bluuebunny/crossref_metadata_2025_split huggingface dataset:
An Experimental Study of Liquid Flooding in Vertical Large Scale Rectangular Channel With the Counter-Current Flow of Air and Water Film
Authors: Du Kashuai , Hu Po , Zhai Shuwei , Yang Xiaojie , Wang Weibo 
Published: October 2018

In the present study, a test facility, called WAFT [1], has been set up to study liquid flooding phenomena on a heated large flat plate with the counter-current air-water flow.
The various thermal hydraulic working condition parameters, especially, temperature and velocity of water, air and oil media, could be obtained by adjust the control devices, such as, heater, valve and frequency regulator. Because of transparent organic glass slab in the test section, the dynamic behavior of water film flow in the different locations of test area can be clearly observed by virtue of high-speed video camera, such as wave pattern change, film rupture, droplet splash and dry spot occurrence.
During the test, a series of tests were performed under the conditions of steel plate heat flux 25 kw/m2, the water film with mass flow rate 0.1 and 0.6 kg/s.m and the temperature 70 and 92 °C, the velocity of air entrance ranging from 6.0 to 11.50 m/s and the inlet air temperature of 45 and 60 °C, respectively.
According to the acquisition data, a modified Wallis liquid flooding predicted relation was proposed considering the influence factor of temperature from large steel plates, air and water film, i.e., the effect of water film evaporation. The results indicate that the mass flow rate of water film and inlet air velocity have a great effect on the emergence of liquid flooding.
Numerical Investigation of Aerosol Collection Efficiency in a Venturi Scrubber
Authors: Goel Paridhi , Nayak Arun K. 
Published: October 2018

In an extreme situation, as happened in Fukushima nuclear power plant, the failure of multiple safety systems may lead to severe core damage and melt relocation. This may be accompanied by production of large amount of steam which may result in the over-pressurization of the containment. Another associated concern is that the exposed core melt is highly radioactive which if exposed to the atmosphere can ruthlessly deteriorate the quality of environment and living beings. The radioactive materials present in the containment can be in vapor form or aerosol form. The containment of a nuclear power plant is therefore the final shielding to prevent the release of radioactive products to the environment. Therefore, the installation of Filtered Containment Venting system (FCVS) is mandatory in a nuclear reactor which actuates passively to depressurize the containment. Additionally it assists in the retention of radionuclides in the containment.
The FCVS consists of venturi scrubbers submerged in a pool of scrubbing liquid along with a demister housed in a scrubber tank. The performance of venturi scrubber is dependent on the interaction of the contaminated air stream from the nuclear reactor with the scrubbing liquid. This represents a multi-field and multi-fluid system. The present analysis investigates this system through a computational framework in which air stream is solved using Eulerian framework while the scrubbing liquid is tracked through Lagrangian framework. The collection efficiency of aerosols is modeled assuming impaction to be the predominant mechanism.
Sensitivity of CTF Solution to Subchannel Window Size
Authors: Mukin Roman , Seidl Marcus , Clifford Ivor , Ferroukhi Hakim 
Published: October 2018

In this work, a so-called mini-core consisting of a 3 × 3 array of 17 × 17 pressurized water reactor (PWR) fuel assemblies (FA) is considered with the aim of identifying the most conservative window size for hot channel analysis of bowed fuel assemblies. Overall, five different mini-core configurations are analyzed: one is the reference case, i.e. without FA displacement and four different cases with diagonal and parallel FA displacements. Rod power maps for these mini-cores were exported from neutronic calculations with CASMO-SIMULATE codes.
Subchannel modelling with COBRA-TF code of all five mini-cores allows one to identify the rod position with a minimum departure from the nucleate boiling ratio (DNBR) and to construct input decks with different rod window sizes around the previously identified rod position. Overall, eight different window sizes are considered: 3 × 3, 5 × 5, 7 × 7, 9 × 9, 11 × 11, 13 × 13, 15 × 15 and 17 × 17. Results of subchannel analysis for a mini-core and different subchannel window configurations are compared with the help of DNBR parameter, which is the ratio between the critical heat flux (CHF) and the actual local heat flux on a rod. An assessment of three different CHF models is applied in this work: Groeneveld CHF look-up table (LUT), W3 CHF correlation, and Doroschuk CHF LUT.
The general conclusion of this work is that for deformed core configurations, an appropriate rod window size needs to be determined to adequately capture the local flow redistribution. For large displacements (the largest displacement considered in this work is 10 mm), the DNBR ratio can drop to one. DNBRs obtained with the W3 CHF correlation give the most conservative results.
CFD Thermal Analysis of ITER Pressure Suppression Tanks
Authors: Basili Lorenzo , Lo Frano Rosa , Olcese Marco , Sekachev Igor , Aquaro Donato 
Published: October 2018

