Mechanical Engineering Journal
Online ISSN : 2187-9745
ISSN-L : 2187-9745
Advance online publication
Displaying 1-49 of 49 articles from this issue
  • Keisuke YAMADA, Jinchen JI
    Article ID: 23-00411
    Published: 2024
    Advance online publication: March 18, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    This paper presents a vibration analysis using the substructure elimination and binding method for vibration systems governed by a one-dimensional wave equation. Coupled vibration analysis has been developed, and the component mode synthesis method is commonly used for dynamic analysis. In the component mode synthesis method, each substructure is formulated, and then coupling between substructures is considered. The component mode synthesis method is a type of modal analysis, and the coupled vibration between vibration systems with different governing equations can be easily formulated. The component mode synthesis method has the problem of increasing the degrees of freedom when the entire structure is complicated and needs to be divided into many substructures. Therefore, the first author proposed methods to analyze the entire vibration system without dividing it into substructures, for example, when a structure is installed inside an acoustic field or when acoustic fields with different media are in contact. These methods have the advantage that only the eigenmodes of the entire acoustic field are used. However, the calculation accuracy has been found to deteriorate because of the discontinuities or non-smooth points in sound pressure and particle displacement at the interface between air and a structure or between two acoustic fields. This study proposed a method to set a virtual elimination region at the interface and then bind the two ends of the virtual elimination region to solve this problem. The analytical model for this method was presented, and a wave equation was derived in this study. Modal analysis was applied to the wave equation. The simulations revealed that the density and bulk modulus of the virtual elimination region should be zero and that its length should be set at 2.5–3.5 times the wavelength of the highest eigenmode of the entire vibration system. To investigate the advantage of low DOFs, the simulation results obtained using the proposed method were compared with those obtained using the component mode synthesis method based on the exact solutions.

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  • Tomomasa FUNAKOSHI, Sou WATANABE, Yoichi ARAI, Toshihiro IWAMOTO, Masa ...
    Article ID: 23-00445
    Published: 2024
    Advance online publication: March 16, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Various types of machine oil are used for analysis and utility equipment, and these organic liquid wastes are stored in nuclear facilities and laboratories due to a lack of appropriate treatment processes. Treatment of organic liquid waste is one of the principal tasks since radiolysis of organic material generates various hazardous products. Perfluoro oil, generally used in vacuum pumps, is difficult to decompose because of its chemical stability. Calcination of fluorine compounds is possible to generate toxic and corrosive gas products. In order to achieve complete mineralization of the organic liquid wastes, the application of a subcritical water reaction was examined. In this study, the effect of introducing a functional group into a perfluoro compound on its decomposition performance was experimentally evaluated. First, we carried out the transformation of perfluorohexane to perfluorohexyl iodide or perfluoroheptanoic acid based on reported procedures. Next, laboratory scale batch-wise decomposition tests with subcritical water on perfluorohexyl iodide and on perfluoro heptanoic acid were carried out. Analyses of degraded organic products remaining in the aqueous phase by atmospheric pressure ionization mass spectrometry and liquid chromatography-mass spectrometry were carried out. The decomposition products of each fluorine compound were identified, confirming that subcritical treatment is a promising treatment method.

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  • Lei SHI, Yihan WANG, Na MA, Hongjun LIU
    Article ID: 23-00460
    Published: 2024
    Advance online publication: March 16, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Spent nuclear fuel (spent fuel), which is used nuclear fuel that has been exposed to radiation usually produced by nuclear reactors in nuclear power plants, is an inevitable product from the development of nuclear energy. Almost all of the fuel content is radioactive, and long systematic process are required for the safety management, which has always been an important global issue. In order to make sure that spent nuclear fuel should be safely managed, different countries developing nuclear power have established a complete policy and legislative system, so as to ensure that the whole process of spent fuel management is systematic, standardized and effective. In developed countries such as France, Russia and Japan, closed-cycle strategy is implemented with industrial-scale reprocessing plant under construction or in operation. At present, China has become the country with the largest scale of nuclear power under construction in the world. There will be a large number of spent nuclear fuel requiring properly and safely managed. The lessons-learning of how developed countries managing spent nuclear fuel arising is important for China. The authors suggest that it is necessary to combine the top-level design to the legal practice, so that there are laws to respect during all steps of spent fuel management, and responsibilities of all parties are clear.

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  • Yue WANG, Xiaohui ZHUO, Qiang CHENG, Zixue LUO
    Article ID: 23-00443
    Published: 2024
    Advance online publication: March 14, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    In order to improve the combustion performance of anthracite coal and solve the problem of resource utilization of distillers’ grains, experiments on the mixing combustion of anthracite coal and distillers’ grains were carried out and the effects of different combustion conditions on their ash characteristics were investigated in order to provide theoretical guidance for the optimization of combustion parameters in the actual operation of power plants. The ash samples were made on a drop tube furnace and a tube furnace, and then the microscopic morphology of the slag was observed by Scanning Electron Microscope-Energy Dispersive Spectrometer (SEM-EDS), while its mineralogical composition was analysed by X-ray diffraction (XRD). The results show that the increase of distillers’ grain mixing ratio, combustion temperature and oxygen concentration will lead to the deterioration of ash slagging characteristics, while the excess air coefficient has little effect on it.

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  • Akira SUAMI, Kousuke MINAMI, Nobusuke KOBAYASHI, Yoshinori ITAYA
    Article ID: 23-00489
    Published: 2024
    Advance online publication: March 14, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    In recent years, the amount of waste clothes has been increasing. And the recycling of the waste clothes would be necessary for material cycles. A new method for separating blended fibers is required, because the separating of fibers is difficult to be tangle fibers in clothes. Recent reports indicate that the addition of an iron-based catalyst in the pyrolysis improves the decomposition rate of cotton, and the TBT (Titanium Butoxide Tetramer) catalyst improves the pyrolysis decomposition of PEs. Consequently, a new method of separating single fiber from blended fibers by catalytic pyrolysis was proposed. In this study, thermo-gravimetric (TG) analysis and a lab scale pyrolysis experiment of cotton and PEs were conducted by using two types of catalysts, FeCl2 and TBT. These catalysts were selected to understand the fundamental phenomena of cotton and PEs pyrolysis. The results showed that the FeCl2 catalyst made the pyrolysis starting temperature of cotton lower and pyrolysis rate increased. Whereas the TBT did not affect the pyrolysis behavior of cotton and PEs. The FeCl2 was the most effective catalyst for controlling the pyrolysis temperature and separating the fibers during pyrolysis. In a lab scale catalytic pyrolysis experiment with actual cotton and PEs, the cotton was decomposed to 90 % of its mass by pyrolysis. While the PEs was not so decomposed, and 90 % of PEs was remained at 633 K pyrolysis temperature. As the verification of thermal effect on the PEs residue, FTIR analysis revealed that the surface functional groups of the PEs raw material were not affected by pyrolysis. This study demonstrated that the catalytic pyrolysis would accelerate the decomposition of only cotton, while PEs could be remained and potentially reused as a raw material.

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  • Akihiro MANO, Takuya SATO, Masakazu ICHIMIYA, Naoto KASAHARA
    Article ID: 23-00515
    Published: 2024
    Advance online publication: March 08, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    After the Great East Japan earthquake, measures for beyond design basis events have become more important. For a component, mitigation of an influence of a failure on its safety performance is required in addition to prevention of the failure. These kinds of measures are reduced the risk of safety performance defined as multiplication of failure frequency and failure consequence regarding impact on safety performance. One of safety analyses regarding such measures is fragility analysis. However, the existing fragility is represented using failure probability, and does not take into account the failure consequence. In this paper, for the purpose of more rational fragility analysis, we propose a new concept of performance-based fragility incorporating failure consequence, which represents the risk of safety performance degradation. In addition, to present the effectiveness, the proposed fragility is quantified by applying to a piping system subjected to excessive earthquakes which is a key issue of beyond design basis event. A piping system subjected to excessive earthquake can fail by ratchet, fatigue, collapse, and break. These failure modes have different occurrence conditions and lead to different consequences about core cooling performance. After the analyses on occurrence probability and failure consequence on each failure modes, the proposed fragility is calculated as the multiplication of them.

