• 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.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.407-412
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• 2014

In this research, we investigated the burnup characteristics and the conversion of fertile $^{232}Th$ into fissile $^{233}U$ in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning $^{232}Th$ fuel (fuel pin 1) and $^{233}U$ fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.535-542
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• 2012

A sodium-cooled demonstration fast reactor with the KALIMER-600 as a reference plant is under design by KAERI. The safety grade decay heat removal system (DHRS), which is important to mitigate design basis accidents, was changed in the reactor design. A loss of heat sink and a vessel leak in design basis accidents were simulated using the MARS-LMR system transient analysis code on two plant systems. In the analyses, the DHRS of KALIMER-600 had a weakness due to elevation of the overflow path for the DHRS operation, while it was proved that the DHRS of the demonstration reactor had superior heat transfer characteristics due to the simplified heat transfer mechanism.

• Journal of the Earthquake Engineering Society of Korea
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• v.11 no.3 s.55
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• pp.73-86
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• 2007

In this paper, a simple seismic analysis model of the KALIMER-600 sodium-cooled fast reactor selected to be the candidate of the GEN-IV reactor is developed. By using this model, the seismic time history analysis is carried out to investigate the feasibilities of a seismic isolation design. The developed simple seismic analysis model includes the reactor building, reactor system,, IHTS piping system, steam generator, and seismic isolators. The dynamic characteristics of the simple seismic model are verified with the detailed 3-dimensional finite element analysis for each part of the KALIMER-600 system. By using the developed simple seismic model, the seismic time history analyses for both cases of a seismic isolation and non-isolation design are performed for the artificial time history of a SSE (Safe Shutdown Earthquake) 0.3g. From the comparison of the calculated floor response spectrum, it is verified that the seismically isolated KALIMER-600 reactor building shows a great performance of a seismic isolation and assures a seismic integrity.

• Journal of the Korean Society of Radiology
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• v.10 no.2
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• pp.117-124
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• 2016

The purpose of this study was tried to evaluate the filling sodium states inside the fuel rod of sodium-cooled fast reactor by digital medical X-ray. We used the diagnostic X-ray generators in digital radiography (DR). This study have found the optimal conditions by changing the effective focal spot size of X-ray tube and post-processing of the DR method with a tungsten edge plate in order to ensure excellent sharpness At this time, the sharpness and resolution were evaluated using the MTF (modulation transfer function). As a result, this study obtained a spatial resolution of 3.871 lp/mm (0.1 MTF), 3.290 lp/mm (0.5 MTF) when implemented the contrast strengthen post-processing in small focal spot. In this research, the result is able to evaluate the level of sodium inside the fuel rod by using the diagnostic X-ray generators in medical digital radiographic images.

• Journal of the KSME
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• v.51 no.12
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• pp.51-54
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• 2011

이 글에서는 제4세대 원자로로 다시 부각되고 있는 소듐냉각고속로(SFR: Sodium-cooled Fast Reactor)의 활용성, 계통설계 구성 및 공학적 안전설비에 대하여 가압경수로(PWR: Pressurized Water Reactor)와의 차이점을 위주로 소개한다.

• Korean Chemical Engineering Research
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• v.55 no.2
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• pp.170-179
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• 2017

We investigated design requirements of feed water drain and hydrogen vent systems for the sodium-water reaction pressure relief system (SWRPRS) of the prototype generation IV sodium cooled fast reactor (PGSFR). We evaluated the areas of the gas vent pipe of the water dump tank and the length of the water drain pipe of the steam generator to rapid drain of the water steam inside the steam generator for the normal and refueling operations, respectively. We also calculated the diameter of the gas vent pipe of the sodium dump tank which met its design pressure.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.47-58
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• 2015

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. Thefsensitivity of cooling time before prior to pyro-processing was studied. As the cooling time sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ${\leq}20%$ in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.187-192
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• 2007

The basic research on fast reactor technology was started in the mid-1960's in China. The emphasis was put on fast reactor neutronics, thermohydraulics, sodium technology, materials, fuels, safety, sodium devices and instrumentation. In 1987, the research turned to applied basic research with the conceptual design of a 60 MW experimental fast reactor as a target. The Project of the China Experimental Fast Reactor(CEFR) with a thermal power 65 MW was launched in 1993. The R&D of fast reactor technology then carried out to serve a design demonstration connected with the different phases of the conceptual, preliminary and detailed design of the CEFR. Recently, three directions of fast rector technology R&D activities have been considered, and some research programs have been developed. They are: (1) R&D related to the CEFR, i.e. experiments to be conducted on the CEFR for its safe operation, (2) R&D related to the projects of a prototype and the demonstration of fast reactors, and(3) advanced SFR technology within the framework of the international cooperation of INPRO and GIF.

• Nuclear Engineering and Technology
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• v.47 no.3
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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.