displacement. Limit case axial compression loading test up to 33 kN was performed to check structural integrity of UCPS (Upper Core Plate Simulator) support frame. Test results show that all test equipment and measurement system have acceptable range of alignment, signal to noise ratio, load carrying capacity limit without loss of integrity. This paper introduces newly constructed fuel assembly's mechanical test facility and summarizes results of functional test for the mechanical test equipment and data acquisition system.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4499-4513
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• 2022

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel-specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM^TM). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ~50 years of dry storage.