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  • Shigeru TAKAYA, Akiyuki SEKI, Masanori YOSHIKAWA, Naoto SASAKI, Xing Y ...
    Article ID: 23-00408
    Published: 2024
    Advance online publication: March 07, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Enhancing the ability to manage abnormal situations is important for improvement of the safety of nuclear power plants. It is necessary to investigate potential risks thoroughly in advance, and prepare countermeasures against the identified risks. In addition, in case of an occurrence of an abnormal situation, plant operators are required to recognize the plant situation promptly and select a suitable countermeasure. However, the human ability to perform it is limited because the number of such abnormal situations in actual nuclear power plants is indefinite. Due to the advent of AI, it becomes possible to compensate for such limitation, by learning abnormal situations and assessing the effectiveness of prepared countermeasures virtually. The present study aims to develop such an AI-based support system for the plant operators to deal with abnormal situations steadily. Although many previous studies about detection of anomalies have been conducted, few studies consider countermeasures, especially against unexperienced abnormal situations. This study develops a novel plant operator support system designed not only to estimate details of anomalies in a plant but also propose countermeasures adaptively by employing several AI technologies of deep neural network and reinforcement learning. A plant simulator is used to prepare training data for the AI system. The combination of the proposed AI-based system and the plant simulator makes it possible to identify abnormal situations unknown to operators and propose countermeasures. The design and performance of the proposed system is illustrated using High Temperature engineering Test Reactor (HTTR) operated in Japan Atomic Energy Agency.

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  • Xiaoran LI, Weifeng WU, Yin ZHANG, Chengyi LI, Jing ZHANG, Chengqiang ...
    Article ID: 23-00458
    Published: 2024
    Advance online publication: March 07, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Different from a gas expander, there is a phase change from liquid to vapor called the flash inside the two-phase expander. But, same with the gas expander, only the vapor phase expands and pushes the rotation of the expander to output power in the two-phase expander. It means that the amount of the vapor is increased during the expansion by the flash. Obviously, the flash rate of the working fluid from liquid to vapor is critical to the expansion power recovery. In this paper, an experiment system with water as the working fluid was established to study the heat transfer characteristics of flash evaporation during the two-phase expansion process. The flash was controlled inside a flash chamber, under the initial temperature from 100 ℃ to 130 ℃, and the initial liquid level height from 110 mm to 175 mm. The pressure evolution of the flash evaporation was limited by the pressure regulating valve. The obtained results showed that the level of the instant liquid superheat was positively associated with the flash rate of evaporation. The temperature drop rate behaved negatively correlated with the instant superheat, while higher the instant superheat resulted in faster liquid temperature reduction. Both the liquid-vapor heat transfer coefficient and the instant flash rate were increased with the increase of instant superheat. Generally, the peak of instant superheats occurred earlier than that of instant flash rate and instant heat transfer coefficient. The instant superheat level was increased and affected by the initial temperature and initial liquid level height.

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  • Shota UEDA, Takahiro ARAI, Masahiro FURUYA, Riichiro OKAWA
    Article ID: 23-00496
    Published: 2024
    Advance online publication: March 06, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    In severe accidents in light-water reactors, the core melt, which falls to the bottom of the containment vessel, is expected to be cooled in the retained water. The particulate debris is generated after the molten material falls from the pressure vessel. For example, the cooling characteristics remain to be further elucidated in a system where particulate debris and structure material coexist and the gas-liquid two-phase flow inside particulate debris. This study investigated air–water two-phase flow inside a particulate bed and near a structure wall using a high-speed camera with a refractive index matching the transparent bed with purified water and a wire-mesh sensor (WMS). Particles with diameters of ø 3, 5, and 10 mm were used and subjected to air–water two-phase flow tests with a superficial gas velocity of 4.0–2.0×103 mm/s and superficial liquid velocity of 0.5–75.3 mm/s. Visualization results of bubble behavior near the wall and inside the bed showed the advection of bubbles from within the bed to the vicinity of the structure wall. This finding explains the higher void fraction measured in the vicinity of the structure wall using the WMS. Analysis of the void fraction distribution based on the theoretical models indicated that the higher void fraction near the structure wall observed in the case with larger particle size may be attributed to the slip velocity between the gas and liquid phases in the particulate bed. The developed techniques and insights gained in this study contribute to a detailed understanding of the thermal-hydraulic phenomena that accompany two-phase flows in particulate debris.

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  • Md. Iqbal HOSAN, Kohei TAKANISHI, Koji MORITA, Wei LIU, Xu CHENG
    Article ID: 23-00423
    Published: 2024
    Advance online publication: March 01, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    The accident at the Fukushima Daiichi Nuclear Power Plant in 2011 led to a core meltdown, resulting in the significant release of radioactive materials into the environment, revealing the urgent need for further in-depth development of Level 2 probabilistic safety assessment technology. To help establish an effective source-term migration evaluation method, this study investigates fission product migration behavior across leak pathways. Specifically, an experimental line is developed, and experiments are performed under conditions that simulate the environmental and flow conditions in containment vessel penetrations and failure locations during a severe accident. The experiments are conducted in narrow circular pipes, which represent the leak pathways in the containment vessel and reactor building, to determine the impact of flow rate, particle size, and flow path size on the decontamination factors. Additionally, a turbulent deposition model that accounts for re-entrainment effects has been developed, and the experimentally obtained decontamination factors are compared with the developed model, as well as a conventional model. The predicted decontamination factors from the present model exhibit similar trends and values to the experimental results.

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  • Toshihiro IWAMOTO, Madoka SAITO, Youko TAKAHATAKE, Sou WATANABE, Masay ...
    Article ID: 23-00444
    Published: 2024
    Advance online publication: March 01, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Nuclear fuel fabrication process has generated a large amount of wastes contaminated with enriched uranium, and those have been accumulated for long time in Japan. These waste hasn’t treated, because there is no protocol to treat and dispose these. Procedure is proposed; 1) Leaching Uranium from the waste, 2) recovering uranium from the leaching solution, 3) treating residue and secondary waste for disposing in the near field region. As concentration of uranium in the leaching solution must be far smaller than that of dissolving solution of the spent fuel, an efficient uranium recovery procedure which generates less secondary wastes than the solvent extraction technology is required. Temperature swing extraction technology is one of promising methods to recover uranium efficiently. In the technology, uranium ions are extracted by extraction, and complex with ligands are separated by a polymer with reversible hydrophilic-hydrophobic properties. Monoamides can be adopted to the procedure. However suitable structure of monoamide for the technology is unclear. Therefore, for deciding appropriate structure of monoamides, these with different alkyl chain structures were synthesized, and applicability of the extractants were evaluated by solvent extraction and temperature swing extraction experiments. In experiments, Ce(IV) was used as a simulant of U(IV). Three monoamides were synthesized, and solvent extraction experiments and Temperature swing extraction experiments were carried out. Monoamides with straight chain showed larger recovery ratios whereas the branching monoamide showed far smaller values on solvent extraction experiments, and separation factors of Ce was good on monoamides with straight chain on temperature swing extraction experiments. Therefore the monoamide with long straight chain can be favorable for uranium recovery.

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  • Yihua DUO, Zili HUANG, Hong XU
    Article ID: 23-00464
    Published: 2024
    Advance online publication: March 01, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    As an aftermath of the Fukushima nuclear accident and its implications for public acceptance of nuclear energy, the nuclear industry has begun to seriously rethink and improve the safety strategies of nuclear power plants. The beyond-design-basis external events, such as the earthquake and tsunami in Fukushima nuclear accident, may lead to an extreme accident which may lead to the failure of the traditional strategies. In the recent decade, the concept of FLEX strategies has been accepted in the nuclear engineering community and attracted a large number of researchers to investigate it for nuclear accident mitigation. The development of FLEX strategies is divided into two steps. One is an uncertainty study of system response, especially the safety injection system. The other is training and testing of machine learning (ML)-based FLEX strategies and to verify the effectiveness of ML algorithms in the development of FLEX strategies for nuclear power plants. This paper focuses on first step uncertainty study of the nuclear accidents by using statistical algorithm to generate a database of incidents and system responses. The conclusion of this paper is useful for analyzing the characteristic of typical nuclear accidents and benefits to the FLEX strategy development, which will be focused on in the near future.

