bluuebunny/crossref_metadata_2025 huggingface dataset:
THERMAL AND HYDRODYNAMIC PERFORMANCE OF A MICROCHANNEL HEAT SINK COOLED WITH CARBON NANOTUBES NANOFLUID
Authors: Nik Mazlam Nik Ahmad Faiz , Mohd-Ghazali Normah , Mare Thierry , Estelle Patrice , Halelfadl Salma
Published: October 2016

The microchannel heat sink (MCHS) has been established as an effective heat removal system in electronic chip packaging. With increasing power demand, research has advanced beyond the conventional coolants of air and water towards nanofluids with their enhanced heat transfer capabilities. This research had been carried out on the optimization of the thermal and hydrodynamic performance of a rectangular microchannel heat sink (MCHS) cooled with carbon nanotube (CNT) nanofluid, a coolant that has recently been discovered with improved thermal conductivity. Unlike the common nanofluids with spherical particles, nanotubes generally come in cylindrical structure characterized with different aspect ratios. A volume concentration of 0.1% of the CNT nanofluid is used here; the nanotubes have an average diameter and length of 9.2 nm and 1.5 mm respectively. The nanofluid has a density of 1800 kg/m3 with carbon purity 90% by weight having lignin as the surfactant. The approach used for the optimization process is based on the thermal resistance model and it is analyzed by using the non-dominated sorting multi-objective genetic algorithm. Optimized outcomes include the channel aspect ratio and the channel wall ratio at the optimal values of thermal resistance and pumping power. The optimized results show that, at high operating temperature of 40°C the use of CNT nanofluid reduces the total thermal resistance by 3% compared to at 20°C and consequently improve the thermal performance of the fluid. In terms of the hydrodynamic performance, the pumping power is also being reduced significantly by 35% at 40°C compared to the lower operating temperature.
Ulama Dan Wacana Islam Dalam Hubungan Intelektual Di Tanah Melayu Dan Indonesia.
Authors: Mohd Nasir Badlihisham , Abd. Rahim Rahimin Affandi , Hamzah Khafidz
Published: October 2016

This article is about the Islamic discourse upon the role of ulama and intellectual relationship in enriching the Muslim civilization in Malaysia and Indonesia. At the outset, it will explore the significant of Malay-Muslim civilization as well as the major role play by ulama in the development of Muslim civilization in Malay Archipelago. Besides, for the specification purpose it will concentrate on several issues pertaining to the main focus, namely; the Terengganu's Batu Bersurat, the Kaum Muda-Tua discourse in 1930s, the dakwah movement in 1970s and the Islamization of knowledge. Finally we may sum up that those intellectual discourses had enrich the Malay-Muslim knowledge resources in coping with the force modernization initiated by western power in the Malay Archipelago.
Steady-State and Stability Analysis of Natural Circulation With Fluids at Supercritical Pressure
Authors: Carbone Francesco , Ambrosini Walter
Published: October 2016

Natural circulation of CO2 at supercritical pressure is addressed in the present paper in order to assess the capabilities of linear and nonlinear analysis in-house codes developed for studying steady-state and transient behaviour of natural circulation loops. In past activities, the models have been developed, taking into account also the effect of the presence of heating structures, and have been equipped with a low diffusion numerical scheme, in order to minimise truncation error effects. In the present work, available heat transfer correlations were introduced in order to properly simulate heat transfer to supercritical fluids, though the present state-of-the-art in the field does not yet allow to reliably simulate relevant effects as the deteriorated of heat transfer occurring when buoyancy effects become dominant. The obtained transient and linear stability analysis codes were applied to published experiments performed at the Bhabha Atomic Research Centre in Mumbai, India, making use of carbon dioxide as working fluid. Both steady-state conditions and transient behaviour were addressed, obtaining interesting data in comparison to experimentally observed behaviour.
Creep Assessment of Large Size High Temperature Components Using Small Creep Test Specimens
Authors: Ali Balhassn S. M.
Published: October 2016

Most of the large components in the thermal, traditional and nuclear power plants such as pressurized vessels and pipes are operating at elevated temperatures. These temperatures and stress are high enough for creep to occur. For variety of reasons many of these power plants are now operating beyond their design life time. It is -known fact that as the high temperature components aged the failure rate normally increases as a result of their time dependent material damage. Further running of these components may become un-safe and dangerous in some cases. Therefore, creep assessment of the high temperature components of these plants is essential for their safe operation. Mainly for economic reasons these components have to be creep assessed as they are in service. However, assessing the creep strength for these high temperature components as they are in service, it can be challenging task, especially when these components are operating under extremely high temperature and/or stress. This paper introduces newly invented, small creep test specimens techniques. These new small types of specimens can be used to assess the remaining life times for the high temperature components, using only small material samples. These small material samples can be removed from the operating components surface, without affecting their safe operation. Two of the high temperature materials are used to validate the new testing techniques.
Development of Risk Assessment Methodology of Decay Heat Removal Function Against Natural External Hazards for Sodium-Cooled Fast Reactors: Project Overview and Volcanic PRA Methodology
Authors: Yamano Hidemasa , Nishino Hiroyuki , Kurisaka Kenichi , Okano Yasushi , Sakai Takaaki , Yamamoto Takahiro , Ishizuka Yoshihiro , Geshi Nobuo , Furukawa Ryuta , Nanayama Futoshi , Takata Takashi
Published: October 2016

