• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3388-3400
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• 2023

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4226-4230
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• 2022

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kW_th) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3549-3558
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• 2023

The jet pump can be used in a test device of a nuclear reactor for high flow amplification as it reduces inlet flow requirement and thereby size of the process components. In the present work, a miniature jet pump was designed to meet high flow amplification greater than 3. Subsequently, experiments were carried out using a test setup for design validation and performance evaluation of the jet pump for different parameters. It was observed that a minimum pressure of 0.6 bar (g) was required for the secondary fluid inside the jet pump to ensure cavitation free performance at high amplification. Spacing between the nozzle tip and the mixing chamber entry point had significant effect on the performance of the jet pump. Variation in primary flow, temperature and area ratio also affected the performance. It was observed that at high flow amplification, the analytical solution differed significantly from experimental results due to very large velocities encountered in the miniature size jet pump.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.603-616
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• 2009

The past two decades were mainly devoted to model validation and computer code verification against global corium experiments, code application to reactor situations, and investigation of the role of melt properties in steam explosion energetics. Corium data were essentially provided by JRC-Ispra in the FARO and KROTOS facilities and by KAERI in the TROI facility. Verification of code applicability to reactor situations was performed essentially in the frame of the international OECD/SERENA programme. The paper makes a synthesis of the findings made during the above-mentioned period and expresses a personal view of the author with respect to the progress made and expected for the resolution of the steam explosion issue for light water reactors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.183-190
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• 1996

Leak-before-break(LBB) approach has been shown to be both cost and risk effective by reducing maintenance cost and occupational exposure when applied to high energy piping in nuclear power plants. For Korean Next Generation Reactor(KNGR) development, LBB is considered for the Main Steam Line(MSL) piping inside containment. Unlike the reactor coolant piping leakages which can be detected by particulate and gaseous radiation monitoring, main steam line leak detection systems must be based on principles that do not involve radioactivity. Ceramics are widely used as humidity sensor materials which can be further developed for nuclear applications. In this paper, we describe the progress in the development of ceramic humidity sensors for use with the main steam lines of KNGR.

• 한국환경과학회지
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• 제11권7호
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• pp.743-748
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• 2002

UV photolysis process is little known in parts of air pollution treatment, so there are not many applications in field. Therefore we have to do more experiment and study application possibility for treatment of VOCs(Volatile organic compounds). To solve these problems, we have been studying for simultaneous application of this technology. It has shown that concentration of TCE and B.T.X., diameter of reactor and wavelength of lamp have effected on decomposition efficiency. Analysis of TCE and B.T.X. concentration was carried out by GC-FID. A cylinderical reactor consisting of a quartz tube and a centrally located lamp(${\psi}25mm$) was used. The length and diameter of reactor were 1800mm, 75mm. It has shown that the generated ozone concentration goes up 250ppm when using 64watt ozone lamp. When using Photolysis process only, the rates of fractional conversion of each material are TCE 79%, Benzene 65%, Toluene 68%, Xylene 76%. This phenomenon can be rationalized in terms of the different bond energy that indicates how easily VOCs species can be decomposed.

• 산업공학
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• 제13권4호
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• pp.751-758
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• 2000

Situation awareness is defined as a person's perception of the elements of the environment within a volume of time and space, the comprehension of their meaning and the projection of their status in the near future. Situation awareness is important in attempting to evaluate human behavior in operating complex systems such as aircraft, air traffic control, and nuclear power plant systems. From the literatures this study reviews the relationship between situation awareness and numerous individual, system and environmental factors, and also reviews the methodologies for the empirical measurement of situation awareness applicable to Korea Next Generation Reactor (KNGR) design project. Attention, working memory, workload, stress, system complexity, and automation are presented as critical factors limiting operator's situation awareness. Mental models and goal-directed behavior are hypothesized as important mechanisms overcoming these limits. This study summarized hypothesized guidelines for interface design to improve situation awareness of reactor operators. Some of the guidelines should be tested in the KNGR evaluation experiments in the future.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.269-274
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• 2019

$^{64}Cu$ is a favorable radionuclide in nuclear medicine applications because of its unique characteristics such as three types of decay (electron capture, ${\beta}^-$ and ${\beta}^+$) and 12.7 h half-life. Production of $^{64}Cu$ by irradiation $^{nat}Cu$ and $^{nat}CuNPs$ in Tehran Research Reactor was investigated. The characteristics of copper nanoparticles were investigated with SEM, TEM and XRD analysis. The cross section of $^{63}Cu(n,{\gamma})^{64}Cu$ reaction was done with TALYS-1.8 code. The activity value of $^{64}Cu$ was calculated with theoretical approach and MCNPX-2.6 code. The results were compared with related experimental results which showed good adaptations between them.

• 한국자원리싸이클링학회:학술대회논문집
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• 한국자원리싸이클링학회 2005년도 추계정기총회 및 제26회 학술발표대회 고분자리싸이클링기술 특별심포지엄
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• pp.83-96
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• 2005

열가소성수지인 PVC는 우수한 물성을 가지고 있어 다양한 용도로 사용되지만 높은 염소함량으로 인하여 폐기할 때 환경문제를 야기한다. PVC로부터의 탈염소반응이 기타 플라스틱열분해 반응보다 낮은 온도에서 일어나는 점을 이용하여 전처리공정으로의 탈염소반응 연구를 수행하였다. 반응기는 교반능력이 우수한 2축 스크류반응기를 사용하였다. 실험변수는 1차반응기온도, 2차 반응기온도, 혼합플라스틱의 PVC농도, 혼합플라스틱 점도, 공급량, 2차반응기의 스크류회전수이다. 적절한 공정조건하에서 탈염율은 90%이상이었으며 탈염소공정에서 배출되는 염소가스를 물에 흡수하여 염산으로 회수가 가능하였다. 염소 물질수지를 취하여 스크류반응기 전후의 염소 흐름을 분석하였다.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.