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  • Kenichi MABUCHI, Kazuya TATSUMI, Reiko KURIYAMA, Kazuyoshi NAKABE
    Article ID: 23-00524
    Published: 2024
    Advance online publication: February 25, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    We developed a technique which can increase the yield of one-to-one particle encapsulation by applying the dielectrophoretic particle alignment technique using boxcar-type electrodes. Dielectrophoretic force generated by the boxcar-type electrodes accelerate and decelerate the particles periodically as they flow in the electrode region. Further, the dielectrophoretic force is turned on and off at constant frequency. The force exerted on the particle periodically over space and time can align them in the streamwise direction with even interval. In this study, the boxcar-type electrodes were installed in the microchannel in the region upstream of the flow-focusing channel in which the water-in-oil droplets were generated. By adjusting the on-off period of the applied voltage generating the dielectrophoretic force to the period of the droplet generation, each particle could be separately encapsulated in the droplets. The principle of particle alignment using periodic force was first described based on a one-dimensional model. The flow structure and the characteristics of the droplet generation in the flow-focusing channel was then discussed in relation to the surface tension of the fluids and the wettability of the wall. We measured the velocity distribution of the particles flowing in the boxcar-type electrode region to evaluate the effects of the droplet generation on the motion of the particles and the alignment performance. The results showed that the particle could be aligned in the fluctuating flow caused by the droplet generation, and each particle can be encapsulated in different droplets. This was further demonstrated by measuring the probability function of the droplets containing specific number of particles, which showed that 100% yield of one-to-one particle encapsulation can be achieved under the investigated condition of particle number density of 0.4. Moreover, the throughput increased 46% compared to the case of having the particles supplied randomly.

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  • Yutaka NAKANO, Katsunori CHO, Takamasa HASE, Yuki MATSUMURA, Hiroki TA ...
    Article ID: 23-00389
    Published: 2024
    Advance online publication: February 20, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    In a printer fuser assembly, unpleasant noise caused by friction-induced vibrations can occur owing to friction between the pad and the inner surface of the fixing sleeve. However, the vibration mode of the fuser assembly involved in the generation of friction-induced vibration has not yet been identified, and several aspects of the noise-generation mechanism remain unexplained. This study aimed to clarify the mechanism of noise generation due to friction-induced vibration in the printer fuser assembly and propose countermeasures against this noise. The relationship between the tendency of noise occurrence, friction characteristics of the fixing sleeve, and the generation mechanism of noise generation was investigated by examining the actual operating vibration modes of the fixing sleeve and pressure roller during noise generation. The results confirmed that the fixing sleeve is the main vibration source during noise generation and that the fixing sleeve’s vibration modes in the radial and circumferential directions are coupled. The effect of negative velocity gradient characteristics on the critical friction coefficient at the onset of frictional self-excited vibration caused by mode coupling was examined using a 2-degree-of-freedom lumped mass model. The friction characteristics between the sleeve and pad and the actual operating vibration mode of the sleeve during noise generation revealed that the noise generation mechanism is characterized by mode-coupling type self-excited vibration and that the noise can occur at low friction due to the negative gradient characteristic of the friction coefficient. Finally, the countermeasure was proposed against noise based on the characteristics of self-excited vibration caused by mode coupling. The authors found that placing the spacer on the trailing side and increasing the surface pressure distribution on the trailing side could control the friction-induced noise without changing the negative velocity gradient characteristics of the friction coefficient.

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  • Yang XUE, Yoji SHIBUTANI
    Article ID: 23-00455
    Published: 2024
    Advance online publication: February 20, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Adhesive bonding is the preferred method for joining dissimilar materials owing to the low weight of adhesives. To evaluate the application of adhesive bonding to multi-materialization, in this study, we used conical cup testing as an integrated formability test method to assess the forming limit of bonded multi-layered metal sheets, comparing them with the equivalent unitary sheets. The conical cup testing equipment used in this study was designed and produced based on JIS Z 2249, with a viewing port that can be used to observe the deformed bottom of the specimens during the conical cup test. Multi-layered specimens composed of aluminum (Al) and steel (steel plate cold-rolled commercial (SPCC) steel) layers were produced using acrylic and epoxy adhesives, and single-(material)-layered specimens composed of layers of the same material were also produced. In addition, the unitary specimens of the materials without any bonding were examined as reference materials. The experimental results showed that the forming limits of the single-(material)-layered Al specimens (pure aluminum A1050 and aluminum alloy A5052) were nearly identical to those of the unitary specimens. Therefore, the use of the adhesive improved the formability of the Al sheet to that of the corresponding unitary specimen with twice the thickness. The results of the A5052-SPCC multi-layered specimen bonded by acrylic adhesive were compared with those of the SPCC unitary specimen, and its forming limit increased by 35% with a notable 33% reduction in weight, demonstrating the enhanced performance of the multi-layer structure. The reason for the improvement in formability may be the suppression of the abrupt increase in surface roughness, which otherwise leads to plastic instability.

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  • Chikara KONNO, Mami KOCHIYAMA, Hirokazu HAYASHI
    Article ID: 23-00386
    Published: 2024
    Advance online publication: February 16, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Activation cross-section libraries for the ORIGEN and ORIGEN-S codes have been generated from the latest Japanese nuclear data library JENDL-5 and JENDL Activation Cross Section File for Nuclear Decommissioning 2017 (JENDL/AD-2017) for activation calculations in nuclear facility decommissioning. The ORIGEN activation cross-section libraries of the 200- and 48-group structures were generated with the AMPX-6 code, while the ORIGEN-S activation cross-section libraries with a MAXS format of the 199-group structure were done with the PREPRO2018 code. In order to verify the generated ORIGEN and ORIGEN-S activation cross-section libraries, activation calculations for Japan Power Demonstration Reactor (JPDR) were carried out in detailed with the DORT, ORIGEN and ORIGEN-S codes, the generated activation cross-section libraries, the ORIGEN and ORIGEN-S bundled activation cross-section libraries, and a JPDR partial model. The following comparisons were performed: 1) the ORIGEN calculation results with the generated activation cross-section libraries and bundled ones, 2) the 200-group and 48-group ORIGEN calculations, 3) the ORIGEN-S calculation results with the generated activation cross-section libraries and bundled ones, and 4) the ORIGEN and ORIGEN-S calculations with the JENDL-5 activation cross-section libraries. Most of the differences of the calculation results were less than 20%, which demonstrated that the libraries were generated adequately. The generated libraries have been released with additional programs from the JAEA website.

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  • Sou WATANABE, Youko TAKAHATAKE, Kenta HASEGAWA, Ichiro GOTO, Yasunori ...
    Article ID: 23-00461
    Published: 2024
    Advance online publication: February 16, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Japan Atomic Energy Agency is developing extraction chromatography technology to recover MA(III) from spent nuclear fuel. Developments in the extraction chromatography system especially focusing on safety and stable operation are required for practical application of the technology. This paper discusses the main tasks that have to be challenged preferentially based on achievements obtained by previous studies and potential MA(III) recovery process flow. For the safety assessment, removal of fine particles, optimization in adsorbent for a reduction in pressure drop of the column, degradation behavior of the adsorbents, and recycling of the adsorbents are topics for development. Enhancement in resistance against radiation and acid of material in pump and development of automated operation system is required to maintain product quality during long time operation. Those studies are carried out simultaneously in a research project of 2019-2024, and new technologies such as a filter with porous media, porous adsorbents for low-pressure drop column, used adsorbent recycling, radiation resistant pump, and automated valve switching system using online analysis are developed in this project. A demonstration of the total system with a large-scale apparatus will be carried out at the end of the project.

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  • Yohei MORIYA, Kenichi IHARA, Hiroki NAKAMURA, Hitoshi NOJIMA, Satoshi ...
    Article ID: 23-00414
    Published: 2024
    Advance online publication: February 07, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    An importance for risk management has come to be re-recognized in Japan after the accident of the Fukushima Daiichi nuclear power plants. A Probabilistic Risk Assessment (PRA) fulfills a significantly important role in the framework of the risk management. Technical adequacy of PRA model and high quality of data are required in order to utilize PRA for such risk-informed activities. Therefore, Japanese utilities launched the project for enhancing the quality of PRA models and are upgrading PRA models to comply with international standards. Chugoku Electric Power Company (Chugoku) is also upgrading the internal events at power PRA model including both the level 1 and level 2 (except the source term analysis) and enhancing PRA quality in order to utilize risk insight for improving nuclear safety. The aim of the PRA upgrade is to reflect international state-of-the-practice approaches and to meet ASME/ANS PRA standard (ASME/ANS 2013) requirements (Capability Category II). Our PRA upgrading process is divided into 3 phases: “Phase I”, “Phase II” and “As-is”. This paper reports the detail of PRA upgrade implemented in Phase I and Phase II such as initiating event refinement, accident sequence re-evaluation, detailed fault tree (FT) analyses, modeling severe accident measures and updating human reliability analysis (HRA).

