• Nuclear Engineering and Technology
• /
• 제54권10호
• /
• pp.3897-3908
• /
• 2022

High-fidelity nuclear data libraries and neutronics simulation tools are essential for the development of fast reactors. The IAEA coordinated research project on "Neutronics Benchmark of CEFR Start-Up Tests" offers valuable data for the qualification of nuclear data libraries and neutronics codes. This paper focuses on the verification and validation of the CEFR start-up modelling using OpenMC Monte-Carlo code against the experimental measurements. The OpenMC simulation results agree well with the measurements in criticality, control rod worth, sodium void reactivity, temperature reactivity, subassembly swap reactivity, and reaction distribution. In feedback coefficient evaluations, an additional state method shows high consistency with lower uncertainty. Among 122 relative errors in the benchmark of the distribution of nuclear reaction, 104 errors are less than 10% and 84 errors are less than 5%. The results demonstrate the high reliability of OpenMC for its application in fast reactor simulations. In the companion paper, the influence of cross-section libraries is investigated using neutronics modelling in this paper.

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

• 한국생산제조학회지
• /
• 제8권4호
• /
• pp.123-129
• /
• 1999

Although type 316 stainless steel is widely used such as in reactors of petrochemical plants and pipes of steam power plants and s attracting attention as potential basic material for the fast breeder reactor structure alloys in nuclear power plants and is attracting attention as potential basic material for the fast breeder reactor structure alloys in nuclear power plants the effect of precipitates which form during the long term exposure at service temperature on creep properties is not known sufficiently. In this study to investigate the creep properties and the influence of prior aging on the microstructure to form precipitates specimens were first solutionized at 113$0^{\circ}C$ for 20 minutes and then aged for different times of 0 hr, 100 hrs, 1000 hrs and 2200 hrs at 75$0^{\circ}C$ After heat treatments tensile tests both at room temperature and $650^{\circ}C$ and constant load creep ruptuere tests were carried out.

• 한국환경보건학회:학술대회논문집
• /
• 한국환경보건학회 2003년도 Challenges and Achievements in Environmental Health
• /
• pp.208-212
• /
• 2003

This study was performed to examine the effects of chaff as bulking materials on temperature, pH, weight and volume reduction and salinity in aerobic composting of food wastes. Volume ratios of food wastes to chaff in reactor Control, Ch-1, Ch-2, Ch-3 and Ch-4 were 4:0, 4:1, 4:2, 4:3 and 4:4, respectively. Reactors were operated for 24 days with 1 hour stirring by 1 rpm and 2 hours aeration per day. The lowering of the volume ratio of food wastes to chaff resulted in the ascending of the highest reaction temperature and the elongation of the high temperature reaction period. The lowering of the volume ratio of food wastes to chaff resulted in the more fast time of pH ascending. The lowering of the volume ratio of food wastes to chaff resulted in the more fast consistency in the weight and volume reduction rates. Salinities were condensed by reaction days. The final salinity of Control was 2.79%, and the final range of salinities of chaff mixtures was 2.18 - 2.37%.

• Nuclear Engineering and Technology
• /
• 제38권2호
• /
• pp.129-138
• /
• 2006

The Gas-cooled Fast Reactor (GFR) is one of the six reactor concepts selected within the framework of the Generation IV initiative and is the reference concept for the Commissariat $\grave{a}$ l'Energie Atomique $(CEA^1)$. Two reactor unit sizes have been considered: 600 MWth and 2400 MWth. As far as thermal-hydraulics is concerned, reactor decay heat removal (DHR) proves to be a major issue. The CEA has conducted exploratory design studies to address this issue and a reference solution for the 600MWth reactor has been recommended.

• 한국연소학회:학술대회논문집
• /
• 한국연소학회 2012년도 제44회 KOSCO SYMPOSIUM 초록집
• /
• pp.231-233
• /
• 2012

Circulating Fluidized Bed (CFB) is a technically and economically proven technology for boiler systems and large CFB coal boilers are making inroads into the domestic power boiler market. For biomass gasification, it is also considered as a very promising technology for commercial. Due to the lack of experiences of a large scale CFB gasifier, however, any large scale CFB gasifiers are hard to in Korea in spite of fast-growing demand of domestic market. In this study, a 3 $MW_{th}$ CFB gasifier was developed for biomass gasification. The CFB gasifier consists of interconnected fast and bubbling fluidized bed reactors including unique features for in-situ tar removal. Various numerical and experimental approaches will be presented such as basic modeling works, investigation of hydrodynamics with a cold model, computational particle fluid dynamics and experiments in the 3 MWth gasifier.

• Nuclear Engineering and Technology
• /
• 제43권4호
• /
• pp.335-342
• /
• 2011

As part of the Department of Energy's Fuel Cycle Research and Development Program an electrochemical technology employing molten salts is being developed for recycle of metallic fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. This technology has been deployed for treatment of used fuel from the Experimental Breeder Reactor II (EBR-II) in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory. This process is based on dry (non-aqueous) technologies that have been developed and demonstrated since the 1960s. These technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including preparation of associated high-level waste forms.

• Nuclear Engineering and Technology
• /
• 제52권1호
• /
• pp.51-58
• /
• 2020

Despite the recent growth of interest in molten salt reactor technology and the crucial role which heat transfer plays in the design of power reactors, specific studies on the design of heat exchangers for the Molten Salt Fast Reactor have not yet been performed. In this work we deliver a preliminary but quantitative analysis of the intermediate heat exchangers, based on reference design data from the SAMOFAR H2020-Euratom project. Two different promising reference technologies are selected for study thanks to their compactness features, the Printed Circuit and the Helical Coil heat exchangers. We present preliminary design results for each technology, based on simplified design tools. Results highlight the limiting effects of the compactness constraints imposed on the fuel salt inventory and the allowed size. Large pressure drops on both flow sides are to be expected, with negative consequences on pumping power and natural circulation capabilities. The small size required for the flow channels also represents possible fabrication issues and safety concerns regarding channel blockage.

• Nuclear Engineering and Technology
• /
• 제51권5호
• /
• pp.1272-1278
• /
• 2019

A lumped kinetic modeling platform is developed to investigate the coupled nuclear/thermo-fluid features of the closed natural circulation loop in a low power lead cooled fast reactor. This coolant material serves a reliable choice with noticeable thermo-physical safety characteristics in terms of natural convection. Boussienesq approximation is resorted to appropriately reduce the governing partial differential equations (PDEs) for the fluid flow into a set of ordinary differential equations (ODEs). As a main contributing step, the coolant circulation speed is accordingly correlated to the loop operational power and temperature levels. Further temporal analysis and control synthesis activities may thus be carried out within a more consistent state space framework. Nyquist stability criterion is thereafter employed to carry out a sensitivity analysis for the system stability at various power and heat sink temperature levels and results confirm a widely stable natural circulation loop.

• Nuclear Engineering and Technology
• /
• 제56권2호
• /
• pp.437-443
• /
• 2024

The development and testing of inspection equipment is necessary for the safe deployment of advanced nuclear reactors. One proposed advanced reactor design is Westinghouse's lead-cooled fast reactor (LFR). In this paper, the process of achieving adequate wetting for an ultrasonic under-lead viewing system is discussed and results presented. Such a device would be used for inspection in the molten lead core during reactor outages. Wider tests into the wetting of various materials in molten lead at microscale were performed using electron microscopy. The possible mechanisms and kinetics for materials wetting in lead, particularly stainless steel and nickel, are proposed and discussed.