bluuebunny/crossref_metadata_2025_split huggingface dataset:
Study on the Impact Limiter Design in Spent Fuel Transfer Cask in Nuclear Power Plants
Authors: Hao Yuchen , Li Yue , Wang Jinhua , Wu Bin , Wang Haitao 
Published: October 2020

Abstract
In nuclear power plants, the amount of spent fuel stored on-site is limited. Therefore, it is necessary to be shipped to off-site storage or disposal facilities regularly. The key risk in the transfer of spent fuel involves a release of radiation that could cause harmful effects to people and the environment. Transfer casks with impact limiters on both ends are always employed to ensure safe containment of radioactive materials, which should be verified by the 9 meters drop test onto an unyielding surface according to IAEA SSR-6. In this paper, we focus on the influence of the impact-limiter parameters, including geometry dimensions and mechanical properties, on the results of drop events to achieve an optimized approach for design. The typical structure of impact limiter is bulk energy-absorbed material wrapped by thin stainless-steel shells. Compared to traditional wood, foam has advantages of isotropy and steady quality. In this paper, theoretical and numerical methods are both adopted to investigate the influence of impact limiters during hypothetical accidental conditions for optimizing buffer influence and protecting the internal fuel components. First of all, a series of polyurethane foam is selected according to the theoretical method, because its mechanical property is related to density. Therefore, using explicit finite element method to investigate the influence of parameters of foam in impact limiter. These discrete points from the above result can be utilized to establish damage curves by date fitting. Finally, a design approach for spent fuel transfer cask is summarized, to provide a convenient formula to predict the damage and optimize structure design in drop condition. Furthermore, this design approach can be applied in the multi-module shared system of SNF, which can contain different fuel assemblies.
Preventing Nuclear Fuel Material Adhesion on Glove Box Components Using Nanoparticle Coating
Authors: Segawa Tomoomi , Kawaguchi Koichi , Ishii Katsunori , Suzuki Masahiro , Tachihara Joji , Takato Kiyoto , Okita Takatoshi , Satone Hiroshi , Suzuki Michitaka 
Published: October 2020

Abstract
To minimize retention of nuclear fuel materials in glove box components and curtail the external exposure dose, plutonium and uranium mixed oxide powder adhesion prevention technology involving nanoparticle coating of the acrylic panels of the glove box is developed. The nanoparticle coating reduces the van der Waals force between alumina particles and the acrylic test piece surface because of formation of nano-sized rugged surfaces. The nanoparticle coating reduces the minimum adhesion force normalized by the particle diameter between the uranium dioxide particle and the acrylic test piece surface, for the smallest particle of about 5 μm associated with desorption, this minimum adhesion force reduced to about 5%. The nanoparticle coating also lowers the adhered plutonium and uranium mixed oxide powder amounts on the acrylic test piece to about 10%. This study reveals that applying the nanoparticle coating to the acrylic panels of the glove box prevents adhesion of nuclear fuel materials. This method effectively reduces the retention of nuclear fuel materials in the glove box, lowers the external exposure dose, and improves the visibility of the acrylic panels.
The Choice of MOSFET Manufacturing Technique Used in Emergency Response Robot
Authors: Yining Liu , Renze Wang , Yapeng Yang , Jiangang Zhang , Ning Wang , Zongyang Feng , Linsheng Jia , Boning Liang 
Published: October 2020

Abstract
For the aim of helping the development of robots used in Radiological Emergency Planning and Preparedness, the Total Ionizing Dose (TID) effects on the threshold voltage shift (ΔVth) of different kinds of Metal-Oxide-Semiconductor Field Effect Transistor (MOSFET) with different geometry and different scaling technology was compared. The different gate width and length dependent between bulk Complementary Metal-Oxide-Semiconductor Transistor (CMOS) process and nanowire (NW) MOSFET as well as higher and lower technology node is noticed. The reason of this difference is explained from the aspects of Radiation Induced Narrow channel effect (RINCE) and Radiation Induced Short channel effect (RISCE). It is found that some studies in recent years have corrected the influence of negative bias temperature instability (NBTI) when considering radiation effects. The TID effects on ΔVth of several kinds of new devices such as MOSFETs with new layout geometry as well as Ge-channel and GaN channel MOSFETs are described which can be investigated more deeply.
Simulations of Experiments on Isothermal Containment Atmosphere Mixing Caused by Vertical Injection
Authors: Krpan Rok , Tiselj Iztok , Kljenak Ivo 
Published: October 2020

