bluuebunny/crossref_metadata_2025_split huggingface dataset:
Sensitivity Analysis for Dynamical Response of Reactor Coolant System Based on OPTIMUS
Authors: Yanli Yuan , Xianhui Ye , Lijuan Li , Feng Yuan 
Published: October 2020

Abstract
The sensitivity analysis of the dynamical response of reactor coolant system to the input parameters is an important precondition for the design optimization. In this paper, the sensitivity of the dynamical loads at the nozzles of the equipment under seismic conditions is analyzed with an integrated platform called OPTIMUS, taking the stiffness of the dampers in the steam generator and the main pump as the input variables. The key parameters of the reactor system are usually different from the design value due to the calculation error, random and other uncontrollable errors in the manufacturing process and installation process. In a nuclear power project, the measured stiffness values of the dampers on the steam generator and the main pump in the manufacturer are deviated from the requirements in the equipment specification, and it is necessary to evaluate the influence of the deviation on the dynamical response analysis of the reactor system.
According to the traditional method, it is necessary to establish the models of the reactor coolant system for nonlinear analysis according to the different stiffness of the dampers, and then the calculation results are compared by EXCEL. In this paper, the sensitivity analysis of output parameters which are the loads at the nozzles of the equipment to the input parameters which are the stiffness of the dampers on the steam generator and pump is realized by OPTIMUS, which is a kind of integration platform. Not only can ANSYS simulation calculations be carried out automatically on the OPTIMUS, but also the output data can be processed rapidly automatically, and the influence of manufacturing deviation of the stiffness of the dampers on the dynamical response of the reactor coolant system can be analyzed quantitatively in the above-mentioned problems, and the data support is provided for the determination of the design variables for subsequent optimization analysis.
Study on the Influence Factors of Heat Transfer Coefficient of Supercritical Water Under Different Circulation Modes
Authors: Xu Peng , Zhou Tao , Zhang Jialei , Chen Juan , Fu Zhongguan 
Published: October 2020

Abstract
There are many factors that can affect the heat transfer coefficient (HTC) of supercritical water in forced and natural circulation. The correlation between the factors with the HTC under different circulation modes has an important influence on the reactor core design. By extracting the experimental data of supercritical water in forced circulation and natural circulation, the grey correlation model was used to analyze the relational degree between these factors with HTC. The results show that: Under the condition of forced circulation, there is a positive correlation between the inlet temperature, mass flow velocity, the thickness of the grid body with the HTC of supercritical water, and the order is: mass flow velocity &gt; inlet temperature &gt; the thickness of the grid body; there is a negative correlation between the pressure, heat flux with the heat transfer coefficient of supercritical water, and the order is: pressure &gt; heat flux. Under the condition of natural circulation, there is a positively correlation between heating power, inlet temperature and circulation flow rate with HTC, and the order of magnitude is: circulation flow rate &gt; heating power &gt; inlet temperature; diameter and pressure are negatively correlated with heat transfer coefficient, and the order of magnitude is: pressure &gt; diameter. In the two circulation modes, mass flow rate is an important factor affecting the heat transfer capacity of supercritical water, while the effect of heat flux on the heat transfer coefficient is contrary.
Image Calibration Based on Dynamic No-Load Data for the 60Co Gantry-Movable Dual-Projection Radiography Inspection System
Authors: Guangchao Li , Peng Cong , Xiaojing Guo , Liu Chen , Minzi Ni , Yuewen Sun , Zhongwei Zhao 
Published: October 2020

