• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2427-2434
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• 2022

Series of unequal quantity Nd/Ce co-doped ceramic nuclear waste forms, (Gd, Nd)₂(Zr, Ce)₂O₇, were prepared to tailor its ordered pyrochlore or disordered fluorite structure. The phase transition, microtopography, and elemental composition of the ceramic samples were systematically investigated, especially the effect of order-disorder structure on the chemical stability. It was confirmed that unequal quantity of Nd/Ce could synchronously replace the Gd/Zr-sites of Gd₂Zr₂O₇. And the phase transition of order-disorder structure could be successfully tailored by regulating the average cationic radius ratio of (Gd, Nd)₂(Zr, Ce)₂O₇ series. The elements of Gd, Nd, Zr, and Ce are uniformly distributed in the ordered or disordered structures. The MCC-1 leaching results showed that (Gd, Nd)₂(Zr, Ce)₂O₇ pyrochlore ceramic nuclear waste forms had excellent chemical stability, whose elements' normalized leaching rates were as low as 10^-4-10^-7 g·m^-2·d^-1 after 7 days. In particular, the chemical stability of disordered structure was superior to that of ordered structure. It was proposed that the force constant and the closest packing were changed with the structure transformation resulting the chemical stability difference.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.666-673
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• 2022

The optimization of a transmission type bremsstrahlung conversion target was carried out with Monte Carlo code FLUKA for intense pulsed electron beams with electron energy of several hundred keV for maximum photon fluence. The photon emission intensity from electrons with energy ranging from 300 keV to 1 MeV on tungsten, tantalum and molybdenum targets was calculated with varied target thicknesses. The research revealed that higher target material element number and electron energy leads to increased photon fluence. For a certain target material, the target thickness with maximum photon emission fluence exhibits a linear relationship with the electron energy. With certain electron energy and target material, the thickness of the target plays a dominant role in increasing the transmission photon intensity, with small target thickness the photon flux is largely restricted by low energy loss of electrons for photon generation while thick targets may impose extra absorption for the generated photons. The spatial distribution of bremsstrahlung photon density was analyzed and the optimal target thicknesses for maximum bremsstrahlung photon fluence were derived versus electron energy on three target materials for a quick determination of optimal target design.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1295-1300
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• 2022

There has been increased interest in researching risk perception of radiation to implement successful risk communication, particularly given the recent worldwide nuclear policy movement regarding nuclear energy. This study aimed to investigate characteristics of risk perception among residents living near normally operating nuclear power plants in South Korea by identifying factors associated with risk perception. A survey was conducted with face-to-face interviews for 1200 residents aged 20e84 years by gender- and age-stratified random sampling. Risk perception was associated with trust perception in nuclear safety, but was not highly correlated with benefit perception for utilizing nuclear power. Relatively high risk perception was observed in women, older age groups, and residents not having experience of nuclear-related education or work. This association remained after adjusting for other factors including benefit perception, trust perception, and psychological distress. In addition to these individual characteristics, risk perception was also associated with a residential district's own unique context, indicating that a strategy of risk communication should be developed differently for residents facing nuclear-related circumstances. Given that risk perception can be changed, depending on social values such as safety culture and economic setting, further studies are required to understand the changing characteristics of radiation risk perception.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.589-600
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• 2022

Investigations on the molten-pool sloshing behavior are of essential value for improving nuclear safety evaluation of Core Disruptive Accidents (CDA) that would be possibly encountered for Sodium-cooled Fast Reactors (SFR). This paper is aimed at synthesizing the knowledge from our recent studies on molten-pool sloshing behavior with solid particles conducted at the Sun Yat-sen University. To better visualize and clarify the mechanism and characteristics of sloshing induced by local Fuel-Coolant Interaction (FCI), experiments were performed with various parameters by injecting nitrogen gas into a 2-dimensional liquid pool with accumulated solid particles. It was confirmed that under different particle-bed conditions, three representative flow regimes (i.e. the bubble-impulsion dominant, transitional and bed-inertia dominant regimes) are identifiable. Aimed at predicting the regime transitions during sloshing process, a predictive empirical model along with a regime map was proposed on the basis of experiments using single-sized spherical solid particles, and then was extended for covering more complex particle conditions (e.g. non-spherical, mixed-sized and mixed-density spherical particle conditions). To obtain more comprehensive understandings and verify the applicability and reliability of the predictive model under more realistic conditions (e.g. large-scale 3-dimensional condition), further experimental and modeling studies are also being prepared under other more complicated actual conditions.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2720-2727
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• 2022

