• Nuclear Engineering and Technology
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• 제45권7호
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• pp.951-960
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• 2013

Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that $UO_2-PuO_2$ (MOX) is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the $UO_2$ fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of $H_2O$ moderated $UO_2-PuO_2$ (MOX) lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

• 한국컴퓨터정보학회논문지
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• 제10권5호
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• pp.179-186
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• 2005

본 논문에서는 유전자 알고리즘을 이용한 rural Postman problem with Time windows(RPPTW) 해법을 위해 유전자 알고리즘에 사용되는 교차 연산자를 제안하고, 기존의 교차 연산자와 비교한다. RPPTW는 다중목적 최적화 문제로서, Rural Postman Problem(RPP)에 서비스 시간 제한을 위한 시간 윈도우(Time Windows)를 두고 제한된 시간 내에 서비스를 받을 수 있도록 구성된 문제이다. 따라서, RPPTW는 주어진 시간 내에 서비스를 받으면서 최소 비용으로 라우팅을 하는 다중 목적 최적화 문제이다. 다중 목적 최적화 문제인 RPPTW를 해결하기 위해서는 Pareto-optimal 집합을 구해야 한다. Pareto-optimal 집합은 각 목적값들의 우수성을 비교할 수 없는 집합이다. 본 논문에서는 12개의 임의로 생성된 문제들에 대해 3개의 교차 연산자를 사용하여 실험을 하여 그 결과를 비교하였다. 본 논문에서 사용된 교차 연산자들은 PMX(Partially Matched Exchange), OX(Order Exchange), 그리고 본 논문에서 제안한 MOX(Modified Order Exchange)이다. 각 문제들에 대한 실험 결과를 통해서 RPPTW를 위한 교차 연산자 중에 본 논문에서 제안한 MOX방법이 효율적임를 알 수 있었다.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.499-508
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• 2011

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in $UO_2$ fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS's precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code's prediction. The database consists of the $UO_2$ irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and $UO_2$ fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

• 대한약침학회지
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• 제2권1호
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• pp.13-25
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• 1999

This study was undertaken to determine whether Juglandis Semen (JAS) extraction exerts protective effect against oxidant-induced inhibition of $Na^+$-pump activity in cerebral cortex. $Na^+$-pump activity was estimated by measuring ouabain-sensitive oxygen consumption. The oxygen consumption significantly inhibited by 1 mM t-butylhydroperoxide (tBHP), which was prevented by addition of 2% JAS extraction. The oxygen consumption was increased by an increase in $Na^+$ concentration from 5 to 100mM, $K^+$ concentration from 0.5 to 10 mM, and $Mg^{++}$ concentration from 0.2 to 5 mM. These changes in ion concentrations did not affect the inhibitory effct of tBHP and protective action of JAS on oxygen consumption. tBHP(1mM) produced a significant increase in lipid peroxidation in cerebral cortex, which was prevented by 2% JAS extraction. These result suggest that JAS exerts protective effect against tBHP-induced inhibition of $Na^+$-pump activity in the cerebral cortex, this effect may be due to by an antioxidant action.

• Nuclear Engineering and Technology
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• 제30권6호
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• pp.541-554
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• 1998

A computer code COSMOS has been developed based on the CARO-D5 for the thermal analysis of LWR UO$_2$ and MOX fuel rod under steady-state and transient operating conditions. The main purpose of the COSMOS, which considers high turnup characteristics such as thermal conductivity degradation with turnup and rim formation at the outer part of fuel pellet, is to calculate temperature profile across fuel pellet and fission gas release up to high burnup. A new mechanistic fission gas release model developed based on physical processes has been incorporated into the code. In addition, the features of MOX fuel such as change in themo-mechanical properties and the effect of microscopic heterogeneity on fission gas release have been also taken into account so that it can be applied to MOX fuel. Another important feature of the COSMOS is that it can analyze fuel segment refabricated from base irradiated fuel rods in commercial reactors. This feature makes it possible to analyze database obtained from international projects such as the MALDEN and RISO, many of which were collected from refabricated fuel segments. The capacity of the COSMOS has been tested with some number of experimental results obtained from the HALDEN, RISO and FIGARO programs. Comparison with the measured data indicates that, although the COSMOS gives reasonable agreement, the current models need to be improved. This work is being performed using database available from the OECD/NEA.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.576-581
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• 2010