This paper describes mainly volcanic probabilistic risk assessment (PRA) methodology development for sodium-cooled fast reactors in addition to the project overview. In the volcanic PRA, only the effect of volcanic tephra (ash) was taken into account because there is a great distance between a plant site assumed in this study and volcanos. The volcanic ash could potentially clog air filters of air-intakes that are essential for the decay heat removal. The degree of filter clogging can be calculated by atmospheric concentration of ash and tephra fallout duration and also suction flow rate of each component. The atmospheric concentration can be calculated by deposited tephra layer thickness, tephra fallout duration and fallout speed. This study evaluated a volcanic hazard using a combination of tephra fragment size, layer thickness and duration. In this paper, each component functional failure probability was defined as a failure probability of filter replacement obtained by using a grace period to a filter failure limit. Finally, based on an event tree, a core damage frequency was estimated about 3 × 10−6/year in total by multiplying discrete hazard probabilities by conditional decay heat removal failure probabilities. A dominant sequence was led by the loss of decay heat removal system due to the filter clogging after the loss of emergency power supply. A dominant volcanic hazard was 10−2 kg/m3 of atmospheric concentration, 0.1 mm of tephra diameter, 50–75cm of deposited tephra layer thickness, and 1–10 hr of tephra fallout duration.
Experimental Study of Steam Pressure Suppression by Condensation in a Water Tank at Sub-Atmospheric Pressure
Authors: Mazed Dahmane , Lo Frano Rosa , Aquaro Donato , Del Serra Daniele , Sekachev Igor , Orlandi Francesco
Published: October 2016

The Vacuum Vessel Pressure Suppression System (VVPSS), a key safety system of the ITER plant, is designed to protect the Vacuum Vessel (VV) from over pressure occurring in the case of LOCA (Loss Of Coolant Accident) or other pressurizing accidents such as LOVA (Loss Of Vacuum Accident). The steam condensation in the Suppression Tanks (main elements of the VVPSS system), occurs at sub-atmospheric pressure.

The steam condensation, at pressures equal or greater than the atmospheric, has been numerically analyzed and experimentally investigated in the past in order to optimize the design of the pressure suppression system of boiling water nuclear reactors.

However, very limited experimental data is available concerning the steam condensation in a water tank at sub-atmospheric pressure. In order to analyze the steam condensation in these operating conditions, an experimental study, funded by ITER Organization, is conducted at the Department of Civil and Industrial Engineering (DICI) of University of Pisa.

The tests analyze the condensation of saturated or superheated steam at sub-atmospheric pressures (4.2 kPa and slightly above the water vapour saturation pressure), and pool temperature up to 50°C at several heights of water head.

The experimental facility, to perform this study, has been set up with a significant scaling factor regarding the full size installation at ITER. In this paper the experimental rig, the conditions of the experiments, and the test matrix are presented. The temperature and pressure measurements with details of the data acquisition system are described.

The tests were performed at different patterns of the sparger exit holes (1, 3 and 9) and for three steam mass flow rates per one hole. The results show very high efficiency of condensation for all examined conditions.

Finally, a comparison between the condensation regimen at sub-atmospheric and at atmospheric pressure is discussed.
Experimental Investigation of Static Characteristics of Swing and Tilting Disc Check Valves
Authors: Marcinkiewicz Jerzy , Adamkowski Adam , Lewandowski Mariusz , Janicki Waldemar
Published: October 2016

The presented experimental investigations of static characteristics of check valves are a contribution to work leading to improve check valve models in 1D network programs. Two designs of check valves were tested: a swing disc and a tilting disc check valve. For each type of these valves static characteristics for the two directions of water flow were performed.

Following physical quantities were measured during the experiments: the angular position of the valve closing element (disc), water discharge, pressure losses caused by the check valve, moment of the hydrodynamic forces acting on the valve disc and pressure at selected locations of the flow system of the test rig.

The paper describes the test rig designed for developing the static characteristics of the check valves and presents the results of tests for the characteristics of the swing and tilting disc check valve under steady flow conditions. These characteristics are presented as time-averaged parameters measured during the tests. Following dimensionless quantities are calculated basing on these parameters:

- the flow rate coefficient,

- the coefficient of the moment of hydrodynamic forces acting on the valve disc,

- the local loss coefficient of the valve.

