bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research and Application to Risk-Informed NPP Events Trend Analysis Method
Authors: Shijun Chen , Sifan Zhang , Peifeng Chen , Kuan Zhang 
Published: October 2017

The method of the event trend analysis is a so-called analyzed process that will begin with a statistics on a certain characteristic for a set of events over a period of time, and then identify and analyze its variation trend and the reason for its existence so as to develop the corresponding corrective actions. this paper combined with both the traditional event trend analysis and probabilistic safety analysis methods, develops a set of risk-informed trend analysis techniques applicable to the domestic NPPs. This method aims to highlight the significance of the events by placing larger weight on the abnormal incidents of high risk significance, and use the strategy of level management to control the common cause events so that the plant will keep a watchful eye on and give priority to eliminate these events.
In addition, this paper also provides an application case to illustrate the practical use of this method. It is shown that some plants have already obtained several benefits through adopting this method into their event analysis programs. In reality, it will benefit the application of this method into subsequent NPPs event trend analysis process and provide reference and assistance for the safety operation of the nuclear power plants.
Modeling of QUENCH 10 Experiment on Air Ingress Using RELAP/SCDAPSIM Mod 3.5
Authors: Kaliatka Tadas , Uspuras Eugenijus , Allison Chris M. 
Published: October 2017

The most important accident management measure to terminate a severe accident transient in a light water reactor is the injection of water to cool the uncovered degraded core. Several experimental and analytical investigations have been made on the subject of zirconium oxidation in the steam and air environment during the last 15 years. The significant group of experiments are the experiments, performed in QUENCH facility. QUENCH-10 experiment was the first experiment from the all QUENCH experiment series with air ingress.
In this article, the RELAP/SCDAPSIM mod 3.5 code version and implemented different oxidation correlations were used to analyze the air ingress experiment QUENCH-10 with the purpose to perform the code and model assessment. The results of analysis will be used in the future computer code improvement and to increase the competence of code users.
Effect of Mesh Resolution and Buoyancy-Driven Flows on the k–ε Turbulence Models Using the HYDRAGON Code
Authors: Saeed Muhammad , Yu Jiyang , Xianping Zhong , Hou Bingxu , Zhang Xiaolong , Ghazanfar Mahmood Ahmad , Guangyin Liu , Yang Chunzhen 
Published: October 2017

Hydrogen may be released by Fuel and Cladding interaction with the steam at very high temperatures into the containment during a severe accident at the nuclear power plant (NPP). Locally, high hydrogen concentration may be achieved that might probably result in detonation or fast deflagration and challenges the reliability of the containment. The mixing and distribution of hydrogen is a serious safety issue for scientists to preserve the structural reliability of the containment. Although the Three Miles Island (TMI) accident in 1979 was the initiator to study the production and buildup of hydrogen in the containment. Subsequently, the hydrogen explosion in the Fukushima Dai-ichi NPP accident (2011), modeling of the gas behavior turn out to be a significant topic in the nuclear safety analysis.
Computational fluid dynamics (CFD) codes can be used to investigate the hydrogen distribution in the containment during accidental scenarios and predicts the local hydrogen concentration in different regions of the containment vessel. In such a way, the associated risk of hydrogen safety can be determined, and safety assessment and procedures can be measured. The current paper presents the results of systematic work done by using the HYDRAGON code, developed by Department of Engineering Physics, Tsinghua University, to study (I) Mesh sensitivity (II) Buoyancy-driven flows analysis, for various turbulence models i.e., a standard k-ε (SKE) model, a renormalization k-ε (RNG) model and a realizable k-ε (RLZ) model and, to demonstrate the HYDRAGON code thermal-hydraulic simulation capability during a severe accident at the NPP. The HYDRAGON code simulation results were compared to the published data performed by Jordan et al., and it has been observed that the simulation results obtained by refined mesh have given generally satisfactory results. However, the global effect of the buoyancy term in the turbulence equations on the HYDRAGON code simulated results is very small.
Public Perception of Nuclear Power Risk in China and Methods to Improve Public Acceptance
Authors: Shen Jian , Liu Mingye 
Published: October 2017

