• Nuclear Engineering and Technology
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• v.13 no.3
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• pp.153-161
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• 1981

The dynamic behavior of the Ko-ri Unit 1 nuclear reactor following some credible and postulated accidents has been analyzed to a certain extent by means of neutronics and temperature equations formulated in terms of point reactor model. In general, the result of numerical calculation is harnessed to be incorporated in more elaborate models so as to predict transient behavior in a reliable mode as a part of accident analysis. It is shown in the case of power response upon an uncontrolled withdrawal of rod cluster control assembly at hot full power that the point reactor kinetics model proves to be good enough to reproduce the generic features described in the final safety analysis report of the Ko-ri Unit 1.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.267-276
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• 2017

In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium) utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the $^*SIMULATE$ module of the Reactor Fueling Simulation Program (RFSP) code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the $^*INSTANTAN$ module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the $^*INSTANTAN$ module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

• Nuclear Engineering and Technology
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• v.49 no.6
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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.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3966-3978
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• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.30-35
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• 2022

In a nuclear reactor, gamma radiation is the dominant energy deposition in non-fuel regions. Heat is generated upon gamma deposition and consequently affects the mechanical and thermal structure of the material. Therefore, the safety of samples should be carefully considered so that their integrity and quality can be retained. To evaluate relevant parameters, an in-core gamma thermometer (GT) was used to measure gamma heating (GH) throughout the operation of the McMaster nuclear reactor (MNR) at four irradiation sites. Additionally, a Monte Carlo reactor physics code (Serpent-2) was utilized to model the MNR with the GT located in the same irradiation sites used in the measurement to verify its predictions against measured GH. This research aids in the development of modeling, calculation, and prediction of the GH utilizing Serpent-2 as well as implementing a new GH measurement at the MNR core. After all uncertainties were quantified for both approaches, comparable GH profiles were observed between the measurements and calculations. In addition, the GH values found in the four sites represent a strong level of radiation based on the distance of the sample from the core. In this study, the maximum and minimum GH values were found at 0.32 ± 0.05 W/g and 0.15 ± 0.02 W/g, respectively, corresponding to 320 Sv/s and 150 Sv/s. These values are crucial to be considered whenever sample is planned to be irradiated inside the MNR core.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2540-2549
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• 2022

For a pressurized water reactor power plant, the reactor coolant pump (RCP) is a kernel component. And for a canned motor RCP, the rotor system's properties determines its safety. The liquid coolant inside the canned motor RCP fills clearance between the metal shields of rotor and stator, forming a lengthy clearance flow. The influence of inlet preswirl on rotordynamic coefficients of clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally investigated in this work. A quasi-steady state computational fluid dynamics (CFD) method has been used to investigate the influence of inlet preswirl. A vertical experiment rig has also been established for this purpose. Rotordynamic coefficients on different inlet preswirl ratios (IR) are obtained through CFD and experiment. Results show that the cross-coupled stiffness of the clearance flow would change significantly with inlet preswirl, but other rotordynamic coefficients would not change significantly with inlet preswirl. For the case of clearance flow between the stator and rotor cans, influence of inlet preswirl is not so significant as the IR is not large enough.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.344-350
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• 2021

As the High Flux Advanced Neutron Application Reactor (HANARO) has been recently required to support new R&D relevant to future nuclear systems requiring a much higher neutron fluence, the development of irradiation capsule technology for long-term irradiation testing was performed in three steps (3, 5, 10 dpa). At first, several design improvements of a standard capsule were suggested based on a failure analysis of the capsule and successfully applied for irradiation testing at HANARO at up to eight reactor operation cycles equivalent to 3 dpa. Based on a schematic stress analysis of the vulnerable parts of the previous capsule, an optimized design of the capsule was made for 5 dpa irradiation. The newly designed capsule was safely out-pile tested up to 450 days, which was equivalent to 5 dpa irradiation in the reactor. The test results were submitted to the Reactor Safety Review Committee of HANARO and irradiation testing for 5 dpa was approved. The capsule was also successfully out-pile tested to evaluate the possibility of irradiation testing for 10 dpa. For a higher neutron fluence exceeding 10 dpa, new capsule technologies, including a new capsule that has a different bottom design and neutron flux boosting capsule, were also suggested.

• Journal of the Institute of Electronics and Information Engineers
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• v.49 no.9
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• pp.279-286
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• 2012

Recently, design of Universal Soft Controller(USC) has been applied to the advanced control room for nuclear power plant. USC is software-based manual control means to control safety components as well as non-safety components in the highly-integrated control room. Therefore, design feature of USC is essential for the implementation of a single workstation in the advanced control room. The traditional control room is replaced by computer-driven consolidated operator interfaces. Considering our design has further reduced the probability of USC spurious signals by requiring two distinct operator control actions to generate any control signal. The reality of USC does not increase the probability of reactor trip because the probability of spurious USC signal is negligible. Universal Soft Control represents a significant evolution in nuclear I&C/HSI System. USC integrates the indicators and controls from multiple divisions into a single integrated visual display unit(VDU) based HSI(Human System Interface). In order to prevent adverse influence on safety function performance from USC failure, ESFAS signals are applied to safety components or functions. In addition, safety manual switches have priority over USC's signals. Therefore, spurious USC signals can be momentarily blocked by selecting a soft control command from the safety VDU.

• Journal of the Korean Institute of Gas
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• v.2 no.2
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• pp.1-11
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• 1998

A quantitative risk assessment and consequence analysis for PBL(Poly Butadiene Latex) reaction processes were performed. As a result of the Quantitative risk assessment, for the accident probability of PBL reactors causing a reaction runaway, was calculated as $9.197{\times}10^{-5}/yr$ The most important factor that affected the accident probability of PBL reactor was the relief device. When the reactor exploded, peak overpressure at the target point was $5.066{\times}10^5(Pa)$ and the range of effects windows to be broken occurred in almost all of the factory areas. The maximum radius of effect was 27m, in which workers could be die by the direct for eardrum damage was calculated at 77m. When the PBL reactor exploded, the extent of structural damage to buildings was calculated from the center of the explosion to a range of 52m. The results of the study's assessment have provided a direction for facility's improvement as well as effective safety investment.