• Journal of the Korean Society for Nondestructive Testing
• /
• v.24 no.2
• /
• pp.171-179
• /
• 2004

The design concepts of in-service inspection and maintenance are very important for the reactor system design of the nuclear power plant. The strategy of in-service inspection and maintenance should be reflected in the mechanical system design for the verification of the operability of liquid metal reactor KALIMER. In this paper the fundamental approaches of the in-service inspection and maintenance of the KALIMER are established to ensure the safety and reliability of the reactor system. The general method and requirement of the in-service inspection and maintenance for the reactor system and components are proposed and described to satisfy the intents of the ASME Section XI Division 3 and the design characteristics of KALIMER.

• Journal of Energy Engineering
• /
• v.11 no.2
• /
• pp.122-129
• /
• 2002

The analytical models and algorithm of the SPIKE code, which has been developed by KAERI's KALIMER team to investigate the sodium-water reaction phenomena in the liquid metal reactor, were introduced with its verification calculation results. The sodium water reaction of KALIMER IHTS was evaluated. Early stage of the sodium-water reaction consists of wave and mass transfer regimes. The pressure variations were independent of specific design features in the wave transfer regime. However in the mass transfer regime, the pressure variations were strongly dependent on cover gas volume and rupture disk set pressure. The early stage SWR analysis showed that the KALIMER IHTS with an appropriate cover gas volume and rupture disk set pressure had enough margin to its design pressure.

• Proceedings of the Earthquake Engineering Society of Korea Conference
• /
• 1999.10a
• /
• pp.220-227
• /
• 1999

The Purpose of this paper is to evaluate the buckling strength of conceptually designed KALIMER reactor vessel. For evaluation of the buckling load buckling load the design equations and the finite element analysis are used. In finite element method the eigenvalue buckling analysis nonlinear elastic buckling analysis using snap-through buckling method and nonlinear elastic-plastic buckling analysis are carried out. the calculated buckling loads of KALIMER reactor vessel using the finite element method are in well agreement with those of the design equations. From the calculated results of buckling load in KALIMER rector vessel it is shown that the plasticity of vessel materials significantly affects the buckling load but the initial imperfection has little effects, In checking the limits of bucking load of KALIMER reactor vessel using the ASME B & PV Section III. Subsection NH the non-seismic isolation design can not satisfy the buckling limit requirements but the seismic isolation design can sufficiently satisfy the requirements.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.11 no.3 s.55
• /
• pp.73-86
• /
• 2007

In this paper, a simple seismic analysis model of the KALIMER-600 sodium-cooled fast reactor selected to be the candidate of the GEN-IV reactor is developed. By using this model, the seismic time history analysis is carried out to investigate the feasibilities of a seismic isolation design. The developed simple seismic analysis model includes the reactor building, reactor system,, IHTS piping system, steam generator, and seismic isolators. The dynamic characteristics of the simple seismic model are verified with the detailed 3-dimensional finite element analysis for each part of the KALIMER-600 system. By using the developed simple seismic model, the seismic time history analyses for both cases of a seismic isolation and non-isolation design are performed for the artificial time history of a SSE (Safe Shutdown Earthquake) 0.3g. From the comparison of the calculated floor response spectrum, it is verified that the seismically isolated KALIMER-600 reactor building shows a great performance of a seismic isolation and assures a seismic integrity.

• Journal of computational fluids engineering
• /
• v.7 no.3
• /
• pp.44-52
• /
• 2002

An analytic methodology was developed for free surface motions between liquid metal coolant and cover gas in order to calculate the phenomena of gas entrainment in hot pool surface through IHX EMP and reactor core. The methodology was setup by applying the first order VOF convection model to CFX4 general purpose fluid dynamics analysis code. The methodology was validated by applying it to an experimental apparatus designed for free surface motions of KALIMER reactor. The distributions of free surface calculated by the present methodology were almost coincident with the experimental data. The developed methodology was applied to the KALIMER reactor of full power operating condition. The shapes of the free surface were nearly uniform. From the results, it was found that the altitude of the free surface from the IHX inlet nozzle of KALIMER reactor is high enough not to affect to free surface motions of generating gas bubbles from the turbulent shear flows such as hydraulic jump and water falls.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05b
• /
• pp.592-597
• /
• 1996

