• 한국안전학회지
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• 제29권6호
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• pp.166-171
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• 2014

Nuclear power plants are equipped with the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS) to improve safety on the normal operation. In the event of the design basis accident (DBA), a various of post accident monitor(PAM)systems support to provide important details (e.g. Containment pressure, temperature and pressure of reactor cooling system and core exit temperature) to determine action of main control room (MCR). Operator should be immediately activated for the accident mitigation with the information. Especially, core exit temperature is a critical parameter because the operating mode converts from normal mode to emergency mode when the temperature of core exit reaches $649^{\circ}C$. In this study, uncertainty which was caused by exterior environment, characteristic of thermocouple/connector and accuracy of calibrator/indicator was evaluated in accordance with ANSI-ISA 67.04. The square root of the sum of square (SRSS) methodology for combining uncertainty terms that are random and independent was used in the synthesis. Every uncertainty that may exist in the hardware which is used to measure the core exit temperature was conservatively applied and the associative relation between the elements of uncertainty was considered simultaneously. As a result of uncertainty evaluation, the channel statistical allowance (CSA) of single channel of core exit temperature was +1.042%Span. The range of uncertainty, -0.35%Span ($-4.05^{\circ}C$) ~ +2.08%Span($24.25^{\circ}C$), was obtained as the operating criteria of core exit temperature.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.249-256
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• 2007

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modem fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with $380^{\circ}C$ core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around $500^{\circ}C$, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.92.2-92
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• 2001

• Nuclear Engineering and Technology
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• 제40권2호
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• pp.117-126
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• 2008

In this paper, a new reactor core design is proposed on the basis of a mixed core concept consisting of a thermal zone and a fast zone. The geometric structure of the fuel assembly of the thermal zone is similar to that of a conventional thermal supercritical water-cooled reactor(SCWR) core with two fuel pin rows between the moderator channels. In spite of the counter-current flow mode, the co-current flow mode is used to simplify the design of the reactor core and the fuel assembly. The water temperature at the exit of the thermal zone is much lower than the water temperature at the outlet of the pressure vessel. This lower temperature reduces the maximum cladding temperature of the thermal zone. Furthermore, due to the high velocity of the fast zone, a wider lattice can be used in the fuel assembly and the nonuniformity of the local heat transfer can be minimized. This mixed core, which combines the merits of some existing thermal SCWR cores and fast SCWR cores, is proposed for further detailed analysis.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2002년도 학술대회지
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• pp.237-240
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• 2002

In the present study, we perform LES of turbulent flow and temperature fields in a circular impinging jet at Re=5000 for two cases of H/D=2 and 6 (H denotes the distance between the jet exit and flat plate, and D does the diameter of the jet exit). In the case of H/D=2, the regular vortical structures observed in free jet do not exist because of the smaller distance than the potential core. The Nusselt number on the wall is largest at $r/D{\cong}10.67$ where vortex rings Impinge. At $r/D{\cong}1.5{\~}2.0$, the vortex rings induce the secondary vortices, resulting in a secondary peak in the Nusselt number there. In the case of H/D=6, the vortex rings change into three-dimensional vortical structures and the small-scale vortices impinge on the flat plate. The increase of turbulent intensity due to small-scale vortices results in the largest Nusselt number at the stagnation point.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.3255-3267
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• 2024

This paper presents a high-fidelity simulation of the Organization for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) 3D whole-core Watts Bar benchmark using the UNIST in-house STREAM3D (Steady State and Transient Reactor Analysis code with Method of Characteristics) neutronic code. The benchmark encompasses various whole-core exercises, including single physics problems, multiphysics simulations, and depletion problems. When comparing parameters during the zero-power physics tests, including ITC, DBW, CRW, and criticality tests, STREAM3D results indicate a strong agreement with the measured data and KENO-VI. The comparison with the MC21/CTF code in 3D HFP BOC condition demonstrated strong agreement, with only a 0.42% difference in the normalized radial power distribution, a 0.38 K difference in the RMS of the assembly coolant exit temperature, and a mere 4 ppm difference in CBC.

• 에너지공학
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• 제24권2호
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• pp.154-166
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• 2015

The in-core instrumentation measures core power distribution and coolant temperature in local regions of the core in pressurized water reactors. The installation types are distinguished by the designs of routing paths that exit either through reactor bottom mounted instrument nozzles or through reactor top mounted instrument nozzles. Although each type has unique advantages, it is generally known that top mounted design is more competitive with respect to emphasizing nuclear safety issues and ability to cope with severe accidents. The international nuclear vendors have provided various types of reactors with top mounted design. Nuclear power reactors in Korea, however, only have been designed to be applicable to the use of bottom mounted design, and it has been pointed out that the capabilities of Korean reactors against severe accidents should be further enhanced. The paper deals with technical issues on reactor internal and external design, in-core instrumentation, support assembly, sealing mechanism with nozzles, handling, and analytical issues in order to establish the ways of development.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집B
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• pp.267-271
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• 2000

The local heat transfer rate of an axisymmetric submerged air jet impinging on normal to a heated flat plate was investigated experimentally with varying solidity of mesh screen. The mean velocity and turbulent Intensity profiles of streamwise velocity component were measured using a hot-wire anemometry. The temperature distribution on the heated flat surface was measured with thermocouples. The screen installed in front of the nozzle exit(behind of 35mm) modify the jet flow structure and local heat transfer characteristics. For higher solidity screen, turbulence intensity at core lesion is high and increases the local heat transfer rate at nozzle-to-plate spacings(L/D<6). For larger nozzle-to-plate spacings(L/D>6), however, the turbulent Intensities of all screens tested in this study approach to an asymptotic curve, but the small mean velocity at the core region reduces the local heat transfer rate for high solidity screens.