• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1665-1670
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• 2023

It is shown that in the IBR-2M reactor by the end of the reactor cycle, changes in dynamics are observed associated with a strong weakening of the fast power feedback (PF), as a result of which the reactor becomes oscillatorily unstable. After each week of zero-power operation the negative changes in reactor dynamics disappear and the stability of the reactor is restored. Thus, the reactor undergoes cyclic changes in the oscillatory instability. The correlation between of a fast PF and a slow PF is experimentally observed, which makes it possible to almost completely eliminate the cyclic component of instability by changing the control mode of rods of the control system.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 Proceedings of the 4th Korea-China Joint Workshop on Nuclear Waste Management
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• pp.220-237
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• 2004

China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basic strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan(2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before 2010 respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor(CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached the top namely 57m above the ground. More than one hundred components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000$m^2$ floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which is only under consideration up to now. Some important technical selections have been settled, but its design has not yet started.

• 한국전산유체공학회지
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• 제16권2호
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• pp.94-101
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• 2011

The fast pyrolysis characteristics of lignocellulosic biomass are investigated for a bubbling fluidized bed reactor by means of computational fluid dynamics (CFD). To simulate multiphase reacting flows for gases and solids, an Eulerian-Eulerian approach is applied. Attention is paid for the primary and secondary reactions affected by gas-solid flow field. From the result, it is scrutinized that fast pyrolysis reaction is promoted by chaotic bubbling motion of the multiphase flow enhancing the mixing of solid particles. In particular, vortical flow motions around gas bubbles play an important role for solid mixing and consequent fast pyrolysis reaction. Discussion is made for the time-averaged pyrolysis reaction rates together with time-averaged flow quantities which show peculiar characteristics according to local transverse location in a bubbling fluidized bed reactor.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.134-140
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• 2016

The Control Rod Assembly (CRA) controls the reactor power by adjusting its position in the reactor core during normal operation and should be quickly inserted into the reactor core by free drop under scram condition to shut down chain reactions. Therefore, the drop time of the CRA is one of important factors for the safety of the nuclear reactor and must be experimentally verified. This study presents the drop performance test of the CRA which has been conceptually designed for the Proto-type Generation IV Sodium-cooled Fast Reactor. During the test, the CRA was free dropped from a height of 1 m under different flow rate conditions and its drop time was measured. The results showed that the drop time of the CRA increased as the flow rate increased; the average drop times of the CRA were approximately 1.527 seconds, 1.599 seconds and 1.676 seconds at 0%, 100% and 200% of design flow rates, respectively.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1096-1108
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• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1071-1082
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• 2016

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2835-2843
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• 2023

After an emergency shutdown of a lead-bismuth fast reactor, thermal stratification occurs in the upper Plenum, which negatively impacts the integrity of the reactor structure and the residual heat removal capacity of natural circulation flow. The research on thermal stratification of reactors has mainly been conducted using an experimental method, a system program, and computational fluid dynamics (CFD). However, the equipment required for the experimental method is expensive, accuracy of the system program is unpredictable, and resources and time required for the CFD approach are extensive. To overcome the defects of thermal stratification analysis, a high-precision full-order thermal stratification model based on CFD technology is prepared in this study. Furthermore, a reduced-order model has been developed by combining proper orthogonal decomposition (POD) with Galerkin projection. A comparative analysis of thermal stratification with the proposed full-order model reveals that the reduced-order thermal stratification model can well simulate the temperature distribution in the upper plenum and rapidly elucidate the thermal stratification interface characteristics during the lead-bismuth fast reactor accident. Overall, this study provides an analytical tool for determining the thermal stratification mechanism and reducing thermal stratification.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1213-1224
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• 2024

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1024-1031
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• 2019

The Molten Salt Fast Reactor (MSFR) with its liquid circulating fuel and its fast neutron spectrum calls for a new safety approach including technological neutral methodologies and analysis tools adapted to early design phases. In the frame of the Horizon2020 program SAMOFAR (Safety Assessment of the Molten Salt Fast Reactor) a safety approach suitable for Molten Salt Reactors is being developed and applied to the MSFR. After a description of the MSFR reference design, this paper focuses on the identification of the Postulated Initiating Events (PIEs), which is a core part of the global assessment methodology. To fulfil this task, the Functional Failure Mode and Effect Analysis (FFMEA) and the Master Logic Diagram (MLD) are selected and employed separately in order to be as exhaustive as possible in the identification of the initiating events of the system. Finally, an extract of the list of PIEs, selected as the most representative events resulting from the implementation of both methods, is presented to illustrate the methodology and some of the outcomes of the methods are compared in order to highlight symbioses and differences between the MLD and the FFMEA.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2474-2482
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• 2023

The development of fast reactors with complex designs and operation status requires more accurate and effective simulation. The Monte-Carlo method can generate multi-group cross-sections in arbitrary geometry without approximation on resonances treatment and leads to good results in combination with diffusion codes. However, in previous studies, the coupling of Monte-Carlo generated multi-group cross-sections (MC-MGXS) and transport solvers has shown relatively large biases in fast reactor problems. In this paper, the main contribution to the biases is proved to be the neglect of the angle-dependence of the total cross-sections. The flux-moment homogenization technique (MHT) is proposed to take into account this dependence. In this method, the angular dependence is attributed to the transfer cross-sections, keeping an independent form for the total sections. For the MET-1000 benchmark, the multi-group transport simulation results with MC-MGXS generated with MHT are improved by 700 pcm and an additional 120 pcm with higher order scattering. The factors that cause the residual bias are discussed. The core power distribution bias is also significantly reduced when MHT is used. It proves that the MCMGXS with MHT can be applicable with transport solvers in fast reactor analysis.