The above-mentioned coefficients were determined for disc angular position of selected values (opening rate) and interpolated within the full range of openings for both directions of flow through the valve.

The static behaviors of the tested valves are discussed and compared with each other.
Handling and Conditioning Techniques of Activated Chips for the Decommissioning of the E. Fermi Nuclear Power Plant
Authors: Di Pietro Massimo , Pistelli Sergio , Garneri Eugenio , Lo Frano Rosa , Ciolini Riccardo
Published: October 2016

In this paper a new technique to handle solid radioactive materials inside a liquid matrix is presented. The conceptual design of the device profits of the experience and know-how gained in decontamination procedures. The proposed system makes use of an ejector for the suction of a water-highly radioactive swarf mixture from the purifiers pool of the Italian E. Fermi NPP and moving it in a suitable container for the subsequent conditioning. A dedicated circuit with an ejector to demonstrate the feasibility of the method was realized. A minimum inlet flow was found to have swarf suction. No swarf packaging conditions were observed in the performed tests, demonstrating the feasibility of the method.
Real-Time State Identification of Boiling Water Reactors Using Relevance Vector Machines
Authors: Alamaniotis Miltiadis , Cappelli Mauro
Published: October 2016

Automated state identification systems facilitate reactor monitoring and control of nuclear systems by consolidating information collected by deployed sensors. In the current paper, we present the use of relevance vector machines (RVM) for real-time state identification of boiling water reactors (BWR). In particular, RVM models utilize the incoming signals of interest and identify in real time the state of the BWR either as normal or as one of the transition states. Each of the RVM models is assigned to a single signal; it receives the measured value at each instance and outputs the identified BWR state. The state that has been designated by the majority of the signals is displayed to the human operator as the identified BWR state. The proposed methodology is applied and tested on a set of signals taken from the FIX-II experimental facility that is a scaled representation of a BWR.
The iROCS Approach to Mitigating Beyond-Design-Basis External Events
Authors: Kim Jaewhan , Park Soo-Yong , Ahn Kwang-Il
Published: October 2016

An extended loss of all electric power occurred at the Fukushima Dai-ichi nuclear power plant by a large earthquake and subsequent tsunami. This event led to a loss of reactor core cooling and containment integrity functions at several units of the site, ultimately resulting in large release of radioactive materials into the environment. In order to cope with beyond-design-basis external events (BDBEEs), this study proposes the iROCS (integrated, RObust Coping Strategies) approach. The iROCS approach is characterized by classification of various plant damage conditions (PDCs) that might be impacted by BDBEEs and corresponding integrated coping strategies for each of PDCs, aiming to maintain and restore core cooling (i.e., to prevent core damage) and to maintain the integrity of the reactor pressure vessel if it is judged that core damage may not be preventable in view of plant conditions. The plant damage conditions considered in the iROCS approach include combinations of the following conditions of the critical safety functions: (1) an extended loss of AC power, (2) an extended loss of DC power (loss of the monitoring and control function at control rooms), (3) a loss of RCS inventory, and (4) a loss of secondary heat removal. From a case study for an extreme damage condition, it is shown that candidate accident management strategies should be evaluated from the viewpoint of effectiveness and feasibility against extreme damage conditions of the site and accident scenarios of the plant.
Research on Manufacturing Technology of China Experimental Fast Reactor Fuel Rod
Authors: Hu Yu , Wang Jianhong
Published: October 2016

CEFR, which is the acronym of China Experimental Fast Reactor, as the fourth generation advanced nuclear power system, plays a very important role in the development of nuclear industry in the future. Currently, as a kind of study, has not yet formed a fuel rod and fuel assembly manufacturing company in China. Realizing the localization of fast reactor fuel rod, fuel assembly and equipment has a very positive meaning to China’s nuclear industry development. At the same time for the CJNF’s development, promoting the manufacture level and scientific research technology of product, the research plays an extremely important role.

In this paper, a fuel rod extrusion pit device and a fuel rod cladding pipe wire equipment have been designed according to the characteristic of fast reactor components. The fuel rod extrusion pit device to ensure pit deepness is 1.8–2.0mm, the radius of pit is 3mm, three pits into each other 120°, deviation is ±5°, the distance is 450.4±0.3mm. And the fuel rod cladding pipe wire equipment meet the product technical requirements for the pitch of 100±5mm, each distance deviation shall not be more than ±15mm, wire tension force for 100±20N. Massive tests on the extrude pits, press plug, load pellet, wire and spot welding, fuel rods ring welding and other important process step have been conducted and develop a set of optimized production plan.

