• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.244-252
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• 2021

Wind plays an important role in cases of unexpected radioactive pollutant dispersion, deciding distribution and concentration of the leaked substance. The accurate prediction of wind has been challenging in numerical weather prediction models, especially near the surface because of the complex interaction between turbulent flow and topographic effect. In this study, we investigated the characteristics of atmospheric dispersion of radioactive material (i.e. ¹³⁷Cs) according to the simulated boundary layer around the HANARO research nuclear reactor in Korea using the Weather Research and Forecasting (WRF)-Mesoscale Model Interface (MMIF)-California Puff (CALPUFF) model system. We examined the impacts of orographic drag on wind field, stability calculation methods, and planetary boundary layer parameterizations on the dispersion of radioactive material under a radioactive leaking scenario. We found that inclusion of the orographic drag effect in the WRF model improved the wind prediction most significantly over the complex terrain area, leading the model system to estimate the radioactive concentration near the reactor more conservatively. We also emphasized the importance of the stability calculation method and employing the skillful boundary layer parameterization to ensure more accurate low atmospheric conditions, in order to simulate more feasible spatial distribution of the radioactive dispersion in leaking scenarios.

• Nuclear Engineering and Technology
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• v.40 no.7
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• pp.607-614
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• 2008

Since pipes with wall-thinning defects can collapse at fluid pressure that are lower than expected, the collapse moment of wall-thinned pipes should be determined accurately for the safety of nuclear power plants. Wall-thinning defects, which are mostly found in pipe bends and elbows, are mainly caused by flow-accelerated corrosion. This lowers the failure pressure, load-carrying capacity, deformation ability, and fatigue resistance of pipe bends and elbows. This paper offers a support vector regression (SVR) model further enhanced with a fuzzy algorithm for calculation of the collapse moment and for evaluating the integrity of wall-thinned piping systems. The fuzzy support vector regression (FSVR) model is applied to numerical data obtained from finite element analyses of piping systems with wall-thinning defects. In this paper, three FSVR models are developed, respectively, for three data sets divided into extrados, intrados, and crown defects corresponding to three different defect locations. It is known that FSVR models are sufficiently accurate for an integrity evaluation of piping systems from laser or ultrasonic measurements of wall-thinning defects.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2112-2124
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• 2023

With the rise of economic and safety standards for nuclear reactors, new concepts of Gen-IV reactors and modular reactors showed more complex designs that challenge current tools for reactor physics analysis. A Monte Carlo (MC) two-step method was proposed in this work. This calculation scheme uses the continuous-energy MC method to generate multi-group cross-sections from heterogeneous models. The multi-group MC method, which can adapt locally-heterogeneous models, is used in the core calculation step. This calculation scheme is verified using a Gen-IV modular lead-based fast reactor (LFR) benchmark case. The influence of homogenized patterns, scatter approximations, flux separable approximation, and local heterogeneity in core calculation on simulation results are investigated. Results showed that the cross-sections generated using the 3D assembly model with a locally heterogeneous representation of control rods lead to an accurate estimation with less than 270 pcm bias in core reactivity, 0.5% bias in control rod worth, and 1.5% bias on power distribution. The study verified the applicability of multi-group cross-sections generated with the MC method for LFR analysis. The study also proved the feasibility of multi-group MC in core calculation with local heterogeneity, which saves 85% time compared to the continuous-energy MC.

• Journal of the Korean Society of Safety
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• v.36 no.5
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• pp.86-91
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• 2021

Calculating the scrutable core damage frequency (CDF) of nuclear power plants is an important component of the seismic probabilistic safety assessment (SPSA). In this work, a simple approach is developed to calculate CDF from minimal cut sets (MCSs) with non-rare events. When conventional calculation methods based on rare event approximations are employed, the CDF of industry SPSA models is significantly overestimated by non-rare events in the MCSs. Recently, quantification algorithms using binary decision diagrams (BDDs) have been introduced to prevent CDF overestimation in the SPSA. However, BDD structures are generated from a small part of whole MCSs due to limited computational memory, and they cannot be reviewed due to their complicated logic structure. This study suggests a simple approach for scrutinizing the CDF calculation based on whole MCSs in the SPSA system analysis model. The proposed approach compares the new results to outputs from existing algorithms, which helps in avoiding CDF overestimation.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1002-1012
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• 2024

