• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.286-294
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• 2017

The events at the Fukushima Daiichi Nuclear Power Station in March 2011 point out, among other matters, that concurrent accidents at multiple units of a site can occur in reality. Although site risk has been deterministically considered to some extent in nuclear power plant siting and design, potential occurrence of multi-unit accident sequences at a site was not investigated in sufficient detail thus far in the nuclear power community. Therefore, there is considerable worldwide interest and research effort directed toward multi-unit site risk assessment, especially in the countries with high-density nuclear-power-plant sites such as Korea. As the technique of probabilistic safety assessment (PSA) has been successfully applied to evaluate the risk associated with operation of nuclear power plants in the past several decades, the PSA having primarily focused on single-unit risks is now being extended to the multi-unit PSA. In this paper we first characterize the site risk with explicit consideration of the risk associated with spent fuel pools as well as the reactor risks. The status of multi-unit risk assessment is discussed next, followed by a description of the emerging issues relevant to the multi-unit risk evaluation from a practical standpoint.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.27-37
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• 2020

The Fukushima Daiichi accident in 2011 revealed some vulnerabilities of existing Nuclear Power Plants (NPPs) under extended Station Blackout (SBO) accident conditions. One of the key Severe Accident Management (SAM) strategies developed post Fukushima accident is the In-Vessel Retention (IVR) Strategy which aims to retain the structural integrity of the Reactor Pressure Vessel (RPV). RELAP/SCDAPSIM/MOD3.4 is selected to predict the thermal-hydraulic response of APR1400 undergoing an extended SBO. To assess the effectiveness of the IVR strategy, it is essential to quantify the underlying uncertainties. In this work, both the epistemic and aleatory uncertainties are considered to identify the success window of the IVR strategy. A set of in-vessel relevant phenomena were identified based on Phenomena Identification and Ranking Tables (PIRT) developed for severe accidents and propagated through the thermal-hydraulic model using Wilk's sampling method. For this work, a Systems Engineering (SE) approach is applied to facilitate the development process of assessing the reliability and robustness of the APR1400 IVR strategy. Specifically, the Kossiakoff SE method is used to identify the requirements, functions and physical architecture, and to develop a design verification and validation plan. Using the SE approach provides a systematic tool to successfully achieve the research goal by linking each requirement to a verification or validation test with predefined success criteria at each stage of the model development. The developed model identified the conditions necessary for successful implementation of the IVR strategy which maintains the vessel integrity and prevents a melt-through.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.67-77
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• 2020

Traditionally nuclear thermal hydraulic and nuclear safety has relied on numerical simulations to predict the system response of a nuclear power plant either under normal operation or accident condition. However, this approach may sometimes be rather time consuming particularly for design and optimization problems. To expedite the decision-making process data-driven models can be used to deduce the statistical relationships between inputs and outputs rather than solving physics-based models. Compared to the traditional approach, data driven models can provide a fast and cost-effective framework to predict the behavior of highly complex and non-linear systems where otherwise great computational efforts would be required. The objective of this work is to develop an AI algorithm to predict the peak fuel cladding temperature as a metric for the successful implementation of FLEX strategies under extended station black out. To achieve this, the model requires to be conditioned using pre-existing database created using the thermal-hydraulic analysis code, MARS-KS. In the development stage, the model hyper-parameters are tuned and optimized using the talos tool.

• Journal of Radiation Protection and Research
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• v.47 no.2
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• pp.77-85
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• 2022

