bluuebunny/crossref_metadata_2025 huggingface dataset:
Global Initiative for Asthma Strategy 2021
Authors: Reddel Helen K. , Bacharier Leonard B. , Bateman Eric D. , Brightling Christopher E. , Brusselle Guy G. , Buhl Roland , Cruz Alvaro A. , Duijts Liesbeth , Drazen Jeffrey M. , FitzGerald J. Mark , Fleming Louise J. , Inoue Hiromasa , Ko Fanny W. , Krishnan Jerry A. , Levy Mark L. , Lin Jiangtao , Mortimer Kevin , Pitrez Paulo M. , Sheikh Aziz , Yorgancioglu Arzu A. , Boulet Louis‐Philippe
Published: October 2021

The Global Initiative for Asthma (GINA) Strategy Report provides clinicians with an annually updated evidence‐based strategy for asthma management and prevention, which can be adapted for local circumstances (e.g., medication availability). This article summarizes key recommendations from GINA 2021, and the evidence underpinning recent changes.

GINA recommends that asthma in adults and adolescents should not be treated solely with short‐acting β2‐agonist (SABA), because of the risks of SABA‐only treatment and SABA overuse, and evidence for benefit of inhaled corticosteroids (ICS). Large trials show that as‐needed combination ICS–formoterol reduces severe exacerbations by ≥60% in mild asthma compared with SABA alone, with similar exacerbation, symptom, lung function, and inflammatory outcomes as daily ICS plus as‐needed SABA.

Key changes in GINA 2021 include division of the treatment figure for adults and adolescents into two tracks. Track 1 (preferred) has low‐dose ICS–formoterol as the reliever at all steps: as needed only in Steps 1–2 (mild asthma), and with daily maintenance ICS–formoterol (maintenance‐and‐reliever therapy, “MART”) in Steps 3–5. Track 2 (alternative) has as‐needed SABA across all steps, plus regular ICS (Step 2) or ICS–long‐acting β2‐agonist (Steps 3–5). For adults with moderate‐to‐severe asthma, GINA makes additional recommendations in Step 5 for add‐on long‐acting muscarinic antagonists and azithromycin, with add‐on biologic therapies for severe asthma. For children 6–11 years, new treatment options are added at Steps 3–4.

Across all age groups and levels of severity, regular personalized assessment, treatment of modifiable risk factors, self‐management education, skills training, appropriate medication adjustment, and review remain essential to optimize asthma outcomes.
Managing patients with developmental co‐ordination disorder in dentistry: Developing an online resource for dental professionals by a review of the literature
Authors: Nash Julie , Woolley Stephen
Published: October 2021

Aim: To produce an online resource for dental professionals, advising them on ways to manage patients with Developmental Coordination Disorder (DCD).

Method and Results: Literature search into the management of patients with DCD, and how to produce a high‐quality leaflet using specific keywords. Using online databases, such as PubMed, the Cochrane Database and an internet search engine, an online resource in printable leaflet form was produced following a pilot readability assessment and review by those who work with individuals with DCD and a Special Care Dentistry special interest group. From the assessment tools used, the resource scored well in terms of readability and comprehension. The resource also received positive and constructive feedback from colleagues and those who work with individuals with DCD.

Conclusion: An online resource was produced for dental professionals, although further evaluation is required on whether it will be useful to the profession. The literature review suggests the need for more research to be carried out on the association between DCD and oral health, and how dental professionals can manage those with DCD within a general dental practice.
Enhancing the inpatient ward learning environment: A practical guide
Authors: Carter Adam Nicholas , Gifford Albert , Lyons Amy Victoria , Milton Joseph Sebastian , Streeter Hamish , Carter Matthew
Published: October 2021

Inpatient ward placements are rich with practical learning opportunities for medical students; however, many such opportunities go overlooked and underused. Newly qualified doctors often feel underprepared for work on wards; improving student experience on ward placements can address this. Although this requires an active effort from both students and clinical educators, it is not arduous and has mutual benefit: improving medical education while simultaneously reducing staff workload. Here, we present a guide for both teachers and students highlighting three key areas of ward learning that may be improved: strategies to drive active learning, integration of students into the ward team and underutilised resources to develop a patient‐centred approach.
Pre‐dam valley reconstruction based on archival spatial data sources: Methods, accuracy, and 3D printing possibilities
Authors: Pacina Jan , Cajthaml Jiří , Kratochvílová Darina , Popelka Jan , Dvořák Vojtěch , Janata Tomáš
Published: October 2021