The KALIMER core, of which nuclear design is largely governed by inherent safety and reactivity control issues, is fueled with metallic fuel, and the initial core will be loaded with 20% enriched Uranium metal fuel. KALIMER safety design objectives include the accommodation of unprotected, ATWS events without operator action, and without the support of active shutdown, shutdown heat removal, or any automatic system without damage to the plant and without jeopardizing public safety. The transient analysis of the core designs has been focused on severe events to assess the margins in the design, and ATWS events are the most severe events that must be accommodated by the KALIMER design. The ATWS performance has been evaluated for the preliminary initial core design of KALIMER with a particular emphasis on the inherent negative reactivity feedback effects, including the Doppler, sodium density, fuel axial expansion, core radial expansion, and control rod driveline expansion. Results show that the Uranium metal core design meets the temperature limits with margin.

• Proceedings of the Korean Operations and Management Science Society Conference
• /
• 2002.05a
• /
• pp.561-568
• /
• 2002

한국원자력연구소에서 개발중인 액체금속로인 KALIMER는 경수로나 증수로와 근본적으로 설계가 상이하므로 PSA 방법에 대한 새로운 접근방법을 개발해야 한다. 액체금속로 KALIMER에 대한 확률론적 안전성 평가 방법 (PSA, Probabilistic Safety Assessment) 관련 연구는 초기 사건의 도출 및 빈도계산 방법과 주요 계통의 신뢰성 예비 평가에 대한 것이다. 초기 사건이란 원전에 과도 현상을 유발하여 발전소 정지를 초래하는 모든 비정상 사건을 의미하는 것으로 PSA에서 사건 수목을 구성하는 데 기본이 되는 정보이다. 액체금속로는 기존의 경수로 및 중수로와는 전혀 다른 설계를 갖고 있으므로 액체금속로 특유의 초기 사건을 도출하는 방법 및 이들 초기 사건의 빈도를 계산하는 방법에 대한 연구를 수행하였다. KALIMER 주요 계통의 신뢰성 예비 평가를 수행하기 위하여 확률론적 안전성 평가에서 계통분석기법으로 널리 이용되는 고장수목분석의 절차와 방법에 대한 방법론을 선정하여 PSA 방법론을 개발하였다.

• Journal of Energy Engineering
• /
• v.13 no.2
• /
• pp.101-111
• /
• 2004

The main purpose of a liquid metal reactor core thermal-hydraulic design is to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power distribution in the core. The thermal-hydraulic design procedure consists of the coolant flow distribution to the sub-assemblies, the coolant/fuel temperature calculations and detailed subchannel analysis. This paper describes the LMR core thermal-hydraulic design methodology and summarizes the major design and analysis results of KALIMER breeder and breakeven cores and subassemblies. KALIMER is a 150 MWe rated (392 MWth) heterogeneous core with U-TRU-Zr ternary alloy fuel and sodium coolant.

• Nuclear Engineering and Technology
• /
• v.41 no.6
• /
• pp.785-796
• /
• 2009

A KALIMER-600 concept which is a type of sodium-cooled fast reactor, has been developed at KAERI. It uses sodium as a primary coolant and is a pool-type reactor to enhance safety. Also, a supercritical carbon dioxide ($CO_2$) Brayton cycle is considered as an alternative to an energy conversion system to eliminate the sodium water reaction and to improve efficiency. In this study, a simplified model for analyzing the thermodynamic performance of the KALIMER-600 coupled with a supercritical $CO_2$ Brayton cycle was developed. To develop the analysis model, a commercial modular modeling system (MMS) was adopted as a base engine, which was developed by nHance Technology in USA. It has a convenient graphical user interface and many component modules to model the plant. A new user library for thermodynamic properties of sodium and supercritical $CO_2$ was developed and attached to the MMS. In addition, some component modules in the MMS were modified to be appropriate for analysis of the KALIMER-600 coupled with the supercritical $CO_2$ cycle. Then, a simplified performance analysis code was developed by modeling the KALIMER-600 plant with the modified MMS. After evaluating the developed code with each component data and a steady state of the plant, a simple power reduction and recovery event was evaluated. The results showed an achievable capability for a performance analysis code. The developed code will be used to develop the operational strategy and some control logics for the operation of the KALIMER-600 with a supercritical $CO_2$ Brayton cycle after further studies of analyzing various operational events.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2001.05a
• /
• pp.183.2-183
• /
• 2001