Background: After the Fukushima Daiichi Nuclear Power Station (FDNPS) accident, a model was developed to estimate the external exposure doses for residents who were expected to return to their homes after evacuation orders were lifted. However, the model's accuracy and uncertainties in parameters used to estimate external doses have not been evaluated. Materials and Methods: The model estimates effective doses based on the integrated ambient dose equivalent (H^*(10)) and life patterns, considering a dose reduction factor to estimate the indoor H^*(10) and a conversion factor from H^*(10) to the effective dose. Because personal dose equivalent (H_p(10)) has been reported to agree well with the effective dose after the FDNPS accident, this study validates the model's accuracy by comparing the estimated effective doses with H_p(10). The H_p(10) and life pattern data were collected for 36 adult participants who lived or worked near the FDNPS in 2019. Results and Discussion: The estimated effective doses correlated significantly with H_p(10); however, the estimated effective doses were lower than H_p(10) for indoor sites. A comparison with the measured indoor H^*(10) showed that the estimated indoor H^*(10) was not underestimated. However, the H_p(10) to H^*(10) ratio indoors, which corresponds to the practical conversion factor from H^*(10) to the effective dose, was significantly larger than the same ratio outdoors, meaning that the conversion factor of 0.6 is not appropriate for indoors due to the changes in irradiation geometry and gamma spectra. This could have led to a lower effective dose than H_p(10). Conclusion: The estimated effective doses correlated significantly with H_p(10), demonstrating the model's applicability for effective dose estimation. However, the lower value of the effective dose indoors could be because the conversion factor did not reflect the actual environment.

• The Transactions of the Korean Institute of Electrical Engineers P
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• v.63 no.2
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• pp.61-68
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• 2014

Function of battery bank stores energy for DC load in general, and DC power system of the nuclear power plant is used to supply DC loads for safety- featured instrumentation and control such as inverter, class 1E power system control and indication, and station annunciation. Class 1E DC power system must provide a power for the design basis accident conditions, and adequate capacity must be available during loss of AC power and subsequent safe shutdown of the plant. In present, batteries of Class 1E DC power system of the nuclear power plant uses lead-acid batteries. Class 1E batteries of nuclear power plants in Korea are summarized in terms of specification, such as capacity, discharge rate, bank configuration and discharge end voltage, etc. This paper summarizes standards of determining battery size for the nuclear power plant, and analyzes duty cycle for the class 1E DC power system of nuclear power plant. Then, battery cell size is calculated as 2613Ah according to the standard. In addition, this paper analyzes performance test results during past 13 years and shows performance degradation in the battery bank. Performance tests in 2001 and 2005 represent that entire battery cells do not reach the discharge-end voltage. Howeyer, the discharge-end voltage is reached in 14.7% of channel A (17 EA), 13.8% of channel B (16 EA), 5.2% of channel C (6 EA) and 16.4% of channel D (19 EA) at 2011 performance test. Based on the performance test results analysis and size calculation, battery capacity and degradation by age in Korearn nuclear power plant is discussed and would be used for new design.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.62 no.3
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• pp.300-305
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• 2013

LOOP(Loss of Offsite Power) and SBO(Station Blackout) events have been occurring in nuclear power plants should be reviewed and be controlled on important electrical equipments by professional engineer to prevent and to safety improvement from safety assessment and reliability analysis report. LOOP and SBO occasionally happened by internal or external causes. This paper contained that LOOP frequency in the United States NPPs and in the domestic NPPs have compared and analyzed data by the past lessons and probabilistic statistics. Additionally will be installed MG(Mobile Generator) according to the lessons of Fukushima nuclear accident in Japan, which CDF(Core Damage Frequency) and LOOP frequency have reconsidered. And this paper proposed to reduce reliability criteria using PSA(Probabilistic Safety Analysis).

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2183-2195
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• 2020

A mathematical model for the flowrate and rotation speed of RCP during idling was established. The numerical calculation method and dimensionless method were used to analyze the flow, head, torque and pressure and speed changes under idle conditions. Regularity, using the Q criterion vortex identification judgment method combined with surface flow spectrum morphology analysis to diagnose the vortex dynamic characteristics on RCP blade. On impeller blade, there is two oscillations in the pressure ratio on pressure surface in blade outlet region. The velocity on the suction surface is two times more oscillating than the inlet of blade, and there is an intersection with the velocity ratio curve on pressure surface. On blade of guide vane, the pressure ratio increases along the inlet to outlet direction, and the speed ratio decreases with the increase of idle time. There is a vortex that rotates counterclockwise on the suction surface, and the streamline on the suction surface of blade is subjected to the entrainment and blocking action of the vortex creates a large reverse flow in the main flow region. There are two vortices at the outlet of guide vane suction side and the vortices are in opposite directions.