Archival spatial data sources (maps and aerial photographs) allow reconstruction and 3D visualization of landscapes that have been altered by human activity. This article is dedicated to the reconstruction of the 300 km‐long pre‐dam valley of the Vltava River (Czech Republic, Europe) inundated by nine reservoirs built between 1930 and 1992. We used methods based on archival aerial photographs and old maps to reconstruct pre‐dam georelief, which is exceptional because the total area subject to the reconstruction is 1,670 km2. We found that old maps are preferred for terrain reconstruction. The map series “State Map Derived 1:5,000” released between 1950 and 1959 was chosen as the most suitable data source for the pre‐dam valley reconstruction because this map series covers the entire area of interest with elevation information. Based on the processed maps, more than 26,000 km of contour lines were derived by semi‐automatic vectorization. The resulting pre‐dam digital elevation model (DEM) was created using a combination of interpolation techniques. We demonstrated that the methods used for georelief reconstruction under standard conditions should be used carefully while working with such large areas. Geostatistical analysis was performed to verify the accuracy of the resulting DEM. The differences between the model and LiDAR surveyed data were analyzed from three points of view: spatial autocorrelation, normality, and elevation stability. We tested several 3D printing methods for producing a realistic 3D model with a texture applied. The computer numerical control (CNC) milling method was chosen to produce 3D models of the three reservoirs at scales of 1:8,000 and 1:6,000.
Innovating but still poor: The challenges of regional development in regions with mature industries
Authors: Marques Pedro , Barberá‐Tomás David
Published: October 2021

The persistence of patterns of decline or stagnation in developed countries has led to a renewed interest in the study of less developed regions. This renewed interest is informed by a variety of approaches, which range from studies on the unequal distribution of economic power and resources to analysis of regional endowments, such as human capital or quality of institutions. However, a persistent misconception in some of this literature is that the primary factor explaining levels of development is innovation happening within firms. In particular, this narrative fails to consider the role of national and international agents and processes, the impact of value chain management on value appropriation, and the challenges to diversification within peripheral regions. This paper will contribute to this debate by developing these matters theoretically and by drawing on an analysis of three wine regions in Portugal.
The signature of the disaster: Witness‐ness in death camp and tsunami survivor testimony
Authors: Carter‐White Richard , Doel Marcus A.
Published: October 2021

The prominence of witness testimony in a range of contemporary political events is reflected in interdisciplinary efforts to theorise the act and genre of witnessing. Notably, this literature has framed the occurrence of errors, contradictions, and other instances of "problematic" speech in testimony as attesting to the traumatic quality of disastrous events. In this paper we extend this line of reasoning by recasting the fundamental quality or "witness‐ness" of disaster survivor testimony outside a logic of representational correspondence. Instead, drawing on the philosophy of Maurice Blanchot, we suggest that the disorienting features of testimony can be interpreted as the disruptive influence or inscription of the disaster itself on the recollections of survivors; a certain "writing of the disaster." Furthermore, we suggest that different disasters disrupt the testimonies of its survivors in unique ways, thus imprinting a signature that betrays the material and psychological character of the event. The "witness‐ness" of survivor testimony is therefore argued to dwell not in its representational accuracy, but in the distinctive, signature ways that it disorients the search for a coherent accounting of the disaster. We explore this proposition first in relation to Nazi death camp survivor testimony, before exploring this approach in the very different testimonial context of the 2011 tsunami in Tohoku, Japan. In the wake of these readings, we argue that the concept of the signature has potential not only for broadening the repertoire of testimonies admissible in the study of disaster, but also for investigating the societal impacts and "countersignatures" of disasters more generally.
Nature-based interventions for bereavement care: a scoping review protocol
Authors: Cleary Anne , Dean Julie , Pollock Danielle , McDaid Lisa , Boyle Fran
Published: October 2021

Objective:

The objective of this scoping review is to understand the types of nature-based interventions currently used in bereavement care and to map the theories of change explaining how such interventions may support bereavement care.

Introduction:

Bereavement can be associated with adverse psychosocial health outcomes. Nature-based interventions have been shown to have positive health and well-being outcomes, and offer diverse and flexible design and delivery options. However, this is an emerging field, and there is a need to explore the literature on the different types of nature-based interventions used in bereavement care and to identify the underpinning theories of change.

Inclusion criteria:

This review will include studies or reviews of participants who have experienced bereavement following the death of a loved one, and who have taken part in a nature-based intervention (eg, animal-assisted therapy, care farming, therapeutic horticulture, community gardens). Considering all nature contexts and geographic locations, this review will include studies that explore the role of nature-based interventions in bereavement care.

Methods:

Searches will be conducted in eight databases: Web of Science, Scopus, PubMed, PsycINFO, CINAHL, Social Science Database (ProQuest), PTSDpubs (ProQuest), and Research Library: Health and Medicine (ProQuest). Manual searches of reference lists of included full-text articles and keyword searches of prominent journals will be conducted. Data on the type of nature-based intervention, target sample, health outcomes, measures and methods used, and theory of change will be extracted from selected articles. The data extraction results will be presented in tabular format along with a narrative summary.
Mechanical Testing of Additively Manufactured Superalloy Lugs
Authors: Roychowdhury Sushovan , Karlsson Henrik , Henriksson Björn , Carlson Pher-Ola
Published: October 2021