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  • Nhut Vu LUU, Kunihisa NAKAJIMA
    Article ID: 23-00446
    Published: 2024
    Advance online publication: February 07, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Cs distribution is crucial for decommissioning Fukushima Daiichi Nuclear Power Station (1F). Several experimental studies confirmed Cs retention on stainless steels by performing chemical reactions at high temperatures (typically > 800°C), but the Cs retention on concrete, used in large quantities in light water reactors, is not fully understood. This study demonstrated that Cs might have been deposited and retained on concrete structures where the temperature was not so high during the 1F accident. Results showed that the CsOH/concrete interaction at ~200°C occurred in water-insoluble Cs–(Al, Fe)–Si–O deposits and water-soluble phases, i.e., Cs carbonate hydrate and possibly Cs2SiO3 if Al and Fe are absent. CsOH might be trapped on concrete by chemical reaction with CaCO3 to form Cs2CO3 hydrate and with aluminosilicate and SiO2 (quartz) to form Cs–Al–Si–O and Cs–Si–O deposits, respectively. This output could help elucidate the trapping mechanism that caused extremely high radioactivity on concrete shield plugs at 1F and develop an effective decommissioning practice for concrete structures.

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  • Hodaka MATSUZAKI, Akira YOSHIDA, Yoshiharu AMANO
    Article ID: 23-00465
    Published: 2024
    Advance online publication: February 07, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Japanese geothermal power plants are expected to be significant renewable energy sources owing to the abundance of geothermal sources in Japan. The plant capacity factor in geothermal power plants is low. One reason is frequent sudden pressure drops in production well corresponding to the change of subsurface condition. To obtain a stable steam quantity, it is necessary to observe the subsurface conditions in real-time and perform appropriate operations. The use of a model to predict steam enthalpy in real-time has potential to monitor changes in subsurface conditions and contribute to the composition of plant operational strategies. However, training a model requires a large amount of data. The purpose of this study is to evaluate the effectiveness of transferring the knowledge of a pretrained model for predicting steam enthalpy in one plant to another plant with limited data. This study proposes a methodology based on the combination of the temporal fusion transformer (TFT) architecture and transfer learning (TL). This approach represents a novel way to share knowledge from a pretrained model based on historical data from a plant, which helps reduce the need for large amounts of data when dealing with a new plant. A pretrained TFT model (PM) enables the prediction of rapid steam enthalpy decreases in the source plant. Transfer learning using a PM was confirmed to enhance the performance of steam enthalpy prediction in another plant compared to using a model without pretraining. The effectiveness of transfer techniques has the potential to contribute to improving the operational efficiency of geothermal power plants. The transfer learning strategies proposed in this study heavily rely on the similarity of the source data. In the future, we aim to compute data correlations between plants and the effectiveness of transfer learning.

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  • Tomohiko YAMAMOTO, Tomoyoshi WATAKABE, Masashi MIYAZAKI, Shigeki OKAMU ...
    Article ID: 23-00393
    Published: 2024
    Advance online publication: February 02, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    A sodium-cooled fast reactor (SFR) considers adopting 3-dimensional seismic isolation devices for withstanding seismic loads not only horizontal but also vertical direction. A seismic isolation device consists of a laminated rubber bearing and horizontal oil dampers for horizontal direction, coned disc springs and vertical oil dampers for vertical direction, respectively. And also, in order to make horizontal and vertical motion independent, sliding elements are adopted. In order to investigate the performance of each component and the feasibility of integrated system for SFR, the experiments such as load-displacement tests, vibrating tests, etc., to each component of seismic isolation devices and seismic response analysis are carried out. As those experimental results, the mechanical characteristics of each component and the devices are grasped, then it is demonstrated that components and devices have expected performances to reduce the seismic loading within the design range. As the analytical results of seismic response, it is indicated that this 3-dimesional seismic isolation device and system can reduce the seismic response on horizontal and vertical simultaneously. Based on the analytical studies and experimental results, the feasibility of newly developed 3-dimensional seismic isolation system is obtained and the prospect of practical application of 3D seismic isolation system for fast reactor is implied in this paper.

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  • Mitsuhiro AOYAGI, Tohru MAKINO, Hiroshi OHKI, Akihiro UCHIBORI, Yasush ...
    Article ID: 23-00459
    Published: 2024
    Advance online publication: February 02, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    The SPECTRA code has been developed as an integrated safety analysis tool for comprehensive analyses of in- and ex-vessel phenomena during a severe accident (SA) in a sodium-cooled fast reactor (SFR). The individual modules, such as sodium fire and sodium-concrete reaction, have been implemented into SPECTRA to simulate specific phenomena in an accident of SFRs. However, SPECTRA has no capability for the overlapped event involving sodium fire and sodium-concrete reaction in the same compartment because the sodium pool and the floor concrete are modeled in each module independently. The capability of SPECTRA is enhanced by integrating the analyses of sodium pool fire and concrete ablation for overlapped events of the ex-vessel phenomena. A new shared module for the sodium pool and floor concrete region, named “PFC module” was introduced in the previous study. The sodium fire module is connected to the PFC module in this study. The integrated model is validated through the benchmark analysis of the F7-1 pool fire experiment. The calculation results of the temperature and combustion rate show good agreement with the experimental result. A demonstration analysis is also conducted for an overlapped event of the ex-vessel phenomena including both concrete ablation and sodium pool fire. SPECTRA can simulate a reasonable heat and mass transfer behavior during the overlapped event.

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  • Tomoyoshi WATAKABE, Takahiro OKUDA, Satoshi OKAJIMA
    Article ID: 23-00395
    Published: 2024
    Advance online publication: February 01, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    A three-dimensional seismic isolation system has been developed for the conceptual design of a sodium-cooled fast reactor (SFR) in Japan. Between the reactor building equipped with the isolation system and the turbine building with no isolation system, there are main steam crossover piping systems, in which a large displacement would occur, and this is a challenge for developing an SFR with the isolation system. Existing studies of the seismic design of crossover piping were performed on the piping, considering the conditions of light water reactors (LWRs). However, because the SFR operates at elevated temperatures compared with LWRs, the design of crossover piping systems must comply with the design code for elevated temperature components. It must be also considered that the advanced material, Mod.9Cr-1Mo steel, is employed as a material of the crossover piping in the SFR. In this study, we performed seismic evaluation using an example of a crossover piping layout to refer to existing codes on piping. The evaluation results and insight obtained from existing dynamic failure tests of piping components show the knowledge on seismic evaluation of the crossover piping systems in the SFR with the three-dimensional isolation system to achieve the SFR with the three-dimensional isolation system.

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  • Baigong WANG, Huan MA, Fengqi SI
    Article ID: 23-00385
    Published: 2024
    Advance online publication: January 31, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    This paper proposed a novel cooling system that combines a phase change material (PCM) with an indirect dry cooling system to enhance exhaust steam condensation of the Rankine cycle for a coal-fired power plant (CFPP), especially in high ambient temperature conditions. A PCM heat storage tank (PCMHST) is embedded in series between the natural draft dry cooling tower (NDDCT) and condenser to realize peak and valley shifting of cooling capacity, and RT35HC is selected as the PCM. During discharging process of the PCMHST, RT35HC is chilled by low-temperature circulating water; during charging process of the PCMHST, the circulating water is cooled in the PCMHST and NDDCT. This paper first reveals the effects of cooling load distribution in the novel cooling system on the CFPP operating economy during charging and discharging process. In addition, the PCMHST in series before the tower can reduce the circulating water temperature more efficiently than that in series behind the tower. Results show that integrating a PCMHST can reduce the coal consumption of the CFPP by 5325 kg within a complete back pressure regulation process.

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  • Kazushi MIYATA, Shuichi UMEZAWA
    Article ID: 23-00402
    Published: 2024
    Advance online publication: January 27, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    The heater method is a clamp-on type flow measurement method that can be applied to metal pipes in high-temperature environments and is useful for determining the adequate flow rate in existing pipes for the stable operation of thermal power plants. Specifically, a ring-shaped heater is attached to the outer of an existing pipe, and the outer surface temperature of the pipe near the heater is measured to calculate the fluid flow rate in the pipe. In this study, as the heat transfer coefficient upstream of the heater was found to be constant in the flow direction by computational fluid dynamics (CFD) simulation results, the heat conduction equation was solved to express the correlation between the outer surface temperature and the heat transfer coefficient inside the pipe, upstream of the heater. A correlation equation was then developed for the heat transfer coefficient in the pipe upstream of the heater using CFD simulation results. In addition, to improve the accuracy of the flow rate calculation, a condition equation was incorporated to determine whether the heat transfer rate calculated using the heat transfer coefficient based on the outer surface temperature is consistent with the heating power of the heater. Experimental data confirm that the equations can be used to calculate the fluid flow rate accurately and indirectly in the pipe from the outer surface temperature of the pipe upstream of the heater.