Fast reactor fuel rods structure have a enormous difference with the past other types of products have, the control of the input power for welding and an increase in the pressure of sealed welding chambers have been employed and avoided the problem of weld inflatable and weld porosity successfully. Results showed that the study on manufacturing of fast reactor fuel rod achieves success and realize the localization of fast reactor fuel rod in CJNF.
Thermodynamic Performance Analysis of Megawatts Portable Nuclear Power System Based on Regenerative Brayton Cycle
Authors: You Ersheng , Shi Lei
Published: October 2016

Nuclear energy is a challenging and ambitious choice for space power system in contrast to solar and chemical fuel. It is able to realize high power and long operating time simultaneously to meet the need of potential applications. Aiming at the thermodynamic performances of the regenerative Brayton cycle with two-stage compression, the paper is objective to get a set of reasonable and competitive operating parameters for the design of the space nuclear power system. Thermodynamic process calculation is applied to analyze the relations of cycle efficiency and influence factors including compression ratio, gas temperature at cold side and hot side, recuperator efficiency, system pressure. The mass estimate model is established to calculate total mass and specific mass of the system with the variation of such design parameters. The calculating results using MATLAB code show that the optimal compression ratio of single compressor varies between 1.2 and 2 along with the other parameters. Either decreasing the cold side temperature or increasing the hot side temperature contributes to enhance the cycle efficiency to about 50%. When the recuperator efficiency changes from 60% to 98%, an ideal heat exchange efficiency, the efficiency corresponding to the optimal compression ratio increase from 35.8% to 52%. But the total mass will also rise from 9.1 tons to 29 tons. It is concluded that the system with cold side and hot side temperature of 450 K and 1300 K, recuperation efficiency of 80% is capable to obtain the maximum cycle efficiency of 36% and the system mass of 10.2 tons. Supposing a space nuclear power system with thermal power of 5 MW, the specific mass is only 5.8 kg/kWe, which indicates obvious technical and economic advantages.
Numerical Simulation of Subcooled Flow Boiling Using a Bubble Tracking Method
Authors: Okawa Tomio , Miyano Naoki , Kaiho Kazuhiro , Enoki Koji
Published: October 2016

The process of bubble nucleation in subcooled flow boiling was visualized using a high speed camera to show that the bubble size can be significantly different between the nucleation sites. However, the bubble size is usually assumed constant in the numerical simulation of subcooled flow boiling. To explore the effect of the bubble size distribution on the void fraction in subcooled flow boiling, numerical simulations were performed using a bubble tracking method in which the size and position of each bubble are calculated individually using a Lagrangian coordinates. In the present simulation, the void fraction was greater when the bubble size distribution was taken into consideration. Since the bubble tracking method requires many correlations, further improvement is necessary. The present numerical results however indicate that the bubble size distribution should be taken in to consideration to evaluate the void fraction in subcooled flow boiling accurately.
Experimental Study of Helix-Finned Surface for Nucleate Pool Boiling
Authors: Chen Kailun , Yan Changqi , Kurwitz Cable , Cheng Kun , Bian Haozhi
Published: October 2016

This research presents results from experimental investigations on helix-finned surface fabricated by a 3D printing technique to evaluate boiling heat transfer performance. The experiments were conducted in saturated water at atmospheric pressure. To the author’s knowledge, this is the first attempt that helical pin fins are employed in thermal management. The boiling curve of the enhanced surface was characterized by a much lower wall superheat at the same heat flux compared with plain surfaces. High-speed visualization was used to display instantaneous bubble behaviors such as the bubble departure frequency, which was obtained from analyzed images. It was observed that the helix-finned surface had higher bubble departure frequencies compared to plain surfaces and an earlier onset of nucleate boiling was noticed. It is concluded that the surface roughness and micron level cavities produced by the 3D printing technique on the helix surface are key factors to enhance boiling heat transfer. With the experience gained, dimension optimization of helical structure should be studied further to meet the needs of increased heat dissipation rate.
Study on the Break Accidents of the HTR-PM Primary Loop
Authors: Lang Minggang , Sun Ximing , Zheng Yanhua
Published: October 2016

In thermal hydraulics designing and safety analysis of the High Temperature gas-cooled Reactor-Pebble Bed (HTR-PM), the THERMIX code was used to study the behavior of helium in the primary coolant system. Once the helium leaks out of the primary loop through a break on the pressure boundary or an inadvertent open relief valve, it is difficult to simulate the conditions of the room where the release occurred with THERMIX. In this paper, the latest version of RELAP5/MOD4 was used to simulate the behavior of the helium released to the containment rooms. A RELAP5/MOD4 input deck of the HTR-PM, consisting of the core, the primary coolant system, the secondary loop and the containment, was developed and evaluated in this paper. Based on the model, this paper simulated the accidents consequences of large breaks or small breaks near the inlet or the outlet of the helium circulator located inside the steam generator pressure vessel. The calculating results illustrate that the temperature of the helium flowing into the reactor building through the break was no more than 280°C even after an un-isolating large break. The analysis shows that the systems function to scram the reactor and to monitor the core temperature and pressure after accidents would not be affected by breaks.
Numerical Simulation of Different Sizes Missing Pellet Surface Effects on Thermal-Mechanical Behaviors in Nuclear Fuel Rods
Authors: Tang Changbing , Jiao Yongjun , Li Wenjie , Qing Tao , Miao Yifei , Chen Ping
Published: October 2016

