• Nuclear Engineering and Technology
• /
• v.52 no.5
• /
• pp.937-946
• /
• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.203-210
• /
• 2018

As the modernization of the nuclear instrumentation system progresses, research reactors have adopted digital wide-range neutron power measurement (DWRNPM) systems. These systems typically monitor the neutron flux across a range of over 10 decades. Because neutron detectors only measure the local neutron flux at their position, the local neutron flux must be converted to total reactor power through calibration, which involves mapping the local neutron flux level to a reference reactor power. Conventionally, the neutron power range is divided into smaller subranges because the neutron detector signal characteristics and the reference reactor power estimation methods are different for each subrange. Therefore, many factors should be considered when preparing the calibration procedure for DWRNPM channels. The main purpose of this work is to serve as a reference for performing the calibration of DWRNPM systems in research reactors. This work provides a comprehensive overview of the calibration of DWRNPM channels by describing the configuration of the DWRNPM system and by summarizing the theories of operation and the reference power estimation methods with their associated calibration procedure. The calibration procedure was actually performed during the commissioning of an open-pool type research reactor, and the results and experience are documented herein.

• Nuclear Engineering and Technology
• /
• v.54 no.10
• /
• pp.3897-3908
• /
• 2022

High-fidelity nuclear data libraries and neutronics simulation tools are essential for the development of fast reactors. The IAEA coordinated research project on "Neutronics Benchmark of CEFR Start-Up Tests" offers valuable data for the qualification of nuclear data libraries and neutronics codes. This paper focuses on the verification and validation of the CEFR start-up modelling using OpenMC Monte-Carlo code against the experimental measurements. The OpenMC simulation results agree well with the measurements in criticality, control rod worth, sodium void reactivity, temperature reactivity, subassembly swap reactivity, and reaction distribution. In feedback coefficient evaluations, an additional state method shows high consistency with lower uncertainty. Among 122 relative errors in the benchmark of the distribution of nuclear reaction, 104 errors are less than 10% and 84 errors are less than 5%. The results demonstrate the high reliability of OpenMC for its application in fast reactor simulations. In the companion paper, the influence of cross-section libraries is investigated using neutronics modelling in this paper.

• KSCE Journal of Civil and Environmental Engineering Research
• /
• v.41 no.5
• /
• pp.505-512
• /
• 2021

A 3D model was developed to calculate the UV intensity of a submerged-type UV disinfection reactor. Numerical experiments were conducted by inputting the design factors of an open channel-type disinfection reactor and a pipe-type disinfection reactor that were installed in an actual sewage treatment plant. The following data were obtained: The average UV intensity of the installed open channel-type reactor and pipe-type reactor was 7.87 mW/cm² and 13.09 mW/cm², respectively; the UV dose reflecting the UV irradiation time and taking into account attenuation effects such as mixing imbalance, lamp aging, temperature, and fouling, was expected to be 21.1 mJ/cm² and 24.8 mJ/cm², respectively, and these values are 5 % and 24 % higher than the target UV dose of 20 mJ/cm², respectively. By using the UV3D model, the optimal lamp position, which maximizes the average UV intensity without changing the size of the disinfection reactor or lamp output power, can be found. In this case, by only adjusting the lamp position, the average UV intensity can be increased by 0.9 % for the open channel-type and 0.5 % for the pipe-type, respectively. A better average UV intensity can be obtained by model simulation. By adjusting the horizontal and vertical ratio of the open channel-type reactor and by moving the lamp position, the average UV intensity can be increased by 7.4 % more than the present case.

• Proceedings of the Korean Institute of IIIuminating and Electrical Installation Engineers Conference
• /
• 2004.11a
• /
• pp.101-104
• /
• 2004

The past harmonic filter was reactor and Zig-zag. However neutral reactor over heating become an issue. If take advantage of Zig-zag TR line and Neutral line Open Delta mode TR to disappear TR sounds and harmonic by the help of experiments. The disappear harmonic have not relation of neutral line electric current quantity

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1604-1615
• /
• 2023

In an open-pool type research reactor with a downward forced flow in the core, pipes can be under sub-atmospheric pressure because of the large pressure drop at the reactor core in the atmospheric pool. Sub-atmospheric pressure can result in air inflow into the pipe from the pressure difference between the atmosphere and the inside of the pipe, which in a postulated pipe break scenario can lead to the breakdown of the cooling pump. In this study, a plant-scale experiment was conducted to study air inflow in large piping systems by considering the actual operational conditions of an advanced research reactor. The air inflow rate was measured, and the entrained air was visualized to investigate the behavior of air inflow and flow regime depending on the pipe break size. In addition, the developed drift-flux model for a large vertical pipe with a diameter of 600 mm was compared with other correlations. The flow regime transition in a large vertical pipe under downward flow was also studied using the newly developed drift-flux model. Consequently, the characteristics of two-phase flow in a large vertical pipe were found to differ from those in small vertical pipes where liquid recirculation was not dominant.

• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.1002-1012
• /
• 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
• /
• v.42 no.1
• /
• pp.125-130
• /
• 2010

Sipping measurements were implemented at the Open Pool Australian Light water reactor (OPAL) to test for failure in reactor fuel elements. Fission product released by the fuel element into the pool water was measured using both High Purity Germanium (HPGe) detection via samples and a NaI(Tl) detection in-situ with the sipping device. Results from two fuel elements are presented.

• The Transactions of the Korean Institute of Electrical Engineers
• /
• v.33 no.8
• /
• pp.299-309
• /
• 1984

This paper proposes the adapter for interrupting capacity augmentation of circuit breaker which can be applied in case of shortage in a existing circuit breaker's interrupting capacity due to utility system extension. The adapter utilizes two winding type of reactor instead of single winding type of reactor and the control of 2ry circuit is excuted by a triac interlocked with the system protective relays actuation so as to cut out the reactor by short circuit of the 2ry winding in normal situation and to cut in the reactor by open circuit of the 2ry winding in abnomal situation such as short circuit accident. As a result of the theoritical analysis and experiment, it is proved that the adaptor can reduce the voltage crop and iron loss due to the reactor signigicantly in normal system condition and do a role of reactor upon the power system accident.

• 제어로봇시스템학회:학술대회논문집
• /
• 2001.10a
• /
• pp.97.2-97
• /
• 2001

This work is concerned with the improvement of the productivity and the product quality in the polymerization reactors by using model-on-demand predictive control(MoDPC). This technique is applied to a continuous styrene polymerization reactor and a semibatch methyl methacrylate (MMA)/vinyl acetate(VAc) copolymerization reactor. The regress is constructed with the most influential variables the conversion and the jacket inlet temperature for the styrene polymerization reactor, and the free volume and the reactor temperature for the MMA/VAc copolymerization reactor through open loop operations. From the simulation results for setpoint tracking and disturbance rejection problems, it is demonstrated that the MoDPC shows ...