• Nuclear Engineering and Technology
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• 제56권3호
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• pp.785-792
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• 2024

This article provides an overview of the design methodology used to develop a conceptual set of reactivity control mechanism of a micro reactor based on the U-Battery. The U-Battery is based on remote deployment and therefore it is favourable to provide a long fuel lifecycle. This is achieved by implementing a high fissile loading content, which proves challenging when considering reactivity control methods. This article follows the design methodology used to overcome these issues, with an emphasis on a new concept of a moveable moderator which utilises the size of the U-Battery as a small reduction in moderation provides a significant reduction in reactivity. The latest work on this project sees the moveable moderator investigated during a depressurised loss of forced coolant accident, where a reduction of moderator volume increases the maximum fuel temperature experienced. The overall conclusion is that the maximum fuel temperature is not significantly increased (4 K) due to the central reflector region relatively lower volumetric heat capacity compared to that of whole core. However, a small temperature increase is observed immediately after the transient due to the central reflector removal because it reaches energy equilibrium with the fuel region faster.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.838-843
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• 1998

Risk importance measures are widely wed to rank risk contributors in risk-based applications. Typically, Fussell-Vesely (F-V) importance and risk achievement worth (RAW) are used in the component importance raking for the reliability centered maintenance (RCM) analysis of safety system in nuclear power plants (NPPs). This study was performed as part of feasibility study on RCM for domestic NPPs, which is focused on the component importance ranking approach the maintenance recommendation. The approach of modulizing faulting tree basic events was applied in the simplification process of the PSA model and the validity of the approach was evaluated As a result of the case study, this paper included the importance and the maintenance recommendations for the safety-related equipments associated with safety injection and containment spray in large loss of coolant accident sequences.

• 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.

• 열처리공학회지
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• 제17권4호
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• pp.216-222
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• 2004

Study was focused on the quenching embrittlement property of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment in terms of high temperature oxidation and phase transformation. Property in LOCA condition of advanced cladding that contained Nb element was also investigated. Claddings were oxidized at given temperature and given time followed by water quenching. The results showed that ${\beta}$ phase which formed at quenching stage has an influence on cladding property. In case of advanced cladding, Nb retards cladding oxidation, thus enhances quenching resistance.

• Nuclear Engineering and Technology
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• 제15권2호
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• pp.98-109
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• 1983

2차원 다군 확산 이론에 의한 원자로 동특성 방정식의 해를 구하기 위해서 two-step alternating direction explicit method를 도입하였다. Alternating direction implicit method의 특별한 경우로써 이 방법의 정확도 및 안전성을 해석하였다. 이 방법의 타당성을 시험하기 위해서 TWIGL 전산조직에 사용한 implicit difference method와 비교하여 두 방법의 결과가 일치함을 알았다. 이 방법을 이용하여 가압경수형 원자로(PWR)의 제어봉 삽입시의 중성자 신속의 시간변화와, CANDU-PHW 원자로의 가상된 냉각재상실 사고시의 중성자 신속의 시간변화를 계산하여 이들 원자로의 제어능력을 확인하였다.

• Nuclear Engineering and Technology
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• 제11권3호
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• pp.195-202
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• 1979

In this paper, a cause-consequence chart is applied to evaluate the probability that the containment spray system in a nuclear power plant may not be woring properly, given a demand for spryaing at loss of coolant accident (LOCA). It is shown how the diagram provides a basis for calculating two probability measures for malfunctioning of this system, in case the test policy of the system is taken into account, i.e., average probability that the containment spray cannot be established, and average probability that the containment spray is established : spray stops before the required operating time is over.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.328-333
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• 1996

Boric acid concentrations of the refueling water storage tank (RWST) and the accumulators for Ulchin 1&2 (UCN 1&2) are increased to meet the post loss of coolant accident (post-LOCA) shutdown requirement for the extended fuel cycles from 12 months to 18 months. To maintain long term cooling (LTC) capability following a LOCA, the switchover tine is examined using BORON code to prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results show that, at 8 hours after the initiation of LOCA. the emergency core noting system (ECCS) should be manually realigned to the simultaneous recirculation mode from the cold leg recirculation mode.

• Nuclear Engineering and Technology
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• 제30권1호
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• pp.47-57
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• 1998

Boric acid concentrations of the refueling water storage tank and the accumulators for Westinghouse 3-loop type plants are increased to meet the post loss of coolant accident shutdown requirement for the extended fuel cycles from 12 months to 18 months. To maintain long term cooling capability following a LOCA, the switchover time is examined using BORON code to prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results show that hot leg recirculation switchover times are shortened to 7.5 hours from 24 hours after the initiation of LOCA for Kori 3&4 and 8 hours from 18 hours for Ulchin 1&2, respectively. The How path in the mode J for Kori 3&4 is recommended to realign to the simultaneous recirculation of both hot and cold legs from the cold leg recirculation, as done by Ulchin 1&2.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.475-479
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• 1995

During the hypothetical loss of coolant accident in the nuclear power plant the emergency core cooling water could not penetrate to the reactor core when the steam flow rate from the reactor core exceeds CCFL (Countercurrent flow limitation). The CCFL generated by earlier investigators are developed under the steady gas flow. However the flow instability in the reactor loop could generate oscillatory steam flow, hence their applicability under oscillating flow should be investigated. In this work, an experimental investigation of countercurrent flow in the vertical flow channel has been conducted under oscillatory gas flow. Pulsation of gas under oscillatory flow disturbs the flow pattern significantly and prevents flooding (CCFL) when its minimum value is less than the threshold gas flow rate value.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2006년도 춘계학술대회논문집
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• pp.246-249
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• 2006

Finite element vibration analysis of the trial 5x5 partial fuel assembly in the still water was performed using the substructure method. ANSYS software was used as a finite element modeling and modal analysis tool. The calculated natural frequencies of the partial fuel assembly were more consistent with the experimental results for the identical test model compared to the much larger solid model. This modeling technique can be utilized for the fuel assembly dynamic behavior analysis under normal operation, seismic and loss-of-coolant-accident analysis.