• Nuclear Engineering and Technology
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• 제41권6호
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• pp.785-796
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• 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.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• 대한기계학회논문집A
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• 제37권5호
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• pp.665-671
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• 2013

제 4 세대 소듐냉각 고속로에는 중간열교환기(IHX), 붕괴열제거 열교환기(DHX), 공기 열교환기(AHX), 핀형 소듐-공기 열교환기(FHX) 및 증기발생기(SG)를 포함한 다양한 열교환기들이 설치된다. 본 연구에서는 STELLA-1 시험루프에 설치된 소듐-공기 열교환기인 AHX 와 SELFA 시험루프에 설치될 핀형(finned) 소듐-공기 열교환기인 FHX 등 2 기의 열교환기 설계에 대해 3D 상세 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준을 따라 크리프-피로 손상평가를 수행하였다. 손상 평가결과 AHX와 FHX는 의도하는 크리프 피로 손상 하중 하에서 구조 건전성을 유지하는 것으로 확인되었다.

• 대한기계학회논문집B
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• 제34권11호
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• pp.991-998
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• 2010

소듐냉각 고속로 내부기기 배치 변경에 의한 초기냉각 성능변화를 검토하기 위하여 중간열교환기의 수직배치가 다른 3개의 원자로를 대상으로 COMMIX-1AR/P 코드를 활용한 다차원 해석을 수행하였다. 원통좌표계의 중심축을 기준으로 원주방향의 1/4 부분만을 모델링하고 정상상태 및 과도상태 분석을 수행하여 IHX 수직배치 변화가 초기 냉각 특성에 미치는 영향을 분석하였고, DHX를 통한 후기 냉각 모드 개시 시점에 미치는 영향도 분석하였다. 분석 결과 IHX 수직배치 상승은 원자로 풀내부 자연 순환 유량을 증가시켜 초기 냉각과정에서 노심 최고 온도의 급격한 상승을 방지할 수 있으며, 초기냉각 성능을 향상시키기 위한 관성회전차의 가용설계재원의 범위도 확대시킨다. 또한 IHX 수직배치 상승은 후기냉각모드에 큰 영향을 주지 않으면서 초기냉각성능의 향상에 기여할 수 있을 것으로 사료된다.

• 대한기계학회논문집A
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• 제38권8호
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• pp.831-836
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• 2014

본 연구에서는 한국원자력연구원 내에 설치될 예정인 소듐시험 시설인 SELFA(Sodium Thermal-hydraulic Experiment Loop for Finned-tube Sodium-to-Air heat exchanger) 내에서 정상상태 가동온도가 $510^{\circ}C$의 고온 압력용기인 팽창탱크에 대해 고온 건전성 평가를 수행하였다. 팽창탱크에 대해 3 차원 유한요소해석에 기초하여 고온설계 기술기준인 ASME Section III Subsection NH 와 프랑스의 RCC-MRx 코드를 따라 크리프-피로 손상평가를 수행하였다. 평가결과 팽창탱크는 크리프-피로 설계 과도 하중 하에서 구조적 건전성을 유지하는 것으로 나타났다. 316L 스테인리스강 재질의 동 압력용기에 대해 정량적 코드 비교 분석을 수행하였다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2009년도 학술논문요약집
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• pp.51-52
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• 2009

Recently, several industrial countries using the fission energy have given attention to the Gen-IV SFR (sodium-cooled fast reactor) for achieving sustainable nuclear energy systems. In this context, an SFR is currently developed at the design concepts study stage in the Republic of Korea [Kim & Hahn 200909]. The sustainability of systems means economic, environment-friendly, proliferation-resistant, and safer systems. More specifically, this sustainability can be accomplished in terms of resource recycling and radioactive waste reduction. In the present work, the objective of fuel cycle cost modeling is to identify the impact of various conceptual options as a cost reduction measure for the Gen-IV SFR at the design concepts study stage. It facilitates the selection of several reasonable fuel cycle pathways for the future Gen-IV SFR from an economic viewpoint.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3207-3216
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• 2021

Three-dimensional structures of a vortical flow field and heat transfer characteristics in a partially blocked 7-pin fuel assembly mock-up of sodium-cooled fast reactor have been investigated through a numerical analysis using a commercial computational fluid dynamics code, ANSYS CFX. The simulation with the SST turbulence model agrees well with the experimental data of outlet and cladding wall temperatures. From the analysis on the limiting streamline at the wall, multi-scale vortexes developed in axial direction were found around the blockage. The vortex core has a high cladding wall temperature, and the attachment line has a low cladding wall temperature. The small-scale vortex structures significantly enhance the convective heat transfer because it increases the turbulent mixing and the turbulence kinetic energy. The large-scale vortex structures supply thermal energy near the heated cladding wall surface. It is expected that control of the vortex structures in the fuel assembly plays a significant role in the convective heat transfer enhancement. Furthermore, the blockage plate and grid spacer increase the pressure drop to about 36% compared to the bare case.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.932-941
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• 2021

