bluuebunny/crossref_metadata_2025 huggingface dataset:
Inverse Heat Conduction Problem in Estimating NPP Pipeline Performance
Authors: Cancemi Salvatore A. , Lo Frano Rosa
Published: October 2020

Most of today’s operating nuclear plants that were originally designed for 30 or 40-year life are facing the long-term operation issues. Therefore, it is of meaningful importance to assess the time-dependent degradation and the ageing of the relevant nuclear systems, structures, and components (SSCs) because of resulting loss of structural capacity.

In this framework, the inverse method is implemented starting from temperatures at an accessible boundary, which are measured through a monitoring system. The reconstruction technique uses the elaborated signal provided by the monitoring system to determine temperature at inaccessible surface: this is the so-called inverse heat transfer problem (IHTP).

The inverse space marching method is applied.

Analytical and numerical studies are performed taking into account thermal transient conditions in order to determine thermal loads. In particular, the developed code demonstrates to be able to reconstruct temperature and stress profiles in any section of the pipe with a good accuracy.

In addition, the thermal loads obtained suggest that the investigated transient condition is not able to jeopardise the integrity of NPP, confirming the possibility of the plant extension of life.
Nuclear Material Measurement Based on Fast Neutron Multiplicity Counter
Authors: Su Xianghua , Zhang Quanhu , Hou Suxia , Li Sufen , Yang Jianqing , Hou Linjun , Chen Qifan , Xu Zhichun
Published: October 2020

Fast neutron multiplicity counting (FNMC) analysis method, as a new non-destructive analysis method for nuclear materials, plays an increasingly important role in the measurement of nuclear material properties. Based on the derivation of the FNMC analytical equation of Pu material, the method of solving the sample parameters was given. By analyzing the mechanism of interaction of neutrons and matter, the model used by Geant4 (version 10.4) software was determined, and a set of three-layer, fast neutron multiplicity counters with six liquid scintillation detectors per layer was constructed. Using the fast neutron multiplicity counter to analyze the measured parameters, the detection efficiency variation was less than 0.4% within the 150g sample mass range, and the PuO2 fluctuation was less than the metal Pu. By studying the detection efficiency and the multiplicity counting rate as a function of sample mass, within the 150g sample mass range, both basically meet the model assumptions of the FNMC analytical equation. The metal Pu and PuO2 samples were set separately, and the FNMC analysis equation was solved. When the sample mass was within 150g, the sample mass solution deviation was less than 10%. The results show that the built-in fast neutron multiplicity counter can better measure Pu sample properties.
Simulation and Construction of Fast Neutron Multi-Counter Based on Liquid Flash
Authors: Chen Qifan , Zhang Quanhu , Hou Suxia , Li Sufen , Yang Jianqing , Su Xianghua , Hou Linjun , Xu Zhichun
Published: October 2020

As an important attribute authentication technology, the Fast Neutron Multiplicity Counting analysis method has the potential to solve the shortcomings of existing measurement methods. In this paper, the Geant4 10.4 program based on Monte Carlo method is used to simulate the geometric model, physical process and primary behavior, and simulate the detection system composed of six liquid flash probes in three different layouts to achieve the measurement of sample properties. Pu samples were measured by using different layout detection systems. By comparing the detection efficiency and the multiplicity count rate, it was determined that the No. 2 layout was used as the final detection system, and the detection efficiency was about 3.9%. The variation of the detection efficiency of the simulated sample in the radial direction is obtained. The uniform detection efficiency zone of the system is ±3 cm from the center of the cavity. The detection efficiency deviation is less than 0.12%, the mass deviation is less than 5%, and the deviation variance is less than 10%, and the detection system meets the requirements for sample measurement. Studies have shown that the built detection system has the ability to measure sample properties and can be used in the next step of the study.
Study on the Impact Limiter Design in Spent Fuel Transfer Cask in Nuclear Power Plants
Authors: Hao Yuchen , Li Yue , Wang Jinhua , Wu Bin , Wang Haitao
Published: October 2020

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

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

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

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

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

As an essential method for security inspection in nuclear facilities, digital radiography has the ability to find hidden contraband efficiently. However, the images obtained by current scanning digital radiography system can be degraded by several factors, such as statistical noise and response time of detectors. At high scanning speed, the statistical noise and vibration of the system deteriorates the quality of images. In addition, the reduction of image quality will influence the accuracy of image observation and recognition. To meet the demand of detection efficiency and quality, it is necessary to guarantee the quality of images under high scanning speed.