Additively manufactured parts, in spite of their many advantages, face substantial challenges on the path toward certification. This challenge is more pronounced in the quality-demanding aviation industry, where the safety considerations are paramount. A major reason for this challenge is the lack of history associated with additively built parts compared to the traditional cast and wrought components. In assessing the structural integrity of cast and wrought components, material properties obtained from laboratory coupon tests are routinely applied for design calculations of large components. However, for additive manufacturing (AM) parts, questions remain over transferability of properties over multiple length scales due to possible variations in material chemistry, microstructure, and defect. In this work, this issue is investigated by conducting mechanical tests on small simplified lugs of nickel-based superalloy Haynes 282. The lugs are produced by the laser powder bed fusion process. After appropriate heat treatment and machining operations, the lugs are subjected to strength, low cycle fatigue, and crack propagation tests. Multiple tests are carried out in order to assess repeatability. Design calculations are performed to assess whether the test results can be predicted with standard methods. The results in this work generate confidence in predictable, repeatable behavior of the AM built lugs. Continuation of this approach over larger length scales has the potential to build enough confidence so that additively manufactured parts can be used in load-bearing structural elements of the aircraft engine.
Alternative Absorber Materials for Control Rods in ALFRED
Authors: Guo Hui , Jin Xin , Feng Kuaiyuan , Gu Hanyang
Published: October 2021

The next-generation reactors require improved safety performance and longer cycle length, which initiate the research on alternative absorber materials. In this context, potential absorber materials including borides (B4C, HfB2, and ZrB2), rare earth oxides (Eu2O3, Gd2O3, Sm2O3, and Dy2TiO5), metals/alloys (Hf and AIC), and metal hydride (HfHx) were compared in a large sodium fast reactor. The design of control rods for Generation-IV fast reactors strongly depends on the core characteristics. In this paper, some alternative absorbers are assessed in a lead fast reactor ALFRED using depletion capability in the Monte-Carlo particle transport code OpenMC. Results show that the ALFRED reference control rod design with B4C largely satisfies the shutdown and operation requirements. 60% 10B enriched HfB2 and HfH1.18 can replace the operation part of the reference design. In the future, the safe operating life of B4C and HfB2 should be assessed taking into account the irradiation-induced swelling, temperature margin, and gas release. HfH1.18 has a limited and local influence on the core power distribution. Eu2O3 has little loss on the absorption ability after 5 cycle irradiation. This oxide absorber satisfies the shutdown function even with only half control rod insertion, while its critical insertion depth at beginning of the cycle should be increased to realize reactivity compensation function.
Research on Setting Alarm Thresholds of Gaseous Effluent Radiation Monitoring From Nuclear Power Plants in China
Authors: He Wei , Jiang Jing , Xu Chen , Lei Qiang , Xu Chunyan , Liu Xinhua , Wang Yu , Xie Feng
Published: October 2021

During the past 30 years, several types of nuclear power plants have gained importance in China; these primarily include CPR1000, Russian advanced pressurized water reactor (WWER), passive advanced pressurized water reactor (AP1000) and European advanced pressurized water reactor (EPR). The treatment technologies for the radioactive solid, liquid, and gaseous wastes adopted by various types of nuclear power plants are different; consequently, emission control requirements and specifications for gaseous and liquid effluents also vary accordingly. Especially, the alarm thresholds of gaseous effluent radiation monitoring from nuclear power plants are highly concerned by the regulatory agencies since they determine the upper limit of the emission amount of gaseous radioactivity under normal operation and provide the alarm timely in case of accidents. In this paper, the setting method and basis of the radiation monitoring alarm thresholds for gaseous effluents from various nuclear power plants in China are examined, and the calculation method for the alarm thresholds is compared and analyzed. The unified setting principles and methods for pressurized water reactors (PWRs) are proposed, which can serve as a reference for setting and reviewing the radiation monitoring alarm thresholds of the gaseous effluents from the new nuclear power plants in China, and supply technical support for the improvement of the related regulations and standards for the nuclear power plants in future.
Sensitivity Study on TRISO Fuel Failure Probability Evaluation for HTGR
Authors: Li Jian , She Ding , Shi Lei
Published: October 2021

The tri-structural isotropic (TRISO) fuel particles are utilized by High Temperature Gas-cooled Reactor (HTGR), which is important for the safety of HTGR. Recently, a new TRISO fuel performance code called FART is developed by the Institute of Nuclear and New Energy Technology (INET). FRAT employs more delicate models involving gas pressure buildup, mechanical stress analysis and the temperature and burnup effects. Furthermore, mechanical stress and failure probability are influenced by complex factors, such as geometric parameters, material properties and irradiation parameters. In addition, material properties and irradiation parameters are affected by temperature, fast neutron fluence, burnup, anisotropic factor and other parameters. In this paper, a comprehensive sensitivity study on influences of particle stress and failure probability from geometric, material and irradiation parameters of particle coating layers was performed. Results show that the calculated TRISO fuel failure probability is greatly affected by the input parameters. This work lays a ground for uncertainty quantification of FRAT code.
A Graph-Based Scene Understanding Approach for Ensuring Proper Use of Personal Protective Equipment at the Decommissioning Site of Fukushima Daiichi Nuclear Power Station
Authors: Chen Shi , Demachi Kazuyuki
Published: October 2021

