• Nuclear Engineering and Technology
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• v.17 no.1
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• pp.16-24
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• 1985

A computer model is developed capable of simulating the transient behavior of a pool-type liquid metal-cooled fast breeder reactor (LMFBR). The model, SIMFARP, is a fast running computer code which may be used to simulate the loss of power to any pump(s), a complete loss-of-forced cooling, and the natural circulation behavior. Eight governing equations are derived and a Runge-Kutta algorithm is applied to integrate the eight differential equations. The developed computer program is applied to two cases; loss of electric power to any pump(s), and loss of all external electric supply power without scram in Super-Phenix-I.

• Transactions of the Korean Society of Mechanical Engineers
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• v.13 no.3
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• pp.490-499
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• 1989

These days the closed cycle gas turbine attracts considerable attention due to : (1) The possibility of directly coupling the closed cycle gas turbine with a high temperature gas cooled reactor ; (2) the economical use of dry coolers to reduce the thermal charge of the environment ; and (3) the reduction of pollution and energy consumption, by replacing the domestic hearth by a central heating and power station. In this paper, we selected the optimal cycle from the characteristic of thermodynamic cycle for the optimal design of closed CO$_{2}$ gas turbine cycle usuable in nuclear energy power plant. Also the effects of between the parameters and thermal efficiency were investigated by computer simulation. These results and design data will be added to basics in optimal designing closed CO$_{2}$ cycle gas turbine plant.

• Journal of the Korean Society of Visualization
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• v.9 no.4
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• pp.63-68
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• 2011

The flow-field of a liquid-metal system is very important for the safety analysis and the design of the steam generator of liquid-metal fast breeder reactor. Dynamic neutron radiography (DNR) is suitable for a visualization and measurement of a liquid metal flow and a two-phase flow in a metallic duct. However, the three dimensional DNR techniques is not enough to obtain the velocity information in the wide channel up to now. In this research, a high speed DNR technique was applied to visualize the heavy liquid-metal flow field in the narrow channel with the HANARO-beam facility. The images were taken with a high frame-rate neutron radiography at 250 fps and analyzed with a Particle Image Velocimetry(PIV) method. The images were compared with the results of the commercial CFX code to study the feasibility of DNR technique for the measuring the heavy liquid-metal flow field. The PIV images could discern the turbulent vortex flow in the two-dimensional narrow channel.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.264-268
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• 2008

An assessment of creep-fatigue crack behavior is required to ensure the structural integrity for high temperature components such as fast breeder reactor structures or thermal power plant components operating at an elevated temperature. In this study, an evaluation of creep-fatigue crack growth has been carried out according to the French assessment guide of the RCC-MR A16 for austenitic stainless steel structures. The assessment procedures for creep-fatigue crack growth in the recent version of the A16 (2007 edition) have been changed considerably from the previous version (2002 edition) and the material properties (RCC-MR Appendix A3) have been changed as well. The impacts of those changes on creep-fatigue crack growth behavior are quantified from the assessments with a structural model. Finally the assessment results were compared with the observed images obtained from the structural tests of the same structural specimen.

• Nuclear Engineering and Technology
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• v.39 no.6
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• pp.683-690
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• 2007

Because of increasing concerns regarding global warming and the longevity of oil and gas reserves, the importance of nuclear energy as a major source of sustainable energy is gaining recognition worldwide. To make nuclear energy truly sustainable, it is necessary to ensure not only the sustainability of the fuel supply but also the sustained availability of waste repositories, especially those for high-level radioactive waste (HLW). From this perspective, the effort to maximize the waste loading density in a given repository is important for easing repository capacity problems. In most cases, the loading of a repository is controlled by the decay heat of the emplaced waste. In this paper, a comparison of the decay heat characteristics of HLW is made among the various fuel cycle options. It is suggested that, for a future fast breeder reactor (FBR) cycle, the removal and burning of minor actinides (MA) would significantly reduce the heat load in waste and would allow for a reduction of repository size by half.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.19-24
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• 2004

A number of creep data were collected and filed for type 316LN stainless steels through literature survey and experimental data produced in KAERI. Using these data, polynomial equations for predicting creep life were obtained for Larson Miller (L-M), Qrr-Sherby-Dorn (O-S-D) and Manson-Haferd (M-H) parametric methods. In order to find out the suitability for them, the relative standard error (RSE) and standard error of estimate (SEE) values were obtained by statistical process of creep data. The O-S-D parameter showed better fitting to creep-rupture data than the L-M or the M-H parameters, and the three parametric methods did not generate the large difference in the SEE and the RSE values.