To remove insoluble fission products, which could possibly cause reactor instability and significantly reduce heat transfer efficiency from primary system of molten salt reactor, a helium bubbling method is employed into a passive molten salt fast reactor. In this regard, two-phase flow behavior of molten salt and helium bubbles was investigated experimentally because the helium bubbles highly affect the circulation performance of working fluid owing to an additional drag force. As the helium flow rate is controlled, the change of key thermal-hydraulic parameters was analyzed through a two-phase experiment. Simultaneously, to assess the applicability of numerical model for the analysis of two-phase flow behavior, the numerical calculation was performed using the OpenFOAM 9.0 code. The accuracy of the numerical analysis code was evaluated by comparing it with the experimental data. Generally, numerical results showed a good agreement with the experiment. However, at the high helium injection rates, the prediction capability for void fraction of helium bubbles was relatively low. This study suggests that the multiphaseEulerFoam solver in OpenFOAM code is effective for predicting the helium bubbling but there exists a room for further improvement by incorporating the appropriate drag flux model and the population balance equation.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2628-2641
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• 2023

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosols suspended in the containment will settle onto the wall or sump water through the natural deposition mechanism, thereby reducing atmospheric radioactivity. Aiming at the low accuracy of the aerosol model in the ISAA code, this paper improves the natural deposition model of aerosol in the containment. The aerosol dynamic shape factor was introduced to correct the natural deposition rate of non-spherical aerosols. Moreover, the gravity, Brownian diffusion, thermophoresis and diffusiophoresis deposition models were improved. In addition, ABCOVE, AHMED and LACE experiments were selected to validate and evaluate the improved ISAA code. According to the calculation results, the improved model can more accurately simulate the peak aerosol mass and respond to the influence of the containment pressure and temperature on the natural deposition rate of aerosols. At the same time, it can significantly improve the calculation accuracy of the residual mass of aerosols in the containment. The performance of improved ISAA can meet the requirements for analyzing the natural deposition behavior of aerosol in containment of advanced PWRs in severe accident. In the future, further optimization will be made to address the problems found in the current aerosol model.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.713-718
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• 1998

Henry-Fauske critical flow model was incorporated into TRAC-PF1 to correct some errors in the original TRAC-PFI critical flow model. Henry-Fouske mode1 was numerically implemented and tested against steady-state steam-water experimental data. The model was incorporated into TRAC-PFI and code assessment against Marviken Critical Flow Tests 15 and 24 was carried out. Calculations using RELAP5/MOD3 were also made for comparison. Ten cases were calculated each test and sensitivity study on nodalization as well as critical flow or model was performed Stand-alone numerical model test and code assessment were done for verification and validation of code modification. Calculation results show that the modified version of TRAC-PF1 has a capability to model critical flow correctly in various conditions.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.241-253
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• 2001

Under severe accidents, the pressure and temperature response has an important role for the integrity of a nuclear power plant containment. The history of the pressure and temperature is characterized by the amount and state of steam/air mixture in a containment. Recently, the heat transfer rate to the structure surface is supposed to be increased by the wavy interface formed on condensate film. However, in the calculation by using CONTAIN code, the condensation heat transfer on a containment wall is calculated by assuming the smooth interface and has a tendency to be underestimated for safety. In order to obtain the best- estimate heat transfer calculation, we investigated the condensation heat transfer model in CONTAIN 1.2 code and adopted the new forced convection correlation which is considering wavy interface. By using the film tracking model in CONTAIN 1.2 code, the condensate film is treated to consider the effect of wavy interface. And also, it was carried out to investigate the effect of the different cell modelings - 5-cell and 10-cell modeling - for KNGR(Korean Next Generation Reactor) containment phenomena during a severe accident. The effect of wavy interface on condensate film appears to cause the decrease of peak temperature and pressure response . In order to obtain more adequate results, the proper cell modeling was required to consider the proper flow of steam/air mixture.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.146-146
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• 2002