bluuebunny/crossref_metadata_2025_split huggingface dataset:
Development of Three-Dimensional Distribution Visualization Technology for Boron Using Energy Resolved Neutron-Imaging System (RADEN)
Authors: Abe Yuta , Tsuchikawa Yusuke , Kai Tetsuya , Matsumoto Yoshihiro , Parker Joseph D. , Shinohara Takenao , Ohishi Yuji , Kamiyama Takashi , Nagae Yuji , Sato Ikken 
Published: October 2020

Abstract
Boron carbide is used as a neutron-absorbing material in Fukushima-Daiichi Nuclear Power Station (1F), producing borides that are twice as hard as oxides (such as UO2 and ZrO2). The high neutron absorption of boron affects the evaluation of re-criticality during the process of debris retrieval. Therefore, it is important not only to determine the presence of boron but also to investigate the distribution of boron inside the material in a non-destructive manner during decommissioning. To address the uncertainties in the core material relocation behavior of boiling water reactor (BWR) during a severe accident (SA), solidified melt specimens of a simulated fuel assembly were prepared by plasma heating. If core material melting and relocation (CMMR) specimens can be used to estimate the B distribution in 1F Unit-3, that will provide valuable information in the decommissioning of 1F. To address this, the authors focused on the energy-resolved neutron-imaging system, RADEN, which utilizes a wide energy range, from meV to keV. This is an innovative three-dimensional analysis technology for boride distribution that affects the evaluation of hardness and re-criticality. In the calibration standard samples (ZrxB1-x and FexB1-x), there was a good correlation between boron concentration and the energy-dependence of the cross sections of cold and epithermal neutrons. In the CMMR specimens, boron distribution was confirmed from the contrast difference between cold and epithermal neutrons. In the future, the results of calibration standard samples will be applied to the results of CMMR specimens. With this method, three-dimensional boron distribution will be measured, and the understanding of boride distribution 1F Unit-3 will be improved, which may be reflected in an improved SA code.
Analysis of Reactivity Accidents in Modular HTGRs
Authors: Gorton Jacob P. , Brown Nicholas R. 
Published: October 2020

Abstract
A control rod withdrawal (CRW) is a possible reactivity initiated accident (RIA) in modular high-temperature gas-cooled reactors (mHTGRs). The purpose of this study is to perform a sensitivity analysis of a CRW event in an mHTGR model using the systems code RELAP5-3D with point kinetics feedback to demonstrate the impact of uncertainty in heat transfer and reactor kinetic parameters. The adaptive Sobol decomposition method in the uncertainty quantification code RAVEN is used to perform the sensitivity study and to determine which input parameters have the greatest impact on the figures of merit, which in this case are peak reactor power and maximum fuel temperature. This study addresses a need highlighted by the Nuclear Regulatory Commission (NRC) for transient fuel testing by quantifying the impact of uncertainty in heat transfer and reactor kinetic parameters and by generating potential boundary conditions for transient testing of conventional mHTGR fuel.
Proposal of Inspection Rationalization Method and Application for Sodium Cooled Fast Reactor
Authors: Hiroki Yada , Shigeru Takaya , Yasuhiro Enuma 
Published: October 2020

Abstract
In order to rationalize maintenance for nuclear power plants, it is necessary to develop optimize maintenance plan by considering characteristics of each plant. In sodium-cooled fast reactor, there are constraints on inspections due to the specialty of handling sodium equipment, that is one of the important points when considering rationalization of maintenance. To solve this problem, we proposed a basic concept of maintenance optimization scheme that is a design support tool in order to develop maintenance strategy, based on “system based code (SBC)”. SBC is a concept to optimize the reliability of a nuclear power plant by consideration of all related technical requirements. “ASME Code Case N-875” and “ASME Boiler and Pressure Vessel Code, Section XI, Division 2 (RIM)” based on system based code were already developed as standards for inspection. One of the proposed scheme goals is to make a concrete way of necessary assessment method. Another is to provide several combinations of design and maintenance, and information for owner in order to choose the acceptable combination. In the beginning, we are working to develop the scheme that can be applied to sodium fast reactor as the main concept of next generation reactor. In this context, primary heat transfer system (PHTS) piping of fast reactor was evaluated by the scheme. This piping was chosen because it is major significant component and the inspection have constraint conditions that need preparation work. As a result, design candidate (e.g. single and double wall piping) and inspection candidate (e.g. ultrasonic testing and continues leakage monitoring) combinations along with benefit of each cases were provided.
Determining Airborne Release Fraction From DOT 7A Drums Exposed to a Thermal Insult
Authors: Mendoza Hector , Figueroa Victor G. , Gill Walter , Sanborn Scott 
Published: October 2020

