• 에너지공학
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• 제25권4호
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• pp.124-132
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• 2016

노심보충탱크 상부에 설치되는 유동분사기 형상에 따른 냉각수 주입특성 및 탱크 내에서의 열수력 현상 변화를 파악하기 위한 안전주입배관 2인치 파단 소형냉각재상실사고(SBLOCA) 모의시험이 잔열 및 피동잔열제거계통(PRHRS) 모의 없이 수행되었다. 두 가지 형상의 유동분사기를 설치하고 수행한 각각의 시험은 거의 유사한 초기 및 경계조건에서 수행되었으며, 이로 인해 반복시험에 대한 재현성이 충족되었다고 판단된다. 시험결과는 유동분사기의 종류(본 시험에서는 구멍의 개수에 해당)에 관계없이 유사한 열수력학적 거동을 보였으며, 초기 주입유량 관점에서는 구멍의 개수가 2배인 B형이 A형에 비해 좀 더 우수한 주입 성능을 보였다. 노심보충탱크 격리 밸브가 개방된 후 압력평형배관을 통해 유입되는 고온의 원자로냉각재는 상부 헤더에서 상대적으로 저온인 $50^{\circ}C$ 물과 혼합되면서 증기 응축과 같은 상변화에 의한 압력 변동을 동반하는 다차원 열유동 현상을 일으키게 된다. 이로 인해 초반부 노심보충탱크 주입 유량은 상온운전 조건에서 보다는 작게 되고, 일정시간 경과 후에는 유사한 주입유량 특성을 보였다.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1555-1562
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• 2023

Natural circulation phenomena have been nowadays largely revisited aiming to investigate the performances of passive safety systems in carrying-out heat removal under accidental conditions. For this purpose, assessment studies using CFD (Computational Fluid Dynamics) and also 3D thermal-hydraulic system codes are considered at different levels of the design and safety demonstration issues. However, these tools have not being extensively validated for specific natural circulation flow regimes involving flow mixing, temperature stratification, flow recirculation and instabilities. In the present study, an experimental test case based on a small-scale pool test rig experiment performed by Korea Atomic Energy Research Institute, is considered for code-to-code and code-to-experimental data comparison. The test simulation is carried out using the FLUENT and the 3D thermal-hydraulic system CATHARE-2 codes. The objective is to evaluate and compare their prediction capabilities with respect to the test conditions of the experiment. It was observed that, notwithstanding their numerical and modelling differences, similar agreement results are obtained. Nevertheless, additional investigations efforts are still needed for a better representation of the considered phenomena.

• 해양환경안전학회:학술대회논문집
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• 해양환경안전학회 2003년도 춘계학술발표회
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• pp.125-134
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• 2003

이 논문에서는 새로 개정된 2000 SOLAS 규정에 근거하여 소형선박용 수동식 레이더 리프렉터(PRR-S) 설계에 관해서 기술했다. 이 연구의 설계개념은 PRR-S이 선박요동에 의해서 레이더 반사면적(RCS)이 변동하는 것을 방지하기 위하여 ‘catch rain’ 위치를 유지하도록 한 것이다. PRR-S은 원형판으로 구성한 옥타헤더럴 레이더 리프렉터와 요동방지를 위한 짐발장치로 구성하였다. 설계한 PRR-S는 전파흡수실에서 성능실험을 하였다. 실험결과, 기준으로 정한 Davis Echomaster사의 리프렉터 보다 설계한 PRR-S이 높은 RCS를 나타냈다.

