• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3133-3150
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• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

• 한국산학기술학회논문지
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• 제18권4호
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• pp.50-57
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• 2017

본 사출 금형을 통해 제품 생산 시 사출 금형 내에서 제품 형상을 형성하는 코어는 금형 내부에 틀의 형태로 가공되어 설치되며 이때 부분적인 코어의 형상을 핀에 가공하여 설치하는 부품을 코어 핀이라 한다. 이러한 사출 코어 용 코어 핀은 제품의 소형화 집적화에 따라 그 크기가 마이크로의 크기로 작아지고 있다. 하지만 이를 가공 시 기존의 센터리스 연삭 장치로는 마이크로 사이즈의 피삭재를 고정하여 밀착 시켜주는 장치의 부재로 인해 진동이 발생한다. 이러한 이유로 마이크로 크기의 직경을 가지는 코어 핀의 경우 가공 시 진동에 의해 변형 발생으로 가공 불량률이 매우 높다. 따라서 본 논문에서는 마이크로 크기의 코어 핀을 가공하기 위해 기존의 평면 연삭기에 설치하여 사용이 가능한 소형의 연삭 시스템을 개발 하였다. 이를 이용하여 코어 핀에 대한 연삭 실험을 진행하였으며 표면 거칠기, 진원도, 원통도의 측정을 통해 성능을 검증하였다.

• 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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• 제53권5호
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• pp.1403-1415
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• 2021

The transient capability of a SP₃ based pin-wise core analysis code SPHINCS is developed and verified through the analyses of the C5G7-TD benchmark. Spatial discretization is done by the fine mesh finite difference method (FDM) within the framework of the coarse mesh finite difference (CMFD) formulation. Pin size fine meshes are used in the radial fine mesh kernels. The time derivatives of the odd moments in the time-dependent SP₃ equations are neglected. The pin homogenized group constants and Super Homogenization (SPH) factors generated from the 2D single assembly calculations at the unrodded and rodded conditions are used in the transient calculations via proper interpolation involving the approximate flux weighting method for the cases that involve control rod movement. The simplifications and approximations introduced in SPHINCS are assessed and verified by solving all the problems of C5G7-TD and then by comparing with the results of the direct whole core calculation code nTRACER. It is demonstrated that SPHINCS yields accurate solutions in the transient behaviors of core power and reactivity.

• 한국정밀공학회지
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• 제24권2호
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• pp.41-46
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• 2007

Recently, the study for filling imbalance in thermoplastic polymer has gradually been increased. However, it is hard to find the researches for filling imbalance of thermoplastic elastomer(TPE). The experiment of filling imbalance was conducted for the three kinds of thermoplastic vulcanizes(TPVs) in the mold with geometrically balanced runner system. In this experiment, the effects of the melt temperature and injection speed on the filling imbalance were investigated. To solve the filling imbalance, Runner Core pin(RC pin) in the experimental mold was adopted and it's effects was tested. In this paper, we present that the insert length of RC pin is dependent to each polymers for optimal filling balance.

• Nuclear Engineering and Technology
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• 제35권4호
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• pp.356-368
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• 2003

Material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the updated driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the end of life is predicted to be $68.61\%$ and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is $1.93\%$, satisfying the preliminary design criterion ($3\%$) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1084-1096
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• 2023

On-the-fly(OTF) generation of Spectral Superhomogenization(SSPH) factors is analyzed in the multigroup(MG) whole core calculation employing pin-wise continuous energy(CE) slowing-down solutions. The motivation for the work is to avoid the huge computing time required for the generation of a parametrized SSPH factor library(PSSL) which is used to resolve the angular dependency of MG resonance cross sections, and also to exploit the advantage of flexible choice of a MG structure by using CE slowing-down solutions. Two pin-wise CE slowing-down methods, the equivalent Dancoff cell method and the shadowing effect correction method, are evaluated with the OTF SSPH method. The effectiveness of the OTF SSPH method is examined for various simplified and realistic core problems with various MG structures. It is demonstrated that the computing time overhead of this method is negligible whereas the solution accuracy is considerably enhanced.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2006년도 춘계학술대회 논문집
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• pp.407-408
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• 2006

Almost all injection molds have multi-cavity, which are designed with geometrically balanced runner system in order to made filling balance between cavity to cavity during injection molding. However, filling imbalance has been existed in the geometrically balanced runner system. In this study, we made an experiment and surveyed that are filling balanced variation according to molding condition with thermoplastic vulcanizate (TPV). Also, we conducted experiments in order to know the influence of filling balance for runner core pin (RC pin).

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.225-230
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• 1996

A three-dimensional reactor core simulation code, MASTER has been developed as a part of ADONIS which is the Korean core design package in KAERI. CASMO-3 is used as a precedent lattice code for two-group microscopic cross section and heterogeneous formfunctions. The pin power reconstruction capability of CASMO-3/MASTER was evaluated for a validation and verification Five B&W critical experiments were selected as benchmark problems. These problems included two experiments for CE-type and three for WH-type fuel assemblies. Two of them contained gadolinia rods as burnable absorber. Comparison of the calculated pin power distributions with the measured ones demonstrate that CASMO-3/MASTER can predict the pin power distribution as well as CASMO-3/SIMULATE-3.