Abstract
Fire suppression systems for transuranic (TRU) waste facilities are designed to minimize radioactive material release to the public and to facility employees in the event of a fire. Currently, facilities with Department of Transportation (DOT) 7A drums filled with TRU waste follow guidelines that assume a fraction of the drums experience lid ejection in case of a fire. This lid loss is assumed to result in significant TRU waste material from the drum experiencing an unconfined burn during the fire, and fire suppression systems are thus designed to respond and mitigate potential radioactive material release. However, recent preliminary tests where the standard lid filters of 7A drums were replaced with a UT-9424S filter suggest that the drums could retain their lid if equipped with this filter. The retention of the drum lid could thus result in a very different airborne release fraction (ARF) of a 7A drum’s contents when exposed to a pool fire than what is assumed in current safety basis documents. This potentially different ARF is currently unknown because, while studies have been performed in the past to quantify ARF for 7A drums in a fire, no comprehensive measurements have been performed for drums equipped with a UT-9424S filter. If the ARF is lower than what is currently assumed, it could change the way TRU waste facilities operate. Sandia National Laboratories has thus developed a set of tests and techniques to help determine an ARF value for 7A drums filled with TRU waste and equipped with a UT-9424S filter when exposed to the hypothetical accident conditions (HAC) of a 30-minute hydrocarbon pool fire. In this multi-phase test series, SNL has accomplished the following: (1) performed a thermogravimetric analysis (TGA) on various combustible materials typically found in 7A drums in order to identify a conservative load for 7A drums in a pool fire; (2) performed a 30-minute pool fire test to (a) determine if lid ejection is possible under extreme conditions despite the UT-9424S filter, and (b) to measure key parameters in order to replicate the fire environment using a radiant heat setup; and (3) designed a radiant heat setup to demonstrate capability of reproducing the fire environment with a system that would facilitate measurements of ARF. This manuscript thus discusses the techniques, approach, and unique capabilities SNL has developed to help determine an ARF value for DOT 7A drums exposed to a 30-minute fully engulfing pool fire while equipped with a UT-9424S filter on the drum lid.
Risk-Informed Analysis for Accident Tolerant Fuel in Pressurized Water Reactors
Authors: Ma Zhegang , Parisi Carlo , Davis Cliff , Zhang Sai , Zhang Hongbin 
Published: October 2020

Abstract
This paper presents the research activities performed by Idaho National Laboratory (INL) for the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk-Informed System Analysis (RISA) Pathway, Enhanced Resilient Plant (ERP) Systems research, using the probabilistic risk assessment (PRA) tool SAPHIRE and the deterministic best estimate tool RELAP5-3D for risk-informed analysis. The ERP research supports DOE and industry initiatives by developing Accident Tolerant Fuel (ATF), the Diverse and Flexible Coping Strategy (FLEX), and passive cooling system designs to enhance existing reactors’ safety features (both active and passive) and to substantially reduce operating costs of nuclear power plants (NPPs) through risk-informed approaches to analyze the plant enhancements and their characterization.
The risk-informed analysis used SAPHIRE and RELAP5-3D to evaluate the risk impacts from near-term ATF (FeCrAl and Chromium-coated clads) on a generic Westinghouse three-loop pressurized water reactor (PWR) under the following accident scenarios: station blackout (SBO), loss of feedwater (LOFW), steam generator tube rupture (SGTR), loss-of-coolant accidents (LOCAs), locked rotor transient, turbine trip transient, anticipated transient without scram (ATWS), and main steam line break (MSLB). The RELAP5-3D simulations included the time to core damage, time to 0.5 kilograms hydrogen generation, and total hydrogen generation. The simulation results show there are modest gains of coping time (delay of time to core damage) due to efficacy of the near-term ATF designs in various accident scenarios. The risk benefits on behalf of the core damage frequency (CDF) brought by the ATF designs would be small for most of the scenarios. However, results revealing much less hydrogen being produced at the time of core damage show a clear benefit in adopting ATFs.
Experimental and Numerical Investigation of Thermal Performance of Synthetic Jet Impingement
Authors: Singh Pushpanjay K. , Kothari Rohit , Sahu Santosh K. , Upadhyay Prabhat K. , Singh Shashwat 
Published: October 2020

