bluuebunny/crossref_metadata_2025 huggingface dataset:
A Piezoelectric Droplet Generating Device for Experiment in Successive Droplets Impacting Onto Solid Surface
Authors: Li Jianxin , Zhang Huang , Li Yuzheng , Liu Qianfeng , Bo Hanliang
Published: October 2018

The outcomes of successive droplets impacting onto solid surface of the steam separator in a nuclear power system’s steam generator has a strong effect on the separating efficiency. Due to amounts of influencing factors, experimental research is an important method to study the phenomena of droplet-wall collision. However, because it is hard to generator continuous droplets with controllable uniform size and frequency, experimental studies about successive droplets impacting on solid surface are relative limited. In this study, a novel drop-on-demand (DOD) droplet generator is designed and fabricated based on piezoelectric ceramics, in which successive droplets with a uniform diameter can be generated. Firstly, the structure design of the DOD droplet generator, the setup of the control system and working principle are described in detail in this paper. Then the droplet generating performance of the device under different signal frequency fs, signal amplitude U, duty ratio Dr, and nozzle diameter Dn are investigated experimentally using a high-speed camera at 4000 fps. Finally, the influence of the signal frequency fs, voltage U, duty ratio Dr and nozzle diameter Dn on the diameter of droplet Dd is discussed. A test of successive droplets generated by the device impacting on an aluminum plate is conducted.
Development of Device for Detecting Helium Leak From Canister: Part 1 — Experiment on Temperature Behavior During Gas Leak From Canister of 1/4.5 Scale Cask Model
Authors: Takeda Hirofumi , Goto Masanori
Published: October 2018

A concrete cask has an advantage in cost and period of manufacturing compared with a metal cask. In the metal cask, monitoring of helium pressure between a primary lid and a secondary lid of the cask is required by regulation. On the other hand, in the concrete cask, the lids of a canister are welded and have high sealing performance, so that the monitoring of helium leak from the canister is not required. However, a loss of the sealing performance of the secondary lid, which is caused by stress corrosion cracking (SCC), is concerned in the case of long-term storage by the concrete cask. In the view of this situation, it would be useful to install a helium leak detector in the concrete cask in order to improve the safety of the long-term storage. Thus, we have been developing the detector. The phenomenon that the temperature at the bottom of the canister (TB) increases and the temperature at the top of the canister (TT) decreases during the helium leak from the canister has been confirmed by the experiments in the previous study. We have proposed the performance monitoring by monitoring the temperature difference ΔTBT (= TB-TT) instead of pressure monitoring. This time, to make the installation and maintenance of the helium leak detector easier in consideration of practical use, we proposed a new detection technique using only the temperature of either the lid or the bottom of the canister.

We performed leak tests by using a 1/4.5 scale cask model based on a similarity law of thermal hydraulics. In the experiments, air was used for an inner gas of a canister of the model, and a heat flux of a canister surface had the same value as that of the actual canister surface. In this model, Ra* number can be made to coincide with that of the actual canister.

Besides, Gr* number and Bo* number are almost equal to those of the actual canister. Temperature data at respective canister parts were obtained under condition of canister internal pressure from 6 atm to 1 atm (atmosphere pressure). Also, the new and old leak detection methods were evaluated by using the obtained temperature data.
Study on the Operating Strategy Optimization of Moving to Remote Shutdown Station When Main Control Room Is Un-Inhabitable
Authors: Li Li , Shengtao Zhang , Zhao Xu , Yu Du
Published: October 2018

For PWR, remote shutdown station (RSS) is a redundant control mean to shut down the reactor when main control room (MCR) inhabitation is challenged (e.g. fire, smoke...). Nowadays, due to nuclear power plants control measures were improved with DCS system, a full function DCS RSS was equipped and more essential equipment could be controlled on RSS.

Under operating conditions that prohibit nuclear power plant operators to stay in the main control room, the operators should move to RSS and shutdown the reactor to ensure plant safety following  (RSS strategy for short) to fallback the plant from power operation to cold shutdown. The original operating strategy by nature circulation is no longer the best choice both for operation safety and economy efficiency, and an optimized new strategy should be raised.

Based on the former reason, an optimized operation strategy was raised in this paper. In the optimized strategy, all plant normal standard operation modes were considered as initial conditions, rather than only considering power operation condition in the original one. The fallback mode and fallback strategy for each initial condition was also designed and optimized. To accelerate the depression and heat removal process, a forced circulation operation strategy is adopted when the reactor coolant pumps are available, and less local operation was included by taking advantages of the full function operating measures on RSS. To simplify the whole procedure structure, the operation modules of other general operating procedures are reused.

