• Nuclear Engineering and Technology
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• v.15 no.1
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• pp.41-49
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• 1983

A computer code, BERD (Bayesian Estimation of Reliability Data), has been developed and tested in order to update the data for the reliability analysis of safety related systems in a specific nuclear power plant. The code has been used to derive the plant-specific data for reliability analysis of the auxiliary feedwater system of a pressurized water reactor. The prior information for components selected was taken from the U.S. Reactor Safety Study, WASH-1400, and the operating experiences from published licensee event reports. The results show that the updated data are well fitted to log-normal distribution curves and the error factors are reduced significantly.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.689-696
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• 2012

A small integral reactor SMART (System Integrated Modular Advanced ReacTor), being developed in Korea since late 1990s and targeted to obtaining a standard design approval by the end of 2011, is introduced. The design scope and level for design certification (DC) is well described in the U.S. NRC SECY documents published the early 1990s. However, the documents are valid for a one-step licensing process called a combined operating license (COL) by the U.S. NRC, while Korea still uses a two-step licensing process. Thus, referencing the concept of the SECY documents, we have established the design scope and level for the SMART DC using the contexts of the standard review plan (SRP). Some examples of the results and issues raised during our review are briefly presented in this paper. The same methodology will be applied to other types of reactor under development in Korea, such as future VHTR reactors.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.152-161
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• 2022

We analyzed mitigation strategies for steam generator tube rupture (SGTR) accidents using MARS code under both full-power and low-power and shutdown (LPSD) conditions. In general, there are two approaches to mitigating SGTR accidents: supplementing the reactor coolant inventory using safety injection systems and depressurizing the reactor coolant system (RCS) by cooling it down using the intact steam generator. These mitigation strategies were compared from the viewpoint of break flow from the ruptured steam generator tube, the core integrity, and the possibility of the main steam safety valves opening, which is associated with the potential release of radiation. The "cooldown strategy" is recommended for break flow control, whereas the "RCS make-up strategy" is better for RCS inventory control. Under full power, neither mitigation strategy made a significant difference except for on the break flow while, in LPSD modes, the RCS cooldown strategy resulted in lower break and discharge flows, and thus less radiation release. As a result, using the cooldown strategy for an SGTR under LPSD conditions is recommended. These results can be used as a fundamental guide for mitigation strategies for SGTR accidents according to the operational mode.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.37-41
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• 2020

The Nuclear containment building is a main safety-related structure that performs shielding and conservation functions to prevent highly radioactive materials from leakage to the outside environment in the case of various environmental conditions and postulated accidents. The containment building contains a reactor, steam generator, pressurizer, tank, reactor coolant system, auxiliary system and engineering safety system, and is designed so that highly radioactive materials above the limits specified in 10 CFR 100 do not escape to the outside environment in the case of LOCA(Loss of Coolant Accident) for instance. The containment metal liner plate(CLP) is a carbon steel plate with a nominal plate thickness of 6 mm, which functions as a mold for the wall and dome of the containment building when concrete is filled, fulfills airtightness to prevent leakage of seriously radioactive materials. In recent years, backside corrosion was found on the liner plate in some domestic nuclear power plants. The main cause of backside corrosion was unfilled concrete. In this paper, an inspection technique of assessing filling suitability for CLP backside concrete is developed. Results show that the validity of inspection technique for CLP backside concrete using vibration sensor is successfully verified.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1015-1024
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• 2009

The presence of a non-condensable gas can considerably reduce the level of condensation heat transfer. The non-condensable gas effect is a primary concern in some passive systems used in advanced design concepts, such as the Passive Residual Heat Removal System (PRHRS) of the System-integrated Modular Advanced ReacTor (SMART) and the Passive Containment Cooling System (PCCS) of the Simplified Boiling Water Reactor (SBWR). This study examined the capability of the Multi-dimensional Analysis of Reactor Safety (MARS) code to predict condensation heat transfer in a vertical tube containing a non-condensable gas. Five experiments were simulated to evaluate the MARS code. The results of the simulations showed that the MARS code overestimated the condensation heat transfer coefficient compared to the experimental data. In particular, in small-diameter cases, the MARS predictions showed significant differences from the measured data, and the condensation heat transfer coefficient behavior along the tube did not match the experimental data. A new method for calculating condensation heat transfer coefficient was incorporated in MARS that considers the interfacial shear stress as well as flow condition determination criterion. The predictions were improved by using the new condensation model.

