bluuebunny/crossref_metadata_2025_split huggingface dataset:
Thermal-Hydraulic Investigations of a Horizontal Dry Cask Simulator
Authors: Pulido Ramon J. M. , Lindgren Eric R. , Durbin Samuel G. , Salazar Alex 
Published: October 2020

Abstract
Recent advances in horizontal cask designs for commercial spent nuclear fuel have significantly increased maximum thermal loading. This is due in part to greater efficiency in internal conduction pathways. Carefully measured data sets generated from testing of full-sized casks or smaller cask analogs are widely recognized as vital for validating thermal-hydraulic models of these storage cask designs. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system.
The purpose of this investigation is to produce data sets that can be used to benchmark the codes and best practices presently used to calculate cladding temperatures and induced cooling air flows in modern, horizontal dry storage systems. The horizontal dry cask simulator (HDCS) has been designed to generate this benchmark data and complement the existing knowledge base.
Transverse and axial temperature profiles along with induced-cooling air flow are measured using various backfills of gases for a wide range of decay powers and canister pressures. The data from the HDCS tests will be used to host a blind model validation effort.
FUELPOOL: A Computer Program to Model CANDU Spent Fuel Pool Severe Accident Progression and Consequences
Authors: Song Yong Mann , Jung Jong Yeob , Nijhawan Sunil 
Published: October 2020

Abstract
CANDU PHWR spent fuel pools (SFPs), smaller than a tennis court, contain up to 38,000 or more (49,000 in Wolsong)fuel bundles in geometries not replicated in any other power reactor. Therefore, the phenomenological issues, accident progression pathways and effectiveness of mitigative actions are somewhat different. This requires a dedicated approach in progression and consequence assessments of potential accidents and development of mitigation measures. The SFPs house densely packed fuel bundles stacked in about a hundred vertical stainless steel tray towers, each containing 24 spent fuel bundles in each of the 16 or more (19 in Wolsong) horizontal fish basket style steel trays. Some of theupto 10 year worth of the on-line refuelled bundles in the SFP are at relatively high decay powers as fuel trays are prepped for the towers in near daily basis. In addition, there is a provision (see Figure 1) that a full core of bundles 20 days after being at full power can be transferred to the spent fuel bay at any time. About 4.5m of additional water layer on top of the tray towers provide radiation protection and a healthy margin to small rate of fluid loss.
Fatigue Risk Evaluation of a Pressure Vessel Plug Subject to Flow Induced Vibration
Authors: Wang Robert , Chang Louis , Hurst Tom , Hurst Antony 
Published: October 2020

Abstract
The steam generator (SG) channel head forms part of the reactor coolant pressure boundary and is of high nuclear safety duty. The channel head failure is considered intolerable and there are no reasonably practicable physical provisions available to prevent its failure. Therefore the channel head is classified as either an Incredibility of Failure (IoF) or High Integrity (HI) component and it requires additional analyses and assessments beyond the design code requirements to achieve and demonstrate its structural integrity.
The hydrodynamic effects of the primary coolant in the annulus around the drain plug of the SG channel head are not very well understood, but are exacerbated by high flow rates in its immediate vicinity due to typical design details. Vibration of the drain plug due to coolant flow may result in fatigue induced failure of the channel head drain penetration weld. In the study presented here, random vibration analyses and a fatigue assessment have been carried out for a submerged drain plug in a pressurized water reactor (PWR) SG channel head. A finite element (FE) model of the drain plug submerged in water coolant has been developed. Modal analyses confirmed that the natural frequency of the submerged drain plug is significantly reduced by the large hydrodynamic added mass from the surrounding fluid. The fatigue evaluation undertaken using ASME III fatigue curve concluded that the fatigue life usage due to the vibration of the drain plug is negligible even after an extended plant life. Therefore the coolant flow-induced drain plug vibration is not a threat to the channel head integrity.
Development of the Fuel Behavior Analysis Code for Mechanical Fuel Cladding Failure During Reactivity Insertion Event in PWR
Authors: Higashi Yuma , Murakami Nozomu , Yodo Tadakatsu , Yamamoto Teruhisa 
Published: October 2020