• 전기의세계
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• v.42 no.7
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• pp.3-8
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• 1993

무한 동력을 얻고자 하는 생각은 "기술"이나 "과학"이라는 단어가 사용된 이래 수많은 두뇌들에 의해 얼핏 한번은 떠올려진 이상이었을 것이다. 지금과 같이 자원의 고갈과 위기를 쉴새없이 부르짖는 시기에 "무한동력"이란 존재는 인류의 모든 어려움을 해결할만한 방법이고, 특히 과소비성 현대의 생활에는 절대적인 수단일 것이다. 무한 동력과는 비교조차도 할 수 없는 한참아래쪽에서 서성거리고 있는 에너지 발생원인 액체금속 냉각고속증식로의 존재에 대하여, 언제, 어떻게, 접근해야 할까\ulcorner 하는 고민을 우리는 계속해야 할 필요가 있겠는가의 관한 의문에 대한 대답은 항상 긍정적이어야 하겠다. 이제는 국가적 규모의 연구개발사업이 요구되며 자원빈국이면서 공업선진국을 지향하고 있는 우리나라의 현 상황으로는 아직 실용화되고 있지 않은 LMR 개발기술에 접근할 수 있는 시간적 여지가 남아있다. 따라서 우리특유의 기술소화내지는 토착화의 의지를 결집하여 독작적인 개발계획을 수립하여 적극추진한다면 선두주자들과 어깨를 나란히 하여 우리도 꿈의 원자로인 LMR의 혜택을 누릴 수 있을 것으로 기대한다.자로인 LMR의 혜택을 누릴 수 있을 것으로 기대한다.

• Proceedings of the KIEE Conference
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• 2003.07b
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• pp.906-908
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• 2003

A Linear induction electromagnetic(EM) pump of liquid metal fast breeder reactor(LMFBR) is used for the purpose that the liquid metal of high temperature is transported by EM force. This paper evaluates magnetic flux density necessary for transporting liquid metal, using analytical model of the linear induction EM pump. Using the 2-D finite element method(2-D FEM), magnetic flux density is estimated in consideration of a geometric model, electric parameter, and velocity of liquid metal. From the viewpoint of hydrodynamics, the results can be used for flow analysis of the liquid metal.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1723-1729
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• 2021

Energy and directional distribution of neutrons at the Beloyarsk NPP workplaces is a subject of this study. Measurements of H^*(10) rate and neutron energy distribution were taken at 8 workplaces, which can be divided into three categories: work with spent or fresh nuclear fuel, work with radionuclide neutron sources, work at the rooms adjusted to reactors. The Hp(10) measurements were performed only at 6 out of 8 locations, due to the fact that long term placing of an effective neutron moderator in fresh nuclear fuel storage facility is forbidden. As a result of the research energy and direction distribution of the neutron fields at 8 locations of the Beloyarsk NPP workplaces was obtained. To estimate the accuracy of the H^*(10) rate and Hp (10) measurements the reference values of dose equivalents were calculated using energy and directional distribution. To take into account the difference between the reference values and the measured results site-specific correction factors were calculated.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.267-275
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• 1999

This paper deals with the industrial problem of thermal striping damage on the French prototype fast breeder reactor, Phenix and it was studied in coordination with the research program of IAEA. The thermomechanical and fracture mechanics evaluation procedure of thermal striping damage on the tee-junction of the secondary piping using Green's function method and standard FEM is presented. The thermohydraulic(T/H) loading condition used in the present analysis is the random type thermal loads computed by T/H analysis on the turbulent mixing of the two flows with different temperatures. The thermomechanical fatigue damage was evaluated according to ASME code section 111 subsection NH. The results of the fatigue analysis showed that fatigue failure would occur at the welded joint within 90,000 hours of operation. The assessment for the fracture behavior of the welded joint showed that the crack would be initiated at an early stage in the operation. It took 42,698.9 hours for the crack to propagate up to 5 mm along the thickness direction. After then, however, the instability analysis, using tearing modulus, showed that the crack would be arrested, which was in agreement with the actual observation of the crack. An efficient analysis procedure using Green's function approach for the crack propagation problem under random type load was proposed in this study. The analysis results showed good agreement with those of the practical observations.