• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2262-2275
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• 2022

The Korea Atomic Energy Research Institute (KAERI) studied the sodium-water reaction (SWR) minimized steam generator for the safety of the sodium-cooled fast reactor (SFR), and selected the copper bonded steam generator (CBSG) as the optimal concept. This paper introduces the conceptual design of the CBSG and the development of the CBSG sizing analyzer (CBSGSA). The CBSG consists of multiple heat transfer modules with a crossflow heat transfer configuration where sodium flows horizontally and water flows vertically. The heat transfer modules are stacked along a vertical direction to achieve the targeted large heat transfer capacity. The CBSGSA code was developed for the thermal-hydraulic analysis of the CBSG in a multi-pass crossflow heat transfer configuration. Finally, we conducted a preliminary sizing and rating analysis of the CBSG for the trans-uranium (TRU) core system using the CBSGSA code proposed by KAERI.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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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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• v.48 no.5
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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.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.338-344
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• 2010

A double rotation C-scanning system and a software program Under-Sodium MultiVIEW have been developed for the under-sodium viewing of a reactor core and in-vessel structures of a sodium-cooled fast reactor KALIMER-600. Double rotation C-scanning system has been designed and manufactured by the reproduction of double rotation plug of a reactor head in KALIMER-600. Hardware system which consists of a double rotating scanner, ultrasonic waveguide sensors, a high power ultrasonic pulser-receiver, a scanner driving module and a multi channel A/D board have been constructed. The functions of scanner control, image mapping and signal processing of Under-Sodium MultiVIEW program have been implemented by using a LabVIEW graphical programming language. The performance of Under-Sodium MultiVIEW program was verified by a double rotation C-scanning test in water.

• Nuclear Engineering and Technology
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• v.38 no.7
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• pp.591-618
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• 2006

An integrated R&D program is being conducted to study, qualify, and in some cases, develop materials with required properties for the reactor systems being developed as part the U.S. Department of Energy's Generation IV Reactor Program. The goal of the program is to ensure that the materials research and development (R&D) needed to support Gen IV applications will comprise a comprehensive and integrated effort to identify and provide the materials data and its interpretation needed for the design and construction of the selected advanced reactor concepts. The major materials issues for the five primary systems that have been considered within the U.S. Gen IV Reactor Program-very high temperature gas-cooled, supercritical water-cooled, gas-cooled fast spectrum, lead-cooled fast spectrum, and sodium-cooled fast spectrum reactors-are described along with the R&D that has been identified to address them.

• 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.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.285-290
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• 2007

KALIMER-600 is a sodium cooled fast reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.351-359
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• 2016

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.93-100
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• 2016

Since the sodium-cooled fast reactor is operated in a hostile environment due to the use of liquid sodium as its coolant, advanced techniques for in-service inspection are required to periodically verify the integrity of the reactor. This paper presents the development of in-service inspection techniques for Proto-type Generation IV Sodium-cooled Fast Reactor. First, the 10 m long plate-type ultrasonic waveguide sensor has been developed for in-service inspection of reactor internals, and its feasibility was verified through several under-water and under-sodium experiments. Second, the combined inspection system for in-service inspection of ferromagnetic steam generator tubes has been developed. The remote field eddy current testing and magnetic flux leakage testing can be conducted simultaneously by using the developed inspection system, and the detectability was demonstrated through several damage detection experiments. Finally, the electro-magnetic acoustic transducer which can withstand high temperature and be installable in the remote operated vehicle has been developed for in-service inspection of the reactor vessel, and its detectability was investigated through damage detection experiments.

• Journal of computational fluids engineering
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• v.21 no.3
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• pp.8-14
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• 2016

A numerical analysis is conducted to estimate the core expansion and the energy behaviors induced by a core disruptive accident in a sodium-cooled fast reactor. The numerical formulation based on underwater explosion theory is carried out to simulate the core explosion inside the reactor vessel. The transient pressure, temperature and expansion of the core are examined by solving the equation of state and nonlinear governing equation of momentum conservation in one-dimensional spherical coordinates. The energy balance inside the computation domain is examined during the core expansion process. Heat transfer between the core and the sodium coolant, and the bubble rise during the expansion process are briefly investigated.