Decommissioning of the Fukushima Daiichi nuclear power station (NPS) is an unprecedented effort in which the entire plant was contaminated with radioactive materials from the accident and radiation dose levels were not low. The decommissioning workers in these sites are indicated to wear personal protective equipment (PPE) properly for radiation protection. In response to the difficulties of on-site PPE management in decommissioning site of Fukushima Daiichi NPS, this paper presents a graph-based scene understanding approach that automatically identifies whether workers at a decommissioning site are using PPE properly. The experimental results demonstrate that the high precision and high recall of the model can effectively detect decommissioning workers’ failure use of PPE and can facilitate improved safety inspection and supervision.
Application of High-Precision Assembly Technology for Large Structures by Laser Beam Welding
Authors: Nishiyama Tomoyuki , Kagawa Takashi , Tsubota Shuho , Kimura Masahiro
Published: October 2021

Toroidal Field coils (TF coils), which produce strong magnetic field in ITER, are used in order to confine the high temperature and high density plasmas stably for fusion reaction. The conductor made of Nb3Sn strands transform a superconducting conductor by heat treatment. The superconducting conductor is inserted into the D-shaped Radial Plate (RP), which have the structure with spiral grooves on both side for it, and then, insulating tapes are wound step by step around the superconducting conductor in RP as taking up the superconducting conductor. After insertion of the insulation-treated conductor into RP, Cover Plates (CPs) are put on RP spiral grooves and seal welded. After welding of CPs on RP of both sides, that the D-shaped Double Pancake (DP) is completed. The accuracy affects the TF coil performance and the deviation from ideal shapes and locations of the coils generates so called ‘error field’. In order to start up and confine the plasma stably, the allowable amount of the error field is about a few Gauss (Gauss : 1 × 10-4 T)[1]. Therefore, the dimensional accuracy of the 1/10000 order such as the profile within ±3mm for the circumference length and the flatness within ±3mm in the whole area of the D-shaped DP (14m × 9m) of approx. 37 m is required by ITER design. In addition, the TF coil is a huge component (with a weigh of about 300 ton) to withstand the high electromagnetic force of hundreds of MN.

It is an essential requirement for ITER design to achieve the accuracy of more than 1/10000 even though machining after welding is not allowed. Laser beam welding technology for large structure with thick plate of Mitsubishi Heavy Industries was applied the TF coils to minimize and avoid large welding deformation in the conventional welding process such as TIG welding to meet the ITER requirements.
A Finnish District Heating Reactor: Neutronics Design and Fuel Cycle Simulations
Authors: Leppänen Jaakko , Valtavirta Ville , Tuominen Riku , Rintala Antti , Lauranto Unna
Published: October 2021

The development of a small PWR for district heating applications has been started at VTT Technical Research Centre of Finland, and the pre-conceptual design phase was completed by the end of year 2020. The heating plant consists of one or multiple 50 MW reactor modules, operating on natural circulation at around 120°C temperature. This paper presents the neutronics design and fuel cycle simulations carried out using VTT’s Kraken computational framework. The reactor is operated without soluble boron, which together with low operating temperature and pressure brings certain challenges to the use of control rods and burnable absorber. The reactor core is loaded with 37 truncated AP1000-type fuel assemblies with 2.0–3.0% fuel enrichment and erbium burnable absorber. The resulting cycle length is around 900 days. The results show that the criteria set for stability, reactivity control and thermal margins are fulfilled. More importantly, it is concluded that the new Kraken framework is a viable tool for the core design task.
Application of Best-Estimate Plus Uncertainty Analysis Method in Nuclear Safety Evaluation
Authors: Tian Xinlu , Jing Jianping , Zhuang Shaoxin , Chen Haiying
Published: October 2021

Most of the computer codes for nuclear power plants safety analysis were developed based on conservative methods, which would lead to unnecessary excessive margins and more difficulty for nuclear power plants design and improvement. Using best-estimate plus uncertainty analysis method may reduce or eliminate these unnecessary limits. The NRC had used the code as a basis for safety reviews of the AP series nuclear power plants, driving the development of the best-estimate plus uncertainty analysis methods (BEPU) in the independent audit calculation. It had become an international trend to use BEPU in accident analysis and safety review of nuclear power plants. At present, the calculation method of the best-estimate in the world is used to better understand the accident sequence in the water-cooled reactor, and uncertainty assessment is the necessary supplement to the best-estimate method. The research and application of BEPU method are becoming more and more extensive. Best-estimate method was combined with best-estimation program and the initial and boundary conditions. It accurately described the current safety margin. It was required to consider each impact analysis of the available information and data, including the best-estimate model, the uncertainty calculated in the transient state, the input and the uncertainty of input parameters. Compared with the conservative evaluation method, the BEPU analysis method defined the gap between the calculated result and the real value through the uncertainty analysis, and made a more reasonable evaluation of the safety margin. In the process of nuclear power plant safety review, National Nuclear Safety Administration (NNSA) had adopted the best-estimate plus uncertainty analysis method to review the LOCA accident. In this paper, the four best-estimate plus uncertainty analysis methods are introduced, the source of uncertainty and uncertainty statistic methods are described. Then, the application of best-estimate plus uncertainty analysis method in nuclear safety evaluation are discussed.
Research on Remaining Useful Lifetime Prediction Methods of Main Transformer in Nuclear Power Station
Authors: Li Zikang , Peng Minjun
Published: October 2021

