• Nuclear Engineering and Technology
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• 제49권1호
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• pp.54-70
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• 2017

The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility-the advanced core-cooling mechanism experiment (ACME)-was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups-a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break-were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative effects on the passive core cooling performance caused by nitrogen injection during the SBLOCA transient.

• 구강회복응용과학지
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• 제19권1호
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• pp.27-33
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• 2003

Excessive heat generation at the implant-bone interface may cause irreversible bone damage and loss of osseointegration. The effect of heat generation in vitro at the implant surface caused by abutment reduction with high-speed dental turbine was examined. Titanium-alloy abutments connected to a titanium alloy screw-implant embedded in an acrylic-resin block in a $37^{\circ}C$ water bath were prepared. Temperature changes were recorded via embedded thermocouples at the cervix and apex of the implant surface. Analysis of variance for repeated measures was used to compare seven treatment groups. Fifty seconds of continuous cutting with air and water coolant caused a mean temperature increase of $1.24^{\circ}C$ at apex and $5.77^{\circ}C$ at cervix. Similar intermittent cutting caused increase of $2.50^{\circ}C$ at apex and $1.64^{\circ}C$ at cervix. But, continuous cutting with air coolant caused a mean temperature increase of $6.47^{\circ}C$ at apex and $5.77^{\circ}C$ at cervix. Similar intermittent cutting caused increase of $6.47^{\circ}C$ at apex and $5.77^{\circ}C$ at cervix. Preparation of implant abutment does not lead to detrimental effect on peri-implant tissues provided that adequate cooling. However, without water cooling, extreme overheating could be provoked, reaching the critical temperature that would lead to irreversible bone damage within only a few seconds.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2005년도 춘계 학술발표회 논문집
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• pp.183-189
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• 2005

Nuclear containment structure is the last barrier for being secure from any nuclear power plant accident. Even though the safety requirements of nuclear power plant have been focused on removing accidental situations, nuclear containment structure must reserve the sufficient resisting capacity to any accident because it works as the last barrier. The acceptable nuclear containment structure makes possible to limit the effect of internal accidents and to avoid radioactive release. In this study, to conduct the numerical analysis for the structural safety of a containment structure, loss of coolant accident (LOCA) is considered as the basic accidental load, and Wolsong containment structure is considered as a target structure. The CANDU containment structure, such as Wolsong containment structure, is a prestressed concrete shell structure which has dome and is reinforced with bonded tendons. The evaluation of ultimate pressure capacity was conducted by nonlinear analysis of a prestressed concrete containment structure.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.497-511
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• 2004

A PSV (pressurizer safety valve) popping test carried out in the early phases of a refueling outage may trigger a test-induced LOCA(loss of coolant accident) if a PSV fails to fully close and is stuck in a partially open position. According to a KSNP (Korea standard nuclear power plant) low power and shutdown PSA (probabilistic safety assessment), the failure of a high pressure safety injection (HPSI) accompanied by the failure of a PSV to fully close was identified as a dominant accident sequence with a significant impact on low power and shutdown risks (LPSR). In this study, we aim to investigate and verify a new means for mitigating this type of accident using a thermal-hydraulic analysis. In particular, we explore the applicability of an aggressive cool-down combined with operator actions. The results of the various sensitivity studies performed there will help reduce LPSR and improve Refueling outage safety.

• 기술사
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• 제30권5호
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• pp.82-95
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• 1997

A Study of Improvement in Reactor Thermal Power Measurement Method using Kalman Filter. The objectives of the safety analysis of nuclear power plants are to maintain the surface temperature of fuel and fuel cladding within limit value in case of Loss of Coolant accident (LOCA) so that it ensures the safety and reliability of nuclear power plants. The new technique evaluating the reactor power and improvement of existing plant system increase the safety margin of nuclear power plant operation, and accordingly, economic effect will be anticipated. Hereby, 1 would like to introduce reactor power measurement method using Kalman filter that enables to calculate the reactor power more precisely combining the parameters, for example, turbine output as the 1 st stage pressure of high pressure turbine, and reactor power using energy equilibrium relation. It is expected that the new technique will enhance the accuracy of measurement of reactor power and maintain the reliability of nuclear power operation by increasing operational safety margin, and gain the economic benefit