Abstract
Pellet-Cladding Mechanical Interaction (PCMI) failure is the one of failure mode which must be evaluated in the nuclear fuel safety. PCMI is caused by the mechanical load to cladding due to the fuel pellet expansion. Under the high fuel burnup condition, the fuel cladding may become degraded by embrittlement under the neutron-induced irradiation and hydrogen accumulation due to the waterside corrosion. In order to consider the further deterioration of the material with higher burnup, the evaluation using mechanical indicators, e.g. strain and stress, might be required.
In this study, a transient model is developed, which can mechanistically evaluate the PCMI behavior, in particular, for fuel rods under the higher burnup condition. The model is incorporated in a fuel behavior analysis code and verified by benchmarks with other similar codes. The PCMI predictability of this code is validated using the experimental test data.
Investigation on Velocity Distribution in the Subchannels of Pin Bundle With Wrapping Wire: Evaluation of Reynolds Number Dependence in 3-Pin Bundle
Authors: Aizawa Kosuke , Hiyama Tomoyuki , Nishimura Masahiro , Kurihara Akikazu , Ishida Katsuji 
Published: October 2020

Abstract
A sodium-cooled fast reactor is designed to attain a high burn-up core in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, the deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the flow velocity distribution in a wire wrapped pin bundle. In this study, the detailed flow velocity distribution in the subchannel has been obtained by PIV (Particle Image Velocimetry) measurement using a wire-wrapped 3-pin bundle water model.
The test section consisted of an irregular hexagonal acrylic duct tube and fluorinated resin pins which had nearly the same refractive index with that of water and a high light transmission rate. This enables to visualize the inner subchannel through the outer pins. Flow velocity conditions in the pin bundle were set from 0.036 m/s (Re = 270) to 1.6m/s (Re = 13,500).
From the PIV results, it was confirmed that the normalized flow velocity near the wrapping wire in low Re number condition was decreased relatively compared to that in high Re number condition. In the region away from the wrapping wire, the maximum flow velocity was increased by decreasing the Re number. In particular, the maximum flow velocity was more than twice the cross section average velocity in the laminar flow (Re = 270). Moreover, the PIV measurements by using the 3-pin bundle geometry without the wrapping wire were conducted. From the results, the effect of the wrapping wire on the flow field in the subchannel was understood. There experimental results useful not only for understanding of pin bundle thermal hydraulics but also code validation.
Flow Characteristics of Upward Two-Phase Flows in a Rod Bundle Geometry
Authors: Han Xu , Shen Xiuzhong , Yamamoto Toshihiro , Nakajima Ken , Hibiki Takashi 
Published: October 2020

Abstract
In the present paper, the local two-phase flow parameters were measured with a four-sensor optical probe in adiabatic upward air-water two-phase flows in a 6 × 6 vertical rod bundle with rod diameter of 10 mm, pitch of 16.7 mm, square channel box side length of 100 mm and hydraulic equivalent diameter (DH) of 18.7 mm. The local measurements were performed in an octant triangular region of the rod bundle cross-section at the axial position with height-to-diameter ratio (z/DH) of 149 under a total of 16 flow conditions. The local void fraction, interfacial area concentration (IAC), bubble diameter and bubble velocity were obtained in the four-sensor probe measurements. Both of the measured void fraction and IAC show their radial local distributions with the core-peaking and wall-peaking shapes which are closely linked with the superficial velocity of two phases. The distribution shapes tend to change from the core-peaking to the wall-peaking when the superficial liquid velocity (&lt;jf&gt;) increases and the superficial gas velocity (&lt;jg&gt;) decreases. The measured diameters of local bubbles keep the similar values in the measuring cross-section, increase when the &lt;jg&gt; increases and decrease when the &lt;jf&gt; increases. The bubbles behave with the velocities whose main flow direction component keeps a typical radial power-law (core-peaking) profile and whose cross-sectional velocity components show a significant trend of bubbles migrating from the center to the wall region of the channel box especially under high &lt;jf&gt; conditions. The area-averaged results of void fraction and IAC were obtained by a cross-sectional area-averaging scheme. The resultant area-averaged void fraction and IAC were used to check the void fraction predicting capability of two drift-flux correlations and the IAC prediction performance of two IAC correlations respectively. The applicability of these correlations to the rod bundle geometry was discussed and concluded finally in this paper.
Development of Design Support System for Piping Route and Differential Pressure Flowmeter by Three-Dimensional Fluid Analysis
Authors: Miura Takatsugu , Igarashi Kingo , Hosaka Tomoyuki , Kitagawa Takumi , Yashiki Tatsurou , Itabayashi Yuki , Suzuki Hirotsugu , Nakahara Takashi , Kitamura Junichi , Sano Tadashi 
Published: October 2020