The main transformer of the NPP is the key equipment connects with public power grid and it is also the important electrical equipment to provide power for the power plant electrical equipment during the shutdown of the generator set. In consequence, it’s failure will cause huge consequences and economic losses. Therefore, the research of reaming life prediction technology for nuclear power plants(NPP) is of great significance to the economic operation and safety of NPP.

In this paper, according to DL/T984-2005 and other related literature, it is determined that the aging condition of solid insulating fiber material is the main factor to determine the life prediction of transformer. The main aging mechanism of that is summarized, and the effect of temperature is emphatic introduced. In general, the ambient temperature and load curve are the important factors affecting the aging of transformer insulation, furthermore affecting the life of transformer. Therefore, based on GB/T15164-94 recommended ambient temperature and transformer load, this paper calculates the transformer life loss model.

On this basis, this paper introduces the basic particle filter (PF) and Kalman filter (KF) algorithm flow. Based on the physical formula, a data model is developed to estimate the transformer life loss. Besides, the results are compared and analyzed. The study finally found that there were no significant differences in the accuracy of the predictions. However, considering the fact that the aging of transformer is a monotonic decreasing process, the particle filter based on degradation rate has a slight advantage in life prediction.
Three-Dimensional Numerical Simulation on Transient Natural Circulation Characteristics of DRACS in PLANDTL-DHX Experimental Device
Authors: Chen Zijia , Lu Daogang , Guo Jinsong , Zhang Yuhao
Published: October 2021

The pool-type sodium-cooled fast reactor adopts a new type of decay heat removal method to discharge the core heat to ensure the safety and cool of the nuclear power plant in a long time under the station blackout accident. The natural circulation and forced circulation will lead to different thermal and hydraulic characteristics. In addition, the natural circulation produces different hydrothermal phenomena on the secondary side by operating Direct Reactor Air Cooling System (DRACS). The Plant Dynamics Test Loop (PLANDTL-DHX) experimental device can simulate the flow and heat transfer phenomenon in the core. In the present work, a “modular modeling and integrated calculation method” was developed to conduct three-dimensional numerical modeling and calculation for the PLANDTL-DHX experimental device. Meanwhile, the steady-state verification of PLANDTL-DHX was carried out and compared with the experimental data. On this basis, the transient operating conditions of DRACS were calculated, and the temperature, flow distribution and transient variation characteristics in the key positions such as center subassembly, upper plenum and Intermediate Heat Exchanger (IHX) were preliminarily obtained. Based on the experimental results, the numerical calculation is verified.
ARKADIA – for the Innovation of Advanced Nuclear Reactor Design
Authors: Ohshima Hiroyuki , Asayama Tai , Furukawa Tomohiro , Tanaka Masaaki , Takata Takashi , Enuma Yasuhiro
Published: October 2021

This paper describes the outline and development plan for “Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA),” which the Japan Atomic Energy Agency has begun developing for the purpose of transforming advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source. ARKADIA will realize artificial intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for any challenge that could arise in the design and operation of a nuclear plant, including optimization of safety equipment as well as structures, systems, and components. State-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&D are integrated with AI. In the first phase of development, ARKADIA-Design and ARKADIA-Safety will be constructed individually, with the first target of a sodium-cooled reactor. In a subsequent phase, everything will be integrated into a single entity that is technology inclusive and applicable not only to advanced rectors with a variety of concepts, coolants, configurations, and output levels but also to existing light-water reactors.
A High-Temperature Gas-Cooled Reactor Simulation System and Its Application Based on VPOWER Platform
Authors: Xie Biheng , Ji Chengzhi , Guo Xiaoyu , Han Wenbin , Hao Yisheng , Chen Junyi , Huang Shanfang , Wang Kan , Wei Hongbin , Liang Yanming
Published: October 2021