To validate the effectiveness of the optimized operating strategy, a full scope PWR simulation tool was employed to make thermo hydraulic calculation validation of the reactor response and also the remote control station HMI supporting validation. By simulating the original strategy and the optimized one and related analysis, we found that the optimized strategy is effective, and able to be executed based on the remote control station hardware. By executing the optimized strategy, the unit can fall back to the cold shutdown condition safely and a few hours were saved compared with the original strategy. The optimized strategy had already been implemented on real PWR nuclear power plant.
Safety Analysis Model of DUCG Based on FMEA/FTA
Authors: Zhou Zhenxu , Nie Hao , Dong Chunling , Zhang Qin
Published: October 2018

Failure Modes and Effects Analysis (FMEA) is a useful tool to find possible flaws, to reduce cost and to shorten research cycle in complex industrial systems. Fault Tree Analysis (FTA) has gained credibility over the past years, not only in nuclear industry, but also in other industries like aerospace, petrochemical, and weapon. Both FMEA and FTA are effective techniques in safety analysis, but there are still many uncertain factors in them that are not well addressed until now. This paper combines FMEA and FTA based on Dynamic Uncertain Causality Graph (DUCG) to solve this issue. Firstly, the FMEA model is mapped into a corresponding DUCG graph. Secondly, FTA model is mapped into a corresponding DUCG graph. Thirdly, combine the above DUCG graphs. Finally, users can modify the combined DUCG graph and calculations are made. This paper bridges the gap between FMEA and FTA by combining the two methods using DUCG. And additional modeling power and analytical power can be achieved with the advantages of the combined DUCG safety analysis model and its inference algorithm. This method can also promote the application of DUCG in the system reliability and safety analysis. An example is used to illustrate this method.
The Security Vulnerability Analysis of Nuclear Power Digital Instrument Control Platform NASPIC
Authors: Liu Hua , Yang Xiao-Hua , Liu Zhao-hui , Li Meng , Chen Zhi , Feng Zhigang , Zhao Qing
Published: October 2018

From the general industrial control system to the nuclear power plant control platform, the threat of information security has its own particularity more than continuity. The original dedicated system in general industrial area is gradually replaced by many common protocol, software and equipment. As a result, the security vulnerabilities are more likely to be used illegally. For a specific nuclear power plant digital control platform-NASPIC, the vulnerability analysis of platform is performed. Mainly two aspects of technology and management are to be carried out. For technical aspects, four categories problems-unauthorized execution, unauthorized write, unauthorized reading and reject service-are analyzed. Management is mainly about the lack of management strategy and strategy vulnerability. By analyzing the fragility of the instrument control platform, the key equipments, key channels and key modules are proposed. The qualitative and quantitative rules are deduced for evaluation of NASPIC information security.
Two Phase Flow Behavior During Pool Scrubbing
Authors: Nakamura Yuki , Fujiwara Kota , Kikuchi Wataru , Saito Shimpei , Yuasa Tomohisa , Kaneko Akiko , Abe Yutaka
Published: October 2018

In severe accidents of nuclear power plants, large amounts of gas containing radioactive particles are generated. In the process of gas release into the atmosphere, it is needed to suppress the leakage of radioactive material. The gas is decontaminated by moving radioactive particles from the gas phase to the liquid phase. This effect of capturing particles is called pool scrubbing, and it has been verified great decontamination effect. Therefore, it is extremely important to analyze the effect in evaluating the influence to the environment. But study on its principle is not carried out sufficiently. And also we don’t have enough experimental date to analyze the effect.

The purpose of this study is to clarify the gas-liquid two-phase flow behavior which is important in elucidating the mechanism of pool scrubbing. Particularly, this study focused on the behavior of bubble generation and breakup after being injected from the nozzle and the flow structure of rising bubbles in the still water. Furthermore, we evaluate the validity of the model used in the existing severe accidents analysis code such as the MELCOR by comparing the model with experimental data.

We measure the gas phase jet injected from the upward nozzle inserted to a test water tank. Nozzle diameter, gas phase flow rate, liquid phase temperature, and water depth were used as parameters. Bubble behavior was observed via a high-speed camera. The bubble rising speed, bubble distribution and void fraction were measured by a wire mesh sensor. In previous studies, experiments using non-condensable gas in normal temperature water have been mainly conducted. In order to conduct the experiment under conditions that simulate actual equipment, steam which is a condensable gas was used in this study. Moreover, it is assumed that thermal stratification is formed in the pressure suppression pool during severe accidents. To reproduce this situation, thermal stratification is formed in the test water tank. For bubble behavior and flow phenomenon, the result of using non-condensable gas was compared with that using steam. We consider the influence of formation of a thermal stratification. As described above, the flow phenomenon in the pool scrubbing was visualized and measured. Finally, we discuss the validity of the analysis code by comparing the calculation formula and model in the analysis code with the experiment data.
Core Design Study of Super FBR With Multi-Axial Fuel Shuffling and Different Coolant Density
Authors: Noda Shogo , Sukarman Sukarman , Yamaji Akifumi , Takei Tetsuo , Fukuda Takanari , Ayukawa Arisa
Published: October 2018

