• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.542-547
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• 1997

In the SSC(Super System Code), the reactivity feedback models of the Doppler effect and fuel axial expansion were modified to evaluate the safety performance of the metal-fueled core. The core radial expansion model was developed and implemented into the code as well. The transient analyses have been performed by the modified SSC for UTOP, ULOHS, ULOF/LOHS, and UTOP/LOF/LOHS events for one of the core design options being considered. Analysis results shows that the reactivity feedbacks can provide an inherent shutdown capability in response to key anticipated events without scram. Development of other reactivity feedback models and validation of these models against experimental data would make the SSC suitable for the assessment of the metal-fueled core safety performance.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1747-1755
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• 2021

This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.

• 한국건축시공학회지
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• 제19권3호
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• pp.229-237
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• 2019

본 논문은 암반용 압축강도테스트 해머를 이용하여 초고강도 콘크리트 모의부재 압축강도 실험을 통한 강도추정에 대해서 검토하고자 하였다. 실험결과에 따르면, 본 실험 데이터값을 토대로 기존에 주로 사용되던 강도 추정식을 적용할 경우 각 식마다 차이가 있는 것으로 나타났다. 또한 압축강도 30MPa 이상으로 갈수록 실측 데이터를 과소평가하고 있는 것으로 나타났으며, 모든 강도 영역에서 실측치의 분포범위를 크게 벗어나고 있는 것으로 나타났다. W/B 종류별 타격방향 및 굵은 골재 유무와 상관없이 암반용 테스트 해머가 N형 슈미트 해머보다 높은 상관관계를 나타내었으며, 모르타르가 콘크리트보다 좀 더 높은 상관관계를 나타내었다. 그리고 굵은 골재 유 무에 따른 모의부재 반발도 측정결과 모르타르(2.26%/1.36)의 변동계수와 표준편차가 콘크리트(4.06%/2.5)보다 낮게 나타났으며, 굵은 골재의 치수가 작을수록 변동계수가 작아져 보다 정확한 값을 나타내는 것을 알 수 있다.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.116-124
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• 2019

$M^2LFR-1000$ is a medium-power modular lead-cooled fast reactor, developed by University of Science and Technology of China (USTC), aiming at achieving a reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing the optimum safety and economics. In order to evaluate the safety performance of $M^2LFR-1000$ reactor core, three typical transients are selected from initiating events, which are unprotected transient overpower (UTOP), unprotected loss of offsite power (ULOHS+ULOF) and increase of feedwater flowrate with primary pumps trip (IFW+PLOF). These three transients presented and discussed in this paper are performed with the code Analysis of THermal-hydraulics of LEaks and Transients (ATHLET), which is developed by Gesellschaft $f{\ddot{u}}r$ Anlagen-und Reaktorsicherheit gGmbH (GRS). The results indicate that the $M^2LFR$ is safe enough under these three transients due to the good inherent safety features of the reactor, without human intervention, the reactor will reach a new steady state under UTOP condition.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.489-494
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• 1997

A comparative transient analyses were performed for oxide and nitride cores or a large (3000 MWt), pool-type, liquid-metal-cooled reactor (LMR). The study was focused on three representative accident initiators with failure to scram : the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected fast transient overpower (UFTOP). The margins to fuel melting and sodium boiling have been evaluated for these representative transients. The results show that there is an increase in safety margin with nitride core which maintains the physical dimensions of the oxide core.

• 한국건축시공학회지
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• 제20권1호
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• pp.53-60
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• 2020

본 논문은 동절기 콘크리트 보양 시 투입되는 에너지(전력) 소비량과 콘크리트 발열량 변화를 통해 단열갱폼의 단열성능을 검토하고자 하였다. 실험결과에 따르면, 일반갱폼은 콘크리트 타설 이후 12시간동안 에너지(전력) 소비가 3회 발생하게 된다. 반면 단열갱폼은 콘크리트 타설 후 21시간 동안 에너지(전력) 소비가 발생되지 않았다. 최종 전력 소비량은 일반갱폼이 단열갱폼보다 3.7배 높게 나타나 에너지(전력) 소비에서 단열갱폼의 우수한 성능을 확인할 수 있었다. 발열량 검토결과는 일반갱폼에서 콘크리트 타설 후 외기 온도변화에 따라 발열량이 크게 변하는 것을 알 수 있었다. 하지만 단열갱폼의 경우 프레임 일부에서 미비한 열손실이 발생했을 뿐 콘크리트 타설 직후부터 거푸집 탈형까지 일정한 발열패턴을 보여주고 있었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.645-650
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• 1997

An oxide fuel small size core (1200 MWt) was analyzed in comparison with a large size core (3600 MWt) in order to evaluate the size effects on transient safety performances of liquid-metal reactors (LMRs). in the first part of the study, main static safety parameters (i.e., Doppler coefficient, sodium void effect, etc.) of the two cores were characterized, and the second part of the study was focused on the dynamic behavior of the cores in two representative transient events: the unprotected loss-of-flow(ULOF) and the unprotected transient overpower (UTOP). Margins to fuel molting and sodium boiling have been evaluated for these representative transients. Results show that the small core has a generally better or equivalent level of safety performances during these events.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• 한국건축시공학회지
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• 제20권3호
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• pp.245-252
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• 2020

합성수지 거푸집은 내부식성이 우수한 경량의 고밀도 폴리에틸렌(HDPE)를 재료로 사용한다. 합성수지 거푸집의 전과정 평가를 위하여 ISO FDIS 13352에서 요구하는 시스템 경계를 만족하도록 공정 흐름도를 고려하였다. 이에 따라 고려된 시스템 경계는 Cradle-to- Product shipmen이다. 고려된 시스템 경계에서 투입 에너지원, 사용재료, 운송수단, 제작공정 등으로부터 산정한 전과정 목록(LCI) 데이터베이스를 분석하였다. 합성수지 거푸집의 LCI 데이터 분석으로 부터 환경부의 환경영향평가지수 방법론에 기반하여 분류화, 정규화, 특성화 및 가중치 과정을 거쳐 환경영향평가를 수행하고, 그 결과는 유로폼의 환경영향 평가값과 비교하였다. 실험결과, 전용횟수를 고려한 CO₂ 배출량은 유로폼 대비 2배 이상의 전용성을 갖는 합성수지 거푸집이 약 32 % 가량 낮았다. 이는 합성수지 거푸집 사용은 유로폼 대비 자재 생산을 1/2로 줄일 수 있으며, CO₂ 배출량 저감으로 이어질 수 있다.