• Nuclear Engineering and Technology
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• 제49권1호
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• pp.71-81
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• 2017

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.134-140
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• 2017

A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2005년도 추계학술대회논문집
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• pp.267-270
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• 2005

Finite element model of pressurized water reactor internals were obtained using ANSYS software package to analyze dynamic characteristics. The pressure vessel, hold-down ring, alinement key, core support barrel(CSB), upper guide structure(UGS) and fluid gap were fully modeled using structural solid element(SOLID45) and fluid element(FLUID80) which is one of element types. Also modal analysis using the above finite element model has been performed. As a result, it was found that the fundamental beam mode natural frequency of the CSB were 8.2 Hz, the shell mode one 14.5 Hz. To verify the Finite Element Analysis(FEA), we compare the analysis result with experimental data that is obtained from the plant IVMS(internal Vibration Monitoring System). The experimental results are good agreement with the FEA model.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.924-933
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• 2015

The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional $^{235}U$ burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using $^{233}U$, $^{242}Pu$, $^{150}Nd$, and $^{133}Cs$ as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

• 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.

• Tribology and Lubricants
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• 제19권1호
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• pp.21-25
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• 2003

The 440C stainless steel is used for ball bearings and bevel gears in the control rod drive mechanism for the integral reactor, SMART. The friction characteristics of 400C stainless steel a investigated in sliding motion using the reciprocating tribometer which can simulate the operating conditions of the control rod drive mechanism. Highly purified water is used as lubricant, and the water is heated and pressurized in the autoclave. Friction force on the reciprocating specimens is measured by the load cells and transformed into friction coefficient. It is verified that frictional characteristic of the 440c steel is not drastically changed up to operating temperature and variation of friction coeffcient at operating temperature from room temperature to 160$^{\circ}C$ is within 5%.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.96-99
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• 2014

The aim of the study was to evaluate the water treatment of pressurized water reactor secondary side by the mixed amine of ammonia and ethanolamine, from the standpoint of corrosion control, as compared with all volatile treatment of ammonia. The pressurized water reactor systems have switched a secondary side pH control agent to minimize the corrosion in the moisture separator/reheater and feedwater heater systems and the transport of corrosion products into steam generator. As results of field test, pH was increased in the steam generator and the wet steam area of moisture separator/reheater and the concentration of Fe were decreased by more than 50% as compared with water treatment of ammonia.

• 한국태양에너지학회 논문집
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• 제27권3호
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• pp.45-54
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• 2007

For a conventional natural-circulation type solar water heater, the pressure head is limited by the height between the storage tank and hot water tap. Therefore, it is difficult to provide sufficient hot water flow rate for general usage. This study deals with a design modification of the storage tank to utilize the tap-water pressure to increase hot-water supply Based on fluid dynamic and heat transfer theories, a series of modeling and simulation is conducted to achieve practical design requirements. An experimental setup is built and tested and the results are compared with theoretical simulation model. The storage tank capacity is 240 l and the outer diameter of piping was 15 mm. Number of tube turns tested are 5, 10, and 15. Starting with initial storage tank temperature of $80^{\circ}C$, the temperature variation of the supply hot water is investigated against time, while maintaining minimum flow rate of 10 1/min. Typical results show that the hot water supply of minimum $30^{\circ}C$ can be maintained for 34 min with tap-water supply pressure of 2.5 atm, The relative errors between modeling and experiments coincide well within 10% in most cases.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3418-3422
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• 2007

The neutron imaging technique was used to investigate the water discharge characteristics at PEMFC. Prior to investigation of water discharge characteristics, the linear attenuation coefficient for water at Neutron Radiography Facility (NRF) was calibrated. The feasibility test apparatus was consisted of pressurized air and water in order to simulate the actual operating PEMFC. The feasibility tests have been performed at 1-parallel serpentine type with 100 $cm^2$ active area and different air flow rate (1, 2, and 4 lpm). The total water volume variations at each condition were calculated from the neutron images. The water at channel is well discharged as soon as supplying the pressurized air into the PEMFC. However, because the water at MEA isn't removed the total water volume is constant after 150. Therefore more effective method is needed in order to discharge water at MEA, and the neutron imaging technique is helpful for it.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.589-596
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• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.