bluuebunny/crossref_metadata_2025_split huggingface dataset:
Design Analysis of Radiation Shielding Door in High-Level Waste Treatment Plant
Authors: Shen Jingyi , Chen Zonghuan , Wang Bingheng , Gao Guiling 
Published: October 2021

Abstract
In high-level waste treatment plants, rooms where the highly radioactive sources (such as the vitrification containers) are operated have high radiation protection risk. On the purpose of ensuring the safety of operators, it is necessary to design corresponding shielding walls, windows and shielding doors to guarantee that the dose rate at the personal operating location outside the room meets corresponding control requirements. Compared with walls, the radiation safety designs of the doors and windows are more complicated. In this article, a shielding door of a hot room inside a high-level waste treatment plant is selected. By means of combing the source characteristics and the source location, the gap between the door and door frame as well as the lap joints of the door and the wall are analyzed in order to characterize the influence factor of the dose rate outside the shielding door. The results illustrate that, under the premise that the shielding design requirements are satisfied, the lap joint between shielding wall and door has a significant impact on the outdoor dose field due to the oblique incidence of the radiation ray. Therefore, in the follow-up design of the radiation shielding door, A certain overlap form of the lap joints between the door and the wall need to be satisfied, furthermore, special attention need to be paid to the shielding of radioactive sources with weak self-shielding effect and strong penetration ability, for ensuring the shielding door to the greatest extent radiation protection ability.
Estimation of Vibration Characteristics of Nuclear Facilities Based on Seismic Observation Records
Authors: Yamakawa Kouki , Saruta Masaaki , Moritani Hiroshi , Yamazaki Hiroaki , Nishida Akemi , Kawata Manabu , Iigaki Kazuhiko 
Published: October 2021

Abstract
After the 2011 off-the-Pacific coast of Tohoku Earthquake, regulatory standards for nuclear power plants were tightened. Therefore, a three-dimensional (3D) finite element model to evaluate the local response of the reactor building is currently being developed for seismic response analysis.
In order to refine the 3D finite element model, it is important to verify the correspondence to the seismic observation behaviors. In this study, the authors analyze the basic response characteristics, such as the natural frequencies and modes of the reactor building, and evaluate the effects of the amplitude of the seismic excitation on the response characteristics based on seismic observation records. This is done to clarify the behavior of a reactor building during earthquakes. These analyses will assist in quantitatively evaluating the correlation between the natural frequency of the building and the amplitude of the seismic excitation. Furthermore, the ratios of rotational displacement and displacement caused by building deformation for natural modes are discussed.
Power Control System Design for a Heat Pipe Cooled Reactor
Authors: Sun Haowei , Sun Peiwei 
Published: October 2021

Abstract
Due to the small size, the high-power density and the strong environmental adaptability, the heat pipe cooled reactor has a broad application prospect in the fields of space exploration, underwater vehicle power system and other fields[1]. It will be one of the subversive technologies which will change the future nuclear power development. The safe operation of the heat pipe cooled reactor requires an effective power control system.
3KeyMaster is a comprehensive real-time simulation analysis platform and the control system can be built by using its logic module. The interface program can be developed in its task program, and other program can be integrated in this platform[2]. The system program of the heat pipe cooled reactor has compiled as a dynamic linking library. Through the task program of the 3KeyMaster simulation platform to call the dynamic linking library by using C++ program, the system program is coupled with the 3KeyMaster simulation platform. The modules of 3KeyMaster platform exchange data with the system program by shared memory.
Through the responses of the temperature due to step change in the reactor power obtained by 3KeyMaster, the transfer functions are obtained by using the system identification module in MATLAB & Simulink. The power disturbance is introduced, and simulation is carried out on the 3KeyMaster. The changes of the important parameters are analyzed to obtain its dynamic characteristics. Based on the dynamic analysis on the system responses, the control strategy for the power control system is proposed.
The power control system is implemented in MATLAB & Simulink. The control system parameters are tuned to determine a group of suitable control parameters[3]. Finally, the control performance of power control system is evaluated by simulations.
Using Monte Carlo Method and Adaptive Sampling to Estimate the Limit Surface
Authors: Zhang Lixuan , Zhang Zhijian , Wang He , Zhang Yuhang , Sun Dabin 
Published: October 2021