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.443-453
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• 2015

A hybrid safety injection tank (H-SIT) can enhance the capability of an advanced power reactor plus (APR+) during a station black out (SBO) that is accompanied by a severe accident. It may a useful alternative to an electric motor. The operations strategy of the H-SIT has to be investigated to achieve maximum utilization of its function. In this study, the master logic diagram (i.e., an analysis for identifying the differences between an H-SIT and a safety injection pump) and an accident case classification were used to determine the parameters of the H-SIT operation. The conditions that require the use of an H-SIT were determined using a decision-making process. The proper timing for using an H-SIT was also analyzed by using the Multi-dimensional Analysis of Reactor Safety (MARS) 1.3 code (Korea Atomic Energy Research Institute, Daejeon, South Korea). The operation strategy analysis indicates that a H-SIT can mitigate five types of failure: (1) failure of the safety injection pump, (2) failure of the passive auxiliary feedwater system, (3) failure of the depressurization system, (4) failure of the shutdown cooling pump (SCP), and (5) failure of the recirculation system. The results of the MARS code demonstrate that the time allowed for recovery can be extended when using an H-SIT, compared with the same situation in which an H-SIT is not used. Based on the results, the use of an H-SIT is recommended, especially after the pilot-operated safety relief valve (POSRV) is opened.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1396-1408
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• 2020

Filtered containment venting system (FCVS) is one of the severe accident mitigation systems designed to release containment pressurization to maintain its integrity. The thermal-hydraulic behaviors in FCVSs are important since they affect the operation characteristics of the FCVS. In this study, a representative FCVS was modeled by RELAP5/Mod3.3 code, and the Station BlackOut (SBO) was chosen as an accident scenario. The thermal-hydraulic behaviors of an FCVS during long-term operation with two venting strategies (open-and-close strategy, open-and-non-close strategy) and the sensitivity analysis of important parameters were investigated. The results show that the FCVS can operate up to 250 h with a periodic open-and-close strategy during an SBO. Under the combined effects of steam condensation and water evaporation, the solution inventory in the FCVS increases during the venting phase and decreases during the intermission phase, showing a periodic pattern. Under this condition, the appropriate initial water level is 3-4 m; however, it should be adjusted according to the environment temperature. The FCVS can accommodate a decay heat power of 150-260 kW and may need to feed water for a higher decay heat power or drain water for a lower decay heat power during the late phase. The FCVS can function within an opening pressure range from 450 kPa to 500 kPa and a closing pressure range between 250 kPa and 350 kPa. When the open-and-non-close strategy is adopted, the solution inventory increases quickly in the early venting phase due to steam condensation and then decreases gradually due to the evaporation of water; drying-up may occur in the late venting phase. Decreasing the venting pipe diameter and increasing the initial water level can mitigate the evaporation of the scrubbing solution. These results are expected to provide useful references for the design and engineering application of FCVSs.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2221-2229
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• 2020

After the Fukushima Daiichi nuclear power plant (NPP) accident, level 3 probabilistic safety assessment (PSA) has emerged as an important task in order to assess the risk level of the multi-unit NPPs in a single nuclear site. Accurate calculation of the radionuclide concentrations and exposure doses to the public is required if a nuclear site has multi-unit NPPs and large number of people live near NPPs. So, there has been a great need to develop a new method or procedure for the fast and accurate offsite consequence calculation for the multi-unit NPP accident analysis. Since the multi-unit level 3 PSA is being currently performed assuming that all the NPPs are located at the same position such as a center of mass (COM) or base NPP position, radionuclide concentrations or exposure doses near NPPs can be drastically distorted depending on the locations, multi-unit NPP alignment, and the wind direction. In order to overcome this disadvantage of the COM method, the idea of a new multiple location (ML) method was proposed and implemented into a new tool MURCC (multi-unit radiological consequence calculator). Furthermore, the MURCC code was further improved for the multi-unit level 3 PSA that has the arbitrary number of multi-unit NPPs. The objectives of this study are to (1) qualitatively and quantitatively compare COM and ML methods, and (2) demonstrate the strength and efficiency of the ML method. The strength of the ML method was demonstrated by the applications to the multi-unit long-term station blackout (LTSBO) accidents at the four-unit Vogtle NPPs. Thus, it is strongly recommended that this ML method be employed for the offsite consequence analysis of the multi-unit NPP accidents.