Abstract
Synthetic jet potentially useful in electronics cooling is investigated both numerically and experimentally. In the present study, a confined three dimensional synthetic jet with sinusoidal moving wall is considered. Computations are carried out using the FLUENT software with the coupled user defined function describing the diaphragm movement. In this study the effect of various geometrical parameters influencing the flow field and heat transfer are investigated. The effects of change in orifice geometry (circular, square and rectangular), orifice aspect ratio, and jet-to-plate distance are studied for a given hydraulic diameter. The heat transfer results obtained from the synthetic jet is compared with the continuous jet. An electromagnetic actuator is used as an oscillating diaphragm for the generation of synthetic jet. A stainless steel foil with 0.05 mm thickness is used as the test specimen. The surface temperature of the test specimen is measured by using a thermal imaging technique during synthetic jet impingement and a constant temperature anemometer has been employed for velocity measurement. Tests are carried out for Reynolds number of 5448, varied range of jet-to-plate distance (1–18). The maximum value of the heat transfer coefficient is found to be 16 times more than the heat transfer coefficient for natural convection.
Research on Hydraulic Model Test of Pumping Station Forebay
Authors: Jian Jiale , Wang Fang , Zhang Rongyong , Tang Benjing 
Published: October 2020

Abstract
A physical model with a scale of 1:20 was used to study the hydraulic characteristics of the flow in the forebay area of the pump station in Tian Wan Nuclear Power Plant Project Unit 5&6. The original layout scheme of the pump station forebay is limited by space, the turning radius of the intake gallery is small and the straight line section is short, the uniformity of water distribution is poor after seawater enters the forebay, the flow in the forebay cannot fully diffuse, the mainstream is concentrated, the flow distribution in the inlet section of each channel of the pump station is uneven, and the flow pattern is poor. Firstly, the distribution of water flow and uniformity of water distribution in the pump station forebay under different turning radius of tunnel are compared, and the turning radius of intake corridor is determined by calculating the head loss in different parts of tunnel. A variety of rectification facilities are arranged in the pumping station forebay, including modifying the grid type, setting diversion wall, diversion plate, energy dissipation beam, rectification bottom sill, etc. The recommended scheme is determined by hydraulic model test. The scheme can satisfy the requirement of uniformity of water distribution in the forebay, and the flow pattern along each section of the channel is also good.
Inter-Unit Common Cause Failure Analysis Based on Data From Intra-Unit Cases
Authors: Higo Eishiro , Soga Shota , Miura Hiromichi 
Published: October 2020

Abstract
This study provides an inter-unit common cause failure (CCF) analysis method utilizing the international CCF database. The inter-unit CCF is one of the possible major risks at a site with multiple units. The conventional CCF analysis is based on the CCF database, in which inter-unit CCF events rarely occur in real nuclear power plants. The conventional approach cannot be directly applied to inter-unit CCF cases because of the lack of data. The method ignores the asymmetricity among units, so it may lead to overly conservative outcomes if it is applied to inter-unit CCF. We have proposed a new concept, “commonality factor,” which represents the degree of similarity among components in different units and showed a concept that the inter-unit CCF probability can be expressed in terms of the intra-unit CCF probability with the commonality factor. The commonality factor is approximated based on CCF coupling factors, which tie two or more failures together as a CCF. This study explains how to estimate the commonality factor by analyzing the conventional CCF database and judging commonality among components from different units. The proposed method is demonstrated through two preliminary examples.
Sensitivity Analysis of External Exposure Dose for Future Burial Measures of Decontamination Soil Generated Outside Fukushima Prefecture
Authors: Shimada Asako , Sawaguchi Takuma , Takeda Seiji 
Published: October 2020