Nuclear fuel rods is mainly composed of uranium dioxide pellets and zirconium alloy cladding, there is a gap between pellets and cladding, which is filled with helium. Under the reactor operation conditions, pellets produce a lot of heat by nuclear fission reactions and at the same time also produce lots of radioactive fission products. Cladding serve as the first barrier to accommodate radioactive fission product, needs to maintain its structural integrity under the reactor operation conditions. Cladding stresses can be effectively limited by controlling power increase rates. However, pellet manufacturing defects such as missing pellet surface (MPS), may lead to cladding local stress significantly high to cause cladding failure. Simulating the impact of these defects correctly can help prevent these types of failure. MPS defects are 3D phenomenon, needs 3D modeling method to study the influence of these defects on the cladding .In this paper, stress update algorithm is derived, with the help of ABAQUS (a commercial finite element software), simulated the thermal-mechanical behaviors of the MPS defects fuel rod with a 3D FEM and completed the sensitivity analysis of MPS defects size for the fuel performance. The models included in this simulation, including pellet irradiation swelling (fission gas products induced swelling and fission solid products induced swelling), pellet densification, pellet relocation, pellet thermal expansion, pellet irradiation creep, pellet irradiation hardening, cladding irradiation growth, cladding thermal expansion, cladding thermal creep, cladding irradiation creep, cladding irradiation hardening and gap heat transfer (gas heat conduction, radiation heat transfer and contact heat conduction) etc. Furthermore, considering the effects of irradiation and temperature on the material parameters such as thermal conductivity, specific heat and young’s modulus etc. According to the simulation result, showing that MPS defects have a large impact on the performance of fuel rods, this impact will be more obvious with the size of MPS defects increase. The MPS defects cause larger gap distance between pellet and cladding, higher gap distance causes smaller gap conductance, and then causes elevated temperature at the center of the pellet and in the region of the pellet adjacent to the defect. The cladding temperature is reduced in the area immediately across from the defect, and is elevated in neighboring areas. Meanwhile, MPS defects clearly have a significant effect on stress distribution and maximum stress of the cladding, cause high tensile stresses in the inner surface of the cladding and high compressive stresses on the outer surface of the cladding at the center of the defect. Around the boundaries of the defect, the stresses are reversed, with high compressive stresses on the cladding interior and high tensile stresses on the cladding exterior.
Oxygen Used as an Oxidizer in Acid In-Situ Leach Uranium: From Theory to Practice
Authors: Liao Wensheng , Tan Yahui , Wang Limin , Li Jianhua
Published: October 2016

A typical sandstone uranium deposit, located in the Tuhar basin, was selected to compare the effect of oxygen as the oxidizer with that of hydrogen peroxide. Based on the feasibility study of oxygenation of ferrous and uranium minerals, batch leaching, pressure column leaching and field testing were carried through. The results of feasibility study and laboratory leaching indicate that ferrous ion is inaccessible to being oxidized by pressure oxygen in acidic solutions with pH 2–2.5, and oxygen can oxidize the uranium minerals. Recovery of uranium is proportional to the oxygen pressure. Additionally, the low concentrations of aluminium and ferric ion alleviate the potential precipitation of aluminum and iron significantly. The further field test confirmed the feasibility of oxygen in acid leach. Oxygen has some extent effects of increasing uranium level and considerable effects of anti-precipitation and clogging. In general, oxygen has better applicability in this deposit.
Application of High-Speed OH-PLIF Technique for Improvement of Lean Hydrogen-Air Combustion Modeling
Authors: Katzy Peter , Boeck Lorenz R. , Hasslberger Josef , Sattelmayer Thomas
Published: October 2016

The presented work aims to improve CFD explosion modeling for lean hydrogen-air mixtures on under-resolved grids. Validation data is obtained from an entirely closed laboratory scale explosion channel (GraVent facility). Investigated hydrogen-air concentrations range from 6 to 19 vol.-%. Initial conditions are p = 1 atm and T = 293 K. Two highly time-resolved optical measurement techniques are applied simultaneously: (1) 10 kHz shadowgraphy captures line-of-sight integrated macroscopic flame propagation; and (2) 20 kHz OH-PLIF (planar laser-induced fluorescence of the OH radical) resolves microscopic flame topology without line-of-sight integration. This paper presents the experiment, measurement techniques, data evaluation methods and initial results. The evaluation methods encompass the determination of flame tip velocity over distance and a detailed time-resolved quantification of flame topology based on OH-PLIF images. One parameter is the length of wrinkled flame fronts in the OH-PLIF plane obtained through automated post-processing. It reveals the expected enlargement of flame surface area by instabilities on microscopic level. A strong effect of mixture composition is observed.
Improvement of FAC Maintenance Program Issued From BRT-CICERO™ via CFD Calculations
Authors: Gipon Elodie
Published: October 2016

