bluuebunny/crossref_metadata_2025 huggingface dataset:
Validation of RELAP5/MOD3.4 for Flashing-Induced Instabilities in a Natural Circulation Loop
Authors: Xu Yifan , Peng Minjun , Xia Genglei , Zhao Yanan
Published: October 2020

This paper aims to validate the effectiveness of the widely used Relap5 code in simulating two-phase natural circulation, and its capability to predict flashing-induced instabilities. The RELAP5 code is validated against experimental data from the NMR test facility, which was designed to investigate the flow instability for a BWR-type novel modular reactor (NMR). The simulations by RELAP5/MOD3.4 code had been performed under various conditions by changing system pressure, core inlet subcooling, core inlet flow resistance, and core heat power etc. The flow stability for a certain operating condition could be determined from the time trace profile of the loop natural circulation flow rate. The results showed that the simulated mass flow rate increased with increasing core inlet temperature, reproducing the experimental trend. And the maximum error between the experimental data and the calculated results is within 10%. The predicted natural circulation dimensionless numbers, the phase change number and inlet subcooling number, also had a good agreement with the experimental data. In general, the RELAP5 code is able to simulate flashing-induced instability and density wave oscillations, which occurred in the natural circulation test facility at low pressures. However, flashing tends to be suppressed at a higher pressure (400kPa). And the enlargement of core inlet resistance coefficient can also have a positive impact on natural circulation system stability.
Experimental Study on the Sub-Channel Void Fraction Characteristics of Bubbly Flow in Rod Bundles
Authors: Ren Quan-yao , Pu Zeng-ping , Zheng Mei-yin , Su Min , Chen Ping , Pan Liang-ming , He Hui , He Qing-che
Published: October 2020

The gas-liquid two-phase flow behaviors are always associated with its dynamic void fraction, such as flow resistance, heat transfer coefficient, phase distribution, critical heat flux etc. As regard to the commercial PWR and BWR, rod bundles are the typical geometry, which contains many sub-channels for coolant flowing. In present study, the sub-channel void fraction was measured in 5 × 5 rod bundles with the sub-channel impedance void meter consisting of 12 strip electrodes. Based on the measured void fraction in different sub-channels, the void fraction dynamics, PDF (probability distribution function) and CDF (cumulative distribution function) curves were analyzed to make clear the effect of superficial gas and liquid velocity, flow development and casing tube. The empirical correlation for PDF of dynamic sub-channel void fraction has been developed, which showed good fitness with PDF and CDF curves and satisfying accuracy of averaged void fraction.
Artificial Neural Networks in Condition Monitoring and Fault Diagnosis of Nuclear Power Plants: A Concise Review
Authors: Jiang B. T. , Zhou J. , Huang X. B.
Published: October 2020

Artificial neural networks (ANNs) are recognized for their good properties in solving the non-linear classification problem. Especially, ANNs and their latest advancements in deep learning (DL) are blooming in artificial intelligence (AI) fields in the past few years. They have recently proven their abilities to handle some complex fault diagnosis problems. In the context of these backgrounds, this paper provides a concise review on the applications of ANNs to condition monitoring and fault diagnosis (CMFD) of nuclear power plants (NPPs). Firstly, a brief description of basic principle of ANNs are given. Then, a number of studies reported in both the journals and conferences are reviewed. These studies are divided into two categories according the application types of ANNs: shallow ANNs and deep ANNs. Finally, the conclusions and trends developed in the future are summarized.
Optimal Design for Helium Cooled Solid Breeder Blanket of CFETR
Authors: Nogueron Jose A. , Baorui Zhang , Zhiwei Zhou
Published: October 2020

On the road to achieving fusion energy production at a commercial level, China has proposed the design of a nuclear fusion reactor based on the tokamak configuration. Aimed to produce self-sustained burning plasma and a closed tritium breeding cycle, this device is expected to provide a bridge between ITER and DEMO designs.