Abstract
In the research on the risk-informed safety margin characterization (RISMC) methodology, how to estimate the limit surface is important. Using the reduced Order Models (ROMs) to simulate calculations can obtain results more quickly and estimate the limit surface. For example, we use ROMs instead of Complex simulation model, Parameters that are critical to the safety of nuclear power plants, such as the peak temperature of the fuel cladding, can be calculated relatively quickly. Using Monte Carlo method to analyze nuclear accident is low efficiency and poor accuracy. To get relatively accurate results, a large amount of simulation experiments is needed. Based on adaptive sampling, the samples which will cause failure will be acquired more easily. Adaptive sampling uses the calculation results of the previous step to guide the next step of sampling, which can quickly obtain the samples points near the failure edge.
This article will introduce the definition of the limit surface and use the Monte Carlo method and the adaptive sampling to estimate the limit surface through ROMs. And compare the calculation results of the two methods and the number of samples required. The two methods are verified by a case.
Study on Deteriorated Heat Transfer in Upward Flow of Supercritical Water in a 1-M Vertical Bare Tube
Authors: Dort-Goltz N. , Pioro I. , McKellar J. 
Published: October 2021

Abstract
SuperCritical Water-cooled Reactors (SCWRs) represent potential improvements over traditional water-cooled reactors in many respects, including thermal efficiency. These reactors are still under development, however, thermalhydraulics data needed for their design are lacking. Experimentation is complex and costly. In spite of a large number of experiments in long bare tubes (pipes) cooled with SCW, developing SCWR concepts requires experimental data in bundle geometries cooled with SCW, which are usually shorter and will have smaller hydraulic-equivalent diameters.
As a first step, tests have been conducted by others on heat transfer in short, vertical bare tubes cooled with the upward flow of SCW. The objective of this work is to analyze that collected data with particular attention to the Deteriorated Heat Transfer (DHT) regime. The DHT regime is characterized by reduced Heat Transfer Coefficients (HTCs) and consequently increased wall temperatures. As such, it represents a hazard to the safe operation of a Nuclear Power Plant (NPP).
The results of this analysis indicate that DHT did occur in each of the tests analyzed, often seen as a gradual decrease in HTC along the heated length, but occasionally as a sharp “dip”. The DHT can occur along the heated length, when the bulk-fluid temperature is close to or within the pseudocritical region. The results also confirmed that the Dittus-Boelter correlation does not adequately predict HTCs within the pseudocritical region. Two other applied correlations (Gupta et al. and Mokry et al.) performed better, but neither was able to predict the occurrence of the DHT.
The results of this analysis will be of use to designers and developers of SCWRs, and can help to plan future experiments.
Experimental Study on Edge Blockage Accidents in a Narrow Rectangular Channel
Authors: Yuan Dongdong , Du Weian , He Yuhao , Zhu Jiahong , Zhang Yonghao , Li Chengwei , Tan Sichao , Li Dongyang 
Published: October 2021

Abstract
The blocked accident of plate fuel assembly is one of the serious accidents in the reactor. It will worsen the heat transfer capacity of the coolant and even affect the safe operation of the reactor. Therefore, Particle Image Velocimetry (PIV) technology and the Laser Induced Fluorescence (LIF) technology are used in current paper to investigate the edge blockage accidents in a narrow rectangular channel with a continuous-wave laser and two high-speed cameras. During the experiments, the obstacle has great effect on the flow field and temperature field. It is found that large-scale vortex structure will be generated in the wake downstream of the obstacle. And where the vortices exist, the fluid temperature is also higher, which means that the vortices prevent the heat from escaping in time. Besides, the temperature distribution at different Re numbers is also analyzed. It can provide a validation for numerical simulations.
An Optimized Dynamic Algorithm With Photon Attenuation Coefficient for Path-Planning in Radioactive Environments
Authors: Miyombo Miyombo Ernest , Liu Yongkuo , Ayodeji Abiodun 
Published: October 2021