Abstract
An experiment performed in SPARC experimental facility, was simulated with the computational fluid dynamics code OpenFOAM. The experiment took place in two phases. In the first phase, a helium-air layer was generated, which was then eroded with a vertical air jet injected in the vessel axis during the second experimental phase. A three-dimensional and a quasi-two-dimensional numerical models of a cylindrical vessel were developed and mesh convergence studies were performed. The mixing process was simulated as a single-phase flow with common momentum equation. Included gas species mass fractions are considered as passive scalars and are calculated using the transport equation. However, term describing the molecular diffusion cannot be neglected in our case and had to be added to diffusion equation implemented in default OpenFOAM solver. The k-ε turbulence model with additional buoyancy term implemented in OpenFOAM was used for turbulence modelling. Despite improvement of the physical model and following the Best Practice Guidelines, the results obtained with OpenFOAM CFD code still, at some locations, differ substantially from the experimental results. A modified definition of low Reynolds number eddy viscosity correction function is proposed, which significantly improves the agreement between measurements and calculation results.
Research on the Simulation of Radiographic Imaging Based on Cosmic-Ray Muon Scattering and the Optimization of PoCA
Authors: Hou Lin-jun , Zhang Quan-hu , Huo Yong-gang , Zuo Wen-ming , Chen Qi-fan , Su Xiang-hua , Xu Zhi-chun 
Published: October 2020

Abstract
Muon scattering tomography technology can be of great use in detecting nuclear materials under excellent shielding, thus distinguishing itself from traditional radiography technology. In this study, we firstly introduced the principle of this radiography technology and the rationale of PoCA algorithm used in image reconstruction. Then several experimental models were constructed with the help of Geant4 toolkit, and several sets of simulation experiments were carried out to test the performance of this technology with PoCA algorithm. In order to improve the imaging quality, we designed experiments to find out the cause of distorted image, and then proposed an optimization method. Finally, a quantity was created to present the imaging quality, and several experiments were carried out to validate the effect of this optimization method. The result showed that this method can significantly improve the performance of this algorithm.
An Image Quality Improvement Method Based on Neural Network for Digital Radiography Security Inspection System
Authors: Leng Zhiying , Wang Zhentao 
Published: October 2020

Abstract
As an essential method for security inspection in nuclear facilities, digital radiography has the ability to find hidden contraband efficiently. However, the images obtained by current scanning digital radiography system can be degraded by several factors, such as statistical noise and response time of detectors. At high scanning speed, the statistical noise and vibration of the system deteriorates the quality of images. In addition, the reduction of image quality will influence the accuracy of image observation and recognition. To meet the demand of detection efficiency and quality, it is necessary to guarantee the quality of images under high scanning speed.
Thus, to improve image quality of vehicles’ digital radiography at a certain scanning speed, we proposed an approach (VDR-CNN) to reduce or eliminate image noise, which is a convolutional neural network (CNN) with residual learning. The high-quality images obtained at low scanning speed of system served as the ground-truth image for VDR-CNN, while the low-quality counterpart corresponding to the high scanning speed served as the input. Then, the two images mentioned above constitute a training pair. By training this network with a set of training pairs, the mapping function of promoting image quality will be automatically learned so that the restored image can be obtained from the low-quality counterpart through the trained VDR-CNN. Moreover, this method avoids the difficulty in figuring and analyzing the complicated image degradation model.
A series of experiments was carried out through the 60Co inspection system developed by Institute of Nuclear and New Energy Technology, Tsinghua University. The experimental result shows that this method has attained a satisfying result in denoising and preserving details of images and outperforms BM3D algorithm in terms of both image quality improvement and the processing speed. In conclusion, the proposed method improves the image quality of vehicles’ digital radiography and it is proved better than traditional methods.
Numerical Simulation Method Research on Pellet-Cladding Mechanical Interaction Based on ABAQUS Software
Authors: Tang Changbing , Jiao Yongjun , Li Yuanming , Zhou Yi , Zhang Kun 
Published: October 2020