Flow Accelerated Corrosion (FAC) is very effective for nuclear power plant. This generalized corrosion can lead to the rupture of pipe and in some dramatic cases to casualties. During the last 20 years Electricité de France (EDF) has developed software called BRT-CICERO™ for the surveillance of the carbon steel piping system of its Nuclear Power Plants (NPPs). This software enables the operator to calculate the FAC wear rates by taking into account all the influencing parameters such as pipe isometrics, alloy content, chemical conditioning, design and operating parameters of the steam water circuit (temperature, pressure, etc…). This is a major tool to help operators organize their maintenance and inspections plan.

The algorithms implemented in BRT-CICERO™ are based on tests conducted by EDF R&D, empirical results (national and international feedback), literature reviews and on permanent adjustments based on the operating feedback, via statistical studies. However, for some piping components, from the turbine’s hall, flow dynamics are not optimized and calculated FAC kinetics may be too conservative.

EDF is committed for optimizing and increasing reliability of its maintenance programs to prevent the risk of pipe rupture due to FAC. As in consequence EDF is leading continuous improvement in parameters and calculation algorithms for BRT-CICERO™. Furthermore studies on the geometric characteristics of the pipes were conducted. In BRT-CICERO™ geometric effect of a pipe component (elbow reduction, tees …) is taken into account by considering a factor called “Geo” in the calculation to tune the thickness loss rate according the component type, its characteristics and specific effect on flow mass transfer.

EDF implements finite element analysis software to compute the mass transfer coefficient k and so ascertain the “Geo” coefficient. These computed “Geo” coefficients are compared to those used in BRT-CICERO™. If necessary, current “Geo” coefficients used in BRT-CICERO™ will be adjusted and optimized to improve maintenance programs issued from the software.

The presentation deals with the calculation method used for these studies and some results will be shown on tube and elbows.
The Method of Specific Heat Capacity Measurement for Irradiated Zr-4 Alloy Capsule Material
Authors: Chen Liang , Pang Hua , Song Xiaoming , Liu Lili , Li Wenjie , Tang Changbing
Published: October 2016

Specific heat capacity of irradiated and un-irradiated Zr-4 alloy in the temperature range from 40°C to 500°C was measured by DSC (Differential Scanning Calorimetry) in Hot Cell of NPIC for the first time in China. The irradiated discal specimen was sampled from Zr-4 alloy plate, and the dose was about 18μSv/h. The measurement was proceeded on DSC device (404 F1 type), and the data of specific heat capacity was obtained by relative method made use of Proteus analytical software. Discharge air exhausted to hot cell by vent in DSC. The test result indicated that the test method was confirmed feasible and appropriate, then provided an effective approach for thermal parameter measurement for irradiated specimens.
Parameter Optimization for AP1000 Steam Generator Feedwater Control System Using Particle Swarm Optimization Algorithm
Authors: Wu Shifa , Wang Pengfei , Wan Jiashuang , Wei Xinyu , Zhao Fuyu
Published: October 2016

The U-tube Steam Generator (UTSG) of AP1000 Nuclear Power Plant (NPP) is the crucial component transferring heat from the primary loop to the secondary loop to make steam. The UTSG of AP1000 NPP is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, it is difficult and challenging to well control the water level of AP1000 UTSG by tuning the PID controller parameter in a traditional way, especially when the system is undergoing a sharp transient. To achieve better control performance, the Particle Swarm Optimization (PSO) algorithm was applied for the parameter optimization of the AP1000 UTSG feedwater control system in this study. First, the mathematical model of AP1000 UTSG was established and the objective function was developed with the system constraints considered. Second, the simulation platform was built and then the simulation was conducted in MATLAB/Simulink environment. Finally, the optimized parameters were obtained and the feedwater control system with optimized parameters was simulated against that without optimized. The simulation results demonstrate that optimized parameters of AP1000 UTSG feedwater control system can significantly improve the water level control performance with smaller overshoot and faster response. Therefore, the PSO based optimization method can be applied to optimizing AP1000 UTSG feedwater control system parameters to provide much better control capabilities.
Investigation of Bypass Valve Positions on Performance of Closed Brayton Cycle
Authors: Li Xiao , Yang Xiaoyong , Zhang Youjie
Published: October 2016

