• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.235-241
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• 1995

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor Power Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems.

• 한국초전도ㆍ저온공학회논문지
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• 제6권2호
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• pp.25-28
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• 2004

Three-phase inductive superconducting fault current limiter (SFCL) with DC reactor rated on 6.6 $KV_{rms}/200 A_{rms}$ has been developed in Korea. This system consists of one DC reactor, AC/DC power converter, and a three-phase transformer, which is called magnetic core reactor (MCR). This paper deals with the short-circuit analysis of the SFCL. The DC reactor was the HTS solenoid coil whose inductance was 84mH. The power converter was performed as the dual-mode operation for dividing voltage between the rectifying devices. The short-term normal operation (1 see) and short-circuit tests (2∼3 cycles) of this SFCL were performed successfully. In regular short-circuit test, the fault current was limited as 30% of rated short-circuit current at 2 cycles after the fault. The experimental results have a very similar tendency to the simulation results. Using the technique for the fault detection and SCR firing control, the fault current limiting rate of the SFCL was improved. From this research, the parameters for design and manufacture of large-scale SFCL were obtained.

• Nuclear Engineering and Technology
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• 제32권3호
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• pp.261-273
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• 2000

During an extensive review made as part of the Integrated Diagnosis System project of the Maiden Reactor Project, MOAS (Maryland Operator Advisory System) was identified as one of the most thorough systems developed thus far. MOAS is an integrated on-line diagnosis system that encompasses diverse functional aspects that are required for an effective process disturbance management: (1) intelligent process monitoring and alarming, (2) on-line sensor data validation and sensor failure diagnosis, (3) on-line hardware (besides sensors) failure diagnosis, and (4) real-time corrective measure synthesis. The MOAS methodology was used at the Maiden Man-Machine Laboratory HAMMLAB of the OECD Maiden Reactor Project. The performance of MOAS, developed in G2 real-time expert system shell for the high-pressure preheaters of the NORS process in the HAMMLAB, was tested against a variety of transient scenarios, including failures of the control valves and sensors, and tube leakage of the preheaters. These tests showed that MOAS successfully carried out its intended functions, i.e., quickly recognizing an occurring disturbance, correctly diagnosing its cause, and presenting advice on its control to the operator. The lessons learned and insights gained during the implementation and performance tests also are discussed.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.369-376
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• 2019

A complete solution for a soluble-boron-free (SBF) small modular reactor (SMR) is pursued with a new burnable absorber concept, namely centrally-shielded burnable absorber (CSBA). Neutronic flexibility of the CSBA design has been discussed with fuel assembly (FA) analyses. Major design parameters and goals of the SBF SMR are discussed in view of the reactor core design and three CSBA designs are introduced to achieve both a very low burnup reactivity swing (BRS) and minimal residual reactivity of the CSBA. It is demonstrated that the core achieves a long cycle length (~37 months) and high burnup (~30 GWd/tU), while the BRS is only about 1100 pcm and the radial power distribution is rather flat. This research also introduces a supplementary reactivity control mechanism using stainless steel as mechanical shim (MS) rod to obtain the criticality during normal operation. A further analysis is performed to investigate the local power peaking of the CSBA-loaded FA at MS-rodded condition. Moreover, a simple $B_4C$-based control rod arrangement is proposed to assure a sufficient shutdown margin even at the cold-zero-power condition. All calculations in this neutronic-thermal hydraulic coupled investigation of the 3D SBF SMR core are completed by a two-step Monte Carlo-diffusion hybrid methodology.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1037-1042
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• 2018

The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (${\beta}_{eff}$), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to $7.90{\pm}0.01mk$ which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2556-2563
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• 2021

Metaheuristic algorithms can work well in solving or optimizing problems, especially those that require approximation or do not have a good analytical solution. Particle swarm optimization (PSO) is one of these algorithms. The response quality of these algorithms depends on the objective function and its regulated parameters. The nonlinear nature of the pressurized light-water nuclear reactor (PWR) dynamics is a significant target for PSO. The two-point kinetics model of this type of reactor is used because of fission products properties. The proportional-integral-derivative (PID) controller is intended to control the power level of the PWR at a short-time transient. The absolute error (IAE), integral of square error (ISE), integral of time-absolute error (ITAE), and integral of time-square error (ITSE) objective functions have been used as performance indexes to tune the PID gains with PSO. The optimization results with each of them are evaluated with the number of function evaluations (NFE). All performance indexes achieve good results with differences in the rate of over/under-shoot or convergence rate of the cost function, in the desired time domain.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2395-2406
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• 2023

To study the influence of parameter uncertainty in small pressurized water reactor (SPWR) once-through steam generator (OTSG), the nonlinear mathematical model of the SPWR is firstly established. Including the reactor core model, the OTSG model and the pressurizer model. Secondly, a control strategy that both the reactor core coolant average temperature and the secondary-side outlet pressure of the OTSG are constant is adopted. Then, the uncertainty quantification method is established based on Latin hypercube sampling and statistical method. On this basis, the quantitative platform for parameter uncertainty of the OTSG is developed. Finally, taking the uncertainty in primary-side flowrate of the OTSG as an example, the platform application work is carried out under the variable load in SPWR and step disturbance of secondary-side flowrate of the OTSG. The results show that the maximum uncertainty in the critical output parameters is acceptable for SPWR.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.147-159
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• 2024

Plutonium-238 has always been considered as the one of the promising radioisotopes for space nuclear power supply, which has long half-life, low radiation protection level, high power density, and stable fuel form at high temperatures. The industrial-scale production of ²³⁸Pu mainly depends on irradiating solid ²³⁷NpO₂ target in high flux reactors, however the production process faces problems such as large fission loss and high requirements for product quality control. In this paper, a conceptual design study of producing ²³⁸Pu in a multi-purpose high flux reactor was evaluated and analyzed, which includes a sensitivity analysis on ²³⁸Pu production and a further study on the irradiation scheme. It demonstrated that the target structure and its location in the reactor, as well as the operation scheme has an impact on ²³⁸Pu amount and product quality. Furthermore, the production efficiency could be improved by optimizing target material concentration, target locations in the core and reflector. This work provides technical support for irradiation production of ²³⁸Pu in high flux reactors.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 학술대회 논문집 정보 및 제어부문
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• pp.576-578
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• 2005

This paper deals with the design of power topology for nuclear power plants. Although rod control system is still classified into non-safety class. much attention on its reliability issue has been given so far because of its importance for the stable operation of the reactor in the plant. In terms of technical aspects, proposed design is reviewed to satisfy system requirements. This paper deals with a design of power topology for driving Control Element Drive Mechanism (CEDM) that is used to withdraw or insert control rods in nuclear reactor.

• Journal of international Conference on Electrical Machines and Systems
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• 제3권3호
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• pp.282-288
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• 2014

A large capacity power conversion system constructed by using two or more existing power converters has a lot of flexibility in how the power converters are used. However, at the same time, it has a problem of cross current flows between power converters. The cross current must be suppressed by controlling the system while miniaturizing the combination reactor. This paper focuses on two current control methods of a power conversion system constructed by using two power converters connected in parallel supplying the same power. In order to elucidate the control performance of cross current, each control method which are aimed at controlling cross current and not directly controlling it are examined in simulations.