• 대한조선학회논문집
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• 제59권1호
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• pp.29-38
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• 2022

In offshore installations, fires cause the structure to lose its rigidity and it leads to structural integrity and stability problems. The Passive Fire Protection (PFP) system slows the transfer rate of fire heat and helps prevent the collapse of structures and fatality. Especially, intumescent epoxy coating is widely used in the offshore industry, and not only is the material cost expensive, but it also takes a lot of time and cost for construction. Several studies have been conducted on the efficient application and optimal design of the PFP system. However, the mechanical properties and the strength of the PFP material have not been considered. In addition, researches on the correlation between the thickness of PFP and the structural behavior were insufficient. Therefore, this study aims to analyze the thermal and mechanical effects of the PFP on the structure when it is applied to the structural member. In particular, it is intended to resolve the change in strength characteristics of the structural members as the thickness of the PFP increases.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1125-1139
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• 2023

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS). There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an incontainment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance. In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.395-400
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• 1996

Scoping analyses for the Safety Injection System (SIS) configuration for Korean Next Generation Reactor (KNGR) are peformed in this study. The KNGR SIS consists of four mechanically separated hydraulic trains. Each hydraulic train consisting of a High Pressure Safety Injection (HPSI) pump and a Safety Injection Tank (SIT) is connected to the Direct Vessel Injection (DVI) nozzle located above the elevation of cold leg and thus injects water into the upper portion of reactor vessel annulus. Also, the KNGR is going to adopt the advanced design feature of passive fluidic device which will be installed in the discharge line of SIT to allow more effective use of borated water during the transient of large break LOCA. To determine the feasible configuration and capacity of SIT and HPSI pump with the elimination of the Low Pressure Safety Injection (LPSI) pump for KNGR, licensing design basis evaluations are performed for the limiting large break LOCA. The study shows that the DVI injection with the fluidic device SIT enhances the SIS performance by allowing more effective use of borated water for an extended period of time during the large break LOCA.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2534-2544
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• 2024

This paper proposes a flooding safety system (FSS) and its operation strategy that can provide long-term safety and effective maintenance for modules of small modular reactor (SMR) and metal containment maintained at dried environment during normal operation. During hypothesized accidents, the FSS re-collects the evaporated steam into the common pool by the condenser installed above the common water pool and provides an emergency coolant for the cavities and auxiliary pools. This study suggested that the condensate re-collection strategy using the FSS can effectively delay the depletion of available water in response to the accidents. Without recollection, the achievable grace periods ranged from 44 to 1507 days for six-module and one-module accidents, respectively. However, with a full re-collection (ratio = 1.0), the time to total depletion of emergency coolant was estimated indefinite. Even with a partial re-collection ratio of 0.3, a grace period of 83.5 days could be ensured for a six-module transient. This study reported the effectiveness of condensate re-collection and the FSS as an innovative safety management strategy and system. Employing a condensate re-collection strategy with a high re-collection ratio can enhance the long-term safety and effective convenience of SMR operations and maintenance.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.227-239
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• 2015

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.537-556
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• 2005

A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method fur the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) Is being designed to be able to produce hydrogen by themo-chemical iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.

• 한국안전학회지
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• 제31권5호
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• pp.95-101
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• 2016

This study proposes a multi-scale dynamic system reliability analysis of control system as a method of quantitative evaluation of its performance in probabilistic terms. In this second paper, we discuss the control effect of the viscous damper on the seismic performance of the structure-level failure. Since the failure of one structural member does not necessarily cause the collapse of the structural system, we need to consider a set of failure scenarios of the structural system and compute the sum of the failure probabilities of the failure scenarios where the statistical dependence between the failure scenarios should be taken into account. Therefore, this computation requires additional system reliability analysis. As a result, the proposed approach takes a hierarchial framework where the failure probability of a structural member is computed using a lower-scale system reliability with the union set of time-sequential member failures and their statistical dependence, and the failure probability of the structural system is again computed using a higher-scale system reliability with the member failure probabilities obtained by the lower-scale system reliability and their statistical dependence. Numerical results demonstrate that the proposed approach can provide an accurate and stable reliability assessment of the control performance of the viscous damper system on the system failure. Also, the parametric study of damper capacity on the seismic performance has been performed to demonstrate the applicability of the proposed approach through the probabilistic assessment of the seismic performance improvement of the damper system.