HTR-10GT is a closed Brayton cycle with two-stage compression and heat recuperation. Bypass control was adopted for rapid power regulation and safety protection. A bypass valve could be set between any two positions at different pressure levels, there would be in total 21 setting possibilities, whose regulating behaviors remained to be fully clarified. The dynamic characteristics of these settings were thus analyzed by implementing numerical simulation on the integrated system model. The reactor was modeled with point-kinetics and 1D thermal-hydraulics model; compressor performance calculation was based on digitization of the performance map and Reynolds number correlation; Flügel formula was chosen to calculate turbine performance. The system was modeled with Modelica, and the DASSL code was used to solve the Differential and Algebraic Equations. The results showed that only 6 choices among the 21 remained acceptable for engineering practice. The degree and rate on power output reduction, anti-surge effect and long-term thermal effects at reactor inlet and along recuperator’s metal wall were evaluated and compared among these choices to give out proposals. This research clarified the characteristics of bypass valves at different positions to give a reference on the final design of the control methods, and proposed a combination of valves for multi-valve cooperative regulation.
Trend Analysis of Loss of Offsite Power Events
Authors: Volkanovski Andrija , Ballesteros Avila Antonio , Peinador Veira Miguel
Published: October 2016

This paper presents the results of the trend analysis of Loss of Offsite Power (LOOP) events registered in two reviewed databases. The reviewed databases include the Nuclear Regulatory Commission (NRC) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS). Both databases were screened for the relevant events registered in period 1990 to 2012.

The statistical analysis of the identified relevant LOOP events is done. The analysis includes assessment of the LOOP initiating event frequency, distribution of the events per year in the analysed period and trend analysis of the identified events. The LOOP frequency is calculated for LOOP events registered in NRC LERs subdivided into four types by cause or location: plant centered, switchyard centered, grid related, and weather related. These four LOOP categories are assessed for two modes of operation (critical and shutdown operation).

The number of LOOP events in each year over the analysed period and distribution of events per unit in given year were assessed from the reviewed databases.

Trend analysis of the identified events is performed with the utilization of four trend measures. Analysis is done for events registered during power and shutdown operation and their sum. The obtained LOOP frequency for events registered NRC LERs for critical and shutdown mode is comparable to the one assessed in NUREG/CR-6890.

Decreasing trend is obtained for the LOOP events registered for events in NRC LER database. Different trends are identified during critical and shutdown modes of operation for the events registered in the IAEA database. The sum of the LOOP events reported during critical and shutdown modes in IAEA IRS show no trend.
Research on the Steam-Gas Pressurizer With Non-Condensable Gas
Authors: Ma Xizhen , Jia Haijun , Liu Yang , Wu Lei
Published: October 2016

The thermal-hydraulic characteristics of a steam-gas pressurizer are considered as one of the most important factors which affect the safe operation of the small and medium reactor. Since the steam and non-condensable gas are used to sustain the pressure in the steam-gas pressurizer, the pressurizer would experience very complicated thermal-hydraulic phenomena with the non-condensable gas. A steam-gas pressurizer model was developed to investigate thermal-hydraulic characteristics of a steam-gas pressurizer. The governing equations in the model include the phasic continuity equations, momentum conservation equations and energy conservation equations. The pressurizer model was verified with the results from the insurge experiments performed at Massachusetts Institute of Technology. In addition, a wall condensation model with non-condensable gas was introduced to consider the effect of non-condensable gas on the heat and mass transfer during the steam condensation. The calculations results which were performed in the Relap5 code had a great agreement with the experiment results. It’s indicated that the non-condensable gas played a key role in the pressure transient process in the steam-gas pressurizer and it’s necessary to consider the effect of non-condensable gas in the steam-gas pressurizer model. With different kinds and mass fraction of non-condensable gas, the pressure transient in the steam-gas pressurizer would be significant different. The accuracy of model would be also different with different mass fraction and kinds of non-condensable gas.
A Proposed Model to Describe the Relationship Between Online Burnup Assay and Economy and Safety of Pebble Bed Reactor
Authors: Tang Yushi , Zhang Liguo , Guo Quiju , Cao Jianzhu , Tong Jiejuan
Published: October 2016

Online burnup measurement is a unique feature for pebble bed gas-cooled reactor and the fuel balls undergo a multi-circulation on the basis of the online burnup assay. It is ascertained that the accuracy of the online burnup assay is related with the economy and safety of pebble bed reactor. In the economical perspective, the burnup assay accuracy allow some part of pebbles that are below the burnup limit in the orifice to be discharged out of the core. In the safety view, the burnup assay allow some part of pebbles in the reactor core to exceed the burnup limit. In this paper, a mathematical model is proposed to establish the relationship. The model is implemented based on some reasonable theoretical hypothesis, and the influence of assay accuracy on the reactor safety and fuel cost issues are discussed based on the simulated results given by different assay accuracy. It is ascertained that improvements on burnup assay accuracy could save the fuel cost and improve the PBR economical efficiency as well as reduce the probability of radioactive release due to over-irradiation and enhance the safe reliability of PBR. Further research on the burnup distribution of pebbles in and out of the core and the burnup assay model are expected to provide some implications on proposing reasonable requirements for accuracy of online burnup assay.
Coupled Source Term Estimation Using Both Forward and Backward Method for Nuclear Power Plant Accident
Authors: Sun Sida , Fang Sheng , Liu Yun , Li Hong
Published: October 2016

