• Journal of the Korean Society of Safety
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• v.37 no.4
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• pp.129-138
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• 2022

A new method for injecting cooling water into the Korean research reactor (KRR) in the event of beam tube rupture is proposed in this paper. Moreover, the research evaluates the risk to the reactor core in terms of core damage frequency (CDF). The proposed method maintains the cooling water in the chimney at a certain level in the tank to prevent nuclear fuel damage solely by gravitational coolant feeding from the emergency water supply system (EWSS). This technique does not require sump recirculation operations described in the current procedure for resolving beam tube accidents. The reduction in the risk to the core in the event of beam tube rupture that can be achieved by the proposed change in the cooling water injection design is quantified as follows. 1) The total CDF of the KRR for the proposed design change is approximately 4.17E-06/yr, which is 8.4% lower than the CDF of the current design (4.55E-06/yr). 2) The CDF for beam tube rupture is 7.10E-08/yr, which represents an 84.1% decrease compared with that of the current design (4.49E-07/yr). In addition to this quantitative reduction in risk, the modified cooling water injection design maintains a supply of pure coolant to the EWSS tank. This means that the reactor does not require decontamination after an accident. Thermal hydraulic analysis proves that the water level in the reactor pool does not cause damage to the nuclear fuel cladding after beam tube rupture. This is because the amount of water in the chimney can be regulated by the EWSS function. The EWSS supplies emergency water to the reactor core to compensate for the evaporation of coolant in the core, thus allowing water to cover the fuel assemblies in the reactor core over a sufficient amount of time.

• Journal of Environmental Science International
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• v.17 no.5
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• pp.573-580
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• 2008

This study aims to measure and to analyze the characteristics of thermal environment of the various permeable pavement materials such as a break stone pavement (Green block cubic), soil protection pavement (Soil tector), soil cement pavement and ceramic brick pavement under the summer outdoor environment. The thermal environment characteristics measured in the study includes the changes of surface temperature during the day, and long and short wave radiation of each pavement surface. The experimental condition is based on the data on the hottest temperature (August 9, 2006, $37.1^{\circ}C$) of the year. The albedo was the highest on the break stone pavement(0.8) from 12:00 to 14:00. The albedo of the ceramic brick pavement, a soil tector pavement and soil cement pavement were 0.35, 0.29 and 0.27 from 12:00 to 14:00, respectively. The peak surface temperature and long wave radiation was the highest on the soil protection pavements($56.6^{\circ}C$/627 W/$m^2$). The peak surface temperatures and long wave radiation on the ceramic brick pavement, a stone brick pavement and soil cement pavement were $51.7^{\circ}C$/627 W/$m^2$, $48.8^{\circ}C$/607 W/$m^2$ and $45.9^{\circ}C$/582 W/$m^2$, respectively. The heat environment was better on the break stone pavement than on the other pavements. This is mainly due to the high albedo of the break stone pavement(0.8) while the albedo value of a ceramic brick pavement, a soil tactor pavement and soil cement pavement were 0.35. 0.29 and 0.27. Large heat capacity($2,629kJ/m^3{\cdot}K$) of the stone brick pavements also contributes to this difference. The heat environment was better on the soil cement pavement than the soil tector pavement. This is mainly due to the evaporation of the soil cement pavement while the active evaporation of the soil tactor pavement was not continued after two days from the rainfall event. To improve the thermal environments in the urban area, it is recommended to raise the albedo of the pavements by brightening the surface color of the pavement materials. Further studies on the pavement materials and the construction methods which can enhance the continuous evapotranspiration from the pavements surface are needed.

• Journal of the Korea Concrete Institute
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• v.26 no.3
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• pp.247-254
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• 2014

Strain properties of concrete member which acts as an important factor in the stability of the concrete structure in the event of fire, significantly affected the characteristics of the coarse aggregate, which accounts for most of the volume. For this reason, there are many studies on concrete using artificial lightweight aggregate which has smaller thermal expansion deformation than granite coarse aggregate. But the research is mostly limited on concrete using clay-based lightweight aggregate. Therefore, in this study, the high temperature compressive strength and elastic modulus, thermal strain and total strain, high temperature creep strain of concrete was evaluated. As a result, remaining rate of high-temperature strength of concrete using lightweight aggregate is higher than concrete with general aggregate and it is determined to be advantageous in terms of structural safety and ensuring high-temperature strength from the result of the total strain by loading and strain of thermal expansion. In addition, in the case of high-temperature creep, concrete shrinkage is increased by rising loading and temperature regardless of the type of aggregate, and concrete using lightweight aggregate shows bigger shrinkage than concrete with a granite-based aggregate. From this result, it is determined to require additional consideration on a high temperature creep strain in case of maintaining high temperature like as duration of a fire although concrete using light weight aggregate is an advantage in reducing the thermal expansion strain of the fire.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.393-402
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• 1995

