• Proceedings of the KSME Conference
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• 2000.11b
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• pp.228-233
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• 2000

The Korean Next Generation Reactor(KNGR) is a Pressurized Water Reactor adopting direct vessel injection(DVI) to optimize the performance of emergency core cooling system(ECCS). In a certain accident, however, pressurized thermal shock(PTS) of the vessel due to the sudden contact with the injected cold water is expected. In this paper, an accident of Main Steam Line Break(MSLB) has been numerically investigated with direct vessel injections and an increased volume flow rate in some cold legs. Using FLUENT code, temperature distributions of the fluid in the downcomer and of reactor vessel including the core region have been calculated, together with the distribution of convective heat transfer coefficient(CHTC) at the cladding surface of the reactor vessel. The result shows that some parts of the core region of the reactor vessel have higher temperature gradient expressing higher thermal stress.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2000.06a
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• pp.679-686
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• 2000

Reactor Noise is defined as the fluctuations of measured instrumentation signals during full-power operation of reactor which have informations on reactor system dynamics such as neutron kinetics, thermal-hydraulics, and structural dynamics. Reactor noise analyses of ex-core neutron detector signals have been performed to monitor the vibration modes of reactor internals such as fuel assembly and Core Support Barrel in Yonggwang 3&4 Nuclear Power Plant. A real time mode separation technique have been developed and applied for the analyses. It has been found that the first vibration mode frequency of the fuel assembly was around 2.5 Hz, the beam and shell mode frequencies of CSB(Core Support Barrel) 8 Hz and 14.5 Hz, respectively. Also the analyses data base have been constructed for the continuous monitoring and diagnose of the reactor internals.

• 전기의세계
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• v.22 no.1
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• pp.20-27
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• 1973

The multi-group diffusion theory is applied to the analysis of the currently constructing Kori reactor core which is to be refuelled by 3-region fuel loading pattern and also to the comparative study on a conceptually designed 5-region reactor core, under the condition that, apart from the thermal-hydraulic considerations, all the input data referred to here in are assumed to be identical for both cases. The numerical calculation is carried out for quantitative analysis of the characteristics of the two fuel loading patterns in details, and the calculated results show that, so far as the nuclear aspects are concerned, the characteristics of the 5-region reactor core are proved to be superior to those of Kori's 3-region reactor core in general.

• 유체기계공업학회:학술대회논문집
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• 2001.11a
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• pp.221-226
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• 2001

HANARO, 30MW of research reactor, was installed at the depth of 13m of open pool, The $90\%$ of primary coolant was designed to pass through the core and to remove the reaction heat of the core. The rest $10\%$, of the primary coolant was designed to bypass the core. And the reactor coolant through and bypass the core was inhaled at the top of chimney by the coolant pump to protect that the radiated gas was lifted to the top of reactor pool. But, the part of core bypass coolant was not inhaled by the reactor coolant pump and reached at the top of reactor pool by natural convection and increased the radiation level on the top of reactor pool. To reduce the radiation level by protecting the natural convection of the core bypass flow, the hot water layer (HWL, hereinafter) was installed with the depth of 1.2m from the top of reactor pool. As the HWL was normally operated, the radiation level was reduced to five percent ($5\%$) in comparing with that before the installation of the HWL. When HANARO was operated with higher temperature than the normal temperature of the HWL by operating the standby heater, it was found that the radiation level was more reduced than that before operation. To verify the reason, the heat loss of the HWL was calculated. It was confirmed through the results that the larger the temperature difference between the HWL and reactor hall was, the more the evaporation loss was increased. And it was verified that the radiation level above was reduced more safely by increasing the capacity of heater.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4743-4750
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• 2022

Heat pipe cooled nuclear reactor has been a very attractive technical solution to provide the power for deep space applications. In this paper, a 200 kWe space nuclear reactor power design has been proposed based on the combination of an integrated UN ceramic fuel, a heat pipe cooling system and the Stirling power generators. Neutronics and thermal analysis have been performed on the space nuclear reactor. It was found that the entire reactor core has at least 3.9 $ subcritical even under the worst-case submersion accident superimposed a single safety drum failure, and results from fuel temperature coefficient, neutron spectrum and power distribution analysis also showed that this reactor design satisfies the neutronics requirements. Thermal analysis showed that the power in the core can be successfully removed both in normal operation or under one or more heat pipes failure scenarios.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.575-600
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• 2024

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%-8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ~4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ~6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

• 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.51 no.1
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• pp.54-59
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• 2019

The main objective of nuclear reactor safety is to maintain the nuclear fuel in a thermally safe condition with enough safety margins during normal operation and anticipated operational occurrences. In this research, core flow bypass is studied under the conditions of the unavailability of safety systems. As core bypass occurs, the core flow rate is assumed to decrease exponentially with a time constant of 25 s to new steady state values of 20, 40, 60, and 80% of the nominal core flow rate. The thermal hydraulic code PARET is used through these calculations. Reactor thermal hydraulic stability is reported for all cases of core flow bypass.

• Nuclear Engineering and Technology
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• v.30 no.2
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• pp.164-172
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• 1998

This paper presents the detailed modelling of reactor vessel ledge region for the dynamic analysis of the coupled internals and core model. The dynamic responses due to earthquake and pipe break are calculated using the input motions of reactor vessel taken from Ulchin nuclear power plant units 3 and 4. Two different representations for detailed and simplified models of the RV ledge region are made. The dynamic responses of the reactor internals components are compared between them. Response characteristics are reported and simplified model is suggested for earthquake and pipe break analysis for the future design of the reactor internals.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.