The Super Fast Breeder Reactor (Super FBR) utilizes supercritical light water as coolant, which changes from liquidlike high density state to gas-like low density state continuously in the core without phase change. In the preceding study (Noda et al., 2017), new concept of axially heterogeneous core with multi-axial fuel shuffling was proposed. The core consisted of two layers of mixed oxide (MOX) fuel and two layers of blanket fuel with depleted uranium (DU), which were arranged alternatively in the axial direction. The study showed that, with independent fuel shuffling in the upper part and lower part of the core, breeding performance could be improved by increasing the upper blanket fuel batch number while keeping the fuel batch number of the rest of the core unchanged, because of increased neutron flux in the upper blanket. However, the study did not consider influence of different coolant density histories in the different axial level of the core on the core neutronics. Hence, this study aims to reveal influence of the different coolant density histories through design and analyses of the multi-axial fuel shuffling core with two MOX layers and three blanket layers. The three levels correspond to the coolant density below, around, and above the pseudo-critical temperature. The neutronics calculations are carried out with SRAC 2006 code and JENDL-3.3 nuclear data library. Unit cell burnup calculations based on collision probability method are carried out for 5 different coolant density histories to consider influence of different neutron spectrum on breeding performance of the core. Influence of instantaneous coolant density changes on the core neutronics are considered by coupling core burnup calculations with thermal-hydraulics calculations based on single channel model. Influence of independent fuel shuffling of the upper blanket on the core neutronics (breeding performance and void reactivity characteristics) is investigated, followed by a similar investigation on the lower blanket. The differences between the two schemes are investigated since coolant density histories are greatly different between the upper blanket and the lower blanket.
Research on the Effect of Grid Clamping Failure on Flow Elastic Stability and Vortex Shedding of Fuel Rod
Authors: Huan-huan Qi , Zhi-peng Feng , Nai-bin Jiang , Qian Huang , Xuan Huang
Published: October 2018

Flow elastic stability and vortex shedding were two important mechanisms of the flow induced vibration analysis. Due to the influence of manufacturing process, transportation and irradiation, the clamping action of grid on fuel rods may be invalid. Taking one fuel assemblies as an example, the effects of clamping failure on the natural frequencies, mode shapes, flow elastic stability and vortex shedding were studied. The results show that the effect of the rigid convex support failure on the natural frequency was directly related to the mode shape. The effect of the grid rigid convex failure near the nodes with larger amplitude on the natural frequency was obvious. The velocity of flow at the top and bottom of the fuel rods were larger and the size was comparable, this induced that the rigid convex failure of the top and bottom grids had a significant effect on the ratio of flow elastic stability and the vortex shedding.
Preliminary Design of Unloading Device for MNSR LEU Conversion
Authors: Qian Hao , Yiguo Li , Dan Peng , Xiaobo Wu , Jin Lu , Jingyan Hong , Jinhua Zhang , Mengjiao Wang
Published: October 2018

In order to solve the problem that the current unloading operation will destroy the sealing performance of Miniature Neutron Source Reactor (MNSR) reactor vessel and the tightness can’t be restored, and to meet the application requirements that the original reactor vessel will be reloaded and operated after MNSR LEU conversion, the new unloading device is designed, which can be used without separation of reactor vessel. There has only one fuel assembly in MNSR. When the fuel assembly are unload for MNSR LEU conversion, the cover plate of the pool is removed, the cadmium string is put in, and the neutron detector is placed at first. After removing the drive mechanism and the control rod, and opening the small cover plate at the top of reactor vessel, the fuel assembly can be grabbed and unloaded by unloading tool only through the opening of the small top cover plate. The MNSR spent fuel has very high radioactivity. The auxiliary mechanical device can be used with unloading tools to realize operation in a long distance by lifting and level motion, which is convenient to shield and can reduce the works’ irradiation dose level effectively. Through calculation and analysis, the results show that the structure strength of unloading device is much larger than the actual load to ensure operation safety and reliability. The unloading device is easy to process and operate, and can be used in the practical operation of MNSR LEU conversion or decommissioning at home and abroad to simplify the operation steps and improve the working efficiency.
Analysis of Aerosol Emission and Dispersion During the Laser Cutting of Fukushima Fuel Debris Simulants
Authors: Porcheron Emmanuel , Peillon Samuel , Gelain Thomas , Chagnot Christophe , Journeau Christophe , Roulet Damien
Published: October 2018

