• Proceedings of the Korean Society of Disaster Information Conference
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• 2023.11a
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• pp.221-221
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• 2023

1969년에 발간된 API521 1st edition에서는 Flare Load 저감용으로 적용되는 HIPS (High Integrity Protection System)는 모두 Pressure Safety Valve의 고장확률보다 낮은 SIL 3 (Safety Integrity Level)등급을 적용할 것을 요구하고 있다. Flare Stack 저감용 HIPS는 주로 압축기 출력압력상승, Reboiler Steam 과다주입, 전력공급중단냉각펌프고장 등에 의한 Flare 발생을 예방하기 위한 기능을 가진 SIF (Safety Instrumented Function)로 구성된다. 하지만 2007년도 발간된 API521 5th edition에서는 LOPA (Layer Of Protection Analysis) 분석을 통해 Target SIL을 도출하는 것으로 요구사항을 변경했다. 이에 따라 이번 연구에서는 Flare Load에 가장 큰 영향을 미치는 시나리오 중 대표적인 시나리오를 대상으로 HAZOP(Hazard and Operability Study)과 LOPA분석을 실시해서 Target SIL이 어떻게 도출되는지를 연구했다. Flare Stack에서 Flare를 발생시키는 대표적인 시나리오들에 대해 LOPA분석을 실시한 결과 압축기 출력압력상승은 SIL 2, Reboiler Steam 과다주입은 SIL 3, 전력공급중단은 SIL 0, 냉각펌프고장은 SIL 0로 모두가 SIL 3 가 나오지는 않았다. SIF 설계 시 Target SIL을 만족시키는 것도 중요하지만 운전 시 SIL 등급이 계속 유지되게 하지 위해 인적오류, 시스템적 고장, 하드웨어고장 등에 의해 SIF 기능불능화가 되는 것을 예방하기 위한 기능안전관리시스템 (FSMS)를 적용하는 것도 중요하다.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2013.10a
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• pp.868-869
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• 2013

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1997.10a
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• pp.271-278
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• 1997

The flow of steam through a safety valve vent pipe system in the boiler has been analyzed to provide a design basis of diffuser silencer for attenuating shock-shell and jet noise. Numerical analysis to estimate inner fluid of silencer and noise propagation outside silencer are performed. The distribution curve of fluid information to provide average values about inner fluid of silencer is presented by theoretical analysis.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.812-824
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• 2021

In order to reduce risks from the Steam Generator Tube Rupture (SGTR) accident and to meet safety targets, various measures have been analyzed to minimize the amount of fission product (FP) release. In this paper, we propose an introduction of a Containment Filtered Venting System (CFVS) connected to the steam generator secondary side, which can reduce the amount of FP release while minimizing adverse effects identified in the previous studies. In order to compare the effect of new equipment with the existing strategy, accident simulations using MELCOR were performed. As a result of simulations, it is confirmed that CFVS operation lowers FP release into the environment, and the release fractions are lower (minimum 0.6% of the initial inventory for Cs) than that of the strategy which intends to depressurize the primary system directly (minimum 15.2% for Cs). The sensitivity analyses identify that refill of the CFVS vessel is a dominant contributor reducing the amount of FP released. As the new strategy has the possibility of hydrogen combustion and detonation in CFVS, the installation of an igniter inside the CFVS vessel may be considered in reducing such hydrogen risk.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.813-820
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• 2001

The Safety depressurization System(SDS) of KNGR prevents RCS from overpressurization by discharging high pressure and temperature coolant through the I-sparger into the IRWST during an accident. If IRWST water temperature rise locally, around the sparger, beyond $200_{\circ}$2000 F by the discharged coolant, unstable steam condensation can cause large pressure load on the IRWST wall. To investigate whether this condition can be avoided for the design basis event IOPOSRV(Inadvertent Opening of one Pilot Operated Safety Relief Valve), the flow and temperature distribution of water in the IRWST is calculated by using CFX 4.3 computational fluid dynamic code. According to the results, since pool water temperature does not exceeds temperature limit within 50 seconds after the opening of one POSRV, it can be assured that the integrity of IRWST wall is maintained.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.880-885
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• 2024