The Generation IV Lead-cooled fast reactor (LFR) take the liquid lead or lead-bismuth eutectic alloy (LBE) as the coolant of the primary cooling circuit. Combined with the natural characteristics of lead alloy and the design features of LFR, the system is the simplest and the number of equipment is the least, which reflects the inherent safety characteristics of LFR. The nuclear main coolant pump (MCP) is the only power component and the only rotating component in the primary circuit of the reactor, so the various operating characteristics of the MCP are directly related to the safety of the nuclear reactor. In this paper, various working conditions that may occur in the normal rotation (positive rotating) of the MCP and the corresponding internal flow characteristics are analyzed and studied, including the normal pump condition, the positive-flow braking condition and the negative-flow braking condition. Since the corrosiveness of LBE is proportional to the fluid velocity, the distribution of flow velocity in the pump channel will be the focus of this study. It is found that under the normal pump condition and positive-flow braking conditions, the high velocity region of the impeller domain appears at the inlet and outlet of the blade. At the same radius, the pressure surface is lower than the back surface, and with the increase of flow rate, the flow separation phenomenon is obvious, and the turbulent kinetic energy distribution in impeller and diffuser domain shows obvious near-wall property. Under the negative-flow braking condition, there is obvious flow separation in the impeller channel.

• Archives of Craniofacial Surgery
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• v.23 no.5
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• pp.205-210
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• 2022

Background: Nasal bone fractures are the most common type of facial bone fracture, but are under-studied in adults above 65 years of age. Therefore, we investigated the epidemiology and patterns of nasal bone fractures among older adults in comparison to different age groups. Methods: This retrospective study included 2,321 nasal bone fracture patients who underwent surgery at our hospital from January 2010 to December 2017. The patients were classified by age as preschoolers, school-age children, young and middle-aged adults, and the elderly. We performed pairwise comparisons between elderly patients and each other age group in terms of sex, cause of injury, and fracture type. Results: The 2,321 nasal bone fracture patients included 76 elderly patients (50 men [65.8%] and 26 women [34.2%]). In these patients, the two most common injury causes were falling or slipping down (n= 39; 51.3%) and road traffic accidents (n= 19; 25.0%). According to the Stranc and Robertson classification, the most common force vector was lateral, and plane 2 fractures with lateral forces predominated. Conclusion: The elderly showed similar patterns of nasal bone fractures to those observed in young and middle-aged adults, but significant differences from preschoolers (in the injury vector and plane of fracture) and from school-age children (in the sex ratio and plane of fracture). However, elderly patients presented significantly different epidemiological characteristics compared to the other three groups. Therefore, it is necessary to improve the quality of life of the elderly and prepare for the upcoming super-aged society by taking steps to reduce the incidence and severity of fractures. Possible options for doing so include strengthening individual-level safety factors and expanding the social safety net for the elderly.

• Geomechanics and Engineering
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• v.27 no.6
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• pp.537-550
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• 2021

There are various technical problems need to be solved in the construction process of pre-setting an isolation wall into a double lane in the outburst prone mine. This study presents a methodology that pre-setting an isolation wall into a double lane without a coal pillar. This requires the excavation of two small section roadways to dig a wide section roadway, followed by construction of the separation wall. During this process the connecting lane is reserved. In order to ensure the stability of the separation wall, the required bearing capacity of the isolation wall is 4.66 MN/m and the deformation of the isolation wall is approximately 25 cm. To reduce the difficulty of implementing support the roadway is driven by 5 m/d. After the construction of the separation wall, the left side coal wall is brushed 1.5 m to make the width of the gas roadway reach 2.5 m and the roadway support utilizes anchor rod, ladder beam, anchor cable beam and net configuration. During construction, the concrete pump and removable self-propelled hydraulic wall mold are used to pump and pour the concrete of the isolation wall. In the process of mining, the stress distribution of coal body and isolation wall is detected and measured on site. The results demonstrate that the deformation of the surrounding rock of roadway and separation of roof in the roadway is small. The stress of the bolt and anchor cable is within equipment tolerance validating their selection. The roadway is well supported and the intended goal is achieved. The methodology can be used for reference for similar mine gas control.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3879-3891
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• 2021

A heat pipe residual heat removal system is proposed to be incorporated into the reactor driven subcritical (RDS) facility, which has been proposed by MIT Nuclear Reactor Laboratory for testing and demonstrating the Fluoride-salt-cooled High-temperature Reactor (FHR). It aims to reduce the risk of the system operation after the shutdown of the facility. One of the main components of the system is an air-cooled heat pipe heat exchanger. The alkali-metal high-temperature heat pipe was designed to meet the operation temperature and residual heat removal requirement of the facility. The heat pipe model developed in the previous work was adopted to simulate the designed heat pipe and assess the heat transport capability. 3D numerical simulation of the subcritical facility active zone was performed by the commercial CFD software STAR CCM + to investigate the operation characteristics of this proposed system. The thermal resistance network of the heat pipe was built and incorporated into the CFD model. The nominal condition, partial loss of air flow accident and partial heat pipe failure accident were simulated and analyzed. The results show that the residual heat removal system can provide sufficient cooling of the subcritical facility with a remarkable safety margin. The heat pipe can work under the recommended operation temperature range and the heat flux is below all thermal limits. The facility peak temperature is also lower than the safety limits.