KAERI (Korea Atomic Energy Research Institute) fabricated MOX (Mixed Oxide) fuel pellets as a cooperation project with PSI (Paul Scherrer Institut) for an irradiation test in the Halden reactor. The MOX pellets were fitted into fuel rods that included instrumentation for measurement in IFE (Institutt for Energiteknikk). The fuel rods were assembled into the test rig and irradiated in the Halden reactor up to 50 MWd/kgHM. The irradiated fuel rods were transported to the IFE, where ceramography was carried out. The fuel rods were cut transversely at the relatively higher burn-up locations and then the radial cross sections were observed. Micrographs were analyzed using an image analysis program and grain sizes along the radial direction were measured by the linear intercept method. Radial cracks in the irradiated MOX were observed that were generally circumferentially closed at the pellet periphery and open in the hot central region. A circumferential crack was formed along the boundary between the dark central and the outer regions. The inner surface of the cladding was covered with an oxide layer. Pu-rich spots were observed in the outer region of the fuel pellets. The spots were surrounded by many small pores and contained some big pores inside. Metallic fission product precipitates were observed mainly in the central region and in the inside of the Pu spots. The average areal fractions of the metallic precipitates at the radial cross section were 0.41% for rod 6 and 0.32% for rod 3. In the periphery, pore density smaller than 2 ${\mu}m$ was higher than that of the other regions. The grain growth occurred from 10 ${\mu}m$ to 12 ${\mu}m$ in the central region of rod 6 during irradiation.

• Nuclear Engineering and Technology
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• 제29권4호
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• pp.299-309
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• 1997

The neutronic feasibility of typical Korean three-loop 900MWe class PWR core loaded with mixed oxide fuels for both annual and 18-month cycle strategies has been investigated as a means for spent fuel management. For this study, a method of determining equivalent plutonium content was developed under the equivalence concept which gives the same cycle length as uranium fuel. Optimal plutonium zoning within the MOX assembly was also designed with the aim of minimizing the peak md power. Conceptual core designs hate hen developed for equilibrium cycle with the following variations: annual and 18-month cycle, 1/3 and full MOX loading schemes, and typical and high moderation lattice. The analysis of key core physics parameters shows that in all cases considered satisfactory core designs seem to be feasible, though addition of control rod system and change in Technical Specification for soluble boron concentration are required for full MOX loading in order to meet the current design requirements.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.19-29
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• 2021

This paper presents the validation of the MCS code for critical safety analysis with burnup credit for the spent fuel casks. The validation process in this work considers five critical benchmark problem sets, which consist of total 80 critical experiments having MOX fuels from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The similarity analysis with the use of sensitivity and uncertainty tool TSUNAMI in SCALE was used to determine the applicable benchmark experiments corresponding to each spent fuel cask model and then the Upper Safety Limits (USLs) except for the isotopic validation were evaluated following the guidance from NUREG/CR-6698. The validation process in this work was also performed with the MCNP6 for comparison with the results using MCS calculations. The results of this work showed the consistence between MCS and MCNP6 for the MOX fueled criticality benchmarks, thus proving the reliability of the MCS calculations.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3071-3079
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• 2023

Helium production in the nuclear fuel matrix during irradiation plays a critical role in the design and performance of Gen-IV reactor fuel, as it represents a life-limiting factor for the operation of fuel pins. In this work, a surrogate model for the helium production rate in fast reactor MOX fuels is developed, targeting its inclusion in engineering tools such as fuel performance codes. This surrogate model is based on synthetic datasets obtained via the SCIANTIX burnup module. Such datasets are generated using Latin hypercube sampling to cover the range of input parameters (e.g., fuel initial composition, fission rate density, and irradiation time) and exploiting the low computation requirement of the burnup module itself. The surrogate model is verified against the SCIANTIX burnup module results for helium production with satisfactory performance.

• 원자력산업
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• 제14권9호통권139호
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• pp.44-53
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• 1994