At present, the most influencing factor on nuclear power’s development in China the public’s attitude and acceptance. This paper studies the public perception of nuclear power risk in China, and provides several feasible methods to improve the quality and effectiveness of public perception level. Therefore, the public acceptance of nuclear power can be ameliorated greatly, which will help the development of nuclear power industry.
For decades, the environment pollution has become one of the most serious and urgent problems in China. Since nuclear power has been proved to be a type of low-carbon and environment-friendly energy, striving to develop nuclear power is a good solution to China’s environment issues. However, by the end of 2015, China’s nuclear power’s electricity production share was only 3.03%, which was far below the average level of developed countries. This situation might be partly due to technical and economic reasons, but the essential cause of the restricted development of nuclear power in China is the public perception of nuclear power risk is far from objective and comprehensive, which leads to the public acceptance of nuclear power is not as high as expected.
This paper states that public perception process of nuclear power risk is a dynamic, complex and closed system, which consists of the risk, the transformation of the risk and related information (both truth and rumors), the public perception process and public’s acceptance of the risk, and the public’s actions after receiving the information. The public’s actions often react on the risk perception. Nuclear power risk, unfortunately, is usually magnified, and it makes people become more frightened and oppose nuclear power more seriously.
In order to solve those issues, in this paper, the public risk perception’s characteristics and external influencing factors are studied, a model describes the public perception process of nuclear power risk is developed and analyzed, and the causes of current acceptance level of nuclear power (which is relatively low) are explained. In addition, based on this model, methods to conduct effective risk communication and public education on nuclear power are provided, the future of nuclear power in China can be much better.
Numerical Simulation of Crud Effects on the Flow and Heat Transfer Performance in PWR 17×17 Rod Bundles
Authors: Ren Bing , Gan Fujun , Dang Yu , Zhu Libing 
Published: October 2017

Corrosion products on fuel cladding surface have a significant impact on reactor operation. These types of deposits are defined as Corrosion Residual Unidentified Deposit (CRUD) and are consist of a porous matrix of nickel and iron based oxides deposited on the fuel cladding surface. It is well known that crud deposits may cause potential Crud Induced Localized Corrosion (CILC) risk and Crud Induced Power Shift (CIPS) risk.
The paper presents a Computational Fluid Dynamic (CFD) method of predicting the crud effect on the thermal hydraulic performance. The effect of the crud roughness is mainly considered in the simulation, the flow near the wall of the crud is solved by modifying wall function in the prism layer.
The simulation object is a span of typical 17×17 rod bundle with a mid grid in PWR, all the structures including grid straps, springs, dimples, mixing vanes and welding spots are included. Thicknesses of grid and fuel cladding are considered in order to precisely simulate the fluid-solid conjugate heat transfer. The crud is set to be covered on the full span downstream of the grid. The simulation is based on the CILC risk pre-analysis and the computed information in the mostly likely crud deposit position is used as boundary condition.
Based on the simulation results, the crud effects on the flow characteristics including vortex structures, circulation, turbulent intensity and second flow intensity and the heat transfer characteristics including rod temperature, fluid temperature and heat transfer coefficient are discussed in detail.
Thermal Performance Investigation of Hexagonal Spent Fuel Dry Storage Facilities
Authors: Ren Bing , Ni Chenxiao , Dang Yu , Liu Jiazheng 
Published: October 2017

A new type of dry storage system is designed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI), which can efficiently remove the decay heat of the hexagonal spent fuel assemblies such as VVER fuel assemblies. The dry storage system includes a Ventilated Concrete Cask (VCC) and a Multi-assembly sealed basket (MSB). Decay heat is removed by natural circulation with helium and air, heat conduction and thermal radiation heat transfer.
Thermal performance of the dry storage system has been investigated by two different numerically methods, i.e., the Computational Fluid Dynamics (CFD) method and the lumped parameter method. The CFD method is utilized based on the commercial software STAR-CCM+, and fuel assemblies are modeled as a porous medium characterized by effective conductivity and the permeability and inertial resistance factor, while other geometry including the lids, base plates, inner and outer shell are modeled explicitly with necessary simplifications. The lumped parameter method is utilized based on the system code GOTHIC, the geometry and the fuel assemblies are divided and represented by 44 volumes. The flow of the air and helium are modeled by flow path which connects the related volumes, and the heat transfer between fluid and solid structures are modeled by thermal conductor models. Heat transfer by convection, conduction and thermal radiation is modeled in both of the two methods.
The maximum temperature of spent fuel assembly can be obtained by both of the two methods, which can be a design basis for investigations attempting to improve the performance of the dry storage system. It is found that the simulation results calculated by the lumped parameter method are more conservative than those calculated by the CFD method. Both methods indicate that after the storage of 7.5 years, the dry storage system is able to remove the decay heat from the hexagonal spent fuel assemblies, keeping maximum cladding temperature below the design limit. Besides, detailed flow characteristic are obtained by CFD simulation. Furthermore, effects of MSB normal operating pressure and the ambient temperature are studied.
Research on Estimating Methods and Application of Fuel Rods Defect
Authors: Li Huaibin , Fu Yaru , Mao Lanfang , Mei Qiliang 
Published: October 2017