Abstract
The 60Co gantry-movable dual-projection radiography inspection system is newly developed, aiming to the inspection of vehicles entering and exiting the nuclear facilities. It adopts two Co-60 radioactive sources and two arrays of gas ionization chambers corresponding to the two sources, respectively. They can move synchronously with the gantry driven by the mechanical and control subsystem. So, dual projections could be obtained through one scan from two different directions. Compared to a single projection, the dual projections make it easier to found hidden objects and distinguish whether a dark area is due to overlapping objects or because there are well-shielded prohibited items such as nuclear materials. Therefore, it is helpful to found well-shielded nuclear materials and prevent them from being stolen.
However, problems also come due to the using of two radioactive sources and the moving gantry. For the former one, it will bring about scattering effect between two sources, while for the latter one, the signals of the detectors would fluctuate as the gantry moves, owing to the vibration of collimator and gantry as well as the non-synchronous movement of sources and detectors. So, the radiography projections are needed to be corrected. In response to the second question, the no-load data is repeatedly measured when the gantry is in different positions, then a method of image calibration based on the dynamic no-load data is proposed for this inspection system to replace the correction with average no-load data. Result shows that the corrected no-load image turns smoother, meaning that the method of dynamic correction could effectively improve the radiation image.
Neutronics and Thermal Hydraulics Coupled 3-D Calculation of Pressure-Tube Deformation
Authors: Ryu Eunhyun , Joo Hangyu , Yoo Seungyul , Jung Jongyub 
Published: October 2020

Abstract
Among the various parts in a pressurized heavy-water reactor (PHWR), pressure tubes are of tremendous importance. This is because they withstand extreme both pressure and temperature differences that exist between the Primary Heat Transport System (PHTS) and the moderator. The pressure tubes also contribute to prevention of fission product release from the PHTS to the PHWR plant (together with end fittings and nearby parts including plugs).
When a PHWR is given a 1% derating, half is due to the aging of the pressure tubes. The main concern with pressure tubes is decrease of the safety margin. Most of the reduction comes from the effects caused by radial expansion and axial sagging, which are belong to four major phenomena including the thinning and the elongation. More specifically, the fuel-pin temperature distribution changes for the worse if deformation of the pressure tube occurs. Because there is extreme irradiation inside the core, the tube content is exposed to high temperature and high pressure. Thus, the shape of the pressure tube is deformed as times goes on.
In this paper, using modeling of a deformed pressure tube in three-dimensional space, the effects on the fuel, coolant temperature, and coolant density, were studied quantitatively. This included a neutronics effect explored using coupled neutronics and thermal hydraulics (T/H) calculations. Among the results, only marginal changes of the neutronics effects were observed. The T/H results, which included temperature and density of the fuel and the coolant, were not critical. Through this study, we are now able to determine in new ways, conventional derating values from a pressure tube.
Applicability of Sub-Modelling Technique for Dynamic Analysis of Concrete Structures With Attached Equipment Under Missile Impact
Authors: Sagals Genadijs , Orbovic Nebojsa , Nitheanandan Thambiayah 
Published: October 2020

Abstract
This paper describes the work conducted by the Canadian Nuclear Safety Commission (CNSC) related to the numerical simulations of reinforced concrete (RC) structures under deformable missile impact. The current paper is a continuation of the work conducted in the frame of the OECD/NEA* IRIS (Improving Robustness Assessment Methodologies for Structures Impacted by Missiles) Phase 3 benchmark project.
The concrete mock-up with two simple structures attached, one welded and another bolted, was built and tested at the VTT Technical Research Centre in Espoo, Finland. This mock-up was impacted by three subsequent missiles with varying velocities in order to obtain the damage accumulation. To examine vibration transmission through the mock-up, the simple structures modelling equipment were attached to the rear wall of the structure, while the missile impact was at the centre of the front wall. The parameters of the missiles and the RC structure were selected to ensure a flexible behaviour of the RC target in the impact area with only moderate damages, specifically cracking and permanent deformation without perforation.
The non-linear dynamic behaviour of the reinforced concrete slabs under missile impact was analyzed using the commercial FE code LS-DYNA. A hybrid FE model using both 3-D solid and 2-D shell FE models was developed for the target discretization. Since the ultimate objective of this work is to model the entire structure over long time periods, a simplified combined shell-solid model with distributed (smeared) reinforcement was selected and validated. This model employs solid FE around an impact area and shell FE for the rest of the mock-up.
Detailed modelling of a large RC structure with all equipment attached leads to a very large finite element (FE) model. Therefore, two-level FE modelling using sub-modelling approach was employed: first, analyze the vibrations of a reinforced concrete structure with simplified equipment modelling, and second, analyze in detail the equipment connected to it. This approach assumes uncoupled dynamic behaviour of the structure and the equipment. While the sub-modelling technique is commonly used in static analysis, a special sensitivity analysis was conducted to prove the applicability of sub-modelling for impact analysis.
Finally, the effect of structural damping was examined and the best possible damping was selected. The selected damping values and sub-models resulted in relatively good agreement with the test results for both global (RC mock-up) and local (equipment) behaviour.
Dynamic PRA Based on System Codes Coupling for Passive Safety System in Integral Pressurized Water Reactor
Authors: Mi Yi , Tokuhiro Akira 
Published: October 2020