The structural analysis of the reactor coolant system mainly consist of too fields. The one is the static analysis considering the impact of pressure and temperature built up during normal operation. The other is the dynamic analysis to estimate the impact of postulated events such as the seismic loads or postulated branch line pipe breaks event. Since the most important goal of the RCS structural analysis is to prove the safety of the RCS during normal operation or postulated events, a widely proven theory having enough conservatism is adopted. The load occurring on the RCS during normal operation is considered as the basic design loading condition throughout whole plant life time. The most typical characteristic of the RCS during normal operation is the thermal expansion of the RCS caused by reactor coolant with high temperature and pressure. Therefore, the exact estimation on the thermal movement of the RCS is needed to get more clear understanding on the thermal movement behavior of the RCS. In this study, the general structural analysis concept and modeling method to evaluate the thermal movement of the RCS under the normal plant operation condition are presented. To discuss the validation of the suggested analysis, analysis results are compared with the measured data which ore referred from the standardized 1000 MWe PWR plant under construction.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.114-129
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• 2017

Background: The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. Materials and Methods: The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012-018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. Results and Discussion: The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. Conclusion: The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3 cores is well within the range estimated by the reverse or inverse method, which are summarized in UNSCEAR 2013 report. But this does not necessarily mean that these two approaches are consistent.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.63-77
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• 1994

The objective of this work is to investigate fluid mixing phenomena related to pressurized thermal shock(PTS) in a pressurized water reactor(PWR) vessel downcomer during transient cooldown with direct vessel injection(DVI) using test models. The test model designs were based on ABB Combustion Engineering(C-E) System 80＋ reactor geometry. A cold leg small break loss-of-coolant accident(LOCA) md a main steam line teak were selected as the potential PTS events for the C-E System 80＋. This work consist of two parts. The first part provides the visualization tests of the fluid mixing between DVI fluid and existing coolant in the downcomer region, and the second part is to compare the results of thermal mixing tests with DVI in the other test model. Row visualization tests with DVI have clarified the physical interaction between DVI fluid and primary coolant during transient cooldown. A significant temperature drop was observed in the downcomer during the tests of a small break LOCA Measured transient temperature profiles agree well with the predictions by the REMIX code for a small break LOCA and with the calculations by the COMMIX-1B code for a steam line break event.

• Journal of Ocean Engineering and Technology
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• v.32 no.5
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• pp.310-323
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• 2018

In order to protect lives and prevent large-scale injuries in the event of a fire on a ship or an offshore plant, most classification societies are strengthening their fire resistance designs of relevant cargo holds and accommodation compartments to keep flames from being transferred from a fire point to other compartments. Particularly in critical compartments, where flames should not propagate for a certain period of time, such as the A60 class division, both the airtightness and fire-resistant design of a piece passing through a bulkhead are subject to the Safety of Life at Sea Convention (SOLAS) issued by the International Maritime Organization (IMO). In order to verify the suitability of a fire-resistant design for such a penetrating piece, the fire test procedure prescribed by the Maritime Safety Committee (MSC) must be carried out. However, a numerical simulation should first be conducted to minimize the time and cost of the fire resistance test. In this study, transient thermal analyses based on the finite element method were applied to investigate the heat transfer characteristics of a bulkhead penetration piece for the A60 class compartment. In order to determine a rational bulkhead penetration piece design, the transient heat transfer characteristics according to the variation of design parameters such as the diameter, length, and material were reviewed. The verification of the design specification based on a numerical analysis of the transient heat transfer performed in this study will be discussed in the following research paper for the actual fire protection test of the A60 class bulkhead penetration piece.