Tsinghua University and Neoswise Science & Technology Co. Ltd jointly developed the HTGR accident simulation system (EHTR) relying on the VPOWER platform. Based on Tsinghua University’s research into HTGR, the simulation system can provide users with a reliable HTGR control system that supports the functions of accident insertion, accident evolution, and accident analysis. At the same time, a friendly logical configuration interface is also provided. In nuclear engineering education, this system can help students deepen their understanding of the operation with advanced reactor equipment, which will help them enter the nuclear industry in the future. In this paper, EHTR is used to simulate the loss of feedwater accident. The changes in nuclear power, thermal power, temperature, and mass flow rate of the steam generator are analyzed. The simulation results are in good agreement with the design values, which shows the reliability of EHTR. Finally, the hybrid interactive teaching method combined with virtual simulation experiments and online teaching is introduced and applied in nuclear engineering courses. The teaching feedback shows that through the hybrid interactive teaching, the students have a better understanding of reactor operation knowledge, and stimulate their enthusiasm in the nuclear industry.
Analysis of IP200 Severe Accident Process Response to SBO and Emergency Power Failure
Authors: Zheng Zhenhang , Peng Minjun , Yu Hao , Yang Yang
Published: October 2021

Advanced SMRs such as the integrated pressurized water reactor IP200 use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In this thesis, the IP200 severe accident induced by the SBO and emergency power failure was modeled and analyzed using RELAP5 / SCDAP / MOD3.4 code. Based on the steady state calculation, which agrees well with designed values, the SBO accident for transient calculation is carried out. First, the case of the SBO accident without the passive core cooling system was calculated. The progression and scenario in the RPV was simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. Then, mitigation measures PRHRS and CMT were put in at four different time points when the core is began to uncovered, the core is completely uncovered, hydrogen is began to produced, and the molten pool is formed. The results show that putting in mitigation measures before the accident progresses to the point where the core starts to produce hydrogen can ensure that the core does not melt and avoid hydrogen risk.
Research and Application of Different Seismic Analysis Methods in Nuclear Power Equipment
Authors: Huang Xuan , Xiong Furui , Liu Shuai , Qi Huanhuan , Huang Qian , Zhang Ke
Published: October 2021

The integrity and reliability of nuclear power equipment are directly related to nuclear safety. According to ASME safety classification standard, the supports of pressure vessel, main pump and main cooler in reactor coolant system belong to nuclear safety class I and seismic class I. In order to verify the safety of the supports, according to the relevant regulations, it is necessary to carry out seismic analysis. The response spectrum method is based on the modal results, which can more accurately reflect the dynamic characteristics of the structure, so it is widely used in nuclear power engineering. Response spectrum method includes single point response spectrum method and multi-point response spectrum method. The single point response spectrum method uses the same seismic spectrum as the calculation input, which is suitable for the equipment with single support and small span; while the multipoint response spectrum method inputs different seismic spectra at different constraint points, which is suitable for equipment with multiple supports and large span. In this paper, the above two methods are used to carry out the seismic analysis of the main equipment support of the reactor coolant system, and the application of different seismic analysis methods in nuclear power equipment is studied.
A Numerical Study of Supersonic Film Cooling With Discrete Holes
Authors: Ni Hang , Peng Wei , Wang Jie , Zhu Yinhai , Jiang Peixue
Published: October 2021

Nuclear thermal propulsion is an effective propulsion technology that can meet the needs of deep space exploration. The nozzle of the nuclear thermal propulsion system endures high-temperature and high heat flux environment, while supersonic film cooling is an important cooling method to provide thermal protection. The present study numerically investigated the effect of the blowing ratio and the shock wave on supersonic film cooling with discrete holes. The results show that when the blowing ratio is large, a strong kidney vortex is formed and the high-temperature mainstream enters the cooling layer. Therefore, the film cooling effectiveness at the upstream region near the hole decreases rapidly at a large blowing ratio. For the middle and downstream regions, the increase of blowing ratio gives a larger coolant mass flow rate and thereby improves film cooling effectiveness. In general, film cooling effectiveness decreases with shock wave impingement. For a strong shock wave, the cooling stream shows strong lateral diffusion after the shock wave impinging region. As a result, the lateral average film cooling effectiveness increases first and then decreases. This phenomenon is more obvious at a larger blowing ratio.
Severe Accident Analysis of a Floating Nuclear Power Plant After Station Black Out Accident
Authors: Li Longze , Chao Fei , He Chuan , Tai Yun , Qiu Jinrong , Yao Shiwei , Yang Wen , Wang Jue , Hou Xiaofan
Published: October 2021

