bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Application of AFAL Methodology in Substantiation of Instrument Calibration Intervals Extension in Nuclear Power Plant
Authors: Zhou Ping , Xu Xiajun , Tu Caiqing 
Published: February 2014

This is an Introduction about the AFAL methodology of study the instrument drift characteristics, and the application of AFAL in substantiation of instrument calibration interval extension in nuclear power plant. AFAL methodology main aspects include collecting historical instrument calibration data of nuclear power plant, calculating statistics values such as: sample number, mean, median, standard deviation, then work out the drift tolerance interval of the instrument. Based on analysis of these statistical calculated values, will understand the instrument drift performance.
This article also discusses technical issues associated with the application of AFAL and how to solve them, such as: grouping instrument, sample sizes, outliers detecting and processing, high-confidence deduction etc. Through the study of the instrument drift characteristics, evaluate the performance of instrument, determine the calibration interval can prolong properly.
The application practices of AFAL methodology show, extending instrument calibration interval can support nuclear power plant to achieve the goal of prolonging the fuel cycle, under the nuclear safety precondition. The nuclear power plant can improve the capacity factor of the unit and economic performance.
Scaling Assessment for the Integral Effect Tests: 1 — Closed Loop Natural Circulation
Authors: Cai Xiaoyu 
Published: February 2014

In this paper a method is described for using NOTRUMP models to corroborate the Hierarchical Two-Tiered Scaling (H2TS) methodology that has been used for design of the APEX and SPES-2 test facilities. These facilities were built for the Westinghouse Electric Company to obtain data on the performance of the passive safety systems of the advanced pressurized water reactors.
Similarity between the prototype system and the scaled test facilities is investigated for the closed loop natural circulation phenomena in the postulated small break loss of coolant accident (SBLOCA). The objective of this analysis is to provide a basis that an integral test for the passive safety system will provide valid experimental data for the high-ranked phenomena that may occur during the hypothetical SBLOCA transients. In this way, the experiment will capture the phenomena that would be expected to occur in the plant, and hence provide data that can be used to validate computer code.
Experimental Validation for Specific Thermal Hydraulic Phenomena of SMART
Authors: Chung Young-Jong , Lee Won Jae , Ha Jaejoo 
Published: February 2014

SMART (System-integrated Modular Advanced ReacTor) is an integral light water reactor developed by KAERI for both electricity generation and various applications. SMART can produce 90MWe of electricity and 40,000 tons/day of sea-water desalination, which is deemed suitable for a city population of 100,000. The SMART project aims at standard design approval. The experimental validation for SMART specific thermal hydraulic phenomena is to generate background information required for SMART licensing.
The objective of the experimental validation is to demonstrate the safety and performance of the SMART-specific system and component designs, and to provide a database for the validation of the design tools, such as a system analysis code. From the experimental results, it is proved that SMART was designed properly and the data are used to approve the validation of the system analysis code. Also, the experimental results were submitted to a regulatory body as technical background information of the SMART standard design, and they enhance the licensing possibility of this design.
New Cooling Scheme of First Wall in HCCB TBM and Related Performance Analysis
Authors: Wang Qijie , Cao Qixiang , Zhao Zhou , Feng Kaiming 
Published: February 2014

In the design and analysis for the ITER TBM, a key issue is the cooling of the First Wall. For safety, the First Wall FW must be designed to withstand the thermal radiation load from the plasma and the nuclear heat generated by nuclear reaction, and it should satisfy related stress criteria. For technology, it also should meet a pressure drop limits and other engineering requirement.
The former cooling scheme design showed some shortcoming in temperature distribution uniformity and pressure drop, in order to improve it, a new cooling scheme has been designed and analyzed.
The paper describes the detail of the new cooling scheme of FW, and a series related analysis was conducted to compare with former design. In the preliminary analysis, the new cooling scheme design showed much advantage to the former design.
The working conditions of the analysis have been set to ITER design condition. The performances and the results obtained thermo-hydraulics and stress, are presented.
A Scaling Analysis of a CANDU-6 Moderator Tank Scaled-Down Test Facility
Authors: Rhee Bo W. , Kim H. T. , Park S. K. , Cha J. E. , Choi H.-L. 
Published: February 2014

