• 한국지진공학회논문집
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• 제11권3호
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• pp.73-86
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• 2007

본 논문에서는 제4세대 소듐냉각고속로(Sodium-Cooled Fast Reactor)의 후보 노형으로 선정된 KALIMER-600에 대한 단순 지진해석모델을 개발하고 시간이력 지진응답해석을 수행하여 수평 면진설계(Seismic Isolation) 기술이 적용된 원자로건물의 주요기기 및 구조물에서의 지진응답 성능을 분석하였다. 개발된 단순 지진해석모델은 원자로건물, 원자로시스템, 주요 기기, 중간 열전달계통 배관, 그리고 면진장치를 포함하며 각각은 상세 유한요소해석을 통한 동특성 비교검증을 통하여 정확성을 검증하였다. 안전정지기준 0.3g의 설계인공지진 하중에 대한 시간이력 지진응답해석을 수행하여 면진설계와 비면진 설계조건에 따른 원자로 주요 부위에서의 층응답스펙트럼을 비교분석한 결과 KALIMER-600의 면진성능이 우수한 것으로 나타났다.

• 한국압력기기공학회 논문집
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• 제12권1호
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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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• 제48권5호
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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.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1213-1224
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• 2024

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.

• 청정기술
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• 제25권2호
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• pp.161-167
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• 2019

바이오매스의 급속열분해를 위하여 지난 수십 년간 다양한 형태의 반응기가 개발되었다. 급속열분해 공정의 반응기는 유동층 반응기가 많이 사용되어 왔으며, 최근에는 분사층 반응기를 이용한 바이오매스의 급속열분해 특성에 대한 연구가 다수의 연구자들에 의해 수행되고 있다. 분사층 반응기의 유동화 특성은 입자의 물리적 특성, 유체 제트의 속도, core와 annulus의 구조에 영향을 받으며, 반응기의 기하학적 구조는 분사층 내부의 core와 annulus 구조를 결정하는 주요 인자이다. 따라서 분사층 반응기의 최적설계를 위해서는 열분해 반응에 영향을 주는 인자에 대한 바이오매스의 급속열분해 특성에 대한 연구가 수행되어야 한다. 하지만 분사층 반응기의 기하학적 구조에 의한 바이오매스의 급속열분해 특성은 자세히 연구되지 않았다. 본 연구에서는 분사층 반응기의 원뿔각과 반응 온도 변화에 따른 Jatropha curcas L. seed shell cake의 급속열분해 실험을 수행하여 분사층 반응기의 최적 형상과 반응 온도를 도출하였다. 실험결과, 열분해 오일의 에너지 수율은 반응 온도 $450^{\circ}C$, 분사층 반응기의 원뿔각 $44^{\circ}$에서 63.9%로 가장 높게 나타났다. 그리고 분사층 반응기 내 고체입자의 열전달과 기체상 열분해 생성물의 체류시간은 원뿔각의 영향을 받아 열분해 생성물의 수율 및 열분해 오일의 품질에 영향을 주는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제31권4호
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• pp.408-422
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• 1999

To reduce the fast neutron fluence which deteriorates the RPV integrity, additional shields were assumed to be installed at the outer core structures of the Kori Unit 1 reactor, and its reduction effects were examined. Full scope Monte Carlo simulation with MCNP4A code was made to estimate the fast neutron fluence at the RPV. An optimized design option was found from various choices in geometry and material for shield structure. It was expected that magnitude of fast neutron fluence would be reduced by 39% at the circumferential weld of the RPV, resulting in extension of plant lifetime by 4.6 EFPYs based on the criterion of PTS requirement It was investigated that the nuclear characteristics and thermal hydraulic factors at the internal core were only negligibly influenced by the installation of additional shield structure.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2120-2134
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• 2022

A hypothetical core destructive accident (HCDA) has received widespread attention as one of the most serious accidents in sodium-cooled fast reactors. This study combined recent advantages in numerical methods to realize realistic modeling of the complex fluid-structure interactions during HCDAs in a full-scale sodium-cooled fast reactor. The multi-material arbitrary Lagrangian-Eulerian method is used to describe the fluid-structure interactions inside the container. Both the structural deformations and plug rises occurring during HCDAs are evaluated. Two levels of expansion energy are considered with two different reactor models. The simulation results show that the container remains intact during an accident with small deformations. The plug on the top of the container rises to an acceptable level after the sealing between the it and its support is destroyed. The methodology established in this study provides a reliable approach for evaluating the safety feature of a container design.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2126-2132
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• 2021

Various research groups and private interprises are pursuing the design of a Molten Salt Reactor (MSR) as one of the Generation-IV concepts. In the current work a fast neutron MSR using chloride fuel is analyzed, specially analyzing the power production and neutron flux level in the Intermediate Heat Exchanger (IHX). The neutronic analysis in this work is based on a chloride-fuel MSR with 600 MW thermal power. The core power density was set to 100 MW m-3 with a core H/D [[EQUATION]] 1.0 amd four Intermediate Heat Exchanger (IHX). This leads to a power of 150 MW per IHX; this power is also comparable to the IHX proposed in the SAMOFAR framework. In this work, a preliminary design of a 150 MW helical-coil IHX for a chloride-fueled MSR is prepared and the fission rate, capture rate, and inelastic scatter rate are evaluated.

• Nuclear Engineering and Technology
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• 제32권6호
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• pp.566-594
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• 2000

In most LMFBR(Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER(Korea Advanced Liquid MEtal Reactor) reactor internal strictures is carried out for normal operating conditions which have the operating temperature 53$0^{\circ}C$ and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME Code Case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects.

• 한국전산유체공학회지
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• 제21권3호
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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.