The aim of the paper is to investigate the thermal conditions (temperature distribution, heat losses) in the support system of the Vapour Suppression Tank (VST) of the Vacuum Vessel Pressure Suppression System (VVPSS), a safety important system of ITER fusion reactor, protecting the Vacuum Vessel (VV) against overpressures. The VVPSS includes four VSTs of identical volume and mounted as two stacked assemblies.
The study focuses on the optimization of the design of the thermal insulation of the bottom part of the VST, connecting each two-tank stack to the basement, and also on the identification of the thermal loads at the interface between the tank support and the tank pressure boundary.
A Computational Fluid Dynamics (CFD) analysis of the VST has been performed for four different insulation configurations and considering both steady state and transient loads following accidental conditions.
The results of the analysis are used to provide recommendation on the optimum configuration of the thermal insulation. Measures for minimization of the thermal gradient in the critical area of the joint between the tank hemispherical head and support skirt to limit the thermal fatigue on the welds are also suggested.
Moving From V&V to V&V&C in Nuclear Thermal-Hydraulics
Authors: D’Auria Francesco , Lanfredini Marco 
Published: October 2018

V&V constitutes a powerful framework to demonstrate the capability of computational tools in several technological areas. Passing V&V requirements is a needed step before applications. Let’s focus hereafter to the area of (transient) Nuclear Thermal-hydraulic (NTH) and let’s identify V1 and V2 as acronyms for Verification and Validation, respectively.
Now, V1 is performed within NTH according to the best available techniques and may not suffer of important deficiencies if compared with other technological areas. This is not the case of V2. Three inherent limitations shall be mentioned in the case of Validation in NTH:
1. Validation implies comparison with experimental data: available experimental data cover a (very) small fraction of the parameter range space expected in applications of the codes; this can be easily seen if one considers data in large diameter pipe, high velocity and high pressure or high power and power density. Noticeably, the scaling issue must be addressed in the framework of V2 which may result in controversial findings.
2. Water is at the center of the attention: the physical properties of water are known to a reasonable extent as well as large variations in values of quantities like density or various derivatives are expected within the range of variation of pressure inside application fields. Although not needed for current validation purposes (e.g. validation ranges may not include a situation of critical pressure and large heat flux) physically inconsistent values predicted by empirical correlations outside validation ranges, shall not be tolerated.
3. Occurrence of complex situations like transition from two-phase critical flow to ‘Bernoulli-flow’ (e.g. towards the end of blow-down) and from film boiling to nucleate boiling, possibly crossing the minimum film boiling temperature (e.g. during reflood).
Therefore, whatever can be mentioned as classical V2 is not or cannot be performed in NTH. So, the idea of the present paper is to add a component to the V&V. This component, or step in the process, is called ‘Consistency with Reality’, or with the expected phenomenological evidence. The new component may need to be characterized in some cases and is indicated by the letter ‘C’. Then, the V&V becomes V&V&C.
The purpose of the paper is to clarify the motivations at the bases of the V&V&C.
Thermal Hydraulic Analysis of the CIRCE-HERO Pool-Type Facility
Authors: Gonfiotti Bruno , Barone Gianluca , Angelucci Morena , Martelli Daniele , Forgione Nicola , Del Nevo Alessandro , Tarantino Mariano 
Published: October 2018