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  • Yukihiko OKUDA, Kiyotaka TAKITO, Akemi NISHIDA, Yinsheng LI
    Article ID: 23-00405
    Published: 2024
    Advance online publication: January 27, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    The importance of seismic probabilistic risk assessment has drawn considerable attention for improving the seismic safety of nuclear facilities against earthquakes that exceed the design input ground motion. The realistic seismic response of the equipment or piping in nuclear power plants for fragility assessment in seismic probabilistic risk assessment must be evaluated. In particular, because piping systems have plant-specific complex configurations, the arrangement and stiffness of piping support structures considerably affect the seismic response characteristics of the entire piping system. However, the current seismic design procedure adopts an evaluation method assuming an elastic response. Therefore, this study develops an elasto-plastic response analysis method to estimate the realistic response of piping systems, including piping support structures. The authors have performed loading tests on piping support structures and optimization studies of elasto-plastic property models to develop an elasto-plastic response analysis method for piping systems, including piping support structures. Herein, the applicability of this method is confirmed through a simulation analysis of the previously reported elasto-plastic response of piping support structure loading test. Consequently, a good correlation is found between the test and simulation analysis results for the ductility factor and the damage status. Therefore, the ductility factor is verified to be effective as a damage evaluation index for piping support structures.

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  • Kenta HASEGAWA, Ichiro GOTO, Yasunori MIYAZAKI, Hiromu AMBAI, Sou WATA ...
    Article ID: 23-00407
    Published: 2024
    Advance online publication: January 27, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Japan Atomic Energy Agency has been working on development of extraction chromatography technology for recovery of trivalent minor actinides (MA(III): Am, Cm) from high-level radioactive waste generated in reprocessing of spent fuel. The technology utilizes porous silica particles with about 50μm diameter for support of adsorbents. The particles are coated by styrene-divinylbenzene copolymer, and an extractant for MA recovery is impregnated into the polymer. Pressure drop of the packed column depends on characteristics of the particle (diameter, uniformity and pore size). Large pressure drop of the column is not favorable for safety assessment of the technology although a certain level of the pressure drop is indispensable for excellent separation performance. In this study, we applied a granulation technique using a spray dryer that is widely used in industry, and conducted experiments to find the optimal specifications for silica support particles and conditions for the granulation operation. A basic characterization of the adsorbent prepared from the produced particles was carried out by an adsorption test of simulated high level liquid waste. As a result, it was first found that the uniformity of particle size could be improved by suppressing pulsation during supply of the feed slurry to the spray dryer. Next, by adjusting the viscosity of the feed slurry, it was possible to suppress the generation of particles with undesirable shapes such as hollow particles and depressed particles. It was also confirmed that these particles had a lower pressure drop than conventionally used particles.

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  • Pascal ZAVALETA, Mickaël COUTIN, Thomas GÉLAIN, Jocelyne LACOUE, Phili ...
    Article ID: 23-00463
    Published: 2024
    Advance online publication: January 27, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Fires in nuclear fuel fabrication and reprocessing plants can cause the rupture of the glove box (GB) containment with a risk of dispersion of plutonium dioxide (PuO2) within the facility. Fire safety analyses need to assess the resulting radiological consequences to strengthen the appropriate prevention and protection measures in these plants. To this end, the French Institute for Radiation Protection and Nuclear Safety (IRSN), in partnership with the Japanese Nuclear Regulation Authority (NRA), has carried out since 2019 a research project that aims at assessing the airborne release fraction (ARF) of PuO2 involved in GB fires. This project, named FIGARO (Fires Involving Glove boxes with Aerosol Release Occurrences) follows a progressive approach, that has started with small and medium-scale analytical tests to study separately the various mechanisms involved in the GB fires and the PuO2 airborne release and will end with large-scale GB fire tests to transpose and validate the analytical works for realistic assessments of the ARF of PuO2 both in open and confined conditions. This project also includes the development of a GB fire model in the IRSN SYLVIA and CALIF3S-ISIS softwares and its validation. In addition to presenting the FIGARO program, this paper provides its first outcomes.

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  • Eugene GATETE, Biao SHEN, Akiko KANEKO
    Article ID: 23-00315
    Published: 2024
    Advance online publication: January 24, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    The use of levitated droplets with electrostatic and ultrasonic fields has attracted much attention in the fields of materials development, chemical engineering, droplet-based microfluidics, inkjet printing, and aerospace engineering. To use them properly, it is essential to understand the internal flow of the suspended droplet. The flow inside fluid droplets is visualized with the particle image velocimetry (PIV) method. The fluid that is to be investigated is scattered with small particles that follow the flow well. One problem that occurs when the PIV method is used to measure velocity fields inside droplets is however that light refracts at the droplet surface and the internal flow curvature is distorted. The aim of this research is to improve the accuracy of PIV measurement by correcting the distorted particle images of an acoustically levitated droplet using calibration method. In this study, a simulated droplet with different refractive index and aspect ratio was used to investigate their influence on distortion correction. The circular target plate was also utilized to correct the distortion in a simulated droplet using a calibration method of the python-OpenCV. The experimental results showed that the internal flow curvature can be distorted in two types such as barrel, pincushion distortions, and increase as the refractive index increases. Therefore, correction of the distorted image of particles in the droplet showed good convergence as the aspect ratio decreased.

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  • Bing BAI, Xu HAN, Shi WU, Jin GAO, Xinfu HE, Wen YANG
    Article ID: 23-00483
    Published: 2024
    Advance online publication: January 24, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    The main reason that restricts the increase of fast reactor burnup is the irradiation swelling of the cladding. The reported research shows that the main elements and trace alloy/impurity elements have a great influence on the irradiation swelling. However, the behavior of the above elements in the process of coupling with irradiation defects is complex, and it is difficult to directly measure the relationship between these elements, irradiation defects and microstructure evolution in experiments. The emergence of machine learning and big data mining technology will help to gain new understanding of the impact of irradiation swelling on austenitic stainless steel, so as to find a new type of austenitic stainless steel cladding material resistant to irradiation swelling. Therefore, in this work, about 1000 groups of data such as composition, irradiation conditions and irradiation swelling of austenitic stainless steel are collected, and the data are cleaned and screened for modelling by machine learning. The deep neural network with back propagation is used in this work, and the correlation between alloy composition such as Cr, Ni, Ti and C, irradiation dose and temperature and irradiation swelling of austenitic stainless steel is established. The results show that the addition of a certain amount of Ti and Si can effectively inhibit the irradiation swelling of austenitic stainless steel, but the addition of Ni will aggravate the swelling effect. The addition of Cr, Ni, Ti and Si will increase the swelling inflection point dose, while the addition of C and P will reduce the swelling inflection point dose. Besides, the influence of multi factor coupling such as composition on irradiation swelling of austenitic stainless steel will help to promote the material optimization design of austenitic stainless steel cladding material.

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  • Matic KUNŠEK, Leon CIZELJ, Ivo KLJENAK
    Article ID: 23-00488
    Published: 2024
    Advance online publication: January 24, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Theoretical simulations of dispersed solid particle behaviour inside a scrubbing pool within the bubble rise region are presented. The goal is to evaluate the decontamination factor of the particles during the pool scrubbing process. The basic phenomena of pool scrubbing are described. The setup used for the simulation validation is presented. Then, the boundary and initial conditions of the PECA experiments, which were performed at CIEMAT (Madrid, Spain) and were used for simulations, are presented. The subgrid model for decontamination through transfer of particles from gas bubbles to the surrounding liquid is described. The calculation results are evaluated and compared with the part of the PECA experimental results to which the proposed modelling is applicable.

