scholarly journals Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

2013 ◽  
Vol 2013 ◽  
pp. 1-18 ◽  
Author(s):  
Yeong-il Kim ◽  
Yong Bum Lee ◽  
Chan Bock Lee ◽  
Jinwook Chang ◽  
Chiwoong Choi
Keyword(s):  
Fast Reactor ◽  
Electricity Generation ◽  
Nuclear Power ◽  
Large Scale ◽  
Development Program ◽  
Energy Resources ◽  
Test Facility ◽  
Current Status ◽  
Hydraulic Test ◽  
Design Concepts

Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR) design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

Energy in the 1980s - Operational experience in nuclear power stations

1974 ◽  
Vol 276 (1261) ◽  
pp. 571-586
Keyword(s):  
Nuclear Power ◽  
Nuclear Energy ◽  
Large Scale ◽  
Electrical Power ◽  
Operating Experience ◽  
Operational Experience ◽  
Original Design ◽  
Safety Requirements ◽  
Design Concepts ◽  
Power Stations

From the first self-sustaining nuclear reaction to the present day represents a span of three decades: within that time large-scale generation of electrical power from nuclear energy has become acknowledged as economic, safe and environmentally acceptable. Within the U .K . 10% of electricity consumed is of nuclear origin. Some of the C.E.G.B. reactors have been in service for over 10 years. The operating experience that has been gained shows how the original design concepts have been ultimately developed. Some of the difficulties encountered and the engineering solutions are presented. Operating experience feeds back to the design philosophy and safety requirements for future nuclear plant. In this way a foundation is provided for the further exploitation of what must become a major source of energy in the next decade.

Scientific design of a large-scale sodium thermal–hydraulic test facility for KALIMER—Part I: Scientific facility design

2013 ◽  
Vol 265 ◽  
pp. 497-513 ◽  
Author(s):  
Soon-Joon Hong ◽  
Doo-Yong Lee ◽  
Jae-Hyuk Eoh ◽  
Tae-Ho Lee ◽  
Yong-Bum Lee
Keyword(s):  
Large Scale ◽  
Test Facility ◽  
Facility Design ◽  
Hydraulic Test

Temperature Field Analysis for Freezing Sealing Structure of Sodium Gate Valve in China Experiment Fast Reactor (CEFR)

2013 ◽  
Author(s):  
Lina Hu ◽  
Huajin Yu ◽  
Mingyu Lv
Keyword(s):  
Temperature Field ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Large Scale ◽  
Gate Valve ◽  
Structure Design ◽  
Field Analysis ◽  
Power Station ◽  
Experimental Relation ◽  
Simulation Results

In order to achieve the localization of the large-scale sodium gate valve in fast reactor nuclear power station, this paper use fluid software FLUEN, using two different methods, analysis temperature field of sodium gate valve freezing sealing structure of CEFR. The first method is to establish air model; the second method is to making use of experimental relation to calculate surface coefficient of heat transfer to simulate temperature field of freezing sealing structure of sodium gate valve in CEFR. Compared with measuring results on-site, simulation results of the first method have error within 10%. Therefore, it is feasible that simulating temperature field of freezing sealing structure with this method. That will promote structure design of freezing sealing of sodium gate valve for large fast rector power plant.

Experimental Investigation On the Retention of Soluble Particles by Pool Scrubbing

2021 ◽  
Author(s):  
René Vennemann ◽  
Michael Klauck ◽  
Hans-Josef Allelein
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Structural Integrity ◽  
Water Reservoir ◽  
Test Facility ◽  
Research Centre ◽  
Retention Capacity ◽  
Loss Of Coolant Accident ◽  
Measurement Tools ◽  
Activation Products

Abstract In the late stage of a severe loss-of-coolant accident, the pressure in the containment building of a nuclear power plant could rise beyond the design limits and thus endanger its structural integrity. Therefore, to avoid pressure failure, it may be necessary to perform con-trolled venting of the containment. During the event of an accident, a large amount of fission and activation products are released into the containment as airborne particles (aerosols). These particles are filtered during the venting process, usually with the help of wet filters, in order to keep risks to the surrounding environment to a minimum. Consequently, the knowledge of the retention processes in a water reservoir (pool scrubbing) is of central im-portance for such filtered containment venting systems (FCVS) and for reactor concepts in which water reservoirs are used for pressure reduction (e.g. condensation chamber of a BWR). Investigations on pool scrubbing are carried out in the SAAB test facility at the Juelich Research Centre. SAAB is a unique large scale facility with the ability to perform a great var-iation of experiments using various measurement tools. The influence of numerous parameters, such as the height of the water pool, solubility of aerosols and concentration on the retention capacity, is investigated by means of separate effect studies on both insoluble and soluble particles. This paper gives a detailed overview over the facility and includes example results of the first test series with soluble particles including cesium iodine (CsI).