To support the activities carried out by the ENEA Brasimone Research Center on the Lead-cooled European Advanced Demonstration Reactor (LEADER) project, University of Pisa has improved the RELAP5/Mod3.3 code introducing the Lead and Lead-Bismuth Eutectic (LBE) as coolant fluids. This new version of the STH code was applied to perform several analyses for the steam generator HERO (Heavy liquid mEtal pRessurized water-cOoled tubes) installed in the CIRCE (CIRColazione Eutettico) pool-type facility that uses LBE as working fluid. HERO consists of a bundle of seven Double-Wall Bayonet Tubes (DWBT) with a leakage monitor system, and it was tested in the CIRCE facility to support the development of the steam generator proposed for ALFRED (Advanced Lead Fast Reactor European Demonstrator).
A preliminary analysis was performed comparing the RELAP5/Mod3.3 results against available experimental data. In particular, an isothermal test of the CIRCE-HERO facility allowed to set-up the RELAP5/Mod3.3 nodalization. As a second analysis, a steady state at full power was investigated. This steady-state condition was employed to perform a sensitivity study on the influence of the thermal conductivity of the stainless steel powder characterizing the HERO steam generator. Finally, the analysis of unprotected and protected loss of flow transients, and the analysis of a protected loss of flow with a simultaneous partial Loss of Heat Sink (LOHS) were performed. These last transients were investigated to assess the behavior of the HERO steam generator, especially to verify the possibility to establish a natural circulation capable to safely remove the “decay heat”.
The results of the first analysis (isothermal hydraulic characterization) demonstrated the good capabilities of improved RELAP5/Mod3.3 code to reproduce the forced circulation conditions. The results of the second analysis (steady-state at full power) highlighted the strong influence of the stainless steel powder thermal conductivity on the achieved steady-state conditions. The simulations of the last tests provided the time evolution of the LBE mass flow rate and of the temperatures in different regions inside the pool during unprotected and protected loss of flow transients. In the paper, a critical discussion of the obtained results is performed.
Evolution and Implementation of the Design Extension Conditions (DEC) Concept: Assessment of Selected Events
Authors: Kral Pavel 
Published: October 2018

The paper describes at first the evolution of the Design Extension Condition (DEC) concept, that started in early 1990ies and was accelerated after the Fukushima accident in 2011. Nowadays the concept is elaborated in EUR document [1], WENRA Reference Levels [4], and finally in the IAEA Safety Standards [6][7]. Number of countries is implementing the DEC concept into the national legislation, safety assessment, design modifications and operation of NPPs. Typically, the DEC events are divided to 2 groups: DEC without core melt (DEC-A) and DEC with core melt (DEC-B).
The second part of the paper deals with safety assessment of DEC events with more detailed information about deterministic analysis of DEC-A events. Legislative and methodology basis, acceptance criteria, computer codes and their validation, list of scenarios to be analyzed, examples of analyses performed, and incorporation of results of new analyses into SAR are described step by step.
Influence of Gap Size on Added Mass for Spent Fuel Storage Rack
Authors: Lu Daogang , Liu Yu , Zheng Shu 
Published: October 2018

Free standing spent fuel storage racks are submerged in water contained with spent fuel pool. During a postulated earthquake, the water surrounding the racks is accelerated and the so-called fluid-structure interaction (FSI) is significantly induced between water, racks and the pool walls[1].
The added mass is an important input parameter for the dynamic structural analysis of the spent fuel storage rack under earthquake[2]. The spent fuel storage rack is different even for the same vendors. Some rack are designed as the honeycomb construction, others are designed as the end-tube-connection construction. Therefore, the added mass for those racks have to be measured for the new rack’s design. More importantly, the added mass is influenced by the layout of the rack in the spent fuel pool.
In this paper, an experiment is carried out to measure the added mass by free vibration test. The measured fluid force of the rack is analyzed by Fourier analysis to derive its vibration frequency. The added mass is then evaluated by the vibration frequency in the air and water. Moreover, a two dimensional CFD model of the spent fuel rack immersed in the water tank is built. The fluid force is obtained by a transient analysis with the help of dynamics mesh method.
Feasibility Study of a New Fabrication Method for the Li4SiO4 Pebbles
Authors: Lo Frano R. , Puccini M. , Stefanelli E. , Luppichini M. , Grima C. , Vitolo S. , Aquaro D. , Malquori S. , De Sanctis S. 
Published: October 2018