• Nuclear Engineering and Technology
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• 제32권5호
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• pp.521-529
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• 2000

A modified interfacial heat transfer correlation between a dispersed water droplet and ambient superheated steam is proposed and compared with available experimental data and other correlations. Modified one overcomes the inherent deficiencies of Lee and Ryley's interfacial heat transfer correlation that ignored the effects of steam superheating which can not be neglected especially in the reflood situation of a loss-of-coolant accident. Modified one is represented by (equation omitted) In the present correlation the effect of possible subcooling of a water droplet is not taken into consideration. Comparison of the above correlation with currently available measurement data for a water droplet in high temperature gas flow shows that the proposed one correlates well with the measurement data where the degree of superheating is negligible and considerable.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.775-784
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• 2009

This paper summarizes the tests performed in the ATLAS facility during its first two years of operation (2007${\sim}$2008). Two categories of tests have been performed successfully: (a) the reflood phase of the large-break loss-of-coolant accidents in a cold leg, and (b) the breaks in one of four direct vessel injection lines. Those tests contributed to understanding the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing an evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Several important and interesting phenomena have been observed during the tests. In most cases, the ATLAS shows reasonable accident characteristics and conservative results compared with those predicted by one-dimensional safety analysis codes. A wide variety of small-break LOCA tests will be performed in 2009.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1066-1080
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• 2024

SMART100 has a containment pressure and radioactivity suppression system (CPRSS) for passive containment cooling system (PCCS). This prevents overheating and over-pressurization of a containment through direct contact condensation in an in-containment refueling water storage tank (IRWST) and wall condensation in a CPRSS heat exchanger (CHX) in an emergency cool-down tank (ECT). The Korea Atomic Energy Research Institute (KAERI) constructed scaled-down test facilities, SISTA1 and SISTA2, for the thermal-hydraulic validation of the SMART100 CPRSS. Three separate effect tests were performed using SISTA1 to confirm the heat removal characteristics of SMART100 CPRSS. When the low mass flux steam with or without non-condensable gas is released into an IRWST, the conditions for mitigation of the chugging phenomenon were identified, and the physical variables were quantified by the 3D reconstruction method. The local behavior of the non-condensable gas was measured after condensation inside heat exchanger using a traverse system. Stratification of non-condensable gas occurred in large tank of the natural circulation loop. SISTA2 was used to simulate a small break loss-of-coolant accident (SBLCOA) transient. Since the test apparatus was a metal tank, compensations of initial heat transfer to the material and effect of heat loss during long-term operation were important for simulating cooling performance of SMART100 CPRSS. The pressure of SMART100 CPRSS was maintained below the design limit for 3 days even under sufficiently conservative conditions of an SBLOCA transient.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.937-946
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• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• 한국표면공학회지
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• 제49권4호
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• pp.389-394
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• 2016

The operation method of nuclear power plants is currently changing to high burn-up and long period that can enhance economics and efficiency of the plant. Since nuclear plant operation environment has been becoming severe, the amount of absorbed hydrogen also has increased. Absorbed hydrogen can be fatal securing safety of nuclear fuel cladding in case of Loss of Coolant Accidents(LOCA). In order to examine the impact of hydride on high-temperature oxidation, high-temperature oxidation experiment was performed on normal Zry-4 cladding and on Zry-4 cladding where hydrogen is charged in air pressure steam atmosphere under the $950^{\circ}C$ and $1000^{\circ}C$. According to the results, while oxidation acceleration due to charged hydrogen was not observed prior to breakaway oxidation creation, oxidation began to accelerate in cladding where hydrogens charged as soon as the breakaway oxidation started. If so much hydrogen are charged in the cladding, equiaxial monoclinic phase to unstable of stress is formed and it is presumed that oxidation is accelerated because nearby stress caused a crack in equiaxial phase, and that makes corrosion resistance decline sharply.