The Floating nuclear power plant (FNPP) is a kind of marine nuclear power plant which takes barge as carrier and is fixed in specific sea area by mooring device. A 100 MWt PWR type small modular reactor (SMR) with 2 loops is used in the FNPP. It is designed based on the standard and experience of the traditional onshore nuclear power plant. The engineering safety features in the plant are somewhat different from those in the traditional plants. The special residual heat removal system, the passive residual heat removal, the square steel containment and containment suppression system are designed in the plant. Three severe accident mitigation measures, i.e., the pressurizer (PZR) relief extension, the external vessel reactor cooling (EVRC) and the Passive autocatalystic recombiners (PAR) are set up to actually eliminate the large release of radioactive products. The severe accident induced by station black out (SBO) for the SMR in FNPP is analyzed with MELCOR code in the present work. The MELCOR model were developed for the main systems in the SMR, including the primary and secondary system, the engineering safety features, the containment system, and the severe accident mitigation measures. The severe accident sequence as well as the key phenomena during the accident are analyzed. The simulation results indicated that the core exposed and heat up for the coolant leakage via the pressurizer safety valve, and finally melt with the high pressure and lower pressure injection failure. However, the reactor pressure vessel (RPV) maintained integrity with the mitigation measure, i.e., the external vessel core cooling. The containment also maintained integrity, which prevented the large release of radioactive products to the other cabins and the environment. The work is useful in gaining an insight into the detailed severe accident process involved, and could be used to identify appropriate severe accident management strategies and countermeasures for the FNPP.
100-Gigawatt-Hour Crushed-Rock Heat Storage for Variable Electricity and Heat With Base-Load Reactor Operations
Authors: Forsberg Charles
Published: October 2021

A 100-gigawatt hour (GWh) crushed-rock heat storage system using oil or nitrate salts is proposed to enable base-load 1000-MWe nuclear plants to provide variable electricity and heat to maximize plant revenue. Such a heat storage system has the same capabilities as a large pumped-hydro facility and can provide hourly-to-weekly heat storage. The capital cost goal is $2–4/kWh of heat — more than an order of magnitude below electricity storage technologies (batteries, pumped storage, etc.). Oil (LWR) or nitrate salts (higher-temperature reactors) transfer heat from the reactor to storage and from storage to the non-nuclear power block. The non-nuclear power block output is several times base-load nuclear capacity.

The insulated heat storage structure would have a crushed rock bed 20 m by 250 m by 250 m. The large dimensions minimize the surface area (steel and insulation) and cost per cubic meter of crushed-rock storage. Crushed rock is the lowest-cost heat storage material. Heat is transferred from the reactor to crushed rock by spraying on sequential sections of crushed rock (1) hot oil or (2) molten nitrate salt. The liquid flows by gravity to the drain pan at the bottom of the crushed rock and is returned to the reactor to be reheated. Heat is recovered by spraying cold fluid on the top of the hot crushed rock with hot liquid recovered from the drain pan at the bottom going to the power cycle. The oil or nitrate salt is used for heat transfer — not for heat storage, thus, minimize the quantity and cost of the heat transfer fluid.
Nuclear Air-Brayton Combined Cycles Using Electrically-Heated Conductive Firebrick Heat Storage and Hydrogen for Peak Power
Authors: Forsberg Charles , Stack Daniel , McDaniel Patrick
Published: October 2021

A low-carbon world needs a replacement for the combined-cycle gas turbine to provide variable heat and electricity. We describe a nuclear combined cycle gas turbine (NACC) with a thermodynamic topping cycle where the high-temperature heat is provided by an electrically-heated thermal energy storage system or hydrogen. During base-load operations the NACC compressed air is heated up to 700°C with heat from a high-temperature nuclear reactor to produce electricity. At times of low electricity prices, electricity from the NACC and the grid is converted into high-temperature stored heat in the form of hot firebrick. At times of high electricity prices compressed air is heated up to 700°C by the reactor and then further heated by sending the compressed air through hot firebrick to heat the air up to 1427°C before going through the turbine and heat recovery steam generator to produce added peak electricity. The round-trip electricity-to-heat-storage-to-electricity efficiency is above 70%; that is, if one unit of electricity is sent to heat storage, more than 0.7 units of peak electricity are returned. The enabling technology is a new doped electrically-conductive firebrick that enables conversion of electricity to very-high-temperature stored heat. Stored heat addresses daily variations in demand. Hydrogen can be used for long-term seasonal energy storage but at lower peak electricity efficiency because the efficiency of the electricity-to-hydrogen conversion step is near 75% versus near 100% for electricity-to-heat.
Research on Radioactive Consequence Limits for SSG-30
Authors: Zhao Danni , Pang Zongzhu , Li Ming , Liu Yu
Published: October 2021

After the release of new nuclear safety requirements SSR-2 / 1 (the latest version is 2016 Edition) in 2012, IAEA issued a new guideline SSG-30 “safety classification of structures, systems and components of nuclear power plants” in 2014 proposing a new safety classification method. Subsequently, IAEA published TECDOC-1787 “Application of Safety Classification of Structures, Systems and Components of Nuclear Power Plants” in 2016, which further explained the application of SSG-30. The safety classification method of SSG-30 based on the safety functions and design provisions. An SSC implemented as a design provision should, however, be classified directly because the significance of its postulated failure fully defines its safety class without any need for detailed analysis of the category of the associated safety function. The severity of consequences is divided into three levels (high, medium and low). The high, medium and low consequences are defined by the radiation dose criteria for the public and staff in different plant conditions. In order to classify the structures, systems and components of nuclear power plant, the specific meaning of the definition of high, medium and low “consequences” should be clarified firstly. In this paper, the radioactive dose criteria for the public and staff under different working conditions of nuclear power plant are studied and suggestions are put forward to make SSG-30 more operable when used in nuclear power plant SSCs classification.
Implementation of Solar Salt as Fluid in ASYST4.1 and Validation for a Natural Circulation Loop
Authors: Trivedi A. K. , Novog D. R. , Allison C.
Published: October 2021

