• Nuclear Engineering and Technology
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• 제53권1호
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• pp.111-120
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• 2021

During the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel and steel are solidified into debris particles, which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes the debris particles relocated and the bed flattened, which called debris bed relocation. Because the thickness of debris bed has great influence on the cooling ability of fuel debris in low plenum, it's very necessary to evaluate the transient changes of the shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a large number of debris bed relocation experiments were carried out by improved bottom gas-injection experimental method in this paper. The effects of different experimental factors on the relocation process were studied from the experiments. The experimental data were also used to further evaluate a semi-empirical onset model for predicting relocation.

• 대한기계학회논문집A
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• 제37권10호
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• pp.1251-1259
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• 2013

본 연구에서는 소듐냉각 고속로 붕괴열교환기(DHX)의 고온 설계 및 크리프-피로 손상 평가를 수행하였다. 제 4 세대 소듐냉각 고속로의 능동 및 피동 잔열제거계통에 설치되는 DHX와 한국원자력연구원의 STELLA-1 시험루프에 설치된 DHX에 대해 상세설계 및 3D 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준인 ASME Section III Subsection NH와 RCC-MR 코드를 따라 크리프-피로 손상평가를 수행하였다. 크리프-피로 손상평가 결과에 기초하여 두 설계기준에 대해 비교 분석하고, 설계 기술기준의 보수성 이슈에 대해 토의하였다.

• 한국기계기술학회지
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• 제13권2호
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• pp.145-153
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• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.

• 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.

• 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.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.283-291
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• 2024

Sodium cooled Fast Reactors (SFR) are built with several engineered safety features and hence a severe accident such as a core melt accident is hypothetical with a probability of <10^-6/ry. However, in case of such accidents, the mixture of the molten fuel and structural materials interacts with sodium. This phenomenon is known as Molten Fuel Coolant Interaction (MFCI) and results in fragmentation of the melt due to various instabilities. The fragmented particles settle as a debris bed on the core catcher at the bottom of the reactor vessel, and continue to generate decay heat. Characteristics of the debris particles play a vital role in heat transfer from the bed and need thorough investigation. The size, shape, and physical state of the debris depend on the associated fragmentation mechanism, superheating of the melt, and sodium temperature. Experiments have been conducted by releasing simulated corium, a molten mixture of alumina and iron generated by the aluminothermy process at ~2400 ℃ into liquid sodium, to study the fragmentation phenomena. After the experiment, the fragmented debris was retrieved and the particle size distribution was determined by sieve analysis. The debris was subjected to microscopic investigation for obtaining morphological characteristics. Based on the characteristics of debris, an attempt has been made to assess of fragmentation mechanism of simulated corium in sodium.

• Nuclear Engineering and Technology
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• 제56권9호
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• pp.3483-3490
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• 2024

The initial size distribution and morphological parameters of sodium aerosols are critical in evaluating the accidental suspended aerosol behaviour in Sodium-cooled Fast Reactor (SFR) containment. Mass-based measurements were more familiar in characterizing the sodium aerosols. Real-time number size distribution measurements are carried out in this study. The sensitivity analysis of sodium aerosol effective density (ρ_e) in deriving the actual number size distributions from the measured Aerodynamic Particle Size Distributions (APSD) and predicting suspended aerosol dynamics is presented. Tests are conducted in a 1 m³ chamber at 47 ± 3% RH for different initial mass concentrations (M₀) of 0.1, 1, and 2.9 g/m³. The initial APSDs measured just after the generation completions are observed to be polydisperse with the count median aerodynamic diameter (CMAD) < 1 ㎛. The literature reported ρ_e values of sodium aerosols, 2.27, 1.362, and 0.61 g/cm³ are used to derive mobility equivalent PSDs from APSD in each test. The real-time number concentration decay and size growth for four different PSDs are measured and compared with the estimate using nodal method-based code to ascertain the actual parameters. The validated parameters CMD = 0.66 ㎛, σ_g = 1.96, ρ_e = 1 g/cm³ and χ = 1 are used for improved estimation of sodium aerosol dynamics in Indian SFR containment with M₀ = 4 g/m³ for severe accident scenarios.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.115-125
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• 2016

The prototype SFR has currently been under design by KAERI. The size of its major components is much larger than that of APR1400 and high temperature materials are applied for it. The increased size of components and those specific materials effect on material procurement, manufacturing process and fabrication facilities. The manufacturing methods are studied for Reactor Vessel/Guard Vessel, Control Rod Drive Mechanism, Heat Exchanger, Primary Pump, Reactor Vessel Internals, Steam Generator and In-Vessel Transfer Machine. The proper manufacturing methods are suggested for each component including side forging technology for ultra large forgings of Reactor Vessel to minimize the weld seams on which In-service Inspection should be conducted.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2635-2649
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• 2022

CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect. The neutronic simulation of the start-up experiments conducted at the CEFR have been performed with a deterministic code system RAST-F, which is based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo (MC) code and a multi-group nodal diffusion solver. The RAST-F results were compared against the measurement data. Moreover, the characteristic of neutron spectrum in the fuel rings, and adjacent reflectors was evaluated using different models for generation of accurate nuclear libraries. The numerical solution of RAST-F system was verified against the full core MC solution MCS at all control rods fully inserted and withdrawn states. A good agreement between RAST-F and MCS solutions was observed with less than 120 pcm discrepancies and 1.2% root-mean-square error in terms of k_eff and power distribution, respectively. Meanwhile, the RAST-F result agreed well with the experimental values within two-sigma of experimental uncertainty. The good agreement of these results indicating that RAST-F can be used to neutronic steady-state simulations for small core-size SFR, which was challenged to deterministic code system.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.921-928
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• 2013

A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.