The Breeding Blanket (BB) is one of the key technological challenges to be designed in order to guarantee sufficient tritium production, heat removal capabilities and radiation shielding protection. Two preliminary designs of the Helium-Cooled Solid Breeder (HCSB) blanket have been suggested to be subjected to further investigation. Despite the fact that both designs use the same materials, they present completely different geometrical arrangements. In the present analysis, these two concepts of the HCSB are examined, addressing the critical design issues that affect the performance of the system. A neutronic analysis is performed to calculate the Tritium Breeding Ratio (TBR), which is a crucial parameter to meet tritium production requirements. Based on this investigation, the cooling capacity of the reactor is analyzed with the help of a Computational Fluid Dynamic (CFD) software through a dynamic evaluation. Valuable conclusions can be extracted from the results of this work, which can be referenced for investigations regarding further studies of HCSB blankets.
Study on Rotor’s Vibration Characteristics of Helium Circulator of HTR
Authors: Li Jiahao
Published: October 2020

The helium circulator is used in the 10MW High Temperature Gas-cooled Reactor (HTR-10). The circulator operates at helium condition in order to transport the core heat to the steam generator. Rotor is a key component of the helium circulator. The rotor vibrates during the operation of the helium circulator. The study of the vibration characteristics of the rotor can provide important design parameters for the helium circulator. The effect of the shell on the rotor in the helium circulator is ignored in the usual research and analysis. In this paper, the working process of the rotor considering the shell effect was studied by theoretical analysis, and the vibration characteristics of the rotor considering the shell effect with various types and parameters was studied by finite element method (FEM). More accurate ranges of the rotor parameters which ensure the helium circulator to operate normally were determined. The shell increases the degree of freedom (DOF) of the rotor, which makes the analysis of the motion process of the rotor more complicated.

The results show that the different parameters of the shell on the rotor play an important role in the analysis results of the vibration characteristics. The analysis results considering the shell effect of the rotor are more accurate than the usual analysis results.
Neutronic Study on Conceptual Lithium Fluoride Salt Cooled Fusion Driven System for Actinide Transmutation
Authors: Putra Feryantama
Published: October 2020

The subcritical fission system could be used to transmute actinide elements from spent fuel into thermal power. Since the subcritical system needs an external neutron source, the D-T Fusion reactor was introduced as a fast neutron source to drive the chain reaction. Fusion driven system based on Spherical tokamak with a 1.7 m major radius has been designed. The breeding blanket has been modified to accommodate space for a 4.8 m length oxide fuel rod. Oxide fuel rods were immersed inside the lithium fluoride salt blanket, which acts as a coolant and liquid tritium breeder. The fully liquid blanket design was taken to reduce structure complexity and maximize tritium breeding volume. The neutronic study has been performed to calculate the neutron multiplication factor, actinide burnup in 400 days cycle and theoretical Tritium Breeding Ratio using MCNPX code. The design goal of fusion driven system with a fully liquid blanket is to reach keff value from 0.8 to 0.9 by burning sufficient actinide while breeding Tritium in Fluoride salt coolant. Various lithium fluoride coolant and reflector options have been simulated with the designed geometry. FLiNaBe salt liquid blanket was used as a reference since it could reach a keff value of 0.942 and able to transmute the actinide element in oxide fuel.
Experimental Study of Bubble Behaviour in a Flowing Liquid Layer
Authors: Zhang Zhengzheng , Li Liangxing , Zhang Shuanglei , Saleem Afnan
Published: October 2020

A visualized experimental system is designed and constructed to investigate the bubble dynamic in a flowing liquid layer. Motivated by reducing uncertainties and digging a deep understand on the formation mechanism of boiling bubbles, the bubbles are formed by injecting air through a submerged orifice in our present work, where the influence of thermal physics, nucleation site density and dry spot are stripped. The water flow rate and the air flow rate are in the range of 72–324 ml/min and 0.8–2.0 ml/min, respectively. The bubble formation process in the smooth channel and the rib channel are investigated. The results state that increasing the liquid flow rates lead to the increasing bubble detachment frequency and the decreasing bubble detachment volume. Besides, the larger the liquid flow rate is, the closer the bubble center of mass is to the wall. The rib has a significant influence on the bubble formation process. In the rib channel, it is more difficult for bubbles to detach from the orifice compared that in a smooth channel. Besides, the bubble detachment volume in a rib channel is larger than it in a smooth channel.
Evaluations of TiO2 Deposition on Structure Surfaces and Water Radiolysis for the Corrosive Environment in a Reactor Pressure Vessel
Authors: Mawatari Takashi , Yamamoto Yasushi , Shibasaki Osamu , Hara Takahiro , Horayama Yusuke , Takagi Junichi
Published: October 2020

It is important to avoid a risk of stress corrosion cracking (SCC) of structure materials in nuclear power plants. One of the methods for preventing SCC is titanium-dioxide (TiO2) injection. TiO2 is a photo-catalyst, and injected TiO2 deposits on the surfaces of structure materials, and the deposited TiO2 can decrease the electrochemical corrosion potential (ECP) by irradiation of Cherenkov-light in the reactor. Generally, the ECP is an important indicator about susceptibility to SCC, and ECP-lowering makes a risk of SCC lower. Therefore, initiation and propagation of SCC will be mitigated by the TiO2 injection.