Abstract
An integral Pressurized Water Reactor (iPWR) type SMR was studied featuring Passive Safety Systems (PSSs). Different from active systems, PSSs are easily influenced by system parameters referred to as phenomenological factors such as heat loss, flow friction, oxidation, non-condensable gases and void fraction due to the low driving force of natural circulation. The system parameters also contribute to the uncertainty and dependency of PSS leading to the system unreliability. Thus, efforts are made to improve the reliability of PSS. A classical Probabilistic Risk Assessment (PRA) model describing active systems does not consider time evolution nor event ordering for PSS that dynamic PRA can accommodate. Here we developed and realized coupling between LabVIEW and CAFTA. Isolation Condenser System (ICS) is taken as the benchmark system due to the simple design in single phase without phase change phenomena in order to mainly remove decay heat and secondarily depressurize the reactor pressure vessel (RPV). A classical PRA model of ICS using CAFTA is coupled with real-time simulation of primary loop and ICS in LabVIEW, leading to a dynamic simulation result. The difference in failure probability using dynamic versus classical PRA revealed that for one there are more component demands with different event ordering, such that improved PSS reliability in the iPWR-type SMR designs is possible.
Turbulent Flow Characteristics in a Blocked Exterior Subchannel of a Helically Wrapped Rod Bundle
Authors: Nguyen Thien , Vaghetto Rodolfo , Hassan Yassin 
Published: October 2020

Abstract
Liquid metal fast reactor using sodium as a coolant typically utilize a tightly packed triangular lattice of fuel pins enclosed in a hexagonal duct. During the reactor operation, partial or total flow blockage of coolant channels, may occur at different spatial locations within the fuel assembly, due to potential isolated or combined causes, including collection and accumulation of debris, and cladding deformation. Previous studies have shown that the flow characteristics within the wire-wrapped fuel assembly, i.e., including the interior and exterior subchannels, are very complicated and strongly influenced the flow and heat transfer phenomena between the coolant fluid and fuel rods. It is important to understand and characterize the effects of channel blockage to the flow mixing characteristics in the exterior sub-channels (or bypass channels) of the wire-wrapped fuel bundle.
Texas A&M University has conducted isothermal flow experiments in a wire-wrapped 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the US Department of Energy (DOE). The experimental facility employs matched-index-of-refraction (MIR) techniques and laser diagnostic velocity measurement techniques. In this article, we present the time-resolved particle image velocime-try (TR-PIV) measurements to characterize turbulent flow characteristics in the exterior subchannels of the wire-wrapped fuel bundle, under the presence of a localized total blockage of one of the exterior subchannels. From the obtained TR-PIV velocity fields, turbulent flow characteristics including mean velocity, root-mean-square fluctuating velocity, and Reynold stress, are computed and presented. In addition, spectral analysis to the turbulent velocity fields is performed to investigate the characteristic flow frequencies associated to different flow conditions with the presence of the blockage. Finally, proper orthogonal decomposition (POD) analysis is performed to the velocity snapshots to reveal the dominant flow structures that play important roles in the flow dynamics and heat transfers of the fuel bundle.
Improvement of the Evaluation Accuracy for the Diameter Prediction of the CANDU Pressure Tubes
Authors: Jung Jong Yeob 
Published: October 2020