When fuel rods have defects, the fission products in the fuel rods will come out and enter the reactor coolant through the defects. The release of the fission products will result in the increase of radiation dose, the indeterminacy of the fuel management and will influence the economics and safety of the nuclear power station. Based on the analysis of the typical nuclides activities in reactor coolant, the evaluation of the defect fuel rods can be realized. This paper studied the related analysis around the world and determined the methods to evaluate the number, the defect type (open or tight) and the burn-up of the defect fuel rods.
The evaluation method of this paper can be used to evaluate the defects of fuel rods, and can provide valuable information for the fuel management and dose analysis, and also can be a useful technical support to the operation of nuclear power station.
Simulation of Small Steam Generator Tube Break (SSGTB) in a Small Pressurized Water Reactor (SPWR)
Authors: Cao Huasong 
Published: October 2017

Lots of efforts have been made to Research & Development of Small pressurized water reactors (SPWRs). Steam generator tube break occurs due to wear and corrosion frequently in the reactor. Among the breaks, Small Steam Generator Tube Break (SSGTB) is difficult to detect. Therefore, it is necessary to investigate the features of SSGTB.
A small pressurized water reactor model has been established in this paper by Relap5. The model includes reactor core, pressurizer, steam generator, main coolant pump and auxiliary safety system. The core flow, pressure of pressurizer, core outlet temperature and secondary outlet steam temperature obtained based on steady-state calculation is compared with design data to verify the model correct.
SSGTB is simulated by introducing a small break in the steam generator tube. The important parameters of reactor are recorded and analyzed. The procedure of SSGTB is analyzed and the system response features are summarized.
Bubble Diameter Effects on CFD Simulation for Subcooled Boiling Flow in Rolling Circular Tube
Authors: Xu Sheng , Pan Liang-ming , Yang Yu , Yuan Dewen , Xu Jianjun , Huang Yanping 
Published: October 2017

Eulerian two-fluid model coupled with wall boiling model was employed to calculate the three dimensional flow field and local parameter distribution with different bubble diameter models in circular tube under static and rolling condition. The wall boiling model utilized in this study was validated by Bartolomei experiment data, and a good agreement can be obtained. The calculation results of local void fraction are compared with experiment data to verify the accuracy of the numerical calculation for subcooled boiling flow under rolling condition. The Zeitoun bubble mean diameter model which the most recommended correlation for bubble diameter under low pressure and several fixed bubble diameters are applied to simulate the same condition in low pressure. These results are compared, include the distribution of void fraction, velocity distribution and radial flow induced by rolling motion.
A good agreement with the experimental data has been achieved when Zeition bubble mean diameter and 2 mm fixed bubble diameter are used to describe vapor diameter in static condition. The local void fraction fluctuation has the same period with the rolling motion, and the fluctuation amplitude increases with the increase of rolling amplitude and rolling frequency. The difference shown in rolling condition between calculation results and experimental data demonstrates that better agreement with the experimental data has been achieved in the near-wall region about local void fraction which has bigger fluctuation amplitude. Higher void fraction has gotten using Zeition bubble mean diameter model to describe bubble diameter in subcooled boiling flow, tiny difference has showed in temperature, velocity and radial velocity in different bubble diameter model. Accurate vapor diameter model or method to describe vapor diameter coupled with suitable interphase force model is needed in rolling condition under low pressure to fit the calculation of subcooled boiling better under rolling condition.
Development of Hydrogen Treatment System in Severe Accident: Part 3 — Study of Reactant Diameter Effect on Hydrogen Treatment
Authors: Yamada Akira , Iwaki Chikako , Fujihara Ryoma , Yagyu Motoshige , Okamura Masato , Tanabe Masashi 
Published: October 2017

The hydrogen treatment system has been developed in order to prevent the overpressure of the primary containment vessel (PCV) caused by a large amount of hydrogen generated by the metal-water reaction in severe accidents (SAs) of Light Water Reactor. In previous studies, we evaluated the hydrogen treatment rate using a couple of metal oxides, and confirmed that MnO2, CuO, and Co3O4 were effective for the hydrogen oxidation under the oxygen-free condition, then we selected them as reactants[1].
Although the reactants were granulated with a diameter of 2 mm for application to the system, the hydrogen treatment rate has not been scarcely evaluated for the granulated MnO2 which is expected to treat the hydrogen around 120 °C [2].
Thus, we made the diameter of the granulated MnO2 smaller, and found that the hydrogen treatment was occurred by the granulated MnO2 with a diameter below 1.0 mm. The granules with a diameter below 1.0 mm were also acceptable for the system from the point of view of decreasing the differential pressure (DP). Moreover, the experiments using a test section simulating a reactor of the system had been conducted under the hydrogen condition simulating typical condition of a SA, by loading the granulated CuO with a diameter of 2mm onto the granulated MnO2 with a diameter of 1mm. As a result, the hydrogen treatment was markedly accelerated by supplying enough reaction heat from the granulated MnO2 to the granulated CuO.
A Hybrid Approach to Seismic and Other External Hazards HRA: More Efficient and More Realistic
Authors: Amico Paul , Collins Erin , Cook Elizabeth 
Published: October 2017