This paper suggests a fracture simulation method for SFR metallic fuel pin under accident condition. Two major failure mechanisms - creep damage and eutectic penetration - are implemented in the suggested method. To simulate damaged element, stress-reduction concept to reduce stiffness of the damaged element is applied. Using the proposed method, the failure size of cladding can be predicted in addition to the failure time and failure site. To verify the suggested method, Whole-pin furnace (WPF) test and TREAT-M test conducted at Argonne National Laboratory (ANL) are simulated. In all cases, predicted results and experimental results are overall in good agreement. Based on the simulation result, the effect of eutectic-penetration depth representing failure behavior on failure size is studied.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.772-784
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• 2024

The hexagonal nodal code RENUS has been enhanced to handle irregularly deformed hexagonal assemblies. The underlying RENUS methods involving triangle-based polynomial expansion nodal (T-PEN) and corner point balance (CPB) were extended in a way to use line and surface integrals of polynomials in a deformed hexagonal geometry. The nodal calculation is accelerated by the coarse mesh finite difference (CMFD) formulation extended to unstructured geometry. The accuracy of the unstructured nodal solution was evaluated for a group of 2D SFR core problems in which the assembly corner points are arbitrarily displaced. The RENUS results for the change in nuclear characteristics resulting from fuel deformation were compared with those of the reference McCARD Monte Carlo code. It turned out that the two solutions agree within 18 pcm in reactivity change and 0.46% in assembly power distribution change. These results demonstrate that the proposed unstructured nodal method can accurately model heterogeneous thermal expansion in hexagonal fueled cores.

• Korean Chemical Engineering Research
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• 제54권6호
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• pp.863-870
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• 2016

본 논문의 목적은 소듐냉각고속로(sodium cooled fast reactor, SFR)와 초임계 $CO_2$ Brayton cycle의 연계 시, 원자로 열수송 계통과 동력변환 계통의 압력 경계를 형성하는 회로인쇄형 열교환기의 경계면에 균열이 발생해 고압(약 200 bar)의 $CO_2$가 상압 수준의 액체소듐유로 측에 유입되었을 때의 물리/화학적 현상을 파악하여 열교환기 설계에 활용 가능한 실험 자료를 생산하는 것이다. 열교환기의 소듐-$CO_2$ 경계면 균열 현상은 경계면의 균열 크기에 따라 미세 균열에 의한 소듐유로막힘(plugging) 현상과 상대적으로 큰 균열에 의한 열교환기 재료손상(wastage) 현상으로 나뉜다. Plugging 실험결과, 소듐유로 직경이 3mm일 때 $CO_2$ 주입 즉시 소듐 흐름이 정지한 반면 소듐유로 직경이 5 mm일 때는 유량이 감소되기 시작하는 시점은 3 mm의 경우와 유사하게 $CO_2$ 주입 즉시 나타났지만 소듐의 흐름이 완전히 정지할 때까지는 상대적으로 오랜 시간이 소요되었다. 이러한 실험결과는 실제 열교환기의 소듐-$CO_2$ 경계면에서 미세균열이 발생했을 때, 소듐유로 직경이 3 mm로 좁을 경우 균열 발생과 동시에 해당 소듐유로가 반응생성물에 의해 막혀 해당 유로 외의 유로들로 지속적인 열교환기 운전이 가능하지만, 소듐유로의 직경이 5 mm로 넓어질 경우 소듐유로가 고체생성물에 의해 즉시 막히지 않고 생성물이 소듐유로를 따라 계통 내부를 이동하다 일정 농도 이상이 되어야 소듐유로를 막게 할 것으로 예상할 수 있는 결과이다. Wastage 실험결과, 열교환의 재질(STS316, Inconel600, G91 합금강), 운전온도($400{\sim}500^{\circ}C$), 노즐직경(0.2~0.8 mm), 시편-노즐 거리(2~6 mm)와 무관하게 고압(약 200~250 bar)의 $CO_2$ 분사에 의한 시편의 물리적 손상(erosion) 현상은 발생하지 않았다. 노즐에서의 분사되는 $CO_2$의 분사속도는 마하 0.4~0.7인 것으로 확인되었다. 본 연구의 실험결과는 열교환기 파손 대처 설계에 배경 실험 자료로 활용될 것으로 기대된다.