Abstract
Radiation protection is central to ensure nuclear safety, proper operating conditions, and the prevention of accidents in nuclear facilities. Increasingly, path planning is being used to keep radiation risks as low as reasonably achievable (ALARA). However, most of the approach is sub-optimal, because they rely on exposure and roentgen definition only when considering different types of radiation and materials. Moreover, additional concepts are needed to determine the energy deposited in different body parts. In this study, we present an Optimized Dynamic Algorithm (ODA) for minimum dose path-planning in a radioactive environment based on photon interaction with matter and shortest path algorithm. The proposed ODA can estimate the energy flux per unit area reaching a standard adult from a radiation source using the probability of photon absorption in the air. The approach also utilizes the Compton effect of the energy flux to determine the fraction of energy transferred to electrons in the body. Two radioactive scenarios are investigated to evaluate the performance of the proposed ODA. The significant improvement obtained from the ODA is presented in this paper.
Atomic Insights on Interaction Mechanism of Dislocation With Void/Impurity/Precipitates in BCC Iron
Authors: Azeem M. Mustafa , Yun Di , Zubair Muhammad 
Published: October 2021

Abstract
Ferritic steels are important cladding and structural material for current and advanced reactors containing Fe as a primary compositional matrix element. Three proposed models containing voids, impurity P, and nano oxides embedded in Fe were simulated using molecular dynamics code LAMMPS. The interaction mechanism of dislocation for each model was observed and compared by analyzing its strengthening effect through the stress-strain curve. In considering the interaction of dislocation with voids, it was observed that for all void sizes, dislocation bypass voids following orowan unpinning mechanism without any loop irrespective of any radii, while for P clusters, dislocation is trapped at the interface and leaves it after making a screw dipole shape following orowan mechanism. Moreover, the oxide dislocation interaction in oxide dispersed in iron follow the orowan unpinning mechanism accompanied by loops around oxide. The values of critical unpinning stress for all models were analyzed and compared. Atomic insights on the interaction mechanism indicate that the interaction mechanism is essentially the same for void and a P cluster in Fe i.e. orowan unpinning mechanism. The presence of oxide in Fe matrix during its interaction follows the orowan unpinning mechanism along with a dislocation loop around the oxide and its area is proportional to the size of the oxide. The critical unpinning stress was compared and analyzed for each model.
This work will provide a unique and valuable insight into the strengthening mechanism due to dispersed oxides by providing new parameters for the multi-scale simulation. Our results demonstrate that oxide impedes dislocations and offers hindrance for edge dislocations to saturate further. Furthermore, our simulation results provide atomic insight into the onset of plasticity in iron alloys.
Study on Buckling Strength and Post Buckling Behaviors of Reactor Vessel Lower Heads
Authors: Murohara Masato , Yamazaki Akira , Sato Takuya , Kasahara Naoto 
Published: October 2021

Abstract
As the lessons learned from the Fukushima Daiichi Nuclear Power Plant accident, the importance of controlling the behavior after a failure and mitigating consequences of a failure was recognized. Conventional reactor structural design has been aimed at preventing the occurrence of failure due to Design Basis Events (DBE). This study aims to improve the resilience of the reactor structure under Beyond Design Basis Events (BDBE), such as very high temperatures and excessive earthquakes during severe accidents, by mitigating the consequences after failure.
Optimization of Maintenance Strategy for Sea Water Pumps in Nuclear Plants
Authors: Zhao Ling , Liu Deyi , Zhao Ming 
Published: October 2021

Abstract
Probabilistic Safety Analysis (PSA) methodology is a significant supplement to the deterministic safety analysis in the nuclear power plant. PSA can be used to evaluate the NPP device change, equipment maintenance, in service inspection. The practicability of modifying sea water pump maintenance programme is evaluated in this paper by determining the initial event probability in fault-tree and modifying the PSA calculation model. Based on the evaluation results, the preventive maintenance of sea water pump and 6kv switch can be changed from executed in the plant overhaul to executed in the routine maintenance. This could make a great contribution in optimizing the NPP sea water pump maintenance program.
Pipe Performance in Long Term Operation Framework: Ageing Issues
Authors: Cancemi S. A. , Lo Frano R. 
Published: October 2021