Abstract
A large area of east Japan was contaminated by radiocesium following a nuclear accident at the Fukushima Daiichi Nuclear Power Station. Following decontamination of the soil, external effective dose conversion factors were calculated by changing the volume of decontamination soil, depth of cover soil, and distance of the evaluation point from the decontamination soil. The decrement of the factors with an increase of the distance was larger for the smaller volume of decontamination soil. The factors decrease exponentially with an increase of the depth of cover soil in all cases. When there was no cover soil, annual external exposure doses for residents at 1 m from the repository site and public entry were over 10 μSv/y, even for the smallest size (2m × 2m × 1m) and 50 percentile value of radiation concentration (700 Bq/kg). When the surface was covered by 30 cm of non-contaminated soil, the annual external exposure doses were less than 10 μSv/y for the largest size (200m × 200m × 10m) and 95 percentile concentration (2500 Bq/kg).
Numerical Simulation of the Air-Water Two-Phase Flow Across a 90-Degree Vertical-Upward Elbow
Authors: Qiao Shouxu , Zhong Wenyi , Hao Sijia , Qi Peiyao , Tan Sichao 
Published: October 2020

Abstract
The present study investigates the air-water two-phase flow across a 90-degree vertical-upward elbow with the computational fluid dynamics (CFD) simulation. The Eulerian-Eulerian two-fluid model and the Multi Size Group (MUSIG) model are used to predict the development of the detailed interfacial structures between the two phases. The axial development of the void fraction and the interfacial area concentration are investigated and benchmarked with the experimental data measured using the four-sensor conductivity probe. It is concluded that CFD simulation can predict the characteristics distributions of void fraction and interfacial area concentration and their development downstream of the elbow. The double-peaked void fraction distribution is found to be caused by the secondary flow induced by the elbow. The liquid phase on the outer curvature moves to the inner curvature and forms a double counter rotating vortex, entraining the bubbles to form a double-peaked distribution. The elbow effects become dissipated between 33 and 63 hydraulic diameters. The simulation results of liquid-phase and gas-phase parameters can be used to develop the theoretical two-phase flow models for the elbow region.
Simulation for Predicting Condition of Plant Equipment
Authors: Kashiwase Shoichi , Osaki Kenji , Hatakeyama Makoto , Kaneko Tomokazu 
Published: October 2020

Abstract
Condition-based maintenance (CBM) is a maintenance program that recommends maintenance decisions based on maintenance information by condition monitoring. To apply CBM to industrial systems, it is important to diagnose condition of equipment appropriately based on the knowledge of abnormal condition of the equipment. In nuclear power plant, however, it is quite difficult to obtain the knowledge of abnormal condition. For this reason, simulation is expected to be an effective approach to estimate conditions of equipment instead of data acquisition. In this study, we focused on and built models for a motor operated valve and a pump, since the inspection cost is generally expensive due to the number of them in nuclear power plant. We considered a pump system and analyzed the measurement values in normal and abnormal conditions. Our analytical result reproduces the behavior of the measurement values and shows the characteristic deviation from normal condition in abnormal conditions.
A Series of Investigations Inside of Primary Containment Vessel of Fukushima Daiichi Nuclear Power Station Unit2
Authors: Kobayashi Taiji , Yasuda Toshihiro , Sugiura Tessai , Terai Fujio , Hirose Yukinori , Uranishi Atsuyoshi , Sakamoto Naoya 
Published: October 2020