Thus, to improve image quality of vehicles’ digital radiography at a certain scanning speed, we proposed an approach (VDR-CNN) to reduce or eliminate image noise, which is a convolutional neural network (CNN) with residual learning. The high-quality images obtained at low scanning speed of system served as the ground-truth image for VDR-CNN, while the low-quality counterpart corresponding to the high scanning speed served as the input. Then, the two images mentioned above constitute a training pair. By training this network with a set of training pairs, the mapping function of promoting image quality will be automatically learned so that the restored image can be obtained from the low-quality counterpart through the trained VDR-CNN. Moreover, this method avoids the difficulty in figuring and analyzing the complicated image degradation model.

A series of experiments was carried out through the 60Co inspection system developed by Institute of Nuclear and New Energy Technology, Tsinghua University. The experimental result shows that this method has attained a satisfying result in denoising and preserving details of images and outperforms BM3D algorithm in terms of both image quality improvement and the processing speed. In conclusion, the proposed method improves the image quality of vehicles’ digital radiography and it is proved better than traditional methods.
Numerical Simulation Method Research on Pellet-Cladding Mechanical Interaction Based on ABAQUS Software
Authors: Tang Changbing , Jiao Yongjun , Li Yuanming , Zhou Yi , Zhang Kun
Published: October 2020

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

The work presented in this paper presents new techniques for modeling the combined use of the Reactor Core Isolation Cooling (RCIC) System and Accident Tolerant Fuel (ATF) in a Boiling Water Reactor (BWR). With guidance from Sandia National Laboratories’ Severe Accident Analysis department, a MELCOR BWR model was developed from open source literature. The demonstration shown herein simulates BWR long-term station blackout (LTSBO) conditions with the Nuclear Regulatory Commission’s (NRC) MELCOR severe accident analysis code. By combining state-of-the-art MELCOR modeling practices with new, physics-based RCIC System and ATF MELCOR inputs, this BWR model provides a contemporary platform for BWR severe accident simulations.

The authors are investigating the combined use of the RCIC System and ATF as a means of passively enhancing reactor safety. The benefits of this approach were evaluated by performing simulations using traditional fuel designs (i.e. Zircaloy cladding) and ATF with an iron-chromium-aluminum (FeCrAl) clad under BWR LTSBO conditions. ATF performance was evaluated using severe accident metrics, specifically event sequence timings and the hydrogen production rate from cladding oxidation. Preliminary results show delayed core degradation timelines with less hydrogen production for ATF simulations. Although the results are limited in scope, the presented analysis could easily be expanded to a full-scale uncertainty study that considers a range of severe accident boundary conditions.

This paper describes objective technical results and analysis. Any subjective views or opinions that might be expressed in the paper do not necessarily represent the views of the U.S. Department of Energy or the United States Government.
Thermal Efficiency Optimization of a Modular High Temperature Gas-Cooled Reactor Plant by Extraction Steam Distribution
Authors: Jiang Di , Dong Zhe
Published: October 2020

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

In order to improve the safety of nuclear power plants, it is necessary to make sure measures against their severe accidents. Especially, in the case of a sodium-cooled fast reactor (SFR), there is a possibility of significant energy release due to formation of a large-scale molten fuel pool accompanied by re-criticality in the event of a core disruptive accident (CDA). It is important to ensure in-vessel retention that keeps and confines damaged core material in the reactor vessel even if the CDA occurs.

CDA scenario initiated by Unprotected Loss Of Flow (ULOF), which is a typical cause of core damage, is generally categorized into four phases according to the progression of core-disruptive status, which are the initiating, early-discharge, material-relocation and heat-removal phases for the latest design in Japan. During the material-relocation phase, the molten core material flows down mainly through the control rod guide tube and is discharged into the inlet coolant plenum below the bottom of the core. The discharged molten core material collides with the bottom plate of the inlet plenum. Clarification of the accumulation behavior of molten core material with such a collision on the bottom plate is important to reduce uncertainties in the safety assessment of CDA.