Abstract
The Long-Term Operation (LTO) process involves a full screening of structures, systems, and components (SSCs), for ageing assessment in order to verify their residual safety margin is still acceptable. At today, 46% of the operating Nuclear Power Plant (NPP) has lifetime between 31 and 40 years, while 19% is in operation since more than 40 years. LTO currently represents the highest priority for all the nuclear industry. To propose and plan suitable management strategies, first step is to intensify the efforts for studying phenomena that influence the performance of SSCs and, in turn, may threat the plant safe operation.
This study deals with the investigation of the performance of a primary piping (Class 1 component of a 2nd Generation PWR) subjected to the effects of alteration of material properties as caused by ageing. In this study it is proposed a numerical investigation of a piping characterized by a complex geometry. Numerical analyses were performed by means of MSC©MARC FE code. The (quantitative) influence of ageing and corrosion processes onto bent pipe are so studied. Particularly, corrosion effects generated from an operation of beyond 35 years are considered. The methodology and results may have an influence on future issues about LTO of NPPs.
Development of Ex-Vessel Debris Bed in a Flooded Cavity With Inclined Bottom Structure Under Two-Phase Condition
Authors: Modak Mayank , Park Hyun Sun , Choi Yu Jung , Seo Mi Ro 
Published: October 2021

Abstract
Severe accidents (SA) in Light Water Reactors (LWRs) consist of a sequence of thermal-hydraulic events which leads to situations that progress beyond design-based accidents (DBAs), and involve significant core degradation which threatens the integrity of the containment boundary. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. Here, experimental study was performed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed on an inclined bottom surface. Current study is an extension to the previous work of Kim et al. (2016) in which the previous model for development of an ex-vessel corium debris bed will be modified by incorporating the effect of bottom surface inclinations. This paper reports the experimental part of the study describing the spreading of the debris bed in terms of two-phase flow, debris injection parameters, and bottom inclined surface geometrical parameters is discussed.
Hydrogen Embrittlement in Inconel 718 Produced by Selective Laser Melting
Authors: Macoretta Giuseppe , Beghini Marco , Monelli Bernardo Disma , Valentini Renzo , Aiello Francesco , Moda Mattia 
Published: October 2021

Abstract
Inconel 718 is widely used to produce components subjected to relatively high temperatures and heavy loads. However, this alloy is also employed in aggressive environments promoting the production of hydrogen on the metal surface. Selective Laser Melting (SLM) is an emerging technology for the production of structural components, thanks to its ability to create complex geometries and reduce material consumption. Components produced by SLM are typically characterized by a peculiar microstructure and residual stresses that can affect hydrogen migration and accumulation. While the mechanical properties and fatigue endurance were deeply investigated in recent years, the resistance of SLMed Inconel 718 to the Hydrogen Embrittlement (HE) requires further investigation.
The present paper deals with the effects of the SLM process on reducing the strength and ductility of Inconel 718. Standard tensile tests and slow strain rate tests were carried out for different hydrogen concentrations. Hydrogen content was measured in each specimen after the mechanical test. Fractographic analyses, along with hydrogen diffusion models, were carried out to preliminarily investigate the hydrogen effect on the material strength.
30 CM Horizontal Drop of a Surrogate 17x17 PWR Fuel Assembly
Authors: Kalinina Elena , Ammerman Doug , Grey Carissa , Flores Gregg , Lujan Lucas , Saltzstein Sylvia , Michel Danielle 
Published: October 2021