Abstract
Measurements of the inside diameter of the pressure tubes in CANDU reactors have shown that the diameter has been increasing over time, and this phenomena has been explained as a creep phenomenon which is a kind of aging process of the pressure tube owing to the operating conditions of irradiation by neutron flux, high pressure, and high temperature over the plant life. The diameter expansion of the pressure tube has been regarded as a principle aging mechanism governing the heat transfer and hydraulic degradation within the primary heat transport system of the CANDU reactor. Diametrical expansion results in a reduction of the fuel cooling owing to the increased bypass flow, which increases the possibility of a fuel dry-out and thus limits the operating power of the reactor.
In order to explain the mechanism of the creep phenomena of the pressure tube, traditionally the creep deformation has been modeled as a combination of thermal creep, irradiation creep and irradiation growth. However, this modeling approach is too complex to determine all parameters and constants which are relevant to the equation.
In this research, we proposed a very simple approach for modeling the pressure tube diameter deformation in which the pressure tube diameter was modeled based on the measured data, flux distribution of each fuel channel and temperature variation inside the pressure tube.
New rules were derived to determine the effect of flux and temperature distribution on the diameter expansion based on the measured data of pressure tube diameter. Results from applying the methodology show a dramatic improvement of the prediction accuracy of pressure tube diameter compared to the previous modeling results.
Level 2 PSA Overview of HPR1000 Nuclear Power Plant
Authors: Ziguan Wang , Fang Lu , Benlin Yang , Shi Chen , Lingsheng Hu 
Published: October 2020

Abstract
Risk-informed design approaches are comprehensively implemented in the design and verification process of HPR1000 nuclear power plant. Particularly, Level 2 PSA is applied in the optimization of severe accident prevention and mitigation measures to avoid the extravagant redundancy of system configurations. HPR1000 preliminary level 2 PSA practices consider internal events of the reactor in the context of at-power condition. Severe accidents mitigation and prevention system and its impact on the overall large release frequency (LRF) level are evaluated. The results showed that severe accident prevention and mitigation systems, such as fast depressurization system, the cavity injection system and the passive containment heat removal system perform well in reducing LRF and overall risk level of HPR1000 NPP. Bypass events, reactor rapture events, and the containment bottom melt-through induced by MCCI are among the dominant factors of the LRF. The level 2 PSA analysis results indicate that HPR1000 design is reliable with no major weaknesses.
Direct Numerical Simulation of Fluid Flow in a 5x5 Square Rod Bundle Using Nek5000
Authors: Kraus Adam , Merzari Elia , Norddine Thomas , Marin Oana , Benhamadouche Sofiane 
Published: October 2020

Abstract
Rod bundle flows are commonplace in nuclear engineering, and are present in light water reactors (LWRs) as well as other more advanced concepts. Inhomogeneities in the bundle cross section can lead to complex flow phenomena, including varying local conditions of turbulence. Despite the decades of numerical and experimental investigations regarding this topic, and the importance of elucidating the physics of the flow field, to date there are few publicly available direct numerical simulations (DNS) of the flow in multiple-pin rod bundles. Thus a multiple-pin DNS study can provide significant value toward reaching a deeper understanding of the flow physics, as well as a reference simulation for development of various reduced-resolution analysis techniques. To this end, DNS of the flow in a square 5 × 5 rod bundle at Reynolds number of 19,000 has been performed using the highly-parallel spectral element code Nek5000. The geometrical dimensions were representative of typical LWR fuel designs. The DNS was designed using microscales estimated from an advanced Reynolds-Averaged Navier-Stokes (RANS) model. Characteristics of the velocity field, Reynolds stresses, and anisotropy are presented in detail for various regions of the bundle. The turbulent kinetic energy budget is also presented and discussed.
Experimental Setup for Non-Contact Displacement Measurement of Nuclear Waste via Electromagnetic Waves
Authors: Cappelli Mauro , Surrenti Vincenzo 
Published: October 2020