The general context of the article is related to the development of the laser cutting technique for the fuel debris retrieval on the damaged reactors of Fukushima Dai-ichi. IRSN and CEA are involved in a project, led by ONET, to bring relevant elements to analyze the risk occurred by the dispersion of aerosols emitted by the dismantling operations. Results regarding the aerosols source term characterization emitted during laser cutting of non-radioactive fuel debris simulants were acquired during experiments undertaken on the DELIA cutting laser platform from CEA. IRSN realized aerosol sampling, aerosol size distribution measurement and CFD calculation of aerosol transport and wall deposition. The evaluations performed will enable the Japanese teams responsible for extracting corium from the damaged reactors of Fukushima Dai-ichi to define the best strategies to implement containment, and ultimately to limit the dissemination of radionuclides in the environment.
A Novel More Reliable and Extensible Architecture of Instrumentation and Control Systems
Authors: Zhou Shuqiao , Guo Chao , Li Duo , Huang Xiaojin
Published: October 2018

Digital instrumentation and control (I&C) systems are widely used in many industrial areas. In the recent years, the digitalization process for nuclear power plants has also been moving on rapidly. Full digital I&C systems are now adopted in almost all new constructed nuclear power plants. The architecture of a digital I&C system plays a pivotal role for the safety, reliability and security of the whole nuclear power plant. Moreover, for the advanced small modular reactors, both the reliability and extensibility of I&C systems are especially required.

Therefore, in this paper we propose a new architecture of the digital I&C systems based on the developed computing performance and communication technology. The control units and the data servers in the new proposed architecture are decentralized and working in a mutually redundant and distributed computing/storage way. Thus the architecture is with a flexible extensibility. Moreover, other control units or data servers can take over the functions of a certain number of failed ones. This characteristic benefits the system’s reliability significantly. The reliability of the new architecture is theoretically evaluated and the results demonstrate that it is much higher than that of the traditional architecture of I&C systems.
OECD-NEA Expert Group on Multi-Physics Experimental Data, Benchmarks and Validation
Authors: Valentine Timothy , Ivanov Kostadin , Avramova Maria , Petruzzi Alessandro , Hudelot Jean-Pascal , Rohatgi Upendra , Velkov Kiril
Published: October 2018

High-fidelity, multi-physics modeling and simulation (M&S) tools are being developed and utilized for a variety of applications in nuclear science and technology and show great promise in their abilities to reproduce observed phenomena for many applications. Even with the increasing fidelity and sophistication of coupled multi-physics M&S tools, the underpinning models and data still need to be validated against experiments that may require a more complex array of validation data because of the great breadth of the time, energy and spatial domains of the physical phenomena that are being simulated. The expert group on Multi-Physics Experimental Data, Benchmarks and Validation (MPEBV) of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) was formed to address the challenges with the validation of such tools. The work of the MPEBV expert group is shared among three task forces to fulfill its mandate and specific exercises are being developed to demonstrate validation principles for common industrial challenges. This paper describes the overall mission of the group, the specific objectives of the task forces, the linkages among the task forces, and the development of a validation exercise that focuses on a specific reactor challenge problem.
Creation of an OpenFOAM Fuel Performance Class Based on FRED and Integration Into the GeN-Foam Multi-Physics Code
Authors: Fiorina Carlo , Pautz Andreas , Mikityuk Konstantin
Published: October 2018

The FRED code is an in-house tool developed at the Paul Scherrer Institut for the so-called 1.5-D nuclear fuel performance analysis. In order to extend its field of application, this code has been re-implemented as a class of the OpenFOAM numerical library. A first objective of this re-implementation is to provide this tool with the parallel scalability necessary for full-core analyses. In addition, the use of OpenFOAM as base library allows for a straightforward interface with the standard Open-FOAM CFD solvers, as well as with the several OpenFOAM-based applications developed by the nuclear engineering community. In this paper, the newly developed FRED-based Open-FOAM class has been integrated in the GeN-Foam multi-physics code mainly developed at the École polytechnique fédérale de Lausanne and at the Paul Scherrer Institut. The paper presents the details of both the re-implementation of the FRED code and of its integration in GeN-Foam. The performances and parallel scalability of the tool are preliminary investigated and an example of application is provided by performing a full-core multi-physics analysis of the European Sodium Fast Reactor.
Modeling of Low and High Void Fraction Two-Phase Flow in a Vertical Pipe by Using the Two-Fluid Model
Authors: Yu Shimo , Yan Xiao , Zhang Junyi
Published: October 2018

