• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1171-1180
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• 2009

The internal core baffle structure of a French Pressurized Water Reactor (PWR) consists of a collection of baffles and formers that are attached to the barrel. The connections are done thanks to a large number of bolts (about 1500). After inspection, some of the bolts have been found cracked. This has been attributed to the Irradiation Assisted Stress Corrosion Cracking (IASCC). The $Electricit\acute{e}$ De France (EDF) has set up a research program to gain better knowledge of the temperature distribution, which may affect the bolts and the whole structure. The temperature distribution in the structure was calculated thanks to the thermal code SYRTHES that used a finite element approach. The heat transfer between the by-pass flow inside the cavities of the core baffle and the structure was accounted for thanks to a strong thermal coupling between the thermal code SYRTHES and the CFD code named Code_Saturne. The results for the CP0 plant design show that both the high temperature and strong temperature gradients could potentially induce mechanical stresses. The CPY design, where each bolt is individually cooled, had led to a reduction of temperatures inside the structures. A new parallel version of SYRTHES, for calculations on very large meshes and based on MPI, has been developed. A demonstration test on the complete structure that has led to about 1.1 billion linear tetraedra has been calculated on 2048 processors of the EDF Blue Gene computer.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1995년도 가을 학술발표회 논문집
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• pp.43-49
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• 1995

Sasol Advanced Synthol Reactor was divided into two chambers by grid plate perforated with diffuser holes. The reactor has high stress level beacuse of membrane stress due to internal pressure, thermal stress due to temperature difference and local stress due to structural discontinuity at the juncture of grid plate and shell. Moreover, geometric nonlinear behaviors may appear in the grid plate because of pressure difference between two chambers. In order to survey the stress level and geometric nonlinear behaviors around grid plate, heat transfer analysis, linear static analysis and geometric nonlinear analysis were performed using NISA II developed by EMRC. This paper demonstrates the result of accessment for linear static and geometric nonlinear analysis under various load combinations.

• Tribology and Lubricants
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• 제40권3호
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• pp.97-102
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• 2024

The next-generation Molten Salt Reactor is known for its high safety because it uses nuclear fuel dissolved in high-temperature molten salt, unlike traditional solid atomic fuel methods. However, the high-temperature molten salt causes severe corrosion in internal structural materials, threatening the reactor's safety. Therefore, it is crucial to investigate the high-temperature corrosion resistance and wear performance of materials used in reactors to ensure safety. In this study, the high-temperature corrosion resistances and wear performances of corrosion samples in a NaCl-MgCl2-KCl (20-40-40 [wt%]) molten salt are investigated to evaluate the applicability of economically viable stainless steels, 316SS and 304SS. Hastelloy C276 and a new alloy containing a small amount of Nb are used as reference samples for comparative analysis. The mass loss, mass loss rate per unit volume, and surface roughness of each sample are measured to understand the corrosion mechanisms. Scanning electron microscopy and energy-dispersive spectroscopy analyses are employed to analyze the corrosion mechanisms. Wear tests on the corroded samples are also conducted to assess the extent of corrosion. Based on the experimental results, we predict the lifespans of the materials and evaluate their suitability as candidate materials for molten salt reactors. The data obtained from the experiments provide a valuable database for structural materials that can enhance the stability of molten salt reactors and recommend high-temperature corrosion-resistant materials suitable for next-generation reactors.

• Korean Journal of Chemical Engineering
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• 제35권11호
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• pp.2321-2326
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• 2018

We evaluated the optimal conditions for fluidization of nickel oxide (NiO) and its reduction into high-purity Ni during hydrogen reduction in a laboratory-scale fluidized bed reactor. A comparative study was performed through structural shape analysis using scanning electron microscopy (SEM); variance in pressure drop, minimum fluidization velocity, terminal velocity, reduction rate, and mass loss were assessed at temperatures ranging from 400 to $600^{\circ}C$ and at 20, 40, and 60 min in reaction time. We estimated the sample weight with most active fluidization to be 200 g based on the bed diameter of the fluidized bed reactor and height of the stocked material. The optimal conditions for NiO hydrogen reduction were found to be height of sample H to the internal fluidized bed reactor diameter D was H/D=1, reaction temperature of $550^{\circ}C$, reaction time of 60 min, superficial gas velocity of 0.011 m/s, and pressure drop of 77 Pa during fluidization. We determined the best operating conditions for the NiO hydrogen reduction process based on these findings.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4049-4061
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• 2022