The evaluation model for corrosive environment had been already developed. That “conventional” model was consisted of some analysis such as water radiolysis, ECP and others. However, the model did not consider an effect of TiO2 deposition and was one-dimensional simple model. Therefore, to evaluate ECP distribution in a reactor after TiO2 injection, the “newly developed” model was build: analyses integration, high-resolution and three-dimensional model.

In this paper, the TiO2 deposition and water radiolysis analyses integrated into flow analysis were contained. Both the deposition and the radiolysis depend on flow in the RPV, so that these analyses were built into the computational fluid dynamics (CFD) analysis model for CFD code. The target plant type for the model was 1100 MWe-class BWR-5.

The flow analysis was carried out at steady-state, assuming the TiO2 injection during the rated operation of the plant. Then the TiO2 deposition and the radiolysis analyses were conducted by transient calculations using the result of flow, and the three-dimensional and high resolution results were obtained.
Control Rod Blades Size Reduction Using Underwater Plasma Cutting and its Effects on Boron Carbide Powder Scattering
Authors: Serbouti Yassine , Kurihara Keisuke , Kometani Yutaka , Itagaki Masatoshi , Tatemura Makoto
Published: October 2020

Control rod blades are comprised of a stainless steel sheath, which contains neutron absorber tubes (filled with boron carbide powder). During decommissioning, the first stage of size reduction consists of cutting the connector (bottom portion) of the control rod, while the second stage consists of separating the blades of the control rod by cutting through the tie rod. The last stage consists of segmenting the control rod blades by cutting through absorber tubes.

In this study, the control rod blades segmentation (last stage of size reduction) is investigated using an actual control rod (unused). During the experiments, we used a forming press on the cut locations followed by a plasma arc cutting underwater. The purpose of this cutting technique is to minimize the scattering of boron carbides into water by using the stainless sheath melt to seal the absorber tubes.

After the segmentation, we confirmed the sealing of the absorber tubes by visually examining the cut cross-sections. The water analysis showed that the boron carbide scattering was relatively low (only 0.07% of the total boron carbides was scattered). Finally, we confirmed that the off-gas emission is considerably reduced by using Argon plasma instead of Argon-Hydrogen plasma.
Measurement of Thermal Decomposition Temperature and Rate of Sodium Hydride
Authors: Kawaguchi Munemichi
Published: October 2020

In decommissioning sodium-cooled fast reactors, the operators can be exposed to radiation during dismantling of cold trap equipment (C/T). The C/T is higher dose equipment because the C/T trapped tritium of fission products during the operation to purify the sodium coolant. In this study, thermal decomposition temperature and rate of sodium hydride (NaH) were measured as a fundamental research for development of “thermolysis” process prior to the dismantling.

We measured the thermal decomposition temperature and rate using NaH powder (95.3%, Sigma-Aldrich) in alumina pan with ThermoGravimetry-Differential Thermal Analysis (TG-DTA) instrument (STA2500 Regulus, NETZSCH Japan). The heating rates of TG-DTA were set to β = 2.0, 5.0, 10.0 and 20.0 K/min. The DTA showed endothermic reaction and the TG showed two-steps mass-loss over 580K. This first-step mass-loss was consistent with change of chemical composition of the NaH with heating (NaH → Na+1/2H2). The thermal decomposition temperature and rate were obtained from the onset temperature of the mass-loss and the simplified Kissinger plots, respectively. Furthermore, we set to the thermal decomposition temperature of around 590K, and the mass-loss rates were measured. As a result, over 590K, the thermal decomposition occurred actively, and showed good agreement with the estimation curves obtained by the simplified Kissinger plots. The thermal decomposition rate strongly depended on the heating temperature.
High Resolution 3D Simulation of Melt Jet Breakup Phenomenon Using Multi-GPU-Based Smoothed Particle Hydrodynamics Code and Comparison With Experimental Result
Authors: Park So-Hyun , Jo Young Beom , Kim Eung Soo
Published: October 2020