Several studies have been performed to derive a set of scaling criteria which were thought to be suitable for reproducing thermal-hydraulic phenomena in a scale-down CANDU moderator tank similar to that in a prototype power plant during a full power steady state condition[1,2]. The major variables of interests are moderator flow circulation and temperature inside the moderator tank during a steady state condition. The key phenomena involved include the inlet jet development and impingement, buoyancy force driven by the moderator temperature difference caused by non-uniform heating, and the viscous friction of the flow across the calandria tube array.
In these studies, the governing equations were initially transformed into dimensionless equations based on the representative characteristic values of the basic design such as the time, tank diameter, inlet fluid velocity, and average temperature rise, and 3 dimensionless numbers, Re, Pr, Ar, were identified as those characterizing the key phenomena of the system. The relevant boundary conditions were then identified in a dimensionless form, and the compatibility of keeping these 3 dimensionless numbers, the volumetric heat source distribution, and the boundary conditions in dimensionless forms the same for both the prototype and scale-down tanks were examined, and some of them that are less important are relaxed so as to find a practically implementable set of constraints. The size of the scaled-down moderator tank and corresponding inlet velocity is then found for the available power supply size. As an example, an analysis was performed for a power supply capacity of 500 kW as compared to 100MW for the prototype.
As a way to confirm the validity of the current work two numerical CFD simulations were carried out with the boundary conditions at the inlet and outlet ports, and on the walls of the solid structures, such as the moderator tank and calandria tubes, which were derived from those of the dimensionless scales to check if the moderator flow and temperature patterns of both the prototype reactor and scaled-down facilities are identical or at least similar. A steady-state solution is first obtained for the candu-6 reactor normal operation. Similar simulation was done for the scaled-down facility and results presented. Comparison results are discussed, and the cause of the potential distortion of the scaling owing to practical limitations and possible solutions is finally discussed.
Safety Culture Promoting Activities After Fukushima Accident
Authors: Kurata Satoshi , Suzuki Tetsuya , Sekine Yoshihiro 
Published: February 2014

Nuclear safety culture has been one of the most important aspects for safe and reliable operations of nuclear power plants worldwide.
Promoting nuclear safety culture has been a very important task for Japanese nuclear operators since 2007, when it was introduced into the Safety Regulations.
CEPCO has positively promoted safety culture. Although TEPCO also promoted safety culture, the Fukushima accident occurred in March 2011. The official investigation reports on the accident pointed out that one cause of accident could be safety culture problems. Therefore we should seriously consider effectiveness of safety culture promoting activities which we worked on.
This study presents issues and appropriate improvements for the safety culture promoting activities reflecting the Fukushima accident and our experiences in this five years.
Research on the Stainless Steel Tubes and Pipes Material Standards of Demonstration Fast Reactor Power Plant
Authors: Li Yaping , Song Guangdong 
Published: February 2014

The main characteristics of the sodium pipe system in Demonstration Fast Reactor Power Plant (DFRPP) are high-temperature, thin-wall and big-caliber, which is different from high-pressure and thick-wall of the pressurized water reactor system, and the system is long-term operate in the environment of liquid metal sodium. How to guarantee the reliability of materials in high temperature are most important in material option.
Engineering design depend on the criterion. Material standards are different in different countries, and corresponding construction codes are different too. Comparing the stainless steel pipe material standers at home and abroad and analyzing the material standards’ difference according to different construction codes, a stainless steel pipe material criterion system is put forward in this paper which is applicable for the DFRPP.
Development of PCCSAP-3D Code for Passive Containment: Models of Noncondensable Gases, Aerosols and Fission Products
Authors: Li Ran , Yu Jiyang 
Published: February 2014