In present study, in order to make clear behavior of core melt materials during the CDAs of SFRs, analysis was conducted using the SIMMER-III code for a melt discharge simulation experiment [1] in which low-melting-point alloy was discharged into a shallow water pool. This report shows the validation results for the melt behavior by comparing with the experimental data.
Using Pipe Whip Analysis via the Finite Element Method to Underpin the Delineation Between High and Moderate Energy Lines
Authors: Hurst Antony M. , Bansal Pavitra
Published: October 2020

When a pressurised pipe in a reactor coolant system breaks, it results in hydraulic loads on the reactor containment system and escaping fluid exerts a thrust force on the pipe. The double-ended guillotine break (DEGB) is generally the most onerous loss of coolant accident (LOCA) in design of a reactor coolant system [1]. In addition to the hydraulic loads, the continuing thrust force on the broken end of the pipe generates a rapidly accelerating rotational displacement of the pipe section on the break side of the plastic hinge, the phenomenon called pipe whip. The whipping pipe has the potential to damage objects within the hazard zone therefore must be assessed. Currently a 20bar threshold is used in nuclear power plant (NPP) design of light water reactors (LWRs) to delineate high energy lines (HEL) and moderate energy lines (MEL). The threshold is used as a process to establish the requirement for additional pipe whip assessments to be performed as part of HEL guillotine break analysis. It is currently argued that no such studies are required for MEL. However the basis of using the 20bar threshold for not carrying out pipe whip assessment of MELs is not well understood.

The work presented here provides details of the finite element (FE) analyses undertaken to substantiate the 20 bar threshold used for the differentiation of HEL and MEL. Using the FE analysis method, a range of pipe characteristics, and pressures in the range of 10–50bar have been studied to determine whether a plastic hinge will occur, and whether pipe whip effects will be seen for that case. The FE results have also been used to assess the equivalent plastic strain criteria generally used to define the formation of a plastic hinge and initiation of pipe whip.
Nuclear Regulatory Acceptance of Certified Process Data Reconciliation (CPDR)
Authors: Jansky Andy , Langenstein Magnus
Published: October 2020

The utility industry is currently undergoing a substantial change from an analog to a digital infrastructure. Not only plant performance and utility profit are dependent on accurate plant operational parameters but, more importantly, set safety limits need to be met in order to ensure safe operation of nuclear power plants in particular. Using non-quality-assured process data for operational decisions can result in significant over- or under-power of the plant. In addition, all new technologies such as AI, IIoT, digital twin technology, etc. rely on robust process data as input, putting at risk the significance of the results from the continuing data processing (“garbage in, garbage out”).

One method, certified process data reconciliation, or CPDR, cuts through the vast amount of available process data and generates all relevant process values with the smallest uncertainty possible. 95% of all collected process data can be discarded after introduction of CPDR. With CPDR, plant operation and maintenance can be significantly optimized and utilities can profit by realizing e.g. power recovery and measurement uncertainty recapture (MUR). Because the focus on reconciled instead of measured values constitutes a paradigm shift, the application of CPDR needs to be communicated to Nuclear Regulators. This paper describes the approach and experience of the Regulator acceptance process in various countries around the globe.
Design of an Innovative Moisture Separator Technology for Use in Nuclear Power Plants: Experimental Validation — Part 2
Authors: Hersberger Carsten , Waldvogel Elias , Bopp Joshua , Ribi Beat
Published: October 2020

The Moisture Separator Reheater (MSR) is a key component of Nuclear Power Plants (NPP), both in terms of performance and prevention of erosion/corrosion. Wet steam is usually dried in a MSR by inertial separation of the liquid water using separator elements. Depending on the design of the MSR, the technology of the separator elements contributes significantly to its size and performance.

An innovative concept of separator panels was conceived by means of aerodynamic principles as outlined in part 1 of this paper [1]. Computational Fluid Dynamics (CFD) has been used to understand the working principles of various moisture separating devices. The investigated separator panels are designed to capture the water droplets in a region of flow separation (invisible pockets) within the separator channels.

To characterize the separation performance of these separator panels, a test rig has been developed and built at the University of Applied Sciences Northwestern Switzerland (FHNW). This test rig was then operated at typical MSR operating conditions. To meet the required moisture content and flow conditions, preheated water was injected into the saturated steam flow. In order to measure the residual moisture content after the separation the throttling calorimeter methodology has been adopted.