Abstract
The cladding acts as the first barrier to prevent the release of radioactive fission products, requiring its structural integrity to be maintained throughout the whole operation period of nuclear reactor. Therefore, cladding failure due to PCI (pellet claading mechanical interaction) should be avoided as much as possible in fuel design and operating conditions. At the same time, it is necessary to achieve effective control of the cladding stress by limiting the power growth rate etc. However, in the manufacturing process of fuel rod, the MPS (missing pellet surface) defect is inevitably generated. This defect may lead to a substantial increase in the local stress of the cladding, which in turn exceeds its corresponding stress limit, resulting in cladding failure. Accurate simulation of fuel performance caused by such defects will help prevent such failures. The traditional fuel performance analysis codes are based on a 1.5D analysis framework and cannot handle the local asymmetry problem of fuel such as the MPS defect. In order to accurately simulate the PCI phenomenon caused by the MPS defect, this research establishes a fuel performance analysis code based on the ABAQUS software and this code is suit for the 2D and 3D conditions. Based on the established analysis code, the irradiation-thermal-mechanical behavior of nuclear fuel under typical II transient conditions was studied, and the sensitivity analysis of the influence of different MPS sizes on the local stress of cladding was carried out. The simulation results show that :(1)the mises stress, contact pressure and equivalent creep strain of the cladding may be unevenly distributed due to the MPS defect.(2)the MPS defect will result in a more severe contact pressure on cladding during power transient period, which may lead to failure of cladding and should be prevented. The simulation method established in this research could be very help for the performance analysis for the nuclear fuel rods.
MELCOR Modeling of Combined Accident Tolerant Fuel and Reactor Core Isolation Cooling System Operation
Authors: Faucett Chris , Beeny Bradley , Vierow Kirkland Karen 
Published: October 2020

Abstract
The work presented in this paper presents new techniques for modeling the combined use of the Reactor Core Isolation Cooling (RCIC) System and Accident Tolerant Fuel (ATF) in a Boiling Water Reactor (BWR). With guidance from Sandia National Laboratories’ Severe Accident Analysis department, a MELCOR BWR model was developed from open source literature. The demonstration shown herein simulates BWR long-term station blackout (LTSBO) conditions with the Nuclear Regulatory Commission’s (NRC) MELCOR severe accident analysis code. By combining state-of-the-art MELCOR modeling practices with new, physics-based RCIC System and ATF MELCOR inputs, this BWR model provides a contemporary platform for BWR severe accident simulations.
The authors are investigating the combined use of the RCIC System and ATF as a means of passively enhancing reactor safety. The benefits of this approach were evaluated by performing simulations using traditional fuel designs (i.e. Zircaloy cladding) and ATF with an iron-chromium-aluminum (FeCrAl) clad under BWR LTSBO conditions. ATF performance was evaluated using severe accident metrics, specifically event sequence timings and the hydrogen production rate from cladding oxidation. Preliminary results show delayed core degradation timelines with less hydrogen production for ATF simulations. Although the results are limited in scope, the presented analysis could easily be expanded to a full-scale uncertainty study that considers a range of severe accident boundary conditions.
This paper describes objective technical results and analysis. Any subjective views or opinions that might be expressed in the paper do not necessarily represent the views of the U.S. Department of Energy or the United States Government.
Thermal Efficiency Optimization of a Modular High Temperature Gas-Cooled Reactor Plant by Extraction Steam Distribution
Authors: Jiang Di , Dong Zhe 
Published: October 2020