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  • Reo KAI, Hiroaki WATANABE, Ryoichi KUROSE
    Article ID: 23-00400
    Published: 2023
    Advance online publication: January 12, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Effects of species diffusion models on the laminar burning velocity SL of lean hydrogen-air premixed flame are investigated by performing one-dimensional numerical simulations of lean hydrogen-air premixed flames at an equivalence ratio of 0.5. Maxwell-Stefan (MS) diffusion, mixture-averaged (MA) diffusion, and unity Lewis number diffusion are compared as the concentration diffusion models at a pressure of 0.1 MPa. Moreover, the contribution of the species thermal diffusion is also investigated under three different pressure (P) and unburnt gas temperature (Tu) conditions. Results show that the MA diffusion well reproduces the results of MS diffusion including SL while saving computational cost to three-fourths. The unity Lewis number diffusion overestimated SL by 30 % because of the underestimation of the mass flux of H2. By considering the species thermal diffusion, SL decreases by 6.4 % and 3.0 % under the reference (P = 0.1 MPa and Tu = 300 K) and HPT (P = 2 MPa and Tu = 673 K) conditions, respectively, and increases by 1.3 % under the HP condition (P = 2 MPa and Tu = 300 K). Under the reference and HPT conditions, the species thermal diffusion removes the H radical from the reactive region. This mitigates the chain-branching reaction of H + O2 → OH + O and decreases SL. On the other hand, under the HP condition, the species thermal diffusion supplies the H radical toward the reactive region. This enhances the aforementioned chain-branching reaction and increases SL. Under the HP condition, the recombination reaction of H + O2 + M → HO2 + M is enhanced by a high molar concentration of third body M because of high gas density. The large contribution of this reaction to the consumption of H radicals in the high-temperature region makes peak positions of mass fraction of H and mass flux of H by the species thermal diffusion lower temperature side, which leads to the supply of H radicals to the reactive region and the increase in SL.

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  • Takahiro ARAI, Masahiro FURUYA, Erik De MALMAZET
    Article ID: 23-00475
    Published: 2023
    Advance online publication: January 12, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Ex-vessel molten fuel-coolant interaction is important for evaluating the integrity of the reactor containment vessel during a nuclear power plant accident. When molten metal discharged from the reactor vessel interacts with water in the containment vessel, the oxidation of the melt in water affects the occurrence and magnitude of a steam explosion. This study presents a small-scale experiment in which molten droplets of stainless steel mixed with zirconium for compositions ranging from 0 to 30 wt% zirconium are dropped into a water pool to clarify the oxidation characteristics of the molten droplets in water. The solidified particles with a relatively spherical shape for different metal compositions and initial pool water temperatures are obtained, and the oxygen content of the solidified particles, the thickness of the oxide layer, and the elemental distribution of the solidified particle cross-sections are evaluated. The oxide layer thickness on the surface of solidified particles and the oxygen weight percentage of solidified particles increased with the increase of zirconium content or the increase of initial pool water temperature.

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  • Hiromichi MIURA, Shota SOGA, Eishiro HIGO
    Article ID: 23-00415
    Published: 2023
    Advance online publication: January 11, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    It is crucial to perform a multi-unit probabilistic risk assessment (MUPRA) to identify the risk of a nuclear power station (Hereafter referred to as site) with multiple units. External hazards play a significant role in MUPRA because they simultaneously affect multiple units. In Japan, an earthquake is a critical external hazard in site risk. In this study, we propose a method to assess seismic-induced multi-unit initiating events (SIMUIEs). To utilize insights and results from seismic single-unit probabilistic risk assessment (SUPRA), the proposed method identifies SIMUIEs and estimates their frequencies by extending an initiating event classification tree method (also known as a hierarchical event tree method) commonly used in Japan. The proposed method generates SIMUIEs by combining seismic initiating events considered in seismic SUPRAs. In the SIMUIE generation process, the following factors specific to a multi-unit site are considered: (1) failure of shared equipment among units, (2) cascading effects across units, and (3) correlation of responses and correlation of capacities of structures, systems, and components (SSCs) among units. Then, the SIMUIEs are identified by the combinations considering these factors. In quantification, a logical model is built for each SIMUIE, and they are quantified by considering the dependences among seismic-induced initiating events. In addition, a preliminary evaluation for two nuclear power reactors demonstrates how the proposed method can identify and quantify SIMUIEs.

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  • Motohiro NAKAHATA, Noor SAFFREENA, Akiko KANEKO
    Article ID: 23-00482
    Published: 2023
    Advance online publication: January 11, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Direct contact condensation (DCC) is a phenomenon that occurs when vapor is injected into a subcooled water pool and is essential in many industrial devices. However, the actual phenomenon involves non-condensable gases in the vapor, and there has been insufficient research on how non-condensable gases affect direct contact condensation. In this study, through the detailed observation of the bubble collapse behavior using a high-speed video camera, we aimed to clarify the bubble collapse behavior and heat transfer coefficient of vapor bubbles containing non-condensable gas by DCC. Experiments were compared between DCC under pure vapor conditions and DCC under conditions containing non-condensable gas. Characteristic quantities for each condition were calculated from image analysis of the captured images. The temperature distribution inside the plume was also measured using thermocouples. From these data, the relationship between contraction rate and microbubbles diameter and the average heat transfer coefficient (HTC) were estimated. As a result, the construction rate of the plume containing non-condensable gas was reduced to 4 ~ 20 % of that of pure vapor, and large number of microbubbles were observed under lower air mass fraction with the log-normal distribution in diameter. The average HTC decreased to about 5 ~ 13 % when the non-condensable gas was mixed with about 8 % of air mass fraction. These results were proposed to be because vapor concentration decreases as the plume shrinks and vapor diffusion is suppressed by the mass conservation of non-condensable gas as the vapor condenses.

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  • Hiroshi NAGANUMA, Takehito MORI, Sho WATANABE, Akihiro SAWADA, Taeko G ...
    Article ID: 23-00486
    Published: 2023
    Advance online publication: January 11, 2024
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    In order to decrease ash deposition and hot corrosion in Waste-to-Energy plants, it is necessary to understand mechanisms of the ash deposition and the corrosion, and to develop new solutions. In this study, an adhesion force between an ash pellet and an alloy specimen were firstly measured to investigate the increase mechanisms of the ash deposition and the corrosion, using a tensile testing machine with an electrical furnace. Second, mass loss of the alloy specimens was measured as hot corrosion tests based on Japanese Industrial Standards. Third, phenomenon on the interface was evaluated with production amount of molten salts which were calculated by thermodynamic equilibrium. As a result, the adhesion force and the mass loss of all specimens increased with the interface temperature, and there was exact materials dependency, which indicated that a high chromium content design to improve corrosion resistance was valid also for decrease of the ash adhesion. Moreover, it was observed that mass loss of the alloy specimens with the ash sample collected from an actual boiler had a temperature dependence, and molten salts and accelerated oxidation induced by chlorides attacked the alloy specimens. In addition, generation of HCl gas and molten salts with temperature increase was confirmed with the thermodynamic equilibrium calculations. Judging from these results, starting temperature of the adhesion could be correspond to one which the alloy started to be severely corroded.

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  • Kota FUJIWARA, Yasuo HATTORI, Yuzuru EGUCHI
    Article ID: 23-00403
    Published: 2023
    Advance online publication: December 27, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Tornadoes represent a substantial hazard to nuclear power plant infrastructures, particularly in terms of potential damage caused by wind-borne debris. A comprehensive understanding of the three-dimensional flow structure of tornadoes is essential to assess such hazards. Ward-type tornado-like vortex (TLV) generators have been used to provide experimental insight into the influence of wind loading on potential debris and infrastructures. However, these experiments address challenges arising from highly turbulent and three-dimensional flow structures. Recent advancements in the area of TLVs have seen a surge in numerical analyses. Three-dimensional simulations are capable of reproducing the flow field, although the computational requirements appear to be prohibitive for hazard assessment applications. A numerically proficient model that could reproduce the velocity profile from TLV generators is highly regarded. To achieve this, axisymmetric modeling of TLV in a Ward-type chamber is attempted. This paper uses the VorTECH facility at Texas Tech University, a large-scale Ward-type TLV generator, as a reference case. Through a series of simulations, the critical role of the mesh resolution in the development of the boundary layer is revealed. The axisymmetric model underestimated the experimentally observed pressure at the core center while showing good agreement for the core width and velocity distribution for both single-cell and two-cell TLV flow fields. Through alterations in the domain configuration, including the influence of the inflow, floor roughness, and the presence of a honeycomb rectifier, the effect of such factors on the velocity profiles near the touchdown region was identified. The boundary layer development by the inflow inside the confluence region, the floor roughness of the touchdown region, and flow resistance at the rectifier should be carefully discussed to develop a reproducible axisymmetric TLV model.