Abstract
The 30 cm drop is the remaining NRC normal conditions of transport (NCT) regulatory requirement (10 CFR 71.71) for which there are no data on the response of spent fuel. While obtaining data on the spent fuel is not a direct requirement, it allows for quantifying the risk of fuel breakage resulting from a cask drop from a height of 30 cm or less. Because a full-scale cask and impact limiters are very expensive, 3 consecutive drop tests were conducted to obtain strains on a full-scale surrogate 17 × 17 PWR assembly. The first step was a 30 cm drop of a 1/3 scale cask loaded with dummy assemblies. The second step was a 30 cm drop test of a full-scale dummy assembly. The third step was a 30 cm drop of a full-scale surrogate assembly. The results of this final test are presented in this paper. The test was conducted in May 2020. The acceleration pulses on the surrogate assembly were in good agreement with the expected pulses derived from steps 1 and 2. This confirmed that during the 30 cm drop the surrogate assembly experienced the same conditions as it would have if it had been dropped in a full-scale cask with impact limiters. The surrogate assembly was instrumented with 27 strain gauges. Pressure paper was inserted between the rods within the two long and two short spacer grid spans in order to register the pressure in case of rod-to-rod contact. The maximum observed peak strain on the surrogate assembly was 1,724 microstrain at the bottom end of the assembly. The pressure paper sheets from the two short spans were blank. The pressure paper sheets from the two long spans, except a few middle ones, showed marks indicating rod-to-rod contact. The maximum estimated contact pressure was 4,100 psi. The longitudinal bending stress corresponding to the maximum observed strain value (calculated from the stress-strain curve for low burnup cladding) was 22,230 psi. Both values are significantly below the yield strength of the cladding. The major conclusion is that the fuel rods will maintain their integrity following a 30 cm drop inside of a transportation cask.
Development of Mini-Compact Tension Specimen Fabrication and Test Methods in Hot Cell for Post-Irradiation Examination of Reactor Pressure Vessel Steels
Authors: Naziris Frideriki , Versteylen Casper , Frith Frederick , Bregman Marcel , Kolluri Murthy 
Published: October 2021

Abstract
Most standard surveillance programs for pressurized water reactors (PWR), do not contain enough surveillance specimens for their long-term operation (LTO) beyond the original design life of 40 years. The use of testing methods based on small/sub-sized specimen such as mini-CT (0.16 CT, 4 mm thickness) is considered a promising approach to overcome this limitation. The feasibility of fracture toughness testing using sub-sized CT specimen has been generally demonstrated in Japanese-US lead Round Robin programs for un-irradiated materials. However, application of miniature testing techniques on irradiated materials in hot cells comes along with several technical challenges such as remote controlled machining, application of the extensometers, manipulator sensitivity etc. In addition, the dimensional tolerances of the standard CT specimen defined in the ASTM E1921 standard are defined as a function of specimen width, resulting in severer absolute tolerance for the miniature specimen. This paper presents an in-cell fabrication method of a modified mini-CT specimen and a testing facility suitable for hot cell use. This work is performed at NRG within the framework of the STRUMAT project, funded by the Netherlands Ministry of Economic Affairs. A slightly modified miniature CT specimen is chosen, which allows “outboard” clip gage mounting for direct load-line displacement (LLD) measurement. In addition, a wide notch configuration with a height of 0.6 mm, 0.1 mm wider than specified in the ASTM E1921 standard, is used. This modified specimen geometry facilitates the in-cell fabrication of the mini-CT specimen and eases the remote handling and testing procedure. Three-dimensional finite element analysis validation was carried out to determine the effect of these modifications on the stress state of the specimen. It is demonstrated that the modifications to the mini-CT specimen geometry will not lead to significant deviations in the stress patterns and J-integral values compared to the standard geometry.
Overview of Revisions to the ASME Boiler and Pressure Vessel Code Section VIII Division 3 for the 2021 Edition and Near Future
Authors: Maslowski Adam P. , Sarzynski Melanie , Peters Daniel , Subramanian Kannan 
Published: October 2021

Abstract
This paper provides an overview of the significant revisions to the upcoming 2021 edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section VIII Division 3, Alternative Rules for Construction of High Pressure Vessels, as well as potential changes to future editions under consideration by the Subgroup on High Pressure Vessels (SG-HPV).