The source term information of radioactive release in a nuclear accident is important for nuclear accident classification, radiological consequences evaluation and emergency response. Two major categories of source term estimation techniques are forward method based on the status data of the nuclear reactor and backward method based on environmental monitoring data. Although the forward method is more widely used, it may introduce large uncertainties into the source term estimate due to its subjective parameters and low quality of reactor status data in a severe accident. To reduce these uncertainties, a coupled source term estimation method that combines both forward and backward models, is proposed in this study. The forward part provides a source term forecast model based on Response Technical Manual RTM-96. Under the framework of sequential data assimilation, the backward part iteratively reduces the uncertainties in the source term estimate using environmental monitoring data. Numerical experiments with different uncertainties are performed using the reactor and monitoring point information of real Chinese nuclear power plant and its atmospheric tracer experiments.
Stand-Alone Containment Analysis of the PHÉBUS FPT-1 Test With the ASTEC and the MELCOR Codes
Authors: Gonfiotti Bruno , Paci Sandro
Published: October 2016

The estimation of Fission Products (FPs) release from the containment system of a nuclear plant to the external environment during a Severe Accident (SA) is a quite complex task. In the last 30–40 years several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments, and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state-of-the-art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, a continuous verification and validation work should be carried out.

Therefore, the aim of the present work is to re-analyze the Phébus FPT-1 test employing the ASTEC (F) and MELCOR (USA) codes. The analysis focuses on the stand-alone containment aspects of the test, and three different modellisations of the containment vessel have been developed showing that at least 15/20 Control Volumes (CVs) are necessary for the spatial schematization to correctly predict thermal-hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results, and presents different sensitivity analyses carried out on the aerosols and FPs behavior.
Probabilistic Prediction of Fatigue Life of Pipes Under Turbulent Fluid Mixing
Authors: Costa Garrido Oriol , Cizelj Leon
Published: October 2016

This paper presents a recently proposed analysis of the uncertainties involved in the thermal fatigue assessment of pipes under turbulent fluid mixing. The assumed overriding contributor to the uncertainty of the predicted fatigue life is the variability inherent in the temperature histories of the turbulently mixing fluids, which intrinsically depends on the length of temperature histories considered in the assessment.

The fluid temperature histories are generated with an improved spectral loading approach within the temperature variation space, which includes all physically possible temperature fluctuations and all flow regimes. The results of the presented analysis are compared with the typical sinusoidal method, emphasizing the well-known over-conservatism of this method.

This paper extends the analyses to the initial thermo-mechanical transient of the pipe wall around the mixing fluids and its influence on fatigue life predictions. The high stresses arising at the beginning of the transient yield, generally, conservative life time results. In a separate sensitivity analysis, the impact of the heat transfer coefficient on fatigue predictions is also studied. This is the first attempt to correlate the uncertainties in the physical fluid mixing conditions and heat transfer to the estimated fatigue life.
Genetic Algorithm Optimization of a Dynamic Uncertain Causality Graph (DUCG) for Fault Diagnosis in Nuclear Power Plants
Authors: Zhao Yue , Di Maio Francesco , Zio Enrico , Zhang Qin , Dong Chunling
Published: October 2016

Fault diagnostics is important for the safe operation of Nuclear Power Plants (NPPs). In recent years, data-driven approaches like neural networks, fuzzy and neuro-fuzzy approaches, support vector machine, K-nearest neighbors classifiers and inference methodologies, have been proposed and implemented to tackle the problem. Among these methodologies, Dynamic Uncertain Causality Graph (DUCG) has been proved effective in many practical cases. However, the causal graph construction behind the DUCG is complicated and, in many cases, redundant on the symptoms needed to correctly classify the fault. In this paper, we propose a method to simplify causal graph construction in an automatic way. The method consists in transforming the expert knowledge-based DCUG into a Fuzzy Decision Tree (FDT) by extracting from the DUCG a Fuzzy Rule Base (FRB). Genetic algorithm (GA) is, then, used for the optimization of the FDT, by performing a wrapper search around the FDT: the set of symptoms selected during the iterative search are taken as the best set of symptoms for the diagnosis of the faults that can occur in the system. The effectiveness of the approach is shown with respect to a DUCG model initially built to diagnose 23 faults originally using 262 symptoms of Unit-1 in the Ningde NPP of the China Guangdong Nuclear Power Corporation (CGNPC). Final results show that the FDT whose symptoms and diagnosis strategy has been optimized by GA, can drive the construction of DUCG and lower the computational burden without loss of accuracy in diagnosis.