In the past decades many R&D efforts have been spent in the development of a suitable Li4SiO4 fabrication method (e.g., melt spray process, graphite bed method, capillary-based microfluidic wet method etc.), nevertheless we are still far from an “industrial standard solution”.
The aim of the paper is to develop a new fabrication method capable to produce stable and well-sized pebbles of lithium orthosilicate (Li4SiO4) based on the drip casting. This method is mainly based on the dripping at room temperature, which is novel in the framework of available fabrication processes requiring high temperature: this latter is demonstrated to affect the final product characteristics. It is worthy to remark that the Li4SiO4 is a candidate material for the breeding blanket material of the fusion power reactor. In the paper we will describe the experimental apparatus, designed and built at DICI - University of Pisa with the collaboration of Bitossi Industries, and the procedure adopted in order to produce pebbles of Li4SiO4.
Neutronic Analysis for VVER-440 Type Reactor Using PARCS Code
Authors: Ratti Luca , Mazzini Guido , Ruščák Marek , Giusti Valerio 
Published: October 2018

The Czech Republic National Radiation Protection Institute (SURO) provides technical support to the Czech Republic State Office for Nuclear Safety, providing safety analysis and reviewing of the technical documentations for Nuclear Power Plants (NPPs). For this reason, several computational models created in SURO were prepared using different codes as tools to simulate and investigate the design base and beyond design base accidents scenarios.
This paper focuses on the creation of SCALE and PARCS neutronic models for a proper analysis of the VVER-440 reactor analysis. In particular, SCALE models of the VVER-440 fuel assemblies have been created in order to produce collapsed and homogenized cross sections necessary for the study with PARCS of the whole VVER-440 reactor core.
The sensitivity study of the suitable energy threshold to be adopted for the preparation with SCALE of collapsed two energy-group homogenized cross sections is also discussed. Finally, the results obtained with PARCS core model are compared with those reported in the VVER-440 Final Safety Report.
Evaluation of Local Damage to Reinforced Concrete Panels Subjected to Oblique Impact of Rigid and Soft Missiles
Authors: Nishida Akemi , Nagai Minoru , Tsubota Haruji , Li Yinsheng 
Published: October 2018

Many empirical formulas have been proposed for evaluating local damage to reinforced concrete (RC) structures caused by impacts of rigid missiles. Most of these formulas have been derived based on impact tests normal to the target structures. Up to now, few impact tests oblique to the target structures have been carried out. This study has been conducted with the purpose of proposing a new formula for evaluating the local damage caused by oblique impacts based on previous experimental and simulation results.
In this paper, the results of simulation analyses for evaluating the local damage to a RC panel subjected to normal and oblique impacts by rigid and soft missiles, by using the simulation method that was validated using the results of previous impact experiments. Based on the results of these simulation analyses, the effects of the rigidity of the missile as well as the impact angle on the local damage to the target structures are clarified.
Numerical Impact Simulation of Aircraft Into Reinforced Concrete Walls With Different Thickness
Authors: Hirosaka Kazuma , Takazawa Hidekazu , Miyazaki Katsumasa , Tohyama Norihide , Nouji Hiroyuki , Matsumoto Naomi 
Published: October 2018