The newly designed panels have shown very good separation performance. According to the measurements carried out, a residual moisture content of less than 0.1 % can be guaranteed. The innovative technology, which clearly differentiates the OEM, for who this research was carried out, from its competitors, will allow considerable size and cost reduction as well as opportunities to retrofit existing MSRs.
MELCOR Validation Study on Multi-Room Fire
Authors: Louie David L. Y. , El-Darazi Samir , Fyffe Lyndsey M. , Clark James L.
Published: October 2020

Estimation of radionuclide aerosol release to the environment, from fire accident scenarios, are one of the most dominant accident evaluations at the U.S. Department of Energy’s (DOE’s) nuclear facilities. Of particular interest to safety analysts, is estimating the radionuclide aerosol release, the Source Term (ST), based on aerosol transport from a fire room to a corridor and from the corridor to the environment. However, no existing literature has been found on estimating ST from this multi-room facility configuration. This paper contributes the following to aerosol transport modeling body of work: a validation study on a multiroom fire experiment (this includes a code-to-code comparison between MELCOR and Consolidated Fire and Smoke Transport, a specialized fire code without radionuclide transport capabilities), a sensitivity study to provide insight on the effect of smoke on ST, and a sensitivity study on the effect of aerosol entrainment in the atmosphere (puff and continuous rate) on ST.
Design and Operational Challenges of Switching Working Fluid for a Gen IV Nuclear Powered Closed-Cycle Gas Turbine: Case Studies of Nitrogen and Air, Helium and Argon
Authors: Osigwe Emmanuel O. , Gad-Briggs Arnold , Tukur Nasiru , Pilidis Pericles
Published: October 2020

A unique benefit of using the closed-cycle gas turbine and gas turbomachines employed in the Gen-IV nuclear power plant is the flexibility it offers in terms of working fluid usage. This is so because of the self-containing nature of the closed-cycle gas turbine. To this end, the selection of the working fluid for the cycle operation is driven by several factors such as the cycle performance, system design, and component material compatibility with fluid properties, availability, and many more. This paper provides an understanding of the design and operational challenges of switching working fluids for a nuclear powered closed-cycle gas turbine. Using the plant output power of a simple closed-cycle configuration as a baseline condition, two case studies have been presented in this paper to explore the design and operational challenges of switching working fluids. In the first case study, the fluid was switched from nitrogen to air and in the second case study, helium and argon were utilised. In both cases, using thermodynamic flow relationship, the closed-cycle gas turbine turbomachinery components maps were analysed to understand the operational requirements for switching the working fluids. The paper also provided an insight into the turbomachinery component design considerations for this to be achieved. The overarching results from a thermodynamic perspective showed fluids with similar thermodynamic behaviour could be switched during idle synchronous speed.
Experimental Study on the Thermal Properties and Stability of Hybrid Nanofluids and Evaluation of its Heat Exchange Efficiency
Authors: Xiao Yubai , Zhang Hu , Wu Junmei
Published: October 2020

In recent years, hybrid nanofluids, as a new kind of working fluid, have been widely studied because they possessing better heat transfer performance than single component nanofluids when prepared with proper constituents and proportions. The application of hybrid nanofluids in nuclear power system as a working fluid is an effective way of improving the capability of In-Vessel Retention (IVR) when the reactor is in a severe accident. In order to obtain hybrid nanofluids with excellent heat transfer performance, three kinds of hybrid nanofluids with high thermal conductivity are measured by transient plane source method, and their viscosity and stability are also investigated experimentally. These experimental results are used to evaluate the heat transfer efficiency of hybrid nanofluids. The results show that: (1) The thermal conductivity of hybrid nanofluids increases with increasing temperature and volume concentration. When compared to the base fluid, the thermal conductivity of Al2O3-CuO/H2O, Al2O3-C/H2O and AlN-TiO2/H2O nanofluids at 0.25% volume concentration increased by 36%, 24%, and 22%, respectively. (2) Surfactants can improve the stability of hybrid nanofluids. The Zeta potential value is related to the thermal conductivity of the hybrid nanofluids, and it could be used to explain the relationship between the thermal conductivity of the hybrid nanofluids and the dispersion. It also could provide a reference for subsequent screening of high thermal conductivity nanofluids. (3) The addition of C/H2O can effectively reduce the dynamic viscosity coefficient of hybrid nanofluids. (4) The analysis of heat transfer efficiency of the hybrid nanofluids found that both Al2O3-CuO/H2O and Al2O3-C/H2O have better heat transfer ability than water under certain mixing conditions. This study is conducive to further optimizing hybrid nanofluids and its application to the In-Vessel Retention in severe reactor accidents.
Study on Preparation Process and Properties of High Hydrogen Content Concrete for Neutron Shielding
Authors: Yuanyuan Li , Xiaoqiang Pan , Xiulin Huang , Ying Wu , Jing Yang
Published: October 2020