Fuel Coolant Interaction (FCI), one of the critical phenomena in severe accident, involves a variety of physical phenomena including the interaction between coolant and fuel of high temperature. Especially, the jet break-up of a pre-mixing phase that the bulk of molten fuel breaks into the droplet is important for the accident progression. Understanding the intricate physics of jet break-up is essential to reduce the uncertainties of FCI and to mitigate severe accident. In this study, we have developed Lagrangian-based CFD code (named as SOPHIA) using Smoothed Particle Hydrodynamics (SPH) method, which has an advantage on handling the complicated interfacial behaviors, large deformation and multiphase flow. Furthermore, the SOPHIA code is parallelized on the multi-GPUs to achieve high-resolution and large-scale simulation that enhance the accuracy and practical applicability. Using the multi-GPU based SOPHIA code, this study simulates the benchmark jet breakup experiments in high resolution and three dimensions. The simulation results are compared with the experimental data both qualitatively and quantitatively. As a results, they shows a good agreement, and furthermore, three dimensional high resolution simulation is confirmed to resolve the physical features of jet breakup accurately by taking account into the multi-fluids interactions between jet-pool-air.
Influence of Gas Properties on Gas-Liquid Two-Phase Flow
Authors: Saito Miki , Kanai Taizo , Nishimura Satoshi , Nishi Yoshihisa
Published: October 2020

Understanding the mechanism of fission product (FP) removal by pool scrubbing is essential for improving the prediction accuracy of FP emissions concerning severe accident (SA) in a nuclear power plant. Since FP migrates from a gas-phase to a liquid-phase via a gas-liquid interface, the FP removal efficiency by pool scrubbing is largely affected by the flow regime of gas-liquid two-phase flow. In order to gain a deeper understanding of the influence of gas properties on flow regimes, experiments were performed by injecting helium (He) and nitrogen (N2) gas mixtures of several volumetric ratios through a pool of stagnant water. The result suggests clear effects of gas compositions on gas-liquid two-phase flow, where both void and holdup fractions were found to increase with N2 fraction in the supplied gas. The results were compared with previous studies, and a detailed analysis of bubble characteristics for different compositions of gases was performed using a wire-mesh sensor (WMS). This paper also illustrates further research aspects needed to discuss the effect of its results on FP removal efficiency in a SA, and to acquire comprehensive physics behind such gas property influences on two-phase flow.
Localization of Radiation Sources Using Randomly Distributed Detectors
Authors: Li Ke , Liang Manchun , Yang Jie , Yang Dandan , Shen Hongmin , He Shuijun , Liu Fang
Published: October 2020

With the radiation sources widely used in the field of industry, agriculture, medicine, and resources all over the world, the occurrences of the lost and theft incidents are not unusual. There will be great threats to human health and environment for uncontrolled sources. It can be even worse when they are used by terrorists as weapons in the city, such as RDD, which would disperse the radioactive substances into the environment and have a long-term negative impact on human society. Besides the security control and surveillance of radiation sources, the source-searching method using radiation detectors is also needed to find the lost or stolen sources. In the case of radiation monitoring in a city, a mass of people with detectors are needed to cover the large area, while radiation dosimeters are appropriate for its portability and economy.

In this paper, a simple approach to finding and locating unknown radiation sources, which might be in static or mobile condition, using distributed mobile radiation dosimeters is put forward. This approach is able to estimate the time-dependent sources location and the intensity with the dose rate measurements over time periods.