Two-phase flow is an important and common phenomenon in nuclear reactor systems, and the characteristics of two-phase flow such as heat transfer and pressure drop strongly depend on the radial distribution of void fraction. This paper is presenting the CFD simulation for void fraction radial distribution of mono- and poly-disperse air-water two phase flow using Euler-Euler two-fluid model. Interfacial forces including transverse forces such as lift, wall and turbulent dispersion forces are taken into account, Furthermore the bubble size distribution and bubble break-up and coalescence processes are taken into account in case of a poly-disperse flow by using the S-Gamma model. The sauter mean diameter and interfacial area concentration (IAC) distribution can also be obtained. The simulation results are compared to an experimental database of MT-LOOP test facility (FZD, Germany)[1].
First Attempt to Determine a Critical Heat Flux Correlation for Thermalhydraulic System Codes
Authors: Herer Christophe , Lejosne Antoine
Published: October 2018

Boiling crisis or departure from nucleate boiling is a key phenomenon in heat transfer processes. It appears when a vapor blanket is created at the heated wall and impedes its cooling. Due to either excessive heat flux (Critical Heat Flux - CHF) or high local void (dryout), a significant temperature rise is observed and the clad might be damaged. Thermalhydraulic accident analysis includes therefore the determination of appearance of boiling crisis in order to compute the evolution of the clad temperature during the transient. The timing of departure from nucleate boiling (DNB) provides the corresponding value of residual power and therefore the clad temperature. Boiling crisis cannot be predicted from first principles. Numerous experiments form a huge database from which hundreds of correlations have been derived. As the primary use of boiling crisis correlation is associated with subchannel analysis codes, the developed correlations of interest are constructed with two-fluid mixture local thermalhydraulic variables. This approach imposes that, when used in two-fluid six equation codes, the liquid and vapor variables are combined to compute the two-phase mixture mass velocity and thermodynamic quality to get the CHF. Therefore, it is logical to consider the possibility to set up a correlation directly connecting CHF and two-fluid (liquid and vapor) thermalhydraulic variables (velocities and enthalpies along with pressure and void fraction). Two additional features were considered. The first one was to establish a correlation using data obtained in rod cluster configuration and not in tubes as the ones currently implemented in thermalhydraulic system codes. For this, the measurements carried out at Columbia University on different rod bundle configurations were considered, as being now released by EPRI in the public domain. The second one was to attach to the coefficients of the correlation their uncertainties to allow best estimate plus uncertainties calculations. First calculations showed promising results such as obtaining a correlation with similar qualities as the already existing ones, based on two-phase mixture parameters. However, a lot still remains to be done in order to obtain a reliable correlation to be implemented in thermalhydraulic system codes.
Revamping of a Graduate Radiochemistry Course for Nuclear Forensics Applications
Authors: Drescher Adam , De Luna Brandon , Pasman Marjolein , Haas Derek , Landsberger Sheldon
Published: October 2018

Laboratories in traditional radiochemistry courses typically involve basic and fundamental understanding in solvent extraction, ion exchange, precipitation, etc. procedures. With the increased focus on nuclear forensics in pre- and post-detonation scenarios different skill sets are now required for the student to learn. At the University of Texas we have developed two independent graduate courses in gamma-ray spectrometry and radiochemistry. Currently, we have amalgamated these two courses to 1. better serve our nuclear engineering graduate students, many of which are involved in nuclear forensics and 2. to attract both undergraduate and graduate students from the Chemistry Department. We incorporated gamma-ray spectrometry laboratories with several others which are nuclear forensics related. The seven laboratory sessions include half-life measurement of 137mBa (the daughter produce of 137Cs) and secular equilibrium, basic gamma and beta shielding, and gamma-ray spectrometry calibration, resolution and uncertainty in statistics. These labs have been augmented with four others including uranium fission product identification, 137Cs soil profile with Compton suppression, tritium analysis in water with a liquid scintillation counter and double replacement reaction.
A Monte Carlo Method for Simulating Stochastic Neutron Fields in Criticality Transients
Authors: Xu Qi , Wang Zhe , Xiao Gang
Published: October 2018

In presence of a weak neutron source, the initial growth of neutron population in a supercritical system exhibits a significant stochastic feature, both initiation and burst waiting times are uncertain. As a result, the energy released during criticality excursions is stochastic, obeying a probability distribution. When criticality accidents and pulsed reactor experiments are analyzed, it is important to estimate this kind of stochastic feature, including assessing the initiation probability and then the fission energy probability distribution. Thus a Monte Carlo direct simulation method has been proposed and the corresponding code MES has been developed. By taking random factors during criticality excursions into account in dynamic Monte Carlo simulations, this method is capable of simulating the whole process from source injection to exponential growth of the neutron population, and finally to extinction of the neutron pulse. A set of static initiation probability problems and a figurative criticality excursion problem have been applied to validate this method and MES. Results demonstrate that with the proposed method MES is able to simulate stochastic transient neutron fields in multiplying systems during criticality excursions.
Experimental Research on Steam Jetting and Condensation in Low Sub-Cooled Water
Authors: Dai Chunhui , Liao Mengran , Xiao Qi , Wu Jun , Li Shaodan , Gou Jinlan
Published: October 2018