• Journal of Electrical Engineering and Technology
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• v.9 no.5
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• pp.1471-1481
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• 2014

This paper presents the new power dissipation model of individual switching device in a high-level modular multilevel converter (MMC), which can be mostly used in voltage sourced converter (VSC) based high-voltage direct current (HVDC) system and flexible AC transmission system (FACTS). Also, the voltage balancing method based on sorting algorithm is newly proposed to advance the MMC functionalities by effectively adjusting switching variations of the sub-module (SM). The proposed power dissipation model does not fully calculate the average power dissipation for numerous switching devices in an arm module. Instead, it estimates the power dissipation of every switching element based on the inherent operational principle of SM in MMC. In other words, the power dissipation is computed in every single switching event by using the polynomial curve fitting model with minimum computational efforts and high accuracy, which are required to manage the large number of SMs. After estimating the value of power dissipation, the thermal condition of every switching element is considered in the case of external disturbance. Then, the arm modeling for high-level MMC and its control scheme is implemented with the electromagnetic transient simulation program. Finally, the case study for applying to the MMC based HVDC system is carried out to select the appropriate insulated-gate bipolar transistor (IGBT) module in a steady-state, as well as to estimate the proper thermal condition of every switching element in a transient state.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.187-194
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• 2011

Korea Nuclear Fuel is developing the X-GEN fuel which shows high performance and robust reliability for the worldwide supply. However, the simplified code systems such as CESEC-III which were developed in 1970s are still used in the current Non-LOCA safety analysis of OPR1000 and APR1400 plants. Therefore, it is essential to secure an advanced safety analysis methodology to make the best use of the merits of X-GEN fuel. To accomplish this purpose, the $\b{i}$ntegrated $\b{s}$afety $\b{a}$nalysis $\b{m}$ethodology (iSAM), is developed by selecting the best-estimate thermal-hydraulic code RETRAN. iSAM possesses remarkable advantages, such as generality, integrity, and designer-friendly features. That is, iSAM can be applied to both OPR1000 and APR1400 plants and uses only one computer code, RETRAN, in the whole scope of the non-LOCA safety analyses. Also the iSAM adopts the unique and automatic initialization and run tool, $\b{a}$utomatic $\b{s}$teady-$\b{s}$tate $\b{i}$nitialization and $\b{s}$afety analysis too l (ASSIST), to enable unhandy designers to use the new design code RETRAN without difficulty. In this paper, a brief overview of the iSAM is given, and the results of applying the iSAM to typical non-LOCA transients being checked during the reload design are reported. The typical non-LOCA transients selected are the single control element assembly withdrawal (SCEAW) accident, the asymmetric steam generator transients (ASGT), the locked rotor (LR) accident, and bank CEA withdrawal (BCEAW) event. Comparison to current licensing results shows a close resemblance; thus, it reveals that the iSAM can be applied to the non-LOCA safety analysis of OPR1000 and APR1400 plants.

• International Journal of High-Rise Buildings
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• v.10 no.4
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• pp.265-283
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• 2021

To ensure that fire induced collapse of a building is prevented it is important to understand the sequence of events that can lead to this event. In this paper, the initiation of collapse mechanisms of generic a multi-storey steel frame subjected to vertical and horizontal travelling fires are analysed computationally by tracking the formation of plastic hinges in the frame and generation of fire induced loads. Both uniform and travelling fires are considered. In total 58 different cases are analysed using finite element software LS-DYNA. For the frame examined with a simple and generic structural arrangement and higher applied fire protection to the columns, the results indicate that collapse mechanisms for singe floor and multiple floor fires can be each split into two main groups. For single floor fires (taking place in the upper floors of the frame (Group S1)), collapse is initiated by the pull-in of external columns when heated beams in end bays go into catenary action. For single floor fires occurring on the lower floors(Group S2), failure is initiated (i.e. ultimate strain of the material is exceeded) after the local beam collapse. Failure in both groups for single floor fires is governed by the generation of high loads due to restrained thermal expansion and the loss of material strength. For multiple floor fires with a low number of fire floors (1 to 3) - Group M1, failure is dominated by the loss of material strength and collapse is mainly initiated by the pull-in of external columns. For the cases with a larger number of fire floors (5 to 10) - Group M2, failure is dominated by thermal expansion and collapse is mainly initiated by swaying of the frame to the side of fire origin. The results show that for the investigated frame initiation of collapse mechanisms are affected by the fire type, the number of fire floors, and the location of the fire floor. The findings of this study could be of use to designers of buildings when developing fire protection strategies for steel framed buildings where the potential for a multifloor fire exists.