A series of concrete samples with high hydrogen content were prepared by pressing process with aluminum hydroxide, boron carbide and aluminate cement as raw materials. the mechanical performance of the shielding material was examined by flexural strength tester and compression strength tester. Thermogravimetric analyzer was used to test the weight loss performance of the shielding material, hydrated cement and Al(OH)3. The thermal stability of the concrete at working temperature were investigated by bake the samples at 210°C for 90 hours and at 300°C for 5 hours. The shielding performance against neutron from 241Am-Be source were examined by neutron penetration tests. The test results show that the mechanical properties of the shielding material decrease with the increase of the content of aluminum hydroxide, the compression strength of the shielding material reached above 10MPa. The decomposition temperature of conc is 220°C, the weight loss curve tends to be plateau after baking at 210°C for 20 hours. Most of the crystal water lost after holding at 300°C for 5 hours, but the dimensions of the samples show no obvious change. The shielding performance of the material against neutron from 241Am-Be source is comparable to that of lead-boron polyethylene.
Artificial Neural Network Model for Diagnosing the Performance and the Conditions of Air-Operated Valves
Authors: Kim Taeyun , Chai Jangbom , Lim Chanwoo , Han Ilyoung
Published: October 2020

Air-operated valves (AOVs) are used to control or shut off the flow in the nuclear power plants. In particular, the failure of safety-related AOV could have significant impacts on the safety of the nuclear power plants and therefore, their performances have been tested and evaluated periodically. However, the current method to evaluate the performance needs to be revised to enhance the accuracy and to identify defects of AOV independently of personal skills. This paper introduce the ANN (Artificial Neural Network) model to diagnose the performance and the condition altogether.

Test facilities were designed and configured to measure the signals such as supply pressure, control pressure, actuator pressure, stem displacement and stem thrust. Tests were carried out in various conditions which simulate defects with leak/clogged pipes, the bent stem and so on. First, the physical models of an AOV are developed to describe its behavior and to parameterize the characteristics of each component for evaluating the performance. Secondly, CNN (Convolutional Neural Network) architectures are designed considering the developed physical models to make a lead to the optimal performance of ANN. To train the ANN effectively, the measured signals were divided into several regions, from each of which the features are extracted and the extracted features are combined for classifying the defects. In addition, the model can provide the parameters of maximum available thrust, which is the key factor in periodic verification of AOV with the required accuracy and classify more than 10 different kinds of defects with high accuracy.
Experimental Study of Boiling Characteristics of Seawater After an Accidental Shutdown of the Pump
Authors: Li Yuanjie , Ren Shuai , Pan Chin
Published: October 2020

Water is the most common working fluid using as the coolant for most of nuclear power plants due to its low cost, wide availability, and high heat capacity. In the Fukushima Daiichi accident caused by the tsunami following a powerful earthquake on 11th March 2011, the loss of heat sink occurred. Seawater, which had been considered as an alternative coolant before, was eventually injected into the nuclear reactor for more than one week as an emergency measure. The studies about the characteristics of heat transfer with the seawater as the working fluid have received much more attentions after the Fukushima Daiichi accident. The differences of thermophysical properties such as density, dynamic viscosity, and surface tension, between deionized water and seawater play an important role in the heat transfer. In addition, different boiling behaviors exhibit in the boiling process for two types of working fluid. Compared to pure water, the diameter of bubbles in seawater during pool boiling could be smaller than in water. To improve the safety of a nuclear reactor power plant, mechanisms of the heat transfer using seawater as an alternative emergency coolant should be studied thoroughly. In the present research, experiments of pool boiling in an annulus gap with a riser section after the sudden shutdown of the pump are investigated. Heat transfer characteristic of the heated surface using DI water and 3.5 wt% artificial seawater are described. Boiling behaviors at each fluid are compared and studied.
Research on Cooling Water Source Safety Measures in Water Intake Engineering of Coastal NPP
Authors: Ben-jing Tang , Shu-qiang Hou , Yan-long Wang , Jing Li
Published: October 2020