Steam submerged jetting is an important process in depressurization tank and condenser deaerator tank of nuclear power plant. As the steam contact the liquid water directly, some complicated behaviors such as strong turbulence and phase transition would happens. Especially when the sub-cooling degree is low, the condensation may cause vigorous pressure pulsation and radiation noise, which not only causes noise damage to workers but also affect the safety of the heat exchanger tubes bundle because of vibration transmission. An experiment is proposed to study the complex evolutionary behavior and vibration and noise characteristics of gas-water two-phase flow. The experimental results show that in the case of low subcooling, the mass flow rate of steam has a great influence on gas plume, and, as the mass flow rate increases, the main contribution frequency of noise is gradually increasing from low frequency to high frequency. The researches in this paper can provide the technical basis for the design of the deoxygenation system of condenser in onshore and ship nuclear power plant.
The Effect of Functional Spacers on the Liquid Film Thickness and Dryout in a BWR Fuel Bundle Model
Authors: Bolesch Christian , Robers Lukas , Zboray Robert , Prasser Horst-Michael
Published: October 2018

For the BWRs, the dryout margin is one of the core design limitation factors. Today’s industry standard is to use a large margin to dryout and functional spacer grids with vanes to enhance the heat transfer and to reduce the fraction of entrained droplets. Difficulties for precise measurements under reactor conditions lead to a lack of knowledge on the exact effects of the spacers on the flow and suggest the use of scaled experiments. For this experiment, the goal is to provide high-resolution data for CFD code validation as well as visualizing the effects of functional spacers and the liquid film and potentially the dryout front.

The Dryout Tomography Experiment (DoToX) facility at ETH Zürich is a closed loop experiment for two-phase flow investigations in a fuel bundle model using a modelling fluid. Key aspects are a single undisturbed subchannel and the surrounding four heating rods containing a liquid heating system. This setup allows for a steady state dryout without endangering the structural integrity of the facility and for the 3D reconstruction of the time averaged void distribution within the flow channel by means of an X-Ray and cold neutron Computer Tomography (CT).

In this study we pay special attention to the annular flow in the upper half of the sub channel. We investigate the first results delivered by the facility. Prototypical spacer designs available in the open literature were used. We present the Liquid Film Thickness (LFT) distributions on the walls of the heating rods. Improvements towards the dryout performance as well as drawbacks of the specified spacer design are highlighted.
Heat Transfer to Supercritical Water (Liquid-Like State) Flowing in a Short Vertical Bare Tube With Upward Flow
Authors: Zvorykin A. , Mahdi M. , Popov R. , Barati Far K. , Pioro I.
Published: October 2018

Current Nuclear Power Plants (NPPs) equipped with water-cooled reactors (the vast majority of all NPPs) have relatively low thermal efficiencies within the range of 30–36% compared to those of modern advanced thermal power plants (SuperCritical Pressure (SCP) coal-fired — up to 55% thermal efficiency and combined cycle — up to 62%). Therefore, next generation reactors / NPPs should have higher thermal efficiencies close to those of current thermal power plants.

Around 60 years ago thermal-power industry has moved from subcritical pressures to SCPs with the major objective to increase thermal efficiency. Based on this proven in power industry experience it was proposed to design SuperCritical Water-cooled Reactors (SCWRs), which are one of the six Generation-IV nuclear-reactor concepts under development in selected countries. These days, there are discussions on developing even Small Modular Reactors (SMRs) of SCPs.

In spite of a large number of experiments in long bare tubes (pipes) cooled with SCW, developing SCWR concepts requires experimental data in bundle geometries cooled with SCW, which are usually shorter and will have smaller diameters. However, such experiments are extremely complicated and expensive plus each bundle geometry will have a unique Heat-Transfer (HT) characteristics due to various bundle designs.

Therefore, as a preliminary and a universal approach — experiments in bare tube of shorter heated lengths and of smaller diameters to match heated lengths and hydraulic-equivalent diameters of fuel bundles are required.