Abstract
Spent nuclear fuel and high-level waste from fuel reprocessing are the most hazardous and long-lived radioactive wastes produced by the nuclear fuel cycle, and they must be contained and isolated for thousands of years at an acceptable safety level for human health and the environment. Recently, the authors have proposed a preliminary numerical investigation of possible non-contact techniques for displacement measurement in a geological nuclear waste disposal. In fact, cables used for the detection of measurement signals might be effectively substituted with wireless sensors, thus avoiding all the side effects produced in such a harsh environment (e.g. degradation due to temperature, humidity and radiation conditions). Previous works showed the feasibility of using an electromagnetic approach for the contactless displacement measurement of a canister immersed and suspended in bentonite, a material with a high degree of water retention. The results demonstrated that such an approach is possible provided that some strong requirements are imposed on the sensor design. Sure enough, new sensors must assure reliable transmission through the isolated areas of a repository without affecting the engineered barriers performance and reliable use of energy supply for the measuring equipment in the long term. This work proposes a possible experimental setup where the numerical models developed in previous works can be tested, so to arrive to a prototype sensor to be used in a demonstrator repository.
Large-Scale Rig for the Characterization of DCC at Sub-Atmospheric Pressure
Authors: Lo Frano R. , Pesetti A. , Aquaro D. , Olcese M. 
Published: October 2020

Abstract
The Direct Contact Condensation (DCC) is the main phenomenon characterizing the steam condensation. It plays an important role for the operation of Vacuum Vessel Pressure Suppression System (VVPSS) tanks, particularly for managing the Ingress of Coolant Event (determining fusion reactor overpressurization).
It is safety relevant (key) system of the fusion reactor because by condensing the steam generated during such accident event allows to damp the overpressure.
This paper deals with experimental and theoretical analyses of the DCC at sub-atmospheric pressure. The similitude analysis was elaborated to scale up the experimental results obtained in the reduced scale facility: similitude laws are used for the design of large experimental rig.
Correlations are defined starting from the water temperature and pressure variation already obtained in the small-scale rig. Furthermore, the experimental rig and its main components accordingly designed (and under construction at the University of Pisa) allow to study at large scale the steam condensation. The testing conditions are presented and discussed.
Study of the Ageing Effects on the Lower Head Failure in a PWR Reactor
Authors: Lo Frano R. , Paci S. , Darnowski P. , Mazgaj P. 
Published: October 2020

Abstract
The paper studies influence the ageing effects on the failure of a Reactor Pressure Vessel (RPV) during a severe accident with a core meltdown in a Nuclear Power Plant (NPP).
The studied plant is a generic high-power Generation III Pressurized Water Reactor (PWR) developed in the frame of the EU NARSIS project. A Total Station Blackout (SBO) accident was simulated with MELCOR 2.2 severe accident integral computer code.
Results of the analysis, temperatures in the lower head and pressures in the lower plenum were used as initial and boundary conditions for the Finite Element Method (FEM) simulations. Two FEM models were developed, a simple two-dimensional axis-symmetric model of the lower head to study fundamental phenomena and complex 3D model to include interactions with the RPV and reactor internals.
Ageing effects of a lower head were incorporated into the FEM models to investigate its influence onto lower head response. The ageing phenomena are modelled in terms of degraded mechanical material properties as σ(T), E(T).
The primary outcome of the study is the quantitative estimation of the influence of ageing process onto the timing of reactor vessel failure. Presented novel methodology and results can have an impact on future consideration about Long-Term Operation (LTO) of NPPs.