Abstract
Modular high temperature gas-cooled reactor (MHTGR) is a small modular reactor (SMR) with inherent safety, which is suitable for load following to improve economic competitiveness. The heat regenerative system for MHTGR nuclear power plant, is crucial for the improvement of thermal efficiency. Traditionally, the enthalpy drop distribution method (EDM) is used to study the relationships between thermal efficiency and distribution of extraction steam. However, this strategy is mainly used for off-line design of steam turbine under rated conditions. For load following operation, it is hard to guarantee the extraction steam distribution of EDM due to the highly nonlinear “flowrate-pressure-temperature” coupling of the fluid network. Thus, in this paper, the thermal efficiency is derived analytically based on the steady state model of fluid network. Then the thermal efficiency optimization is cast into a nonlinear programming problem, in which physical constraints can be considered explicitly. The proposed method for extraction steam distribution is of significance for improving the thermal efficiency of normal operation of nuclear power plant.
Validation of Analysis Models on Relocation Behavior of Molten Core Materials in Sodium-Cooled Fast Reactors Based on the Melt Discharge Experiment
Authors: Igarashi Kai , Onuki Ryoji , Sakai Takaaki , Kato Shinya , Matsuba Ken-ichi , Kamiyama Kenji 
Published: October 2020

Abstract
In order to improve the safety of nuclear power plants, it is necessary to make sure measures against their severe accidents. Especially, in the case of a sodium-cooled fast reactor (SFR), there is a possibility of significant energy release due to formation of a large-scale molten fuel pool accompanied by re-criticality in the event of a core disruptive accident (CDA). It is important to ensure in-vessel retention that keeps and confines damaged core material in the reactor vessel even if the CDA occurs.
CDA scenario initiated by Unprotected Loss Of Flow (ULOF), which is a typical cause of core damage, is generally categorized into four phases according to the progression of core-disruptive status, which are the initiating, early-discharge, material-relocation and heat-removal phases for the latest design in Japan. During the material-relocation phase, the molten core material flows down mainly through the control rod guide tube and is discharged into the inlet coolant plenum below the bottom of the core. The discharged molten core material collides with the bottom plate of the inlet plenum. Clarification of the accumulation behavior of molten core material with such a collision on the bottom plate is important to reduce uncertainties in the safety assessment of CDA.
In present study, in order to make clear behavior of core melt materials during the CDAs of SFRs, analysis was conducted using the SIMMER-III code for a melt discharge simulation experiment [1] in which low-melting-point alloy was discharged into a shallow water pool. This report shows the validation results for the melt behavior by comparing with the experimental data.
Using Pipe Whip Analysis via the Finite Element Method to Underpin the Delineation Between High and Moderate Energy Lines
Authors: Hurst Antony M. , Bansal Pavitra 
Published: October 2020

Abstract
When a pressurised pipe in a reactor coolant system breaks, it results in hydraulic loads on the reactor containment system and escaping fluid exerts a thrust force on the pipe. The double-ended guillotine break (DEGB) is generally the most onerous loss of coolant accident (LOCA) in design of a reactor coolant system [1]. In addition to the hydraulic loads, the continuing thrust force on the broken end of the pipe generates a rapidly accelerating rotational displacement of the pipe section on the break side of the plastic hinge, the phenomenon called pipe whip. The whipping pipe has the potential to damage objects within the hazard zone therefore must be assessed. Currently a 20bar threshold is used in nuclear power plant (NPP) design of light water reactors (LWRs) to delineate high energy lines (HEL) and moderate energy lines (MEL). The threshold is used as a process to establish the requirement for additional pipe whip assessments to be performed as part of HEL guillotine break analysis. It is currently argued that no such studies are required for MEL. However the basis of using the 20bar threshold for not carrying out pipe whip assessment of MELs is not well understood.
The work presented here provides details of the finite element (FE) analyses undertaken to substantiate the 20 bar threshold used for the differentiation of HEL and MEL. Using the FE analysis method, a range of pipe characteristics, and pressures in the range of 10–50bar have been studied to determine whether a plastic hinge will occur, and whether pipe whip effects will be seen for that case. The FE results have also been used to assess the equivalent plastic strain criteria generally used to define the formation of a plastic hinge and initiation of pipe whip.
Nuclear Regulatory Acceptance of Certified Process Data Reconciliation (CPDR)
Authors: Jansky Andy , Langenstein Magnus 
Published: October 2020