Current paper provides experimental data obtained in a short (0.6 m) vertical bare tube of a small diameter (6.28 mm) cooled with upward flow of SCW. Analysis of this dataset is also included. Main emphasis of this research is on liquid-like cooling within the possible conditions of future SCWRs and SCW SMRs. Two HT regimes are encountered at these conditions: 1) Normal HT (NHT) and 2) Deteriorated HT (DHT). Conditions at which the DHT regime appeared are discussed.
Estimation of External Dose Volume Correction Factors Based on Neural Network
Authors: Chen Junjun , Qu Jingyuan , Gong Junjun
Published: October 2018

Before the design of the nuclear facility, it is necessary to estimate the dose caused by radioactive material released into the environment during the course of the serious accidents. Semi-infinite hemisphere geometric model is established to estimate the personal external exposure dose outdoors in which the distribution of the nuclides is assumed to be uniform. The exposed staffs are in a limited cubic space when using this model to evaluate the controllability of the main control room. Thus, the volume correction factor is needed to correct the dose, whose traditional expression is f = 352/pow(V, 0.338). The formula cannot satisfy the requirement of higher accuracy due to the neglect of the influence of the shape of geometric model and γ-rays energy. Usually the actual control room is a cube and the γ-rays energies emitted from various nuclides are different.

In order to calculate the accurate volume correction factor of main control room under different geometric conditions, a finite cubic geometric model is established in this paper. The length and width of the model are between 6m and 50 m, the height is between 4m and 6m, and γ-ray energy respectively are 0.05, 0.2, 0.733, 1.2 and 3 MeV, respectively. The effective volume values for different conditions are calculated by the Monte-Carlo program, and 318 groups of results are obtained. The calculated volume dose rate of 360m × 360m × 255m (assuming semi-infinite) cube at 733keV γ-rays energy is taken as a criterion, whose ratio of the other calculation results is the new volume correction factor value. By comparing two volume correction factors, the relative discrepancies are within 3 folds, proving that the calculation result is reasonable and feasible. The new volume correction factor varies with γ-rays energy and the shape of the geometric model.

A neural network model corresponding to the volume correction factor is developed to apply to more cases. 80% of the results are randomly selected as the training set of neural network. The remaining 20% of the result as the test set of the cross-test is to predict the results of the trained neural network, whose relative errors are less than 5%. The neural network model can obtain the volume correction factor under different geometry and γ-ray energy conditions. Finally, a volume correction factor library is established, which can provide a powerful reference to obtain the volume correction factor of the limited space model such as the main control room.
Research on Thermal Efficiencies of Various Power Cycles for GFRs and VHTRs
Authors: Mahdi Mohammed , Popov Roman , Pioro Igor
Published: October 2018

The vast majority of Nuclear Power Plants (NPPs) are equipped with water- and heavy-water-cooled reactors. Such NPPs have lower thermal efficiencies (30–36%) compared to those achieved at NPPs equipped with Advanced Gas-cooled Reactors (AGRs) (∼42%) and Sodium-cooled Fast Reactors (SFRs) (∼40%), and, especially, compared to those of modern advanced thermal power plants, such as combined cycle with thermal efficiencies up to 62% and supercritical-pressure coal-fired power plants — up to 55%. Therefore, NPPs with water- and heavy-water-cooled reactors are not very competitive with other power plants. Therefore, this deficiency of current water-cooled NPPs should be addressed in the next generation or Generation-IV nuclear-power reactors / NPPs.

Very High Temperature Reactor (VHTR) concept / NPP is currently considered as the most efficient NPP of the next generation. Being a thermal-spectrum reactor, VHTR will use helium as a reactor coolant, which will be heated up to 1000°C. The use of a direct Brayton helium-turbine cycle was considered originally. However, technical challenges associated with the direct helium cycle have resulted in a change of the reference concept to indirect power cycle, which can be also a combined cycle.

Along with the VHTR, Gas-cooled Fast Reactor (GFR) concept / NPP is also regarded as one of the most thermally efficient concept for the upcoming generation of NPPs. This concept was also originally thought to be with the direct helium power cycle. However, technical challenges have changed the initial idea of power cycle to a number of options including indirect Brayton cycle with He-N2 mixture, application of SuperCritical (SC)-CO2 cycles or combined cycles.

The objective of the current paper is to provide the latest information on new developments in power cycles proposed for these two helium-cooled Generation-IV reactor concepts, which include indirect nitrogen-helium Brayton gas-turbine cycle, supercritical-pressure carbon-dioxide Brayton gas-turbine cycle, and combined cycles. Also, a comparison of basic thermophysical properties of helium with those of other reactor coolants, and with those of nitrogen, nitrogen-helium mixture and SC-CO2 is provided.
Finite Element Analysis for Fuel Assembly Structural Behavior
Authors: Yoo Youngik , Kim Kyounghong , Eom Kyongbo , Lee Seongki , Yoo Jongsung
Published: October 2018

