• Journal of the Korean Professional Engineers Association
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• v.30 no.5
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• pp.82-95
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• 1997

A Study of Improvement in Reactor Thermal Power Measurement Method using Kalman Filter. The objectives of the safety analysis of nuclear power plants are to maintain the surface temperature of fuel and fuel cladding within limit value in case of Loss of Coolant accident (LOCA) so that it ensures the safety and reliability of nuclear power plants. The new technique evaluating the reactor power and improvement of existing plant system increase the safety margin of nuclear power plant operation, and accordingly, economic effect will be anticipated. Hereby, 1 would like to introduce reactor power measurement method using Kalman filter that enables to calculate the reactor power more precisely combining the parameters, for example, turbine output as the 1 st stage pressure of high pressure turbine, and reactor power using energy equilibrium relation. It is expected that the new technique will enhance the accuracy of measurement of reactor power and maintain the reliability of nuclear power operation by increasing operational safety margin, and gain the economic benefit

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2094-2106
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• 2022

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1721-1728
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• 2019

The advanced node core code CORCA-3D is one of the independent developed codes of NPIC for the nuclear reactor core design. CORCA-3D code can calculate the few-group cross section, solve the 3D diffusion equations, consider the thermal-hydraulic feedback, reconstruct the pin-by-pin power. It has lots of functions such as changing core status calculation, critical searching, control rod value calculation, coefficient calculation and so on. The main theory and functions of CORCA-3D code are introduced and validated with a lot of reactor measured data and the SCIENCE system. Now, CORCA-3D code has been applied in ACP type reactor nuclear cores design.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.96-104
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• 1995

EPRI URD requires that the reactor be capable of accommodating an unintended CEA drop without initiating a trip and operating at a reduced power with ay single CEA fully inserted. YGN 3 and 4 reactors have 12-finger CEAs, and the CPCS will trip the reactor due to their large reactivities when one of them is dropped at a high power. The ABB-CE reactor power cutback system has been proposed to be used against the 12-Finger CEA drop to avoid the reactor trips. The results of study show that the reactor power cutback can prevent the reactor trips of the 12-Finger CEA drop when the CPCS has enough operating thermal margin (more than 9％ for YGN 3&4 Cycle 1). It is noted, however, that the probability of a 12-Finger CEA drop is very low, less than one per 100 reactor years for YGN 3& and System 80$^{+}$ plants.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.329-338
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• 2010

We developed an achievable control logic for the reactor power level during a power maneuvering event and set up some constraints for the control of the reactor power in a conceptual sodium-cooled fast reactor (KALIMER-600) that was developed at KAERI. For simulating the dynamic behaviors of the plant, we developed a fast-running performance analysis code. Through various simulations of the power maneuvering event, we evaluated some suggested control logic for the reactor power and found an achievable control logic. The objective of the control logic is to search for the position of the control rods that would keep the average temperature of the primary pool constant and, concurrently, minimize the power deviation between the reactor and the BOP cycle during the power maneuvering. In addition, the flow rates of the primary pool and the intermediate loop should be changed according to the power level in order to not violate the constraints set up in this study. Also, we evaluated some movement speeds of the control rods and found that a fast movement of the control rods might cause the power to fluctuate during the power maneuvering event. We suggested a reasonable movement speed of the control rods for the developed control logic.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.165-169
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• 2018

An assessment for determining N-16 activity concentrations during the operation condition of Bangladesh Atomic Energy Commission TRIGA Research Reactor was performed employing several governing equations. The radionuclide N-16 is a high energy (6.13 MeV) gamma emitter which is predominately created by the fast neutron interaction with O-16 present in the reactor core water. During reactor operation at different power level, the concentration of N-16 at the reactor bay region may increase causing radiation risk to the reactor operating personnel or the general public. Concerning the safety of the research reactor, the present study deals with the estimation of N-16 activity concentrations in the regions of reactor core, reactor tank, and reactor bay at different reactor power levels under natural convection cooling mode. The estimated N-16 activity concentration values with 500 kW reactor power at the reactor core region was $7.40{\times}10^5Bq/cm^3$ and at the bay region was $3.39{\times}10^5Bq/cm^3$. At 3 MW reactor power with active forced convection cooling mode, the N-16 activity concentration in the decay tank exit water was also determined, and the value was $4.14{\times}10^{-1}Bq/cm^3$.

• KEPCO Journal on Electric Power and Energy
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• v.1 no.1
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• pp.1-8
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• 2015

When the line and switchgear of the substation system are disconnected, ferro-resonance can occur. This happens even if the capacitive reactance and inductive reactance are not equal, which are not common resonance conditions. Resonance conditions vary depending on the busbar configuration environment. Although the damping resistance method applying the existing saturable reactor to cope with ferro-resonance has been successfully applied on site, there can be loss of normal function during long-term operation. The reason is because the rise in the operating frequency of saturable reactors means the saturation number is increased. Therefore, it can no longer function as saturable reactor since the resistor having inadequate capacity is burned out. To address this problem, in this paper, an EMTP-based simulation test was performed by designing and applying a dual reactor method, which adds an extended divergence reactor to the 1st side of the VT. The test result confirms that when the divergence reactor is inserted, the voltage and current values obtained at the 1st side and 2nd side of the VT as well as current values of divergence reactor part were stabilized from the transient phenomena and return to normal values. When compared with existing measures, although this method is similar in adding having a reactor added to a system regarding ferro-resonance, it has the advantage of being able to prevent ferro-resonance in advance since the reactor is added before the system is saturated. In addition, because it does not use damping resistance, it can extend the equipment life and stabilize its operation. Therefore, there are a lot of differences in terms of its operating characteristics and achivement of goal between the conventional method and new divergence reactor method.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3595-3603
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• 2022

Nuclear safety-class DCS is used for nuclear reactor protection function, which is one of the key facilities to ensure nuclear power plant safety, the maintenance for DCS to keep system in a high reliability is significant. In this paper, Nuclear safety-class DCS system developed by the Nuclear Power Institute of China is investigated, the model of reliability estimation considering nuclear power plant emergency trip control process is carried out using Markov transfer process. According to the System-Subgroup-Module hierarchical iteration calculation, the evolution curve of failure probability is established, and the preventive maintenance optimization strategy is constructed combining reliability numerical calculation and periodic overhaul interval of nuclear power plant, which could provide a quantitative basis for the maintenance decision of DCS system.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.203-210
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• 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
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• v.50 no.5
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• pp.648-653
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• 2018

Elimination of soluble boron from reactor design eliminates boron-induced reactivity accidents and leads to a more negative moderator temperature coefficient. However, a large negative moderator temperature coefficient can lead to large reactivity feedback that could allow the reactor to return to power when it cools down from hot full power to cold zero power. In soluble boron-free small modular reactor (SMR) design, only control rods are available to control such rapid core transient. The purpose of this study is to investigate whether an SMR would have enough control rod worth to compensate for large reactivity feedback. The investigation begins with classification of reactivity and completes an analysis of the reactivity balance in each reactor state for the SMR model. The control rod worth requirement obtained from the reactivity balance is a minimum control rod worth to maintain the reactor critical during the whole cycle. The minimum available rod worth must be larger than the control rod worth requirement to manipulate the reactor safely in each reactor state. It is found that the SMR does have enough control rod worth available during rapid transient to maintain the SMR at subcritical below k-effectives of 0.99 for both hot zero power and cold zero power.