• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1769-1785
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• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.

• 대한기계학회논문집
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• 제10권1호
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• pp.7-14
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• 1986

본 연구에서는 가압경수형 원자로심을 모의하는 3*3배열로 된 모의연료 봉다발의 실험장치를 이용하여 재완수과정의 유동특성과 열전달특성을 파악하였으며, 재완수과정중 연료봉의 온도거동을 예측하는 REFLUX코드를 최근 개발된 연구자료를 토대로 수정하여 본실험결과와 비교하였다.

• 한국자동차공학회논문집
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• 제16권1호
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• pp.64-70
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• 2008

Temperatures of engine head and liner depend on many factors such as spray and combustion process, coolant passage flow and engine related structures. To estimate the temperature distribution of engine structure, multi-dimensional computational fluid dynamics (CFD) codes have been mainly adopted. In this case, it is of great importance to obtain the realistic wall temperature distribution of entire engine structure. In the present work, a CFD-FEM coupling methodology was presented to address this demand. This approach was applied to a real large-size marine diesel engine. CFD combustion and coolant flow simulations were coupled to FEM temperature analysis. Wall heat flux and wall temperature data were interfaced between combustion simulation and solid component temperature analysis via translator by a commercial CFD package named FIRE by AVL. Heat transfer coefficient and surface temperature data were exchanged and mapped between coolant flow simulation and FEM temperature analysis. Results indicate that there exists the optimum cell thickness near combustion chamber wall to reasonably predict the wall heat flux during combustion period. The present study also shows that the effect of cell refining on predicting in-cylinder pressure during combustion is negligible. Hence, the basic guidance on obtaining the wall heat flux needed for the reasonable CFD-FEM coupling analysis has been established. It is expected that this coupling methodology is a robust tool for practical engine design and can be applied to further assessment of the temperature distribution of other engine components.

• Nuclear Engineering and Technology
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• 제30권1호
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• pp.40-46
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• 1998

A set of experiments using a 1/10 scale model pump which was manufactured to simulate performance of reactor coolant pump(RCP) of Y.G.N # 3 and 4, was executed in single phase(at atmospheric pressure and room temperature) and near-saturation(300 ~ 600kPa). The pump characteristics in single phase flow was similar to the characteristics of the RCP. The pump characteristic curves at nearly saturated state were correlated in terms of flow coefficient and head coefficient for subcooled temperature using the cavitation number defined as (equation omitted), which can be predicted the cavitation possibility. The pump behavior around the saturated temperature almost consists with single phase behavior until the cavitation occurs(When cavitation occurs. When the flow coefficient is about 0.12), the pump head rapidly degrades. In this situation, subcooled temperature is about 1.8~8$^{\circ}C$ and cavitation number of model pump is 1.0 ~ 1.7.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.573-580
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• 2013

The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The $CANDU^{(R)}$ reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC) and Large Break Loss of Coolant Accident (LBLOCA) events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

• Nuclear Engineering and Technology
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• 제9권3호
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• pp.125-137
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• 1977

천연 산화 우라늄 핵연료와 중수 감속재를 사용하는 월성 1호기에 관한 온도 계수를 조사하였다. 핵연료, 감속재 및 냉각재 온도변화에 따른 중성자의 유효 증배 인자의 변화를 계산하였다. 계산된 결과를 LATREP 전자계산 code에 의한 온도 계수 값들과 비교하였다.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2699-2708
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• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.292-295
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• 2008

In a SCWR (SuperCritical pressure Water cooled Reactor), the coolant temperature initially at below the pseudo-critical temperature at the bottom of a reactor core increases as the coolant flows upward through the sub-channels of the fuel assemblies, and it finally becomes higher than the pseudo-critical temperature when it leaves the reactor core. At certain conditions, heat transfer deterioration occurs near the pseudo-critical temperature and it may cause a drastic rise of the fuel surface temperature resulting a fuel failure. Therefore, an accurate estimation of the heat transfer coefficient is very important for the thermal-hydraulic design of a reactor core. An experiment on heat transfer to the vertically upward flowing $CO_2$ at a supercritical pressure in a circular tube were performed at KAERI. The internal diameter of the test section is 6.32 mm, which corresponds to the hydraulic diameter of a sub-channel in the conceptional design proposed by KAERI. The test range of the mass flux is 285 to 1200 kg/m$^2$s and the maximum heat flux is 170 kW/m$^2$. The inlet pressure is maintained at 8.12 MPa, which is 1.1 times the critical pressure. A new correlation, which covers both the normal and deterioration heat transfer regimes was proposed and compared with the estimations by exiting correlations.

• 오토저널
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• 제15권3호
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• pp.55-67
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• 1993

In this study, the thermal behavior characteristics of the cylinder block of a small 3-cylinder, 4-stroke gasoline engine were analyzed, using the 3-dimensional finite element method. Before numerical analyses were conducted, the performance test and the heat transfer experiment of the engine were carried out in order to prepare the input data for the computations. Engine cycle simulation was performed to obtain the heat transfer coefficient and the temperature of the gas and the mean heat transfer coefficient of coolant. Temperature fields as a result of steady-state heat transfer were obtained and compared with experimental results measured at specific points of the inner and the outer walls of the cylinder block. The thermal stress and deformation characteristics resulting from the nonuniform temperature distributions of the block were investigated. The effects of the thermal behaviors of the cylinder block on the engine operations and the unfavourable aspects of excessive thermal loading were examined on the basis of the calculated results.

• Journal of Advanced Marine Engineering and Technology
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• 제21권3호
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• pp.267-277
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• 1997

In this paper, heat conduction characteristics of the cylinder block of a small 3 - cylinder, 4¬stroke gasoline engine were analyzed using the 3 - dimensional finite element method. Based on the experimental data, the engine cycle simulation was carried out in order to obtain the heat transfer coefficient and the temperature of the gas and the mean heat transfer coefficient of the coolant. Heat transfer data of the gas, which were averaged with respect to exposure time to the wall, were taken as convective boundary conditions corresponding to the operating conditions to obtain the temperature fields of the block. Finally silicon nitride(Si3N4) was taken as the material of the block liner in order to investigate its temperature distribution characteristics and compare the results with the original ones.