In view of the situation that large amounts of marine sewage and organisms enter into the cooling water intake system of a coastal NPP at BeiBu-bay, China, which affects the safe and stable operation of NPP, the optimization research on blockage prevention in the water intake engineering is carried out. Based on the statistical analysis of a large number of NPPs’ cooling water source safety incidents globally, the hydraulic characteristics of shell debris in water-intake open channel are studied first, and the source of shell debris in the drum filter at the water intake pump room is analyzed. Secondly, the water intake layout project scheme is optimized through the wave model test, which can effectively improve the wave cover in the water intake channel. At last, through the wave physical model test, the overall force of the network is studied under the combined action of wind, wave, flow and sewage, and the anchoring foundation which can guarantee the stable operation of the trash-intercepting network is optimized. By carrying out the optimization design of water intake engineering, reducing the waves in channel, reducing the possibility of marine sewage and organisms entering the channel, adding the trash-intercepting network facilities and forming effective cover. It can not only effectively avoid or reduce the impact of marine sewage and organisms on the cooling water source safety of NPPs, but also minimize the loss of marine organisms during the operation of NPPs.
Experimental Investigation of the Transient Pool Boiling Heat Transfer on the Quenching of Vertical Rodlet in Water
Authors: Wang Zefeng , Deng Jian , Qian Libo , Cai Rong , Xiong Jinbiao , Zhong Lei , Ma Yugao
Published: October 2020

Quenching is an important phenomenon in the evaluation of an emergency core cooling system following a hypothetical loss of coolant accident (LOCA) in a nuclear reactor. In the present study, an experimental apparatus is designed and constructed with the purpose of conducting high-temperature transient pool boiling quenching experiments for zirconium (Zr-4) cylindrical test samples. Three thermocouples are inserted in the test sample to investigate the effect of axial distance on the minimum film boiling temperature. The Zr-4 rodlet is heated up to a temperature well above the minimum film boiling temperature (up to 600°C), and then plunged vertically in a quiescent pool of subcooled water. A data acquisition system is used to record the temperature of the embedded thermocouples with time. Data reduction is performed by an inverse heat conduction code to calculate the surface temperature and corresponding surface heat flux. A visualization study with a high-speed camera is conducted to record the quenching behavior on the test sample. It is found that the minimum film boiling temperature decreases with the axial distance, while the CHF temperature is relatively insensitive to the axial distance. The film boiling heat transfer coefficient decreases with surface temperature, and seems to be independent of axial distance. The quench front is observed to originate from the bottom and move upwards. It is found that the quench front velocity remains nearly constant in the lower region of the test sample, and significantly increases in the upper region.
Feasibility Study of a New Fabrication Method for the Li4SiO4 Pebbles
Authors: Lo Frano R. , Puccini M. , Stefanelli E.
Published: October 2020

The production of reliable, stable, and durable breeder blanket material is one of the most important open challenge of fusion reactor technology.

A new low temperature fabrication method for producing lithium orthosilicate (Li4SiO4) pebbles based on the drip casting was developed at DICI - University of Pisa.

The pebbles were produced starting from a suspension of Li4SiO4 precursors prepared by sol-gel method, which was dripped through the experimental device and coagulated into solid spheres in a saline solution. Li4SiO4 pebbles were finally obtained after sintering at high temperature.

The pebbles produced are stable and well-sized (about 1.5 mm diameter). Pebbles have been analyzed in order to determine their physical and mechanical characterization. To the aim, SEM and X-ray diffractometer analyses were performed as well as compression tests by Instron machine. The pebbles obtained from the drip casting have a mechanical strength (about 11 N) comparable with that of the others solid breeder in pebbles form. In addition, if compared to these latter they have same roundness and sphericity.

Results confirmed the feasibility of drip casting forming technique, and provided further useful information for the optimization of the proposed new fabrication process.
Investigation of Turbulent Prt Model and the Segmentation Rule for LBE Turbulent Heat Transfer
Authors: Peiying Li , Jian Deng , Lei Zhong , Libo Qian , Rong Cai , Yugao Ma , Hongxing Yu
Published: October 2020

