• Nuclear Engineering and Technology
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• 제48권4호
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• pp.952-964
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• 2016

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium-water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium-water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

• Environmental Engineering Research
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• 제13권4호
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• pp.177-183
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• 2008

Ex-situ reductive dechlorination of carbon tetrachloride (CT) by iron sulfide in a batch reactor was characterized in this study. Reactor scaled-up by 3.5 L was used to investigate the effect of reductant concentration on removal efficiency and process optimization for ex-situ degradation. The experiment was conducted by using both liquid-phase and gas-phase volume at pH 8.5 in anaerobic condition. For 1 mM of initial CT concentration, the removal of the target compound was 98.9% at 6.0 g/L iron sulfide. Process optimization for ex-situ treatment was performed by checking the effect of transition metal and mixing time on synthesizing iron sulfide solution, and by determining of the regeneration time. The effect of Co(II) as transition metal was shown that the reaction rate was slightly improved but the improvement was not that outstanding. The result of determination on the regeneration time indicated that regenerating reductant capacity after $1^{st}$ treatment of target compound was needed. Due to the high removal rates of CT, ex-situ reductive dechlorination in batch reactor can be used for basic treatment for the chlorinated compounds.

• Nuclear Engineering and Technology
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• 제33권6호
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• pp.566-584
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• 2001

In this paper, the progressive deformation and the creep-fatigue damage for the conceptually designed reactor internals of KALIMER(Korea Advanced Liquid MEtal Reactor) are carried out by using the elastic analysis method in the RCC-MR code for normal operating conditions including the thermal load, seismic load (OBE) and dead weight. The maximum operating temperature of this reactor is 53$0^{\circ}C$ and the total service lifetime is 30 years. Thus, the time- dependent creep and stress-rupture effects become quite important in the structural design. The effects of the thermal induced membrane stress on the creep-fatigue damage are investigated with the risk of the elastic follow-up. To calculate the thermal stress, detailed thermal analyses considering conduction, convection and radiation heat transfer mechanisms are carried out with the ANSYS program. Using the results of the elastic analysis, the progressive deformation and creep-fatigue damages are calculated step by step using the RCC-MR in detail. This paper ill be a very useful guide for an actual application of the high temperature structural design of the nuclear power plant accounting for the time-dependent creep and stress-rupture effects.

• 대한기계학회논문집A
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• 제21권7호
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• pp.1042-1049
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• 1997

LMFBR(Liquid Metal Fast Breeder Reactor) vessel is operated under the high temperatures of 500-550.deg. C. Thus, transient thermal loads were severe enough to cause inelastic deformation due to creep-fatigue and plasticity. For reduction of such inelastic deformations, Y-piece structure in the form of a thermal sleeve is used in LMFBR vessel under repeated start-up, service and shut-down conditions. Therefore, a systematic method for inelastic analysis is needed for design of the Y-piece structure subjected to such loading conditions. In the present investigation, finite element analysis of heat transfer and inelastic thermal stress were carried out for the Y-piece structure in LMFBR vessel under service conditions. For such analysis, ABAQUS program was employed based on the elasto-plastic and Chaboche viscoplastic constitutive equations. Based on numerical data obtained from the analysis, creep-fatigue damage estimation according to ASME Code Case N-47 was made and compared to each other. Finally, it was found out that the numerical predictio of damage level due to creep based on Chaboche unified viscoplastic constitutive equation was relatively better compared to elasto-plastic constitutive formulation.

• Nuclear Engineering and Technology
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• 제38권3호
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• pp.293-300
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• 2006

A liquid metal reactor (LMR) operated at high temperatures is subjected to both cyclic mechanical loading and thermal loading; thus, creep-fatigue is a major concern to be addressed with regard to maintaining structural integrity. The Korea Advanced Liquid Metal Reactor (KALIMER), which has a normal operating temperature of $545^{\circ}C$ and a total service life time of 60 years, is composed of various cylindrical structures, such as the reactor vessel and the reactor baffle. This study focuses on the creepfatigue crack initiation for a cylindrical Y-junction structure made of 316 stainless steel (SS), which is subjected to cyclic axial tensile loading and thermal loading at a high-temperature hold time of $545^{\circ}C$. The evaluation of the considered creep-fatigue crack initiation was carried out utilizing the ${\sigma}_d$ approach of the RCC-MR A16 guide, which is the high-temperature defect assessment procedure. This procedure is based on the total accumulated strain during the service time. To confirm the evaluated result, a high-temperature creep-fatigue structural test was performed. The test model had a circumferential through wall defect at the center of the model. The defect front of the test model was investigated after the $100^{th}$ cycle of the testing by utilizing a metallurgical inspection technique with an optical microscope, after which the test result was compared with the evaluation result. This study shows how creep-fatigue crack initiation for a high-temperature structure can be predicted with conservatism per the RCC-MR A16 guide.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.162.2-162
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• 2001

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• Nuclear Engineering and Technology
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• 제35권6호
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• pp.537-546
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• 2003

The purpose of this paper is to investigate the buckling characteristics of a conceptually designed KALIMER-150(Korea Advanced LIquid MEtal Reactor, 150MWe) reactor vessel and verify the buckling behavior using the reduced scale cylindrical shell structures. To do this, nonlinear buckling analyses using finite element method and evaluation formulae are carried out. From the results, the KALIMER-150 reactor vessel exhibits a dominant bending buckling mode and is significantly affected by the plastic behavior. The interaction effects with the vertical seismic load cause the lateral buckling load to be slightly decrease. From the results of the buckling experiments using reduced scaled cylindrical shell structures, it is verified that the buckling modes such as pure bending, pure shear, and mixed(bending plus shear) mode clearly appear under a lateral load corresponding to the slenderness ratio of cylinder.

• Nuclear Engineering and Technology
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• 제33권2호
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• pp.209-224
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• 2001

The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram.

• Nuclear Engineering and Technology
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• 제33권4호
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• pp.423-440
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• 2001

For sodium cooling liquid metal reactors during the last 25 years, it was most important to verify the safety of the steam generator, which absolutely requires a water leak detection system with fine sensitivity and response. This study describes the structure and leak classification of the HAMMER (Korea Advanced Liquid Metal Reactor) steam generator, compared with other classifications, and explains the effects of leak development. The requirements and experimental situations for the development of the KALIMER acoustic leak detection system (KADS) which detects micro leaks, not intermediate leaks, are introduced. We proposed four frequency bands, 1∼8kHz, 8∼20kHz, 20∼40kHz and 40∼200kHz, split effectively for analyzing the detected acoustic leak signals obtained from the sodium-water reaction model or water model in the mock-up system.