• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1210-1217
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• 2023

The assessment of radiation fields in the lithium loop pipes and dump tank during the operation were performed for International Fusion Materials Irradiation Facility - DEMO-Oriented NEutron Source (IFMIF-DONES) in order to obtain the radiation dose-rate maps in the component surroundings. Variance reduction techniques such as weight window mesh (produced with the ADVANTG code) were applied to bring the statistical uncertainty down to a reasonable level. The biological dose was given in the study, and potential shielding optimization is suggested and more thoroughly evaluated. The MCNP Monte Carlo was used to simulate a gamma particle transport for radiation shielding purposes for the current Li Systems' design. In addition, the shielding efficiency was identified for the Impurity Control System components and the dump tank. The analysis reported in this paper takes into account the radiation decay source from and activated corrosion products (ACPs), which is created by d-Li interaction. As a consequence, the radiation (resulting from ACPs and Be-7) shielding calculations have been carried out for safety considerations.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2404-2411
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• 2024

The GEOUNED code is specifically designed to convert CAD models, defined using the B-rep approach, into MC radiation transport models, defined using the CSG approach, and vice versa from MC to CAD. This code incorporates standard features commonly found in conversion tools, including decomposition, conversion, and automatic void generation. Additionally, it introduces innovative features, mainly in the automatic void generation part, which are described in this article. GEOUNED has demonstrated successful application in highly detailed 3D models used in fusion neutronics, which are known for their complex geometries, particularly those utilized in ITER. The article includes examples showcasing GEOUNED's performance in these challenging models, as well as custom applications that highlight its flexibility in addressing non-standard problems. The code is open-source and utilizes Open CASCADE as the geometry engine, with FreeCAD serving as the Python API.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.85-93
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• 2015

Background: $^{241}Am$, $^{243}Am$, and $^{237}Np$ isotopes are among the most radiotoxic components of spent nuclear fuel. Recently, researchers have planned different incineration scenarios for the highly radiotoxic elements of nuclear waste in critical reactors. Computational methods are widely used to predict burnup rates of such nuclear wastes that are used under fuel matrixes in critical reactors. Methods: In this work, the Monte Carlo N-particle transport code was used to calculate the neutronic behavior of a transuranic (TRU)-bearing CANada Deuterium Uranium 6 reactor. Results: The computational data showed that the 1.0% TRU-containing thorium-based fuel matrix presents higher proliferation resistance and TRU depletion rate than the other investigated fuel Matrixes. The fuel matrix includes higher negative temperature reactivity coefficients as well. Conclusion: The investigated thorium-based fuel matrix can be successfully used to decrease the production of highly radiotoxic isotopes.

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

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.124-129
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• 1998

A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The $^{239}$ Pu and $^{235}$ U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the fled uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%.. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel has shown that it is desirable to increase the $^{239}$ Pu and $^{235}$ U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, il is recommended to have enrichments of 0.45 and 1.00 wt% for $^{239}$ Pu and $^{235}$ U, respectively.

• Nuclear Engineering and Technology
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• 제32권3호
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• pp.214-226
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• 2000

In an effort to assess the performance of KAERI's coupled 3D kinetics - system T/H code, MARS/MASTER, Exercise III of the OECD main steam line break benchmark is solved. The analysis model of the reference plant, TMI-1 - a 2772 MWth B&W plant, consists of three major components: a core neutronics model involving 241$\times$28 neutronic nodes, a vessel 3D T/H model consisting of 374 hydrodynamic volumes, and a 1D system T/H model containing 157 hydrodynamic volumes. The results show that there is a significant amount of flow mixing occurring in the upper and lower plenum regions and the core power distribution evolves to a highly localized shape due to the presence of a stuck rod, as well as the asymmetric flow distribution. It is judged that MARS/MASTER properly captures these drastic 3-dimensional effects. Comparisons with other results submitted to OECD confirm the accuracy of the MARS/MASTER solution.

• Nuclear Engineering and Technology
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• 제13권3호
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• pp.153-161
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• 1981

고리 원자력 1호기에서 일어날 수 있는 가상사고에 의한 동특성 현상이 점근사 원자로 모델에 의한 중성자 및 온도 방정식을 사용하여 해석되었다. 일반적으로 수치해석 결과는 사고해석에 있어서 확실한 동특성 시간전이 현상을 예견하기 위채서는 보다 정밀한 계산모델을 사용해야 된다는 것을 지시한다. 전출력 상태에서 RCCA 인출에 따르는 출력반응의 경우는 점근사 원자로 모델이 고리 1호기의 최종 안정성 분석 보고서의 해석결과와 우수한 일치를 보여줬다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.76-85
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• 2005

The conceptual study of Accelerator Driven System (ADS) had lasted for about five years and ended in 1999 in China. As one project of 'the major state basic research program (973)' in energy domain, which is sponsored by the China Ministry of Science and Technology (MOST), a five years program of basic research for ADS physics and related technology has been launched since 2000 and passed national review last month. CIAE (China Institute of Atomic Energy), IHEP (Institute of High Energy Physics), PKU-IHIP (Institute of Heavy Ion Physics in Peking University) and other institutions are jointly carrying on the research. The research activities are focused on HPPA physics and technology, reactor physics of external source driven sub-critical assembly, nuclear data base and material study. For HPPA, a high current injector consisting of an ECR ion source, LEBT and a RFQ accelerating structure of 3.5MeV has been built. In reactor physics study, a series of neutron multiplication experimental study has been carried out and is being carrying on. The VENUS facility has been constructed as the basic experimental platform for the neutronics study in ADS blanket. It's a zero power sub-critical neutron multiplying assembly driven by external neutron produced by a pulsed neutron generator. The theoretical, experimental and simulation study on nuclear data, material properties and nuclear fuel circulation related to ADS is carrying on to provide the database for ADS system analysis. The main results on ADS related researches will be reported.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.47-54
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• 2010

The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/$cm^3$. The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1250-1258
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• 2017

The multigroup transport theory is the basis for many neutronics modules. A significant point of the cross-section (XS) generation procedure is the choice of the energy groups' boundaries in the XS libraries, which must be carefully selected as an unsuitable energy meshing can easily lead to inaccurate results. This decision can require considerable effort and is particularly difficult for the common user, especially if not well-versed in reactor physics. This work investigates a genetic algorithm-based tool which selects an appropriate XS energy structure (ES) specific for the considered problem, to be used for the condensation of a fine multigroup library. The procedure is accelerated by results storage and fitness calculation speedup and can be easily parallelized. The extension is applied to the coupled code SIMMER and tested on the European Sustainable Nuclear Industrial Initiative (ESNII+) Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID)-like reactor system with different fitness functions. The results show that, when the libraries are condensed based on the ESs suggested by the algorithm, the code actually returns the correct multiplication factor, in both reference and voided conditions. The computational effort reduction obtained by using the condensed library rather than the fine one is assessed and is much higher than the time required for the ES search.