Abstract
In power plants that becoming more compact, it will expend much time and effort to satisfy the requirement for the differential pressure flow measurement according to ISO’s standards. Therefore, it is difficult for engineers in the design phase to completely remove the potential for large errors in flow measurement. This paper presents the 3D fluid analysis system that is a lower cost than the conventional method to confirm the soundness of such measurement in the phase of piping route design.
This system has the function to automatically generate the analysis models from general 3D piping CAD data. The analysis program is written by the open source code to reduce a license fee. Also, this system has the function of calculating the swirl strength along the pipe axis as one of the means for efficiently supporting the design change. In order to verify and validate the analysis system, we analyzed several flow paths, confirmed the response of the swirl strength and flow rate indication value of the differential pressure flowmeter model. The analysis result well simulated the increase or decrease swirl strength in the complex flow path, and fluctuation of the flow rate indication value. Also, the system supports to set the flowmeter in the appropriate position by providing visualization of the swirl strength along the pipe axis. In the flow path analysis in this validation, it took about one month to visualization of the swirl strength along the pipe axis from the generation of the analysis models.
The 3D fluid analysis system collaborative with 3D piping CAD design system has been developed. This system enable to confirm the effects of swirl strength on flow measurement and the soundness of the differential pressure flow measurement at a lower cost in comparison with conventional method.
Development of Safety-Enhanced Fast Reactor by Using Minor Actinide Bearing Internal Blanket
Authors: Fuchita Sho , Takeda Satoshi , Fujimura Koji , Takeda Toshikazu , Fujimata Kazuhiro 
Published: October 2020

Abstract
For a 750MWe sodium-cooled fast reactor core using MOX fuel, safety-enhancement measures have been studied to reduce the risk of core damage under unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents.
As passive measures the followings are considered: 1) adoption of the axial heterogeneous core configuration with sodium plenum and Gas Expansion Modules (GEMs) to lower sodium void reactivity for ULOF, and 2) addition of minor actinides (MAs) as burnable absorber and fertile nuclides to the internal blanket in the inner core to reduce burnup reactivity for UTOP.
In this study, configurations of the safety-enhanced core were optimized based on sensitivity studies as follows.
Firstly, effects of 1) above on the sodium void reactivity were evaluated by changing the inner core height, B-10 content of the upper shield, GEMs, and standby position of the backup control rods, which are the dominant factors of core behavior in the event of ULOF.
Secondly, the effects of 2) above on the burnup reactivity were evaluated by changing the MA content in the internal blanket and the burnup period, which are the dominant factors of UTOP.
Finally, by utilizing sensitivity analysis results, the safety-enhanced core which satisfies the provisional design goals has been developed. This core has negative sodium void reactivity and burnup reactivity less than 1 $.
Development on Simulation Method for Two-Phase Flow in Large Diameter Pipes With 90 Degree Elbows
Authors: Komuro Yoshiteru , Kodama Atsushi , Kondo Yoshiyuki , Tanimoto Koichi , Hibiki Takashi 
Published: October 2020