As stated in the External Events HRA EPRI guidance [Ref. 1], there are inherent differences from internal events PRA that impact the ability of operators to perform actions during an external event.
“Although the process steps and concepts are the same for an external event PRA, a fire PRA and an internal (non-spatial) PRA, there are several key differences needing to be addressed in the development of an external event HRA. Differences include the impact of the initiating event on the site structures, systems, and components (SSCs). Other differences, such as procedures and the impact on timeline development, should also be considered.”
A damage context “hybrid approach” has been developed and utilized for Seismic Probabilistic Risk Assessment (SPRA) that takes advantage of the binning system developed in the EPRI guidance [Ref. 1], while also applying site and context-specific parameters. In addition, the hybrid approach does not provide overly conservative results, like some scaling and factor based methods. This paper describes the methodology of the Seismic and External Hazards (EH) HRA hybrid approach and provides an example of how it has been applied in a SPRA. This approach meets the Capability Category II quality requirements in Part 5 of the ASME/ANS PRA Standard RA-Sa-2009 [Ref. 2].
Visualization Study of Bubble Sliding Characteristics in a Subcooled Flow Boiling Narrow Rectangular Channel Under Natural Circulation Condition
Authors: Yang Kuan , Cao Xiaxin , Yan Changqi , Yang Yongyong , Tian Chunping , Xu Jianjun 
Published: October 2017

Heat transfer enhancement by the motion of the bubbles sliding along the heating surface are wildly reported by many researchers, thus it is of great importance to quantitatively investigate the characteristics of sliding bubbles. A visualization study of subcooled flow boiling of water in a vertical single face-heated narrow rectangular channel under a series of natural circulation working conditions was conducted. Pictures of the bubble sliding behaviors were captured by a high speed camera simultaneously with thermal data. A sequence of digital image processing algorithms were applied to the original picture to extract bubble shape and location information, which post-processing methods were adopted to obtain characteristic sliding parameters (including the distribution of the equivalent sliding bubble diameter and velocity, number density of the sliding bubbles).
It is found that bubbles can be able to nucleate and grow while sliding on the heating plate after the ONB point; the bubble number density, average bubble sliding velocity and the average sliding diameter continue to increase along the test section; heat transfer in the flow channel are significantly enhanced along flow direction with relatively low local void fraction. The average bubble sliding velocity near the inlet is significantly smaller than the sectional average velocity of single-phase fluid of the flow channel, and then it exceeded near the outlet of the test section. The average bubble sliding velocity and diameter increase with increasing heat flux and decreeing local subcooling degree. The equivalent diameter of sliding bubbles and the bubble sliding velocity approximately follow normal distribution. The distribution of the bubble diameter and velocity both cover a wild range. The standard deviations of the probability density function of the sliding bubble diameter and velocity increase with increasing heat flux and decreasing subcooling degree.
Effects of Various Parameters on Supercritical-Pressure Heat Transfer and Limits for Heat Transfer Correlations Obtained in Vertical Bare Tubes
Authors: Maloufi Hakim , Xie Hanqing , Zopf Andrew , Anderson William , Langevin Christian , Pioro Igor 
Published: October 2017

Currently, there is a number of Generation-IV SuperCritical Water-cooled nuclear-Reactor (SCWR) concepts under development worldwide. These high temperature and pressure reactors will have significantly higher operating parameters compared to those of current water-cooled nuclear-power reactors (i.e., “steam” pressures of about 25 MPa and “steam” outlet temperatures up to 625 °C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated, as the steam will be flowing directly to a steam turbine.
In support of developing SCWRs studies are being conducted on heat transfer at SuperCritical Pressures (SCPs). Currently, there are very few experimental datasets for heat transfer at SCPs in power-reactor fuel bundles to a coolant (water) available in open literature.
Therefore, for preliminary calculations, heat-transfer correlations developed with bare-tube data can be used as a conservative approach. Selected empirical heat-transfer correlations, based on experimentally obtained datasets, have been put forward to calculate Heat Transfer Coefficients (HTCs) in forced convective in various fluids, including water at SCPs.
The Mokry et al. correlation (2011) has shown a good fit for experimental data at supercritical conditions within a wide range of operating conditions in Normal and Improved Heat-Transfer (NHT and IHT) regimes. However, it is known that a Deteriorated Heat-Transfer (DHT) regime appears in bare tubes earlier than that in bundle flow geometries. Therefore, it is important to know if bare-tube heat-transfer correlations for SCW can predict HTCs at heat fluxes beyond those defined as starting of DHT regime in bare tubes.
The Mokry et al. (2011) correlation fits the best SCW experimental data for HTCs and inner wall temperature for bare tubes at SCPs within the NHT and IHT regimes. However, this correlation might have problems with convergence of iterations at heat fluxes above 1000 kW/m2.