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  • Jiaying CHENG, Tong ZHU, Biao DENG
    Article ID: 23-00316
    Published: 2023
    Advance online publication: December 26, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Hydrogen is a promising carbon-free fuel, but nitric oxides (NOx) emissions are significantly increased with higher temperature in hydrogen combustion. Internal flue gas recirculation (IFGR) is one of the most effective NOx reduction techniques in boilers. Previous research has widely reported NOx generation with hydrogen enrichments from a chemical kinetic perspective, however, the NOx formation principles in non-premixed methane/air combustion using IFGR with hydrogen addition are still unclear. This work aims to investigate the effects of hydrogen addition on NOx formation in a non-premixed methane low-NOx combustor using IFGR. The Reynolds-averaged Navier-Stokes (RANS) simulations were conducted on a 5 kW non-premixed combustor with IFGR rate of 17.7%, which generated 11.3 mg/m3 of total NOx generation in methane combustion. A series of three-dimensional computational fluid dynamics (CFD) simulations with detailed mechanisms was tested on four hydrogen power fractions. The reaction intensities of the main NOx formation pathways were analyzed on reaction rates. The results show that 70% hydrogen power fractions lead to significantly greater NOx concentration to 677.8 mg/m3, as high as 60 times of the original concentration. With no additional techniques, the original methane low-NOx combustor is only allowed for hydrogen addition of less than 20% power fraction. The NNH route becomes the second dominant pathway in hydrogen-added flames. Influenced by the growing hydrogen contents, the generation of the NNH route is largely generated from 2 mg/m3 to 65 mg/m3. This work connects the hydrogen addition and the NOx formation pathways in non-premixed methane combustion, and highlights the importance of eliminating thermal NOx and the NNH route to achieve low-NOx combustion rather than sole NOx suppression method, which can provide a reference for designing low-NOx techniques.

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  • Filippo BENTIVEGNA, Alberto BECCANTINI, Pascal GALON, Christophe CORRE
    Article ID: 23-00474
    Published: 2023
    Advance online publication: December 22, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Loss Of Coolant Accident causes the propagation of a transient rarefaction wave within the primary circuit that generates a transient pressure load on the baffle surrounding the reactor core. This is the result of nonidentical travel times of the rarefaction wave between the two sides of the baffle: the reactor core on one side and the by-pass between the baffle and core barrel on the other. The two zones have different geometrical characteristics, in particular the perforated reinforcement plates in the by-pass significantly influence wave propagation. Representing these obstacles in numerical simulations of the primary circuit requires the use of simplified models. A study on the accuracy of these models is hereby proposed, through a detailed comparison between numerical simulations realized with the EUROPLEXUS software and experimental results obtained on the MADMAX facility. The numerical models investigated are both quite reliable for the simulation of the case study under consideration. Both models imply a simplified representation of the real geometry of the experimental device, but despite this they can be considered quite valid. Lastly, the Fluid-Structure Interaction calculations enable us to comprehensively assess the phenomenon of cavitation and provide an initial evaluation of the deformations and mechanical stresses experienced by the structure.

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  • Takafumi OKITA, Satoshi TAKEDA, Eiji HOASHI, Takanori KITADA
    Article ID: 23-00494
    Published: 2023
    Advance online publication: December 22, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    In order to reduce Minor Actinides (MA), the transmutation of MA in a fast reactor has been studied in Japan. In the transmutation in the fast reactor, MAs are mixed in the ordinary MOX fuel and they are burned up. For the design of a fast reactor with inherent safety characteristics, it is necessary to secure negative void reactivity and cooling via a natural convection under the situation of the severe accident. Therefore, this study aims to fundamentally design the fast reactor for the MA transmutation with inherent safety characteristics. First, in order to clarify the design range of the core design parameter, the parametric study was conducted using four design parameters determining the core configuration. As the result of this parametric study, the feasible combination of each four parameters satisfying the negative void reactivity was clarified. Among parameters satisfying the negative void reactivity, the condition that results in the maximum reduction of MA amount was obtained. Secondly, the CFD simulation focused on the sodium coolant channel between fuel pins was conducted. In this CFD simulation, as the first step for the design activity, the confirmation of the occurrence condition of a natural convection due to the decay heat after shifting to the transient situation was confirmed. Temperature and velocity distribution were evaluated. And the heat transfer efficiency was summarized using Ra number and Nu number.

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  • Hiroyuki NISHINO, Kenichi KURISAKA, Kenichi NARUTO, Yoji GONDAI, Masay ...
    Article ID: 23-00409
    Published: 2023
    Advance online publication: December 20, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    The effectiveness evaluation of safety measures against severe accident is necessary for restart of experimental sodium-cooled fast reactor Joyo in Japan. These safety measures correspond to those in defense-in-depth (DiD) level 4. In the previous study, an internal event level-1 probabilistic risk assessment (PRA) at power was performed to calculate frequencies of the accident sequences of failure of safety measures in DiD levels 1 to 3, to identify dominant accident sequence groups, and to identify dominant accident sequence for selecting important accident sequences in each accident sequence group, which are needed for implementing the effectiveness evaluation of safety measures in DiD level 4. Based on this, the present study implemented internal event level-1 PRA at power to show quantitatively reduction of those occurrence frequency by the safety measure in the DiD level 4. As a result, the frequency of each accident sequence group decreased significantly, and the total frequency of the accident sequence groups decreased to about 1×10-6 /reactor-year, which is about 1/1000 times the one estimated in the previous study. The protected loss of heat sink was the largest contributor in all the accident groups, and a dominant accident sequence in each accident group was also identified in this study.

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  • Nicolò BADODI, Marica EBOLI, Antonio CAMMI, Alessandro DEL NEVO
    Article ID: 23-00416
    Published: 2023
    Advance online publication: December 20, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    In the framework of the development of fusion energy, one of the most prominent technologies arising to address the issues of tritium breeding and power conversion is the Water-Cooled Lithium-Lead (WCLL). This technology utilizes a molten eutectic alloy of Lithium and Lead which circulates inside the Breeding Blanket (BB) and is irradiated with neutrons to produce tritium. Water is then circulated inside the system to cool the components. The simultaneous presence inside critical areas of the reactor of molten metal alloy and water, at high temperature and pressure, poses significant safety concerns. For this reason, adequate design and analysis techniques are required to ensure the ability of the system to survive and mitigate any possible damage in case of the in-box Loss of Coolant Accident (LOCA), the most critical postulated accidental scenario. With this aim in mind, a novel approach was implemented with the aim of coupling the SIMMER-III code and the ANSYS Mechanical code for the modelling of both the chemical and thermodynamical interactions between water and the alloy, and the resulting effects on the structures. This work presents the status of the coupling technique development and the results of the preliminary validation activities performed against experimental data provided by the LIFUS5 facility operating at ENEA Brasimone Research Centre. The resulting comparison between these data and the codes’ predictions allows a careful evaluation of the errors introduced in each step of the chain. Moreover, it provides confidence in the capacity of the methodology to correctly predict the ability of the structures to withstand incidental loads without suffering extensive damage.

    This work aims at providing engineers with a usable and powerful tool that allows for the safety analysis of WCLL-based components during the early stages of the design phase. This would help save time, and effort and reduce the economic cost that might arise from any undetected issue propagating downstream the design process.

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  • Masaaki TANAKA, Yasuhiro ENUMA, Yasushi OKANO, Akihiro UCHIBORI, Kenji ...
    Article ID: 23-00424
    Published: 2023
    Advance online publication: December 13, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    Japan Atomic Energy Agency (JAEA) is developing an artificial intelligence (AI) aided integrated digital system: “Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)”, to provide the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle. Until 2023, the platform for a common function and subsystems, namely ARKADIA-Design for design study, ARKADIA-Safety for safety assessment, and ARKADIA-knowledge management system (KMS) for the knowledge base, are separately being developed. This paper describes the development concepts of the platform, the progress of the application study of design optimization in ARKADIA-Design, the progress of optimization model development and optimization functionality based on AI technology in ARKADIA-Safety, and the structure of the knowledge base and application of AI technology to ARKADIA-KMS. In subsystems, necessary modules and process integration to obtain an optimal solution in a designated problem of plant design have been advanced. For further development, a strategy for unifying the subsystems with the AI-aided platform into one system, ARKADIA, and the extension of capabilities of the numerical analyses and evaluation technologies required for plant design until 2028 are presented.