For Safety Assisted Engineering works, real-time simulators have emerged as a mandatory tool among all the key actors involved in the nuclear industry (utilities, designers and safety authorities). EDF, Electricité de France, as the leading worldwide nuclear power plant operator, has a crucial need for efficient and updated simulation tools for training, operating and safety analysis support. This paper will present the work performed at EDF/DT to develop a new generation of engineering simulator to fulfil these tasks. The project is called SiRENE, which is the acronym of Re-hosted Engineering Simulator in French. The project has been economically challenging. Therefore, to benefit from existing tools and experience, the SiRENE project combines: - A part of the process issued from the operating fleet training full-scope simulator. - An improvement of the simulator prediction reliability with the integration of High-Fidelity models, used in Safety Analysis. These High-Fidelity models address Nuclear Steam Supply System code, with CATHARE thermal-hydraulics system code and neutronics, with COCCINELLE code. - And taking advantage of the last generation and improvements of instructor station. The intensive and challenging uses of the new SiRENE engineering simulator are also discussed. The SiRENE simulator has to address different topics such as verification and validation of operating procedures, identification of safety paths, tests of I&C developments or modifications, tests on hydraulics system components (pump, valve etc.), support studies for Probabilistic Safety Analysis (PSA). etc. It also emerges that SiRENE simulator is a valuable tool for self-training of the newcomers in EDF nuclear engineering centers. As a modifiable tool and thanks to a skillful team managing the SiRENE project, specific and adapted modifications can be taken into account very quickly, in order to provide the best answers for our users' specific issues. Finally, the SiRENE simulator, and the associated configurations, has been distributed among the different engineering centers at EDF (DT in Lyon, DIPDE in Marseille and CNEPE in Tours). This distribution highlights a strong synergy and complementarity of the different engineering institutes at EDF, working together for a safer and a more profitable operating fleet.

• Nuclear Engineering and Technology
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• v.50 no.3
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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.

• Journal of Energy Engineering
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• v.25 no.4
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• pp.124-132
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• 2016

In order to estimate the effect of flow distributors connected to an upper nozzle of CMT(Core Makeup Tank) on the thermal-hydraulic characteristics in the tank, a simplified 2 inch Small Break Loss of Coolant Accident(SBLOCA) was simulated by skipping the decay power and Passive Residual Heat Removal System(PRHRS) actuation. The CMT is a part of safety injection systems in the SMART (System Integrated Modular Advanced Reactor). Each test was performed with reliable boundary conditions. It means that the pressure distribution is provided with repeatable and reproducible behavior during SBLOCA simulations. The maximum flow rates were achieved at around 350 seconds after the initial opening of the isolation valve installed in CMT. After a short period of decreased flow rate, it attained a steady injection flow rate after about 1,250 seconds. This unstable injection period of the CMT coolant is due to the condensation of steam injected into the upper part of CMT. The steady injection flow rate was about 8.4% higher with B-type distributor than that with A-type distributor. The gravity injection during hot condition tests were in good agreement with that during cold condition tests except for the early stages.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.645-660
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• 1995

The loss of Residual Heat Removal System (RHRS) event during reduced inventory operation for the Korean Standard Nuclear Power Plants (KSNPPS) is simulated by RELAP5/MOD3 and RELAP5/MOD3.1 Tn cases are considered : Base case for an intact Reactor Coolant System (RCS) with no tent and a vent case for an open system. Comparative simulations of base case are peformed by RELAP5/MOD3 and RELAP5/MOD3. 1 computer codes. The results of too simulations are generally in good qualitative and quantitative agreement. However, since the results of RELAP5/MOD3 simulation reveals the deficiency of RELAP5/MOD3 wall heat model, the RELAP5/AOD3.1 computer code is used for the simulation of the vent case. The analysis result of base case show that two steam generators are insufficient to remove decay heat at one day after shutdown, where the RCS is closed. The RCS pressure increased continuously and reached the RCS temporary boundaries design pressure of 0.24 MPa around 4,000 seconds. In the vent case with a flow capacity equivalent to three times the capacity of Pressurizer Safety Valve (PSV), it is shown that the RCS Pressure does not reach 0.24 MPa and core uncovery does not occur until 10,000 seconds. The detailed discussions on the results of this study suggest the feasibility of RELAP5/AOD3.1 as an analysis tool for the simulation of the loss of RHRS event at reduced inventory operation. The results of this study also provide insight for the determination of proper vent capacity.