Correction of Zero-Drift of Differential Pressure Flow Meters of ACME Facility

2014 ◽  
Author(s):  
Mingtao Cui ◽  
Tao Zhang
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Cooling System ◽  
Differential Pressure ◽  
Zero Drift ◽  
Test Facility ◽  
Passive Safety ◽  
Pressure Flow ◽  
Flow Meters ◽  
Core Cooling

ACME facility (Advanced Core-cooling Mechanism Experiment) is a large-scale test facility used to validate the performance of passive core-cooling system under SBLOCA (Small Break Lost of Coolant Accident) for the CAP1400, an upgraded passive safety nuclear power plant of AP1000. To simulate the features of passive safety system properly, DELTABAR, a kind of differential pressure flow meter, is designed to measure different mass flow of ACME. Because of the low pressure loss of DELTABAR, Zero-Drift problem of differential pressure flow meters in ACME is amplified, and some of the measured values are distorted seriously. To minimize the influence of Zero-Drift, analysis on zero-drift phenomenon is made, and a compensation method is proposed. The method is applying to PBL flow meters, and the result shows that the method is applicable.

Westinghouse Advanced Loop Tester (WALT) Update

2008 ◽  
Author(s):  
G. Wang ◽  
W. A. Byers ◽  
M. Y. Young ◽  
Z. E. Karoutas
Keyword(s):  
Nuclear Power ◽  
Hot Spot ◽  
Test Facility ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
System Pressure ◽  
Test Loop ◽  
Zinc Addition ◽  
Water Reactors ◽  
The Impact

In order to understand crud formation on the fuel rod cladding surfaces of pressurized water reactors (PWRs), a crud Thermal-Hydraulic test facility referred to as the Westinghouse Advanced Loop Tester (WALT) was built at the Westinghouse Science and Technology Department Laboratories in October 2005. Since then, a number of updates have been made and are described here. These updates include heater rod improvements, system pressure stabilization, and more effective protection systems. After these updates were made, the WALT system has been operated with higher stability and fewer failures. In this test loop, crud can be deposited on the heater rod surface and the character of the crud is similar to what has been observed in the PWRs. In addition, chemistry in the WALT loop can be varied to study its impact on crud morphology and associated parameters. The WALT loop has been successful in generating crud and measuring its thermal impact as a function of crud thickness. Currently, this test facility is supporting an Electric Power Research Institute (EPRI) program to assess the impact of zinc addition to PWR reactor coolant. Meanwhile, the WALT system is also being utilized by Westinghouse to perform dry-out and hot spot tests. These tests support the industry goal of 0 fuel failures by 2010 set by Institute of Nuclear Power Operations (INPO). Another major goal of the Westinghouse tests is to gain a better understanding of unexpected changes in core power distributions in operating reactors known as crud induced power shifts (CIPS) or axial offset anomalies (AOA).

Feasibility Study on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles for Reduced-Moderation Water Reactors

2004 ◽  
Author(s):  
Akira Ohnuki ◽  
Masatoshi Kureta ◽  
Wei Liu ◽  
Hidesada Tamai ◽  
Hajime Akimoto
Keyword(s):  
Large Scale ◽  
Hydraulic Performance ◽  
Test Facility ◽  
Current Status ◽  
Rod Bundles ◽  
Power Company ◽  
Water Reactors ◽  
Multiple Recycling ◽  
Operation Cycle ◽  
Tight Lattice

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute (JAERI) in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R&D plan and describe the current status focused on an experimental study using large-scale (37-rod bundle) test facility. Steady-state critical power experiments are conducted with the test facility and the experimental data reveal the feasibility of RMWR.

Construction of a High Temperature Grade 91 Sodium Component Test Loop

2010 ◽  
Vol 654-656 ◽  
pp. 416-419
Author(s):  
Hyeong Yeon Lee ◽  
Jae Han Lee ◽  
Tae Ho Lee ◽  
Jae Hyuk Eoh Lee ◽  
Tae Joon Kim ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Fast Reactor ◽  
Large Scale ◽  
Performance Test ◽  
Heat Removal ◽  
Test Facility ◽  
Decay Heat ◽  
Test Loop ◽  
Intermediate Heat Exchanger ◽  
Grade 91

A large scale sodium test facility of ‘CPTL’(Component Performance Test Loop) for simulating thermal hydraulic behavior of the Korean demonstration fast reactor components such as IHX(Intermediate Heat Exchanger), DHX(Decay Heat Removal Heat Exchanger) and sodium pump under development by KAERI is to be constructed. The design temperature of this test loop is 600°C and design pressure is 1MPa. The three heat exchangers are made of Grade 91 steel. Another sodium test facility of the ‘STEF’(Sodium Thermal-Hydraulic Experimental Facility) will be constructed next to the CPTL facility to simulate the passive decay heat removal behavior in the sodium cooled fast reactor. In this paper, the overall facility features of the CPTL and STEF are introduced and preliminary conceptual design of the facilities are carried out.

Space solar power may solve key renewables challenge

2021 ◽  
Keyword(s):  
Renewable Energy ◽  
Nuclear Power ◽  
Large Scale ◽  
Solar Power ◽  
Energy System ◽  
Content Type ◽  
Design Concepts ◽  
Space Solar Power ◽  
Launch Systems ◽  
National Infrastructure

Significance Using satellites to collect solar power in space and transmit it to earth to generate electricity overcomes a serious limitation of terrestrial renewable energy: its intermittency. As such, it could enable transition to a wholly renewable energy system. Impacts Design concepts vary, but generally envisage each satellite generating an amount of electricity on a par with a nuclear power plant. Placing critical national infrastructure in space would both hasten the militarisation of space and add urgency to arms control initiatives. Large-scale funding for SBSP would spur development of supporting technologies and spin-offs, including robotics and launch systems.

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