Random data is introduced to verify the approach. Virtual radiation sources and detectors with random movement are created in the cases and the dose rate measurements are computed accordingly, with which sources location and intensity are estimated, where natural background radiation and the statistical fluctuation of measurements are taken into account. The results of the estimated sources show a good consistency with the assumptions.
Analysis of Transient Corium Pool Structure in the Lower Plenum of Reactor Vessel
Authors: Chen Liang , Pang Hua , Xie Ximing , Zhong Lei , Cai Rong
Published: October 2020

The transient stratification of the corium in the lower plenum and its impact on the heat flux distribution on the outside of reactor vessel is analyzed in this work. A method for predicting the kinetic corium pool structure is proposed, which takes into account both thermo-chemical equilibrium and density evaluation of the corium. The transient stratification of the corium pool formed after a large loss of coolant accident (LLOCA) and a station blackout (SBO) accident of ACP1000 nuclear power plant in China was analyzed by this method. The transient structure of the corium pool was calculated at the moment when the amount of molten materials in the corium pool increased obviously. The results shown that the formation of a three-layer pool is highly possible when a two-layer pool is formed in the previous moment with a heavy metal layer on the bottom and the density of the heavy metal layer at the bottom is greater than the density of the newly added molten material at the next moment. The heat flux on the outside of the vessel wall faced the thin top metal layer and the vessel failure probability of the vessel here are high if a three-layer pool occurred.
Reliability Assessment of NPP System for Risk Management Based on an Information Reasoning Methodology
Authors: Zhu Guixia , Zhou Tianmi , Qian Yugang , Liu Chongchong , Lu Hongxing
Published: October 2020

Risk management (RM) is an effective means of safety assessment and reliability management of nuclear power plant (NPP). During the operation of NPP, the main contents of RM work are as follows:

1. Real-time reliability assessment of NPP system, and evaluate NPP risk, so as to ensure that the NPP system can complete the specified functions to deal with the consequences of system failure.

2. Complete the diagnosis and mitigation of abnormal faults, so as to ensure that the NPP system can return to a safe state as soon as possible in the case of abnormal operation of NPP.

With the continuous improvement of the requirements of online RM, LPSA (Living PSA) technology, as an effective tool of RM, has been developed and provides more powerful support for the safe operation of NPP. In order to evaluate the reliability of NPP system in real time, LPSA model needs to be updated to match the state of the system. Most of the existing LPSA model updating techniques are based on the premise that the system information is completely knowable and the information is accurate. However, in the process of NPP operation, the accessibility of system information and the accuracy of information are difficult to meet the requirements of updating the existing LPSA model.

How to update the LPSA model in the case of incomplete and inaccurate information is one of the difficulties in the application of online LPSA technology. Especially when the NPP is in abnormal conditions, it is more urgent for operators to obtain correct system information in real time to recognize the available state of the plant system.

In fact, the required system information needs to be obtained by the operator through his or her cognitive behavior process, such as the fault diagnosis process. At the same time, this information can be used to help operators complete the diagnosis of system status and the evaluation of system reliability.

In this study, an information reasoning method is proposed to evaluate the probability of various failure modes under the information known to the operator through the modeling and analysis of the cognitive behavior process of the operator. The method can assist the operator to make the correct fault judgment to complete the fault diagnosis and fault mitigation. In addition, the information obtained by the operator in the process of cognitive behavior, and the results of information reasoning can be used to update the LPSA model in real time, so as to obtain the real-time reliability evaluation results of the NPP system.
Sensitivity Analysis for Dynamical Response of Reactor Coolant System Based on OPTIMUS
Authors: Yanli Yuan , Xianhui Ye , Lijuan Li , Feng Yuan
Published: October 2020

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

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 > inlet temperature > 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 > 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 > heating power > inlet temperature; diameter and pressure are negatively correlated with heat transfer coefficient, and the order of magnitude is: pressure > 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

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

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

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

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

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

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

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

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

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

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

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.
Code to Code Validation of SAC-3D Based on EBR-II Benchmark Problem
Authors: Lu Daogang , Lyu Siyu , Sui Danting
Published: October 2020

A lot of previous works indicated that the 3D neutronics calculation module can provide better prediction results in reactor system safety analysis especially under several anticipated transient accidents which with strong spatial influence factors. SAC-3D code is a new version of SAC series reactor system safety analysis codes which was developed by North China Electric Power University[4][5][6][7]. In this version, the neutronics calculation module was developed based on the three-dimensional nodal expansion method. In 2012, IAEA initialed a CRP on benchmark analysis of EBR-II loss of flow without scram test, SHRT-45R benchmark data is one of the benchmark specifications provided by ANL. In the present work, the neutronics benchmark problem of EBR-II SHRT-45R was analyzed with SAC-3D. The neutronics calculation module of SAC-3D was validated by comparing the key results with the benchmark data. The simulation results also agreed well with the results provided by other participants of this CRP.