Spent nuclear fuel (SNF) integrity evaluation related to its handling and transportation for mid-/long-term dry storage is a regulatory requirement. Especially, a drop event is the most fatal failure mode among regulatory conditions. For SNF drop accidents, it is required that the mechanical integrity of the SNF be evaluated using test results or analytic methodologies. The SNF mechanical test, however, takes much time and cost, and there are safety issues related to the release of radioactive materials. Thus, finite element analysis is used as an alternative to the experimental test method to solve this problem. In this study, a three dimensional (3D) finite element model was developed using ABAQUS software to simulate the structural behaviors of a fresh fuel assembly (FA) prior to applying SNF properties because of a lack of SNF test results. Static and dynamic mechanical behaviors were simulated with this model and compared with the fresh FA test results. The analysis results are in good agreement with the test results. Therefore, the analysis model consistent with the test results will be applied to the evaluation of the FA drop integrity reflecting the specific SNF characteristics.
CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor
Authors: Chen Jing , Zhang Dalin , Qiu Suizheng , Zhang Kui , Wang Mingjun , Su G. H.
Published: October 2018

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.
Dynamic Fault Tree Analysis Based on Dynamic Uncertain Causality Graph
Authors: Zhou Zhenxu , Dong Chunling , Zhang Qin
Published: October 2018

Dynamic Fault Tree (DFT) has drawn attention from comprehensive industrial systems in recent years. Many analytical approaches are developed to analyze DFT, such as Markov Chain based method, Inclusion-Exclusion Rule based method, and Sum-of-Disjoint-Product theory based method. Novel methods such as Bayesian Network and Petri Net are also used to solve DFT. However, Basic events are usually assumed unrepairable and are restricted to specific probabilistic distributions. And some methods may suffer from combination explosion. This paper applies Dynamic Uncertain Causality Graph (DUCG) to analyze DFT to overcome the aforementioned issues. DUCG is a newly proposed Probabilistic Graphic Model for large complex industrial systems which allows for dynamics, uncertainties and logic cycles. The DUCG based methodology can be summarized as event mapping, logical mapping, and numerical mapping. This paper studies how to map the PAND, FDEP, SEQ AND SPARE sequential logic gates into equivalent representations in DUCG. With the DUCG representation mode, one can analyze DFT with algorithms in DUCG. Future work will be done on benchmark tests and on software development.
Numerical Simulation of Bubble Dynamic Under Ocean Conditions
Authors: Chong Chen , Mingjun Wang , Wenxi Tian , Suizheng Qiu , Guanghui Su
Published: October 2018

In recent years, nuclear power are widely used in the ship and submarine. Affected by the ocean wave, the bubble dynamic under ocean conditions may be different from stationary condition. In order to investigate the influence mechanism of ocean conditions, the single bubble rising in stagnant liquid under rolling conditions with different frequencies and amplitudes were calculated through CFD method. In present work, the typical ocean conditions such as rolling were realized through dynamic mesh method, this method could simulated the movement of container directly. The Volume-of-fluid (VOF) method was applied to track the interface between liquid and gas phases and the surface tension were calculated by Continuum Surface Force (CSF) method. We can draw the conclusions as follows: (1) the bubble tend to move laterally and periodically under rolling conditions, this may attributed to the additional force caused by rolling motion. (2) the period of lateral movement are in accordance with the rolling period, and the frequency and amplitude of bubble lateral movement may increasing as the decreasing of rolling period. (3) The amplitude of the bubble lateral displacement is proportional to the rolling amplitude. (4)larger bubbles have the higher rising velocity than the small ones, and the larger bubble are easier to break up which may promote the lateral movement.
Research on Simplified Method of Transients Combining and Loading in Fatigue Crack Growth Analysis of Carbon Steel Nuclear Piping
Authors: Liu Zhenshun , Zhen Hongdong
Published: October 2018

The fatigue crack growth analysis of nuclear piping is a nonlinear calculation process. The loading sequence and combination mode could affect the results. How to consider these effects is unclear. Fatigue crack growth analysis includes a large number of nonlinear iterative calculations, and the calculation speed is slow.

This paper selects carbon steel nuclear piping as the research object. Based on the analysis process provided by ASME code XI volume, a simplified analysis method is explored by introducing the reference crack depth a’ and the envelope transient. The simplified analysis method is conservative because it has been proved that the crack growth rate is positively related to the crack size only if the maximum stress intensity factor is greater than 0 and the minimum stress intensity factor is less than 0. The simplified analysis method is proved to be able to significantly improve the calculation speed by comparing the number of iterative calculations in the simplified analysis method and in the conventional analysis method.

The results indicate that the simplified analysis method could provide a conservative way of loading and combining the complex transients and could significantly reduce the number of nonlinear iterative calculations in the process of crack fatigue growth analysis for carbon steel nuclear piping when the maximum stress intensity factor greater than 0 and the minimum stress intensity factor is less than 0.