• Nuclear Engineering and Technology
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• 제51권3호
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• pp.769-775
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• 2019

The tensile strength of irradiated 3C-SiC, SiC with artificial point defects, SiC with symmetric tilt grain boundaries (GBs), irradiated SiC with GBs are investigated using molecular dynamics simulations at 300 K. For an irradiated SiC sample, the tensile strength decreases with the increase of irradiation dose. The Young's modulus decreases with the increase of irradiation dose which agrees well with experiment and simulation data. For artificial point defects, the designed point defects dramatically decrease the tensile strength of SiC at low concentration. Among the point defects studied in this work, the vacancies drop the strength the most seriously. SiC symmetric tilt GBs decrease the tensile strength of pure SiC. Under irradiated condition, the tensile strengths of all SiC samples with grain boundaries decrease and converge to certain value because the structures become amorphous and the grain boundaries disappear after high dose irradiation.

• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1617-1628
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• 2017

In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.803-810
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• 2022

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.207-214
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• 2001

The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.3450-3462
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• 2024

This paper presents a computational framework for drop time assessment of a control element assembly (CEA) under fuel assembly (FA) deformations. The proposed framework consists of three key components: 1) finite element modeling of CEA, 2) fluid-structure interaction to compute drag force, and 3) modeling of frictional contact between CEA and FA. Specially, to accommodate the large motion of CEA, beam elements based on absolute nodal coordinate formulation (ANCF) are adopted. The continuity equation is utilized to calculate the drag force, considering flow changes in the cross-sectional area during the CEA drop. Lastly, beam-inside-beam frictional contact model is employed to capture practical contact conditions between CEA and FA. The proposed framework is validated through experiments under two scenarios: free falls of CEA within FA, encompassing undeformed and deformed scenarios. The experimental validation of the framework demonstrated that the drop time of CEA can be accurately predicted under the complex coupling effects of fluid and frictional contact. The drop times of the S-shaped deformation case is longer than those of the C-shaped deformation case, affirming the time delay due to frictional force. The validation confirms the potential applicability to access the safety and reliability of nuclear power plants under extreme conditions.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2504-2515
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• 2023

In past years the Paul Scherrer Institute (PSI, Switzerland) and the Karlsruhe Institue of Technology (KIT, Germany)) collaborated to develop a model to account for the active role of nitrogen in the air oxidation of a Zircalloy cladding. The "PSI-KIT Nitriding Model for Zirconium based Fuel Cladding" model was implemented at PSI into PSI-MELCOR 1.8.6. In order to make a preliminary evaluation of the effect of the new model on the evolution of full-scale spent fuel pool accidents, one spent fuel pool event was analyzed using the PSI research version of PSI-MELCOR 1.8.6, which includes the nitriding model. To adapt an existing input deck for the calculations, a sensitivity study was conducted to find an optimal nodalization for the analyses. The nitriding model results were compared to those calculated with the MELCOR 1.8.6-PSI without the new nitriding model. The results demonstrate the effect of the nitriding reactions in spent fuel pool accident progression. Moreover, they confirm the impact of ZrN formation during cladding oxidation in air when the oxidation reactions lead to oxygen starvation inside the fuel assemblies. The nitriding reaction led to higher chemical heat generation during the accident and to an earlier failure of the cladding than when the effect of nitrogen reactions was not considered. It should be noted that the nitriding model, as implemented in the PSI version of MELCOR 1.8.6 has not yet been conclusively validated. Thereby the results presented in this paper should be treated as a preliminary demonstration of the capabilities of the model.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.430-438
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• 2021

The spatial distribution of neutron flux or reaction rate was calculated by cell or mesh tally in traditional Monte Carlo simulation. However, either cell or mesh tally leads to the increase of memory consumption and simulation time. In this paper, the function expansion tally (FET) method was developed in Reactor Monte Carlo code RMC to solve this problem. The FET method was applied to the tallies of neutron flux distributions of uranium block and PWR fuel rod models. Legendre polynomials were used in the axial direction, while Zernike polynomials were used in the radial direction. The results of flux, calculation time and memory consumption of different expansion orders were investigated, and compared with the mesh tally. Results showed that the continuous distribution of flux can be obtained by FET method. The flux distributions were consistent with that of mesh tally, while the memory consumption and simulation time can be effectively reduced. Finally, the convergence analysis of coefficients of polynomials were performed, and the selection strategy of FET order was proposed based on the statistics uncertainty of the coefficients. The proposed method can help to determine the order of FET, which was meaningful for the efficiency and accuracy of FET method.

• 방사성폐기물학회지
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• 제9권4호
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• pp.219-230
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• 2011

핵연료주기 분석 연구는 핵연료주기 단계에서 기술들을 분석하고 요건들을 도출하여 국가적 핵연료주기 정책 설정 및 추진을 체계적으로 수행하기 위한 연구이다. 시스템 분석 기술은 대상 시스템의 비교 분석 평가에 활용되며 핵연료주기를 대상으로 하는 경우 각 국가 또는 관심 범위에 따라 다양한 방법이 사용된다. 본 연구에서는 국내 선진 핵연료주기 개발을 위해 필요한 핵연료주기 분석 전략과 함께 이를 위해 사용될 수 있는 분석 기술들을 제시하였다. 핵연료주기 분석은 전략적으로 기술적 분석, 국내외 이해관계, 국가 에너지 프로그램과 연계되어야 한다. 이를 위해 다양한 핵연료주기를 비교하여 제시된 평가 지표에 따라 분석하는 연구는 트레이드 연구 방법을 적용하여 수행할 수 있다. 본 연구를 통한 조사 분석 결과 핵연료주기 분석 전략과 함께 방법적 측면에서 트레이드 연구가 선진 핵연료주기 도출에 활용될 수 있을 것으로 파악되었다. 트레이드 연구에 필수적인 평가지표를 선정하고 각 지표별 핵연료주기에 대한 정보를 얻기 위해서는 기술성숙도 분석 방법과 핵연료주기 시뮬레이터를 활용할 수 있을 것으로 제시하였다. 이들은 핵연료주기의 기술성, 경제성, 환경영향성 등을 비교 평가하여 기술개발을 위한 방향을 제시하고 체계적인 선진 핵연료주기 도출 및 실현에 기여할 것이다.

• 한국세라믹학회지
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• 제44권10호
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• pp.580-584
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• 2007

The ZrC layer instead of SiC layer is a critical and essential layer in TRISO coated fuel particles since it is a protective layer against diffusion of fission products and provides mechanical strength for the fuel particle. In this study, we carried out computational simulation before actual experiment. With these simulation results, Zirconium carbide (ZrC) films were chemically vapor deposited on $ZrO_2$ substrate using zirconium tetrachloride $(ZrCl_4),\;CH_4$ as a source and $H_2$ dilution gas, respectively. The change of input gas ratio was correlated with growth rate and morphology of deposited ZrC films. The growth rate of ZrC films increased as the input gas ratio decreased. The microstructure of ZrC films was changed with input gas ratio; small granular type grain structure was exhibited at the low input gas ratio. Angular type structure of increased grain size was observed at the high input gas ratio.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.384-388
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• 2016

Background: For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. Materials and Methods: In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. Results and Discussion: The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. Conclusion: It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.