Abstract
Two-phase flows are observed in various industrial plants and piping systems. Understanding two-phase flow behaviors such as flow patterns and unsteady void fraction in horizontal and vertical pipes are crucial in improving plant safety. Notably, the flow patterns observed in a large diameter pipe (approx. 4–6 in or larger) are significantly different from those observed in a medium diameter pipe. In a vertical large diameter pipe, no slug flow is observed due to the instantaneous slug bubble breakup caused by the surface instability. Besides, in a horizontal pipe, flow regime transition from stratification of liquid and gas to slug (plug) flow that induces unsteady flow should be taken into account. From this viewpoint, it is necessary to predict the flow regime in horizontal and vertical large diameter pipes with some elbows and to evaluate the unsteady flow regime. In this study, the simulation method based on the two-fluid model is developed. The two-fluid model is considered the most accurate model because the governing equations for mass, momentum, and energy transfer are formulated for each phase. When using the two-fluid model, some constitutive equations should be given in computing the momentum transfer between gas and liquid phases. In this study, several state-of-art constitutive equations of the bubble diameter, the interfacial drag force and non-drag forces such as the lift force and the bubble-bubble collision force, are implemented in the platform of ANSYS FLUENT. The developed simulation method is validated with visualization results and force from an air-water flow at the elbow of the piping system.
Heat Transfer for Supercritical Flow With TRACE
Authors: Spore Jay , Roth Glenn 
Published: October 2020

Abstract
Flow regimes at water pressures above the critical pressure are characterized as supercritical flow. Supercritical flows have no phase change. The heat transfer from the wall to the fluid is single phase (there is no boiling or condensation). Experimental data indicate that for conditions that involve supercritical single-phase heat transfer, the Dittus-Boelter heat transfer correlation can be in significant error.
A pseudo-critical temperature can be defined as a function of pressure for pressures that exceed the supercritical pressure. The pseudo-critical temperature is defined for heat transfer purposes as the temperature at which the specific heat peaks as the pressure is held constant. There is significant variation in fluid properties across the heat transfer boundary layer at temperatures near the pseudo-critical temperature. The large variation in properties is the reason for the failure of the Dittus-Boelter heat transfer correlation.
Comparisons to experimental data indicate that the Mokry heat transfer correlation is a significant improvement over the Dittus-Boelter heat transfer correlation for single phase supercritical heat transfer. The Mokry correlation was chosen to be included into TRACE.
Irradiation Induced Crosslinking in “Green” Polylactic-Acid (PLA) Polymers for Enhanced Strength and Elevated Temperature Applications
Authors: Jiang W. , Bakken A. , Taleyarkhan R. P. 
Published: October 2020

Abstract
This paper presents interdisciplinary (nuclear-mechanical-materials-chemical) engineering technology and results pertaining to use of ionization radiation for tailoring “green”, renewable corn-soy based amorphous and crystalline form polymers for use as low-to-high temperature adhesives. Both amorphous and crystalline form polymer forms of such the Polylactic-Acid (PLA) polymer were studied with and without photon irradiation, alongside with and without cross-linking agent. In order to study and enhance the high-temperature application of PLA as a novel, multi-purpose adhesive, small concentrations of the crosslinking agent triallyl isocyanurate (TAIC) were included into molten amorphous and semi-crystalline PLA cast as glue sticks, followed with Co-60 gamma-irradiation. Bond strength variations were studied in tensile mode at room temperature using the resulting adhesive in between two steel dowels (head-to-head bonded) as well as via shear strength testing at elevated temperatures (50–120°C) under a set pre-load of 222 N. It was found that gamma irradiated samples with TAIC exhibited noticeably improved bonding strength, and importantly, such strength can potentially prevail towards 100°C. These are exciting results which offer potential for application for building construction and safety enhancements especially under fires and similar accidents. Samples without TAIC exhibited significant loss of strength past 90°C. The full paper will discuss details of apparatus, modeling and simulation of irradiation dose delivery, testing protocols results, and future enhancements via hybrid neutron-photon-electron irradiation for utility in variety of industrial applications.
Experimental and Numerical Research on Steam Direct Contact Condensation Process in Automatic Depressurization System of AP1000
Authors: Zhang Yuhao , Feng Li , Qiu Zhimin , Fu Jingpin , Lu Daogang 
Published: October 2020