Abstract
In January 2012, Toshiba Energy Systems & Solutions Corporation inserted a camera into the damaged Primary Containment Vessel (PCV) of Fukushima Daiichi Nuclear Power Station unit 2 for the first time since the Great East Japan Earthquake. Since then, we, as a member of International Research Institute for Nuclear Decommissioning (IRID), has developed several devices and has investigated inside of PCV.
The investigation devices for unit 2 are roughly categorized into three types. For the investigations in early stages, just a camera and sensors with simple support tools were inserted through a small hole. Then small crawler type robots were developed to expand the investigation area to inside of Pedestal which exists the central part of PCV. Since the devices found that there were not enough areas to run around by the robots in the Pedestal, a long telescopic guiding pipe device was chosen to deliver cameras, sensors, and tools into the Pedestal. This type of device took movie of wide area of inside of the Pedestal. An image processing software was also developed to get a high quality and extensive picture from the movie taken by a low resolution camera and insufficient lighting. In February 2019, the other telescopic guiding pipe device finally delivered a camera and fingers like tong to grab deposits fallen on the bottom of pedestal for the first time after the earthquake, and found that some of deposits can be lift up.
Some of projects were implemented as subsidized projects of “Decommissioning and Contaminated Water Management” founded by the Ministry of Economy, Trade and Industry.
Metagenomic Sequencing To Detect Respiratory Viruses in Persons under Investigation for COVID-19
Authors: Babiker Ahmed , Bradley Heath L. , Stittleburg Victoria D. , Ingersoll Jessica M. , Key Autum , Kraft Colleen S. , Waggoner Jesse J. , Piantadosi Anne 
Published: October 2020

Broad testing for respiratory viruses among persons under investigation (PUIs) for SARS-CoV-2 has been performed inconsistently, limiting our understanding of alternative viral infections and coinfections in these patients. RNA metagenomic next-generation sequencing (mNGS) offers an agnostic tool for the detection of both SARS-CoV-2 and other RNA respiratory viruses in PUIs. Here, we used RNA mNGS to assess the frequencies of alternative viral infections in SARS-CoV-2 RT-PCR-negative PUIs (
n
= 30) and viral coinfections in SARS-CoV-2 RT-PCR-positive PUIs (
n
= 45).
Raising the Resurrection plate from an unfolded-slab plate tectonic reconstruction of northwestern North America since early Cenozoic time
Authors: Fuston Spencer , Wu Jonny 
Published: October 2020

AbstractThe configuration of mid-ocean ridges subducted below North America prior to Oligocene time is unconstrained by seafloor isochrons and has been primarily inferred from upper-plate geology, including near-trench magmatism. However, many tectonic models are permitted from these constraints. We present a fully kinematic, plate tectonic reconstruction of the NW Cordillera since 60 Ma built by structurally unfolding subducted slabs, imaged by mantle tomography, back to Earth’s surface. We map in three-dimensions the attached Alaska and Cascadia slabs, and a detached slab below western Yukon (Canada) at 400–600 km depth that we call the “Yukon Slab.” Our restoration of these lower plates within a global plate model indicates the Alaska slab accounts for Pacific-Kula subduction since ca. 60 Ma below the Aleutian Islands whereas the Cascadia slab accounts for Farallon subduction since at least ca. 75 Ma below southern California, USA. However, intermediate areas show two reconstruction gaps that persist until 40 Ma. We show that these reconstruction gaps correlate spatiotemporally to published NW Cordillera near-trench magmatism, even considering possible terrane translation. We attribute these gaps to thermal erosion related to ridge subduction and model mid-ocean ridges within these reconstruction gap mid-points. Our reconstructions show two coeval ridge-trench intersections that bound an additional “Resurrection”-like plate along the NW Cordillera prior to 40 Ma. In this model, the Yukon slab represents a thermally eroded remnant of the Resurrection plate. Our reconstructions support a “northern option” Farallon ridge geometry and allow up to ∼1200 km Chugach terrane translation since Paleocene time, providing a new “tomographic piercing point” for the Baja-British Columbia debate.