• 한국안전학회지
• /
• 제37권4호
• /
• pp.129-138
• /
• 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.

• 한국환경과학회지
• /
• 제17권5호
• /
• pp.573-580
• /
• 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.

• 콘크리트학회논문집
• /
• 제26권3호
• /
• pp.247-254
• /
• 2014

화재시 콘크리트 구조물은 구성재료의 상이한 열적특성으로 인해 강도가 저하하고 동시에 수직부재는 수평부재의 팽창에 의한 모멘트하중을 받아 전단파괴가 발생한다. 따라서 여러가지 화재곡선을 사용한 콘크리트 구조물의 화재시 거동에 대한 연구가 많이 이루어졌지만 주로 온도상승구간에서 발생하는 폭렬특성과 열팽창변형에 관한 연구가 대부분이다. 하지만 고온이 유지될 경우 발생할 수 있는 크리프변형은 화재시 구조물의 안정성에 큰 영향을 미치지만 상대적으로 연구가 미진한 상태이다. 또한 이러한 고온을 받는 콘크리트의 안정성에는 체적의 대부분을 차지하는 굵은골재의 열적특성이 큰 영향을 미치기 때문에 이 연구에서는 화강암계, clay계, clay-ash계 세 종류의 굵은골재를 사용한 콘크리트의 고온 역학적 특성을 평가했다. 그 결과 굵은골재의 성인으로 인한 내부공극 때문에 경량골재를 사용한 콘크리트가 일반골재를 사용한 콘크리트보다 높은 고온강도 및 탄성계수를 나타냈고 열팽창변형과 전체변형의 경우 더 낮은 변형률을 나타내어 온도상승구간에서의 구조적 안정성 측면에서 유리한 것으로 판단되었다. 그러나 고온크리프의 경우 내부공극으로 인해 더 큰 수축량이 발생하기 때문에 내화성능설계시에 이에 대한 추가적인 고려가 필요할 것으로 판단되었다.

• Nuclear Engineering and Technology
• /
• 제27권3호
• /
• pp.393-402
• /
• 1995

원자로냉각재계통의 설계를 위한 구조해석 분야에는 원자로의 정상운전 과정에서 발생하는 유체의 온도와 압력의 변화에 의해 냉각재계통에 발생하는 정적하중해석, 지진과 가상적인 분지관 파단사고에 의해 냉각재계통에 발생하는 동적하중해석분야로 구분할 수 있다. 원자로냉가재계통의 구조해석은 원자력발전소의 안전성 화보 측면을 중시하여 해석시 충분한 여유도를 고려한 보수적인 해석 방법을 원용한다. 지진이나 가상적인 분지관 파단사고에 의한 냉각재계통의 구조해석은 사고시 냉각재계통의 안전성을 유지하는 방어적인 개념으로서 기기의 건전성을 확보하기 위하여 충분한 보수성과 안전여유가 해석시 고려된다 정상운전에 의해 냉각재계통에 발생하는 하중은 원자력 발전소의 상존하는 하중의 개념으로서 냉각재계통의 기본 설계 하중으로 인식된다. 특히 고온 고압의 유체로 인하여 발생하는 냉각재 계통의 열팽창 현상은, 정상운전 하중으로 인하여 나타나는 전형적인 거동으로서, 냉각재계통 구조해석 결과읜 중요한 지표로서 인식된다. 따라서 냉각재계통의 열팽창 현상을 정확히 예측하는 것은 원자로 냉각재계통 구조해석의 가장 중요한 목표중의 하나이다. 본 연구에서는 정상운전 하중에 의한 원자로 냉가재계통의 열팽창 거동을 해석하기 위한 냉각재계통의 모델링 방법과 해석 방법을 제시하였다. 해석 결과의 타당성을 검토하기 위하여 최근 건설 완료 단계에 돌입한 표준형 1000 MWe 급 가압경수로(Pn)의 고온기능시험 (Hot Function Test)과정에서 실측한 자료를 근거로 하여 원자로냉각재계통의 열팽창 거동 해석의 타당성을 입증코자 하였다.

• Journal of Radiation Protection and Research
• /
• 제42권2호
• /
• pp.114-129
• /
• 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
• /
• 제26권1호
• /
• pp.63-77
• /
• 1994

이 연구는 가압경수로의 원자로 다운커머내에서 과도냉각시 직접용기주입에 따른 유체혼합현상을 가압열충격의 견지에서 시험모델을 사용하여 조사한 것이다. 시험모델은 ABB-CE System80＋ 원자로 구조에 근거하여 설계되었다. 이 원자로에 대한 가능성 있는 가압열충격 사고로서 콜드레그 소형파단 냉각재 상실사고와 주중기관 판단 사고가 선정되었다. 시험은 두 부분으로 구성되는데 첫째 부분은 원자로 다운커머에서 직접용기 주입수와 기존냉각재간의 유체혼합을 가시화법에 의하여 시험한 것이고, 둘째 부분은 별도의 시험모델에서 직접용기주입에 따른 열적혼합을 시험한 것이다. 가시화 시험에서는 과도적 냉각기간중 직접용기 주입수와 1차 냉각재간의 물리적 상호작용이 밝혀졌다. 열적혼합시험에서는 소형파단 냉각재 상실사고시 직접용기주입에 의한 심한 냉각현상이 다운커머내서 관찰되었다. 측정된 온도곡선은 소형파단 냉각재 상실사고에 대하여 REMIX 로드, 증기관 파단사고에 대하여는 COM-MIX-1B 코드에 의한 계산과 비교되었다.

• 한국해양공학회지
• /
• 제32권5호
• /
• pp.310-323
• /
• 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
• /
• 제9권5호
• /
• pp.1471-1481
• /
• 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
• /
• 제43권2호
• /
• pp.187-194
• /
• 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.

• 국제초고층학회논문집
• /
• 제10권4호
• /
• pp.265-283
• /
• 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.