With liquid metal like lead-bismuth alloy (LBE) acting as a coolant for nuclear reactors, it is necessary to use a more accurate heat transfer relationship and a more reliable Prt model for the low Pr fluid. Because of the low Pr of liquid metal, the thermal conductivity is more dominant than the momentum transfer, which is quite different from ordinary fluids. In this case, the turbulent Prt can better reflect the heat transfer process. In this study, the Prt = A1+A2/Pr form is selected, and the corresponding coefficients are obtained by the renormalization group analysis method, then corrected by Pr. Furthermore, the applicable range and segmentation rule of the turbulent Prt model are discussed, and the obtained Prt segmentation theoretical model is written into CFD. The result shows that, compared with the previously unmodified model, the radial temperature distribution and Nusselt number (Nu) of the annular and bundle channel obtained by RANS method with the improved Prt model is in good agreement with experimental results, and the deviations are within 5%. It is proved that the turbulent Prt segmentation theoretical model proposed in this study is effective and can represent the heat transfer characteristics of liquid metal from the mechanism.
Development of a Multiphase Particle Method for Melt-Jet Breakup Behavior of Molten Core in Severe Accident
Authors: Wang Zidi , Iwasawa Yuzuru , Sugiyama Tomoyuki
Published: October 2020

In a hypothetical severe accident in a light water reactor (LWR) nuclear power plant, there is a possibility that molten core released from the reactor vessel gets in contact with water in the containment vessel. In this so-called fuel-coolant interactions (FCIs) process, the melt jet will breakup into fragments, which is one of the important factors for a steam explosion, as a potential threat to the integrity of the containment vessel. The particle method could directly and easily capture the large deformed interfaces by particle motions, benefiting from its Lagrangian description and meshless framework. In order to investigate the melt-jet breakup with solidification processes, a multiphase particle method with arbitrary high order scheme is presented in this study. In addition, an interfacial particle shifting scheme is developed to suppress the unnatural particle penetration between different phases. The convergence rate with different order is firstly confirmed by a verification test in terms of both explicit and implicit calculations. Then, a transient heat conduction between two materials is carried out and quite good results are obtained. After that, a rising bubble benchmark is performed to show the feasibility of modelling for deformation and collapse. Improvements of clear interface are indicated compared with previous reported results. Two important multiphase instabilities, namely the Rayleigh-Taylor instability and the Kelvin-Helmholtz instability, are studied since they play important roles during the melt-jet breakup. The results achieved so far indicate that the developed particle method is capable to analyze the melt-jet breakup with heat transfer.
Effects of Pre-Crack Depth and Hydrogen Absorption on the Failure Strain of Zircaloy-4 Cladding Tubes Under Biaxial Strain Conditions
Authors: Li Feng , Mihara Takeshi , Udagawa Yutaka , Amaya Masaki
Published: October 2020

Fuel cladding may be subjected to biaxial tensile stress in axial and hoop directions during pellet-cladding mechanical interaction (PCMI) of a reactivity-initiated accident (RIA). Incipient crack in the hydride rim assisted by the scattered hydrides in the metal phase may lead to failure of the cladding at small hoop strain level during PCMI. To get insight of such phenomenon, biaxial-EDC tests under axial to hoop strain ratios ranging from 0 to 1 were performed with pre-cracked (outer surface) and uniformly hydrided Zircaloy-4 cladding tube samples with final heat-treatment status of cold worked (CW), stress relieved (SR) and Recrystallized (RX). Results showed dependencies of failure hoop strain on pre-crack depth, strain ratio, hydrogen content and final heat-treatment status on fabrication, but no apparent dependencies were observed on the distribution pattern of hydrides (with similar hydrogen contents and hydrides predominantly precipitated in hoop direction) and the heat-treatment process for hydrogen charging. J integral at failure seems to be available to unify the effect of pre-crack depth.
Investigation of Hardening Law on Welding Residual Stress Analysis for Nickel Based Alloy 82 Weld Metal
Authors: Ejiri Mitsuru , Kubota Teppei , Soga Yukihiro , Nishihara Nozomi , Yanagida Nobuyoshi , Katou Hiroomi , Arashiba Tomoyuki , Taniura Wataru
Published: October 2020

There are three types of hardening laws for evaluating welding residual stress with the finite element method (FEM): kinematic hardening law, isotropic hardening law, and combined hardening law that combine these.

The purpose of this paper is to investigate which hardening law is more appropriate for the evaluation of welding residual stress of alloy 82. We first performed two types of welding tests: welding both ends of a plate, and welding the periphery of a disc. We then compared the results of mock-up welding tests with the analysis results of welding residual stress with the kinematic hardening law and combined hardening law.

Both the kinematic hardening law and the combined hardening law showed a welding residual stress distribution close to the results of the mock-up welding tests, but the combined hardening law tended to be closer to the mock-up results. Therefore when it is necessary to confirm the welding residual stress of alloy 82, it is considered appropriate to apply the combined hardening law.