Abstract
In the third generation pressurized water reactor AP1000 plant, the Automatic Depressurization System (ADS) is one of the most important passive safety system. However, the steam Direct Contact Condensation (DCC) microscopic mechanisms are very complicated, which are not very clear yet. Moreover, the high-pressure and high-temperature experiment is very expensive to be conducted for many different test conditions. So in the present work, both the experimental and numerical methods are employed to investigate the steam DCC behavior. The steam DCC experimental bench has been built up, and the key parameters including the flow patterns and steam core temperature distributions are measured to provide validation data for the numerical results. In aspect of the numerical work, CFD simulation on the steam condensation is conducted. The heat and mass transfer process is simulated through the three-dimension commercial software FLUENT 16.0. Some of the key heat and mass transfer correlations are added by User Defined Function (UDF). The key parameters including the condensation steam fraction, temperature, and pressure, etc. are analyzed, which reflect the major heat transfer characteristics. According to the results, the expansion-compression-steam tail could be observed in both the numerical and experimental results. In essential, the steam fraction, temperature, and pressure distributions are determined by the equilibrium and transformation between the thermal dynamic energy and kinetic energy. The results provide working references for the practical ADS steam spraying condensation process in AP1000 reactor.
Corrosion Property of Container Using Hybrid Material for Thermal Decomposition Process of Sulfuric Acid
Authors: Ioka Ikuo , Kuriki Yoshiro , Iwatsuki Jin , Kawai Daisuke , Inagaki Yoshiyuki , Kubo Shinji , Inagaki Yoshiyuki 
Published: October 2020

Abstract
A thermochemical water-splitting iodine-sulfur process (IS process) is one of candidates for the large-scale production of hydrogen using heat from nuclear energy. Severe corrosive environment which is thermal decomposition of sulfuric acid exists in the IS process. To achieve an industrialization of massive hydrogen production system, one of the key factors is the development of structural materials for the severe corrosive environment. A hybrid material with the corrosion-resistance and the ductility had been made by a silicon powder plasma spraying and laser treatment. To confirm the applicability of the hybrid material as the structural material, corrosion tests of the hybrid materials had been performed in 95 mass% and 47 mass% boiling sulfuric acid. The corrosion resistance of specimen in the condition of 95 mass% boiling sulfuric acid had been excellent. This was attributed to the formation of SiO2 on the surface. To confirm the production characteristics as a container using the hybrid material, the container which has a welded part, a chamfer, a curved surface had been experimentally made. A configuration of the container had been 150mm inside diameter, 120mm in height and 6mm in thickness. The substrate of the container made of Hastelloy C276® superalloy had included TIG weld part. To improve the corrosion resistance of the container, pre-oxidation was performed at 800°C for 100 hours in air. There was no detachment of the plasma spraying and laser treated layer on the base metal and the welded part. The pre-oxidized container using hybrid technique was prepared for the corrosion test in boiling sulfuric acid to evaluate the characteristics of the container.
Study on the Liquid Seal Discharge Process in an Over-Pressurized Accident
Authors: Wu Dan , Deng Jian , Du Sijia , Qian Libo 
Published: October 2020

Abstract
In an over pressure accident, one or more pressurizer safety (or relief) valves will open due to the rapid pressure rise process. Once the safety (or relief) valves are open, the liquid seal will be discharged, and this will generate great discharge force to the downstream pipes. Multi-level protection is chosen using pressurizer safety (or relief) valves with different setpoint in most of Nuclear Power Plant, especially in the self-designed Generation-III Nuclear Power Plants. As the over pressure accident progresses, one or more safety (or relief) valves will be open. The downstream pipes will experience one or more times of impacts, which will influence the arrangement of the pipes. The whole discharge process is very complex, and the key influence factors are the pressure rise rate, safety (or relief) valve opening time, liquid seal temperature and volume, and the arrangement of the downstream discharge pipes. In present paper, liquid seal discharge process in an over pressure accident is studied. The pressure rise rate is so fast that three safety (or relief) valves will open one after another, which will generate three impacts on the downstream discharge pipes. It is found that for a specific design of Nuclear Power Plant, well design of the safety (or relief) valve setpoint is very important to the discharge force analysis results.