Aircraft impact analysis is needed for a safety assessment of nuclear power plants. One of the contents which should be analysed for aircraft impact is physical damage of a concrete building and this can be estimated by a numerical simulation. In order to conduct aircraft impact analysis, simulation model which validated by some experimental data needs to be established.
In 1990s, impact test using actual F4 Phantom fighter was conducted at Sandia national laboratory in U.S. and a lot of important experimental data were measured. In this paper, the numerical simulation results for this F4 Phantom impact test are introduced. The relationship between the thickness of the shell of the F4 Phantom simulation model and the deceleration of this model is indicated and the differences of the deceleration between simulation and test results are discussed. In addition, the relationship between fracture strain of the shell of the F4 Phantom simulation model and the destruction mode of this model in simulation is indicated and the differences between the destruction mode of the F4 Phantom between simulation and test results are discussed.
In order to evaluate the physical damage area after the aircraft impact, it is necessary to estimate the aircraft velocity after it perforates the outer concrete wall and to calculate the decrease of the kinematic energy of the aircraft by this perforation. In this paper, several aircraft impact simulations with different concrete wall thickness are conducted and the reduction in kinematic energies of an aircraft by a perforation is estimated. Using these simulation results, the necessary numbers of concrete walls until the impacting aircraft stops is discussed.
WATCH Loop Development and Commissioning Tests
Authors: Byers W. A. , Wang G. 
Published: October 2018

It is essential to understand how crud build-up leads to elevated cladding temperatures and abnormal cladding corrosion for Boiling Water Reactor (BWR) fuel. With a heated rod test loop configuration, the influence of water chemistry on crud build-up and cladding corrosion behavior can be assessed.
A new test loop named the Westinghouse Annular Thermal Crud Hydraulic (WATCH) Loop, has been designed to grow simulated BWR crud under normal BWR operating conditions. WATCH simulates the bottom meter of a BWR core in terms of temperature, flow, steam quality and heat flux. Test results from this loop can be used to benchmark risk assessment tools, which in turn are used to make predictions and minimize or avoid BWR fuel rod failures, consistent with guidelines from the Institute of Nuclear Power Operations (INPO) and the Electric Power Research Institute (EPRI).
In this paper, the Westinghouse Annular Thermal Crud Hydraulic (WATCH) Loop design and development are introduced. The results of a commissioning test and a run producing zinc-rich deposits are also presented.
Beyond vision loss: the independent impact of diabetic retinopathy on vision-related quality of life in a Chinese Singaporean population
Authors: Fenwick Eva K , Man Ryan E K , Gan Alfred Tau Liang , Kumari Neelam , Wong Charlene , Aravindhan Amudha , Gupta Preeti , Wang Jie Jin , Mitchell Paul , Wong Tien Yin , Cheng Ching-Yu , Lamoureux Ecosse Luc 
Published: October 2018

Background/aimsTo assess the independent impact of diabetic retinopathy (DR) on three domains of vision-related quality of life (VRQoL) in a Chinese Singapore population.MethodsThe Singapore Chinese Eye Study (n=3353; 2009–2011) was a population-based, prospective, cross-sectional study conducted at the Singapore Eye Research Institute. The study population included 292 adults with diabetes, with and without DR. DR (better eye) was categorised as presence and absence of any DR; severity of DR (no vision-threatening DR (VTDR); severe non-proliferative DR (NPDR); PDR and/or clinically significant macular oedema and VTDR). Our main outcome was VRQoL which was measured using Rasch-calibrated scores from the ‘Reading’, ‘Mobility’ and ‘Emotional’ domains of the Impact of Vision Impairment questionnaire. The relationship between DR and VRQoL was assessed using multiple linear regression models.ResultsOf the 292 individuals (mean age 61.35 ± 9.66 years; 55.8% male), 201 (68.8%), 49 (16.8%), 20 (6.8%), 22 (7.5%) and 27 (9.2%) had no DR, minimal-mild NPDR, moderate-severe NPDR, PDR and VTDR, respectively. Any DR and VTDR were independently associated with 6% and 12% worse Reading scores and 7% and 18% poorer Emotional well-being, respectively, compared with those without DR. These associations persisted after separate adjustment for visual impairment and presenting visual acuity. No significant difference was found in the Mobility domain between persons with and without DR.ConclusionsWe documented that DR, particularly VTDR, was independently associated with restrictions in Reading and Emotional well-being. Understanding factors underlying the detrimental DR-VRQoL relationship may optimise rehabilitation outcomes for individuals with DR.