TRISO-coated particle fuel is widely used in high temperature gas cooled reactors and other advanced reactors. The performance of coated fuel particle is one of the fundamental bases of reactor safety. The failure probability of coated fuel particle should be evaluated and determined through suitable fuel performance models and methods during normal and accident condition. In order to better facilitate the design of coated particle fuel, a new TRISO fuel performance code named FRAT (Fission product Release Analysis Tool) was developed. FRAT is designed to calculate internal gas pressure, mechanical stress and failure probability of a coated fuel particle. In this paper, FRAT was introduced and benchmarked against IAEA CRP-6 benchmark cases for coated particle failure analysis. FRAT's results agree well with benchmark values, showing the correctness and satisfactory applicability. This work helps to provide a foundation for the credible application of FRAT.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3919-3927
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• 2022

This study presents an optimal design of the coolant system of a non-refueling full-life small reactor by analyzing the space-integrated geometrical and electromagnetic variables of an extra vessel electromagnetic pump (EVEMP) for the circulation of a lead-bismuth eutectic (LBE) coolant. The EVEMP is an ideal alternative to the thermal-hydraulic system of non-refueling full-life micro reactors as it possesses no internal structures, such as impellors or sealing structures, for the transportation of LBE. Typically, the LBE passes through the annular flow channel of a reactor, is cooled by the heat exchanger, and then circulates back to the EVEMP flow channel. This thermal-hydraulic flow method is similar to natural circulation, which enhances thermal efficiency, while providing a golden time for cooling cores in the event of an emergency. When the forced circulation technology of the EVEMP was applied, the non-refueling full-life micro reactor achieve an output power of 60 MWt, which is higher than that achievable via the natural circulation method (30 MWt). Accordingly, an optimized EVEMP for Micro URANUS with a flow rate of 4196 kg/s and developed pressure of 73 kPa under a working temperature of 250 ℃ was designed.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2792-2800
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• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1596-1609
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• 2019

The fuel rod performance has great importance for the safety and economy of an operating reactor. The fuel rod performance analysis code, which considers the thermal-mechanical response and irradiation effects of fuel rod, is usually developed in order to predict fuel rod performance accurately. The FRIPAC (${\underline{F}}uel$ ${\underline{R}}od$ ${\underline{I}}ntegral$ ${\underline{P}}erformance$ ${\underline{A}}nalysis$ ${\underline{C}}ode$) is such a fuel rod performance analysis code that has been developed recently by China Nuclear Power Technology Research Institute Co. Ltd. The code aims at the computational simulation of the Pressurized Water Reactor fuel rod behavior for both steady-state and power ramp condition. A brief overview of FRIPAC is presented including the computational framework and the main behavioral models. Validation of the code is also presented and it focuses on the fuel rod behavior including fuel center temperature, fission gas release, rod internal pressure/internal void volume, cladding outer diameter and cladding corrosion thickness. The validation is based on experimental data from several international projects. The validation results indicate that FRIPAC is an accurate and reliable fuel rod performance analysis code because of the satisfactory comparison results between the experimental measurements and the code predictions.

• 대한금속재료학회지
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• 제50권9호
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• pp.659-667
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• 2012

Alloy 617 is a Ni-base superalloy and a candidate material for the intermediate heat exchanger (IHX) of a very high temperature gas reactor (VHTR) which is one of the next generation nuclear reactors under development. The high operating temperature of VHTR enables various applications such as mass production of hydrogen with high energy efficiency. Alloy 617 has good creep resistance and phase stability at high temperatures in an air environment. However, it was reported that the mechanical properties decreased at a high temperature in an impure helium environment. In this study, high-temperature corrosion tests were carried out at $850^{\circ}C-950^{\circ}C$ in a helium environment containing the impurity gases $H_2$, CO, and $CH_4$, in order to examine the corrosion behavior of Alloy 617. Until 250 h, Alloy 617 specimens showed a parabolic oxidation behavior at all temperatures. The activation energy for oxidation in helium environment was 154 kJ/mol. The SEM and EDS results elucidated a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbides. The thickness and depths of degraded layers also showed a parabolic relationship with time. A normal grain growth was observed in the Cr-rich surface oxide layer. When corrosion tests were conducted in a pure helium environment, the oxidation was suppressed drastically. It was elucidated that minor impurity gases in the helium would have detrimental effects on the high-temperature corrosion behavior of Alloy 617 for the VHTR application.

• 한국분말야금학회:학술대회논문집
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• 한국분말야금학회 2006년도 Extended Abstracts of 2006 POWDER METALLURGY World Congress Part2
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• pp.1161-1162
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• 2006

In order to overcome the recrystallization embrittlement and irradiation embrittlement of Mo, which are major problems for its fusion applications, internally nitrided Mo alloys were prepared by a novel multi-step internal nitriding. Neutron irradiation was performed in the Japan Material Testing Reactor (JMTR). After irradiation, nitrided Mo alloys exhibited $\iota$ ower ductile-brittle transition temperature than irradiated TZM. These results suggested that multi-step internal nitriding was effective to the improvement in the embrittlement by irradiation. Transmission electron microscope observation revealed that TiN particles precipitated by nitriding acted as a sink for irradiation-induced defects.