Various small modular reactor (SMR) designs such as Micro Modular Reactor (MMR), Stable Salt Reactor - Wasteburner (SSR-W) and the Natrium reactor utilize an intermediate molten salt loop for heat transport and thermal energy storage (TES). This technology is well proven for solar thermal plants and uses solar salt (NaNO3-KNO3 in 60:40 ratio by weight) as a heat transfer and heat storage material. The primary advantages are its thermal stability, large temperature range, low cost, high density, and low vapor pressure. For example, MMR is a prismatic block design which has emerged as one preferred concept of gas cooled reactors (GCRs) and uses helium as the primary coolant and the solar salt as intermediate fluid for transporting the heat to TES system. The current thermal hydraulic system codes can model helium as a dry non-condensable gas but in most cases solar salt properties are not available in these codes. Hence there is a lack of integrated analysis capability across the MMR-type system. An effort is made to collect a robust set of thermodynamic and transport properties and heat transfer correlations for solar salt and incorporate them into the code ASYST4.1. It will allow for the simulation of design basis accident which include the dynamics of the primary system, intermediate loop, and TES.

Reliable computational tools are required for thermal hydraulics design and accident analysis of Small Modular Reactors (SMR) including associated experimental validation. One emerging tool is ASYST (Adaptive SYStem Thermal-hydraulics) - ISA (Integral Simulation and Analysis). It is a new code which combines the capabilities of SCDAPSIM/MOD4 and SAMPSON and is being developed jointly by Innovative Systems Software (ISS), USA and Institute of Applied Energy (IAE) of Japan. The thermal hydraulic module, ASYST-THA, replaces the original US NRC-developed RELAP5 code used in RELAP/SCDAPSIM/MOD3.x and the THA code used in SAMPSON, with new system level hydrodynamic options that include multidimensional, multi-fluid models originally developed by ISS and IAE. ASYST represents an internationallevel collaboration on system code development and benefits from the expertise and experiments available world-wide.

This work presents the details of solar salt implementation into ASYST as well as a benchmarking study on Molten Salt Nitrate Natural Circulation Loop (MSNCL). ASYST calculated results are found to closely match the experimental data. This work has been further extended to quantify uncertainties in the calculated heater outlet/inlet temperature and the flow rate for 1600 kW case using Integrated Uncertainty Analysis (IUA) package of ASYST4.1. The margin in the temperatures is about 35 K which confirms that the ASYST can both overpredict or underpredict the experimental temperatures.
Advances in the Development of a Fluid-to-Fluid Similarity Theory for Fluids at Supercritical Pressure: Results From Sensitivity Analyses
Authors: De Angelis A. , Pucciarelli A. , Kassem S. , Ambrosini W.
Published: October 2021

In the present paper, the capabilities of a similarity theory for heat transfer to supercritical pressure fluids are further investigated with respect to past work, widening the range of the addressed operating conditions. In particular, experimental data for water at 25 MPa are here considered; several cases addressed adopting different simulant fluids (i.e. CO2, NH3 and R23), in order to understand possible dependencies of the proposed rationale on the imposed mass flux and the observed heat transfer conditions. The obtained results confirmed the lessons learned from the recent analyses, though highlighting some possible needs for improvements in both the adopted CFD model and the similarity rationale in some specific operating regions.

The aim of the present paper is to show further results in relation to the capabilities of the proposed theory, which seems sufficiently mature to be now considered for validation against experimental data.
Numerical Simulation of Bubble Shape and Departure in Nucleate Pool Boiling at High Superheat
Authors: Bhati Jyoti , Paruya Swapan , Akhtar Farheen
Published: October 2021

In this work, the authors reported the dynamics of bubble shape and departure in nucleate pool boiling at the heated surface at a high superheats of 34 K. The numerical results of bubble shape and departure have also been compared with the CFD-results in literature and with our experimental results. The shape of the bubble is influenced by superheats and Bond number Bo. We have improved our numerical model by including the effect of radial variation of film pressure on microlayer evaporation to accurately compute the bubble shape, bubble diameter and growth time. We see that the approximate analytical solution works much better than the direct numerical integration of Young-Laplace equation for non-spherical bubble at high superheat. The Young-Laplace equation seems to include the growth force for capturing the shape of non-spherical bubbles. Interestingly, we find a train of small bubbles at a low Bo, which coalesce to form a big bubble on increasing Bo. While analysing the forces acting on the bubble surface, we also see that the growth force imparting the liquid inertia becomes important for the hemispherical bubbles.