Abstract
The utility industry is currently undergoing a substantial change from an analog to a digital infrastructure. Not only plant performance and utility profit are dependent on accurate plant operational parameters but, more importantly, set safety limits need to be met in order to ensure safe operation of nuclear power plants in particular. Using non-quality-assured process data for operational decisions can result in significant over- or under-power of the plant. In addition, all new technologies such as AI, IIoT, digital twin technology, etc. rely on robust process data as input, putting at risk the significance of the results from the continuing data processing (“garbage in, garbage out”).
One method, certified process data reconciliation, or CPDR, cuts through the vast amount of available process data and generates all relevant process values with the smallest uncertainty possible. 95% of all collected process data can be discarded after introduction of CPDR. With CPDR, plant operation and maintenance can be significantly optimized and utilities can profit by realizing e.g. power recovery and measurement uncertainty recapture (MUR). Because the focus on reconciled instead of measured values constitutes a paradigm shift, the application of CPDR needs to be communicated to Nuclear Regulators. This paper describes the approach and experience of the Regulator acceptance process in various countries around the globe.
Design of an Innovative Moisture Separator Technology for Use in Nuclear Power Plants: Experimental Validation — Part 2
Authors: Hersberger Carsten , Waldvogel Elias , Bopp Joshua , Ribi Beat 
Published: October 2020

Abstract
The Moisture Separator Reheater (MSR) is a key component of Nuclear Power Plants (NPP), both in terms of performance and prevention of erosion/corrosion. Wet steam is usually dried in a MSR by inertial separation of the liquid water using separator elements. Depending on the design of the MSR, the technology of the separator elements contributes significantly to its size and performance.
An innovative concept of separator panels was conceived by means of aerodynamic principles as outlined in part 1 of this paper [1]. Computational Fluid Dynamics (CFD) has been used to understand the working principles of various moisture separating devices. The investigated separator panels are designed to capture the water droplets in a region of flow separation (invisible pockets) within the separator channels.
To characterize the separation performance of these separator panels, a test rig has been developed and built at the University of Applied Sciences Northwestern Switzerland (FHNW). This test rig was then operated at typical MSR operating conditions. To meet the required moisture content and flow conditions, preheated water was injected into the saturated steam flow. In order to measure the residual moisture content after the separation the throttling calorimeter methodology has been adopted.
The newly designed panels have shown very good separation performance. According to the measurements carried out, a residual moisture content of less than 0.1 % can be guaranteed. The innovative technology, which clearly differentiates the OEM, for who this research was carried out, from its competitors, will allow considerable size and cost reduction as well as opportunities to retrofit existing MSRs.
MELCOR Validation Study on Multi-Room Fire
Authors: Louie David L. Y. , El-Darazi Samir , Fyffe Lyndsey M. , Clark James L. 
Published: October 2020

Abstract
Estimation of radionuclide aerosol release to the environment, from fire accident scenarios, are one of the most dominant accident evaluations at the U.S. Department of Energy’s (DOE’s) nuclear facilities. Of particular interest to safety analysts, is estimating the radionuclide aerosol release, the Source Term (ST), based on aerosol transport from a fire room to a corridor and from the corridor to the environment. However, no existing literature has been found on estimating ST from this multi-room facility configuration. This paper contributes the following to aerosol transport modeling body of work: a validation study on a multiroom fire experiment (this includes a code-to-code comparison between MELCOR and Consolidated Fire and Smoke Transport, a specialized fire code without radionuclide transport capabilities), a sensitivity study to provide insight on the effect of smoke on ST, and a sensitivity study on the effect of aerosol entrainment in the atmosphere (puff and continuous rate) on ST.