• Journal of the Korean Society of Safety
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• v.34 no.1
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• pp.14-20
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• 2019

This paper presents a safety assessment based approach for the safe operation for PCS(Power Conditioning System) of photovoltaic and energy storage systems, applying FTA. The approach established top events as power outage and a failure likely to cause the largest damage among the potential risks of PCS. Then the Minimal Cut Set (MCS) and the importance of basic events were analyzed for implementing risk assessment. To cope with the objects, the components and their functions of PCS were categorized. To calculate the MCS frequency based on IEEE J Photovolt 2013, IEEE Std. 493-2007 and RAC (EPRD, NPRD), the failure rate and failure mode were produced regarding the basic events. In order to analyze the top event of failure and power outage, it was assumed that failures occurred in DC breaker, AC breaker, SMPS, DC filter, Inverter, CT, PT, DSP board, HMI, AC reactor, MC and EMI filter and Fault Tree was drawn. It is expected that the MCS and the importance of basic event resulting from this study will help find and remove the causes of failure and power outage in PCS for efficient safety management.

• Journal of the Korean Society of Safety
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• v.32 no.4
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• pp.7-15
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• 2017

International standards such as IEC-61508 and IEC-61511 require Safety Integrity Levels (SILs) for Safety Instrumented Functions (SIFs) in process industries. SIL verification is one of the methods for process safety description. Results of the SIL verification in some cases indicated that several Safety Instrumented Functions (SIFs) do not satisfy the required SIL. This results in some problems in terms of cost and risks to the industries. This study has been performed to improve the reliability of a safety instrumented function (SIF) installed in hydrodesulfurization reactor heater using Partial Stroke Testing (PST). Emergency shutdown system was chosen as an SIF in this study. SIL verification has been performed for cases chosen through the layer of protection analysis method. The probability of failure on demands (PFDs) for SIFs in fault tree analysis was $4.82{\times}10^{-3}$. As a result, the SIFs were unsuitable for the needed RRF, although they were capable of satisfying their target SIL 2. So, different PST intervals from 1 to 4 years were applied to the SIFs. It was found that the PFD of SIFs was $2.13{\times}10^{-3}$ and the RRF was 469 at the PST interval of one year, and this satisfies the RRF requirements in this case. It was also found that shorter interval of PST caused higher reliability of the SIF.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1037-1051
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• 2024

This paper presents the construction results and design of the UNIST Reactor Innovation platform for small modular reactors as a versatile testbed for exploring innovative technologies. The platform uses simulant fluids to simulate the thermal-hydraulic behavior of a reference small modular reactor design, allowing for cost-effective design modifications. Scaling analysis results for single and two-phase natural circulation flows are outlined based on the three-level scaling methodology. The platform's capability to simulate natural circulation behavior was validated through performance calculations using the 1-D system thermal-hydraulic code-based calculation. The strategies for evaluating cutting-edge technologies, such as the integration of a solid oxide electrolysis cell for hydrogen production into a small modular reactor, are presented. To overcome experimental limitations, the hardware-in-the-loop technique is proposed as an alternative, enabling real-time simulation of physical phenomena that cannot be implemented within the experimental facility's hardware. Overall, the proposed versatile innovation platform is expected to provide valuable insights for advancing research in the field of small modular reactors and nuclear-based hydrogen production.

• Journal of the Korean Society of Safety
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• v.30 no.6
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• pp.48-55
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• 2015

A variety of safety issues were investigated for chemical reactors using a toluene solvent in case of a fire at small to middle scale chemical plants. The issues covered the operation of pressure-relieving valves and the subsequent discharges of the toluene to the atmosphere either directly or through an absorber, which represent the current practice at most small chemical plants. It was shown that the safety valve on the reactor may not operate within about twenty minutes after an external fire breaks out, but, once relieved, the toluene vapor released directly to the atmosphere may form a large explosion range on the ground. It was also shown that if the discharge is routed to an existing absorber used for the scrubbing of volatile organic compounds or dusts, the column may not operate normally due to excessive pressure drops or flooding, resulting in the hazardous release of toluene vapors. This study proposed two ways of alleviating these risks. The first is to ruduce the discharge itself from the safety valve by using adequate insulation and protection covers on the reactor and then introduce it into the circulation water at the bottom of the absorber through a dip linet pipe equipped with a ring-shaped sparger. This will enhance the condensation of toluene vapors with the reduced effluent vapors treated in the packing layers above. The second is to install a separate quench drum to condense the routed toluene vapors more effectively than the existing absorber.

• Journal of Energy Engineering
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• v.24 no.2
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• pp.40-44
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• 2015

Sensitivity on the reactor safety was evaluated for the safety margin and time delay applied to the opening pressure of liquid relief valve(LRV) of the primary heat transport system(PHTS) in the pressurized heavy water reactor(PHWR) type nuclear power plant. Since the LRV is the pressure boundary for the PHTS in the safety analysis, the operating of LRV has a significant effect on the safety analysis results. Therefore it is required during the regulatory review of Wolsong Unit 1 safety analysis to find the safety effect of the application of safety margin and time delay to the LRV opening pressure for the safety analysis of PHTS pressurizing events.