• Nuclear Engineering and Technology
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• v.56 no.3
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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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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• v.53 no.6
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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.

• Journal of the Korea Institute of Building Construction
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• v.19 no.3
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• pp.229-237
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• 2019

This study examines the estimation of strength through a ultra-high strength concrete mock-up specimen using the rock compressive strength test hammer. According to the test result, the commonly used strength estimation formulae showed differences among them when the data of this test were applied. In additional, it show that these formulae underestimated the actual measurements further when the compressive strength was 30MPa or greater and deviated the distribution range of actual measurements in all strength ranges. The rock test hammer showed a higher correlation than type N Schmidt hammer regardless of the direction of hit for each type of W/B and the inclusion of coarse aggregate, and mortar showed a little higher correlation than concrete. As a result, it can be suggested that the coefficient of variation and the standard deviation of the mortar(2.26%/1.36) are lower than those of the concrete(4.06%/2.5), and the smaller the size of the coarse aggregate, the smaller the coefficient of variation and the more accurate the value.

• Nuclear Engineering and Technology
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• v.51 no.1
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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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.

• Journal of the Korea Institute of Building Construction
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• v.20 no.1
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• pp.53-60
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• 2020

This paper aims to examine the insulation performance of insulated gang form by changing the energy (power) consumption and concrete calorific value to assist in concrete protection in cold weather. According to the test results, the general gang form will generate three times the energy (power) consumption for 12 hours after the concrete is poured. In contrast, insulated gang foam consumed no energy (power) for 21 hours after pouring. The final power consumption was 3.7 times higher than that of the general gang form, confirming the improved performance of insulated gang form with regard to energy (power) consumption. The calorific value examination shows that the calorific value changes significantly according to the change of outside temperature after concrete placement in the case of the general gang form. However, in the case of the insulated gang form, only a slight heat loss occurred in the part of the frame, and it showed a constant heating pattern from the concrete casting to the demolding of the mold.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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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.

• Proceedings of the KSME Conference
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• 2008.11b
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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.

• Journal of the Korea Institute of Building Construction
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• v.20 no.3
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• pp.245-252
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• 2020

Synthetic resin formwork is made of lightweight high-density polyethylene(HDPE). This study used a process flow chart that satisfies the system boundary (such as Cradle-to- Product shipmen ) required by ISO FDIS 13352 to evaluate the entire process of synthetic resin foam using. The entire life cycle inventory (LCI) database calculated from input energy sources, materials used, transportation methods, and manufacturing processes at the system boundary was analyzed. Based on the environmental impact assessment index methodology of the Ministry of Environment from the LCI data analysis of synthetic resin formwork, the environmental impact assessment was carried out through classification, normalization, characterization, and weighting process. The experimental results are as follows the amount of CO₂ (carbon) emission considering the number of conversions was about 32% lower than that of the Euroform. This shows that the use of synthetic resin formwork reduces material production by half compared to Euroform and reduces CO₂ (carbon) emissions.