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  • Erina HAMASE, Kazuki KUWAGAKI, Norihiro DODA, Kenji YOKOYAMA, Masaaki ...
    Article ID: 23-00440
    Published: 2023
    Advance online publication: December 13, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    A core design optimization process is developed as part of the design optimization support tool named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) for an efficient and innovative core design process. The process comprises analyses integrated by neutronics, thermal-hydraulics in a fuel assembly, fuel integrity, and plant dynamics for safety assessment. The optimal design parameters are explored using the Bayesian optimization (BO) algorithm to reduce the number of iterative calculations and solve the optimization problem. This study defines a representative problem by identifying the objective functions, constraints, and design parameters for an actual core design based on previous core design experiences to efficiently develop the core design optimization process. Next, a constrained single-objective optimization problem, the simplified representative problem, is solved by the integrated analyses only with neutronics and plant dynamics using the BO algorithm to confirm the applicability of the optimization process for the representative problem. Consequently, the design parameters that optimized the objective function within constraints can be obtained. The optimal solution correlates well with the reference solution. Furthermore, the effectiveness of the optimization process is discussed by comparing an ordinary and a defined core design process.

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  • Shin-etsu SUGAWARA
    Article ID: 23-00375
    Published: 2023
    Advance online publication: December 07, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    The recent trend of nuclear reactor miniaturization may require reconsideration of the existing framework of safety. This study conceptually explores the safety goals for transportable microreactors (TMRs) by focusing on the differences between large light-water reactors (LLWRs), which contain large amounts of hazardous fission products and TMRs. For LLWRs, the safety goals and surrogate goals representing the integrity of the reactor have played a significant role in reducing the negative health effects of radiation exposure in cases of nuclear disasters. Practitioners, notably the operator, have typically been classified as the main users of these goals. However, the innovative feature of TMRs will lead to the reconsideration of the contents and users of safety goals. The size of the radiological consequences of TMR accidents may highlight the need to capture broader consequences other than direct health effects when formulating the top-level goals. Correspondingly, additional surrogates for representing the interplay between the reactor and surrounding areas may be required. Effectively meeting these new goals only by the efforts of licensees may be a challenge; this indicates a need for the local actors wherein the TMRs are deployed to become the extended users of safety goals. Conceptualizing such new framework of microreactor safety goals as “extended safety goals” as an extension of conventional safety goals for LLWRs, the author discusses their implications and challenges.

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  • Kouji HIRAIWA, Rei KIMURA, Satoshi WADA, Tsukasa SUGITA, Kenichi YOSHI ...
    Article ID: 23-00381
    Published: 2023
    Advance online publication: December 07, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    We analyzed the possibility of TRU generation reduction fuel using HALEU as the initial enrichment to reduce the radiotoxicity and decay heat of TRU when using the LWR fuels that can suppress radiotoxicity and decay heat in HLW. Fade-out time was defined as the period required to decay to the level of natural uranium ore, focusing on the radiotoxicity, and decay heat up to one million years, which is generated from the reduced TRU generation type fuel (FORSETI) consisting of uranium fuel of HALEU enrichment. The decay characteristics of TRU nuclides in HLW after plutonium removal by reprocessing and the effect of shortening the fade-out time were mainly discussed. The level of 241Am on both radiotoxicity and decay heat is significant. However, the effect of 239Pu and 240Pu, which are produced as daughter nuclides of 243Am and 244Cm after reprocessing, significantly lengthens the fade-out time to 100,000 years for radiotoxicity and to 40,000 years for decay heat, compared to several thousand years for 241Am alone. When TRU generation reduction fuel is adopted, the production rate of 243Am and 244Cm is reduced by more than 90% when the enrichment is increased from 3.8 wt% to 20 wt%, and the generation of 239Pu and 240Pu as daughter nuclides are also reduced almost correspondingly. As a result, increasing the initial 235U enrichment from 3.8 wt% to 20 wt% reduces the fade-out time of radiotoxicity from 100,000 to 3,000 years and the fade-out time of decay heat from 40,000 to 2,000 years for HLW, respectively by FORSETI-type TRU generation reduction fuel. As described above, it is shown that using TRU generation reduction fuel can significantly accelerate the decay of TRU radiotoxicity and decay heat in HLW without transmutation of TRU.

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  • Narihiro SUZUO, Qiusheng LIU, Makoto SHIBAHARA
    Article ID: 23-00372
    Published: 2023
    Advance online publication: December 03, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    The understanding of the heat transfer process for helium gas cooling in the fusion blanket is important for the development of fusion reactors. This research aims at clarifying the transient heat transfer characteristics of helium gas for turbulent flow in a narrow tube. A circular platinum tube with an inner diameter of 1.8 mm was heated by exponentially increasing heat inputs and cooled by helium gas. The lengths of the tube were 30 mm and 50 mm. The heat transfer coefficients gradually increased for the e-folding time shorter than about 1.5 s. This shows that the heat transfer process is in a transient state for this region. The heat transfer coefficients increased with the increase of flow velocity, and the effect of the flow velocity became weak in the transient state region. By comparing with the experimental results of tube with different lengths, it was obtained that the quasi-steady state heat transfer coefficients for the length of 30 mm were higher than those for the lengths of 50 mm and 90 mm. However, the heated length showed a weak influence for the lengths of 30 mm and 50 mm in the transient heat transfer with a relatively smaller e-folding time. The variation of transient Nusselt number with Fourier number at different Reynolds number describes well the trend for the variation of heat transfer coefficient with e-folding time at different velocities. Based on the experimental data, an empirical correlation was obtained for transient heat transfer at different lengths by using Fourier number.

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  • Kenichi KURISAKA
    Article ID: 23-00377
    Published: 2023
    Advance online publication: November 29, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    This study aims to understand the time-dependent change in the occurrence rate of leak from steam generator (SG) tubes in sodium-cooled fast reactors (SFRs). The target SFRs in the present paper are Phenix in France and BN-600 in Russia. By reviewing publicly available literature that show data from the SFRs, we have investigated the numbers of tube-to-tubeplate welds and tube-to-tube welds, heat transfer areas of tube base metal, operating hours of SGs, dates when SG tube leak occurred, locations of leak, and corrective actions taken after tube leak events, such as replacement of the module, in which a leak occurred. Based on these, we have estimated the time to leak and quantitatively analyzed the time-dependent change of the occurrence rates of SG tube leak for each of the above-mentioned parts by hazard plotting method. The results show the time-dependent change in the rates of both Phenix and BN-600. For Phenix, the rate of leak at tube-to-tube welds shows a significant increase after 40,000 hours. This can be caused by thermal stress repeatedly exerted on the materials. In addition, the decrease in the rate was found, which is probably thanks to improved welding and SG operating conditions. For BN-600, it seems that in many cases, the probable cause of the leak was initial defects that developed to failure during the early stage of reactor operation, and that no special countermeasure was taken in the later stages. Therefore, it would be natural to assume that the rate simply decreased over time.

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  • Kenta KAKITANI, Wataru SUGINO, Yusuke NAKANO, Kenji SATO, Yuichi SHIMI ...
    Article ID: 23-00317
    Published: 2023
    Advance online publication: October 28, 2023
    JOURNAL OPEN ACCESS ADVANCE PUBLICATION

    The replacement of LiOH with KOH for pH control in the primary water of pressurized water reactors (PWRs) is being considered due to the growing cost of enriched 7Li. This study aims to investigate the susceptibility of the primary water stress corrosion cracking (PWSCC) initiation in simulated primary water with KOH. Firstly, the susceptibilities in KOH and LiOH chemistries were compared by conducting uniaxial constant load tests on Alloy X-750 at 360°C. The results showed no significant difference in the time to initiation between the two chemistries. Secondly, the effect of dissolved hydrogen (DH) concentration in the KOH chemistry was examined. The DH concentration of the test water was varied at 5, 30 or 45 ml/kg-H2O. The results showed that the time to initiation of PWSCC was significantly extended under the low DH condition (5 ml/kg-H2O). This observed effect of DH concentration in the KOH environment agrees with the previously reported effect observed in the conventional LiOH environment. To investigate the mechanisms underlying the PWSCC tests, the oxide films on the test specimens were characterized using electron microscopes. The oxide films formed in the KOH and LiOH chemistries did not show significant differences. Additionally, under the low DH condition, the occurrence of selective internal oxidation under the inner oxide film was relatively minor. The results suggest that the use of KOH would not have an adverse effect on PWSCC initiation, and PWSCC initiation can be mitigated with a low concentration of DH in the KOH environment as well as in the LiOH environment.

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