PCCSAC-3D is a code originally developed for AC600 containment thermo-hydraulic analysis. Its validated capabilities include simulating the behaviors of steam-air mixture and liquid water under the unique conditions of an AC600/AP1000 containment after a DBA. The film-tracking model applied gives it the ability to simulate the liquid film both outside and inside the steel containment. Refined with some new models, the new version of the code, named PCCSAP-3D, can cover hydrogen behavior, fission products behavior (in the form of gas and aerosol) and iodine behavior.
In the module of noncondensable gas, diffusion of up to 11 species are taken into consideration. A user-definable recombiner/ignitor model is developed to accommodate different types of hydrogen recombiners and ignitors.
Given the source term as a boundary condition, the fission products model would be able to track up to 64 radio-isotopes after a LOCA. The leakage and spontaneous decay is accounted for all of these nuclides. Besides, the noble gases, gaseous iodine and fission product aerosols are treated separately. There is no removal mechanism of noble gases. Whereas removal mechanisms of radio-aerosols considered include spray, gravitational sedimentation, diffusio-phoresis and thermo-phoresis. A simple model for gaseous iodine comprises organic iodine and elemental iodine, in which the effects of spray and liquid adsorption are treated integrally.
To evaluate the radioactivity consequences of a certain accident, a radioactivity calculation model is brought out to convert the molar concentration or mass concentration of radioactive material into radioactivity concentration.
The new version of PCCSAP-3D code with models aforementioned is preliminarily validated by comparing the simulation results with safety analysis results reported in AP1000 Design Control Document. The accident scenario is set as a design basic LOCA with core melt.
Seismic Analysis of Lead-Bismuth-Cooled Fast Reactor Vessel
Authors: Ma Juan , Huang Mei 
Published: February 2014

This paper is aimed at analyzing the structural responses of a Generation IV heavy-liquid-metal-cooled reactor (lead-bismuth-cooled fast reactor) vessel in the event of earthquake. For a seismic design, the seismic time history response analyses are carried out for both a top support type and a bottom support type. It is found that the bottom support type exhibits the better performance. There is the gap between the reactor vessel and guard vessel being filled with [1] argon. It is indispensable to add bellows at vessel upper end for a bottom support type to achieve the connection and seal between the reactor vessel and guard vessel. This paper is the first attempt to evaluate the effects on structural seismic performance by adopting the equivalent springs to simulate this elastic connection.
For a seismic investigation, there is a key issue that should be focused on, namely: the fluid-structure interaction due to seismic loading. The vessel is filled with a high-density fluid and might lead to severe hydrodynamic pressure significantly in the occurrence of earthquake. It will impair the structural integrity of reactor vessel. In order to study the structure effects of reactor vessel under seismic loading, an appropriate 3-D finite element analysis model has to be set up and the FEM code ANSYS has been implemented.
Study on Operating Criteria of the Supersonic Steam Injector
Authors: Satoh Kyosuke , Abe Yutaka , Kaneko Akiko 
Published: February 2014

Supersonic steam injector (SI) is a passive jet pump with a venturi structure which operates by condensation of the supersonic steam flow in a subcooled water jet. There are advantages that a SI which operates without electric power has high heat-transfer performance and discharges water at high pressure. Therefore the SI is expected to be applied to the safety system which is able to condense steam efficiently and inject water into a core reactor when severe-accident occurs in a nuclear power plant.
According to the previous studies about a SI, there are two conditions about the definition of a SI, one is formation a water jet in the mixing nozzle, the other is stopping drainage from the drain. The 1-dimension analysis model, which the steam condensed completely, at the mixing nozzle, is developed by Narabayashi et al. (1996). Abe et al. (2011) quantified the interfacial behavior of water jet with high resolution observation in the mixing nozzle and evaluated correlation between the interfacial behavior and heat and momentum transfer characteristics. However the operating limit of a SI which is taken account of flow structure at the diffuser is not cleared.
The objective of the present study is to reveal the effects of two-phase flow behavior in the diffuser on the operating limits of the SI. In the present study, static pressure distributions along flow direction were measured in the SI with change about the back pressure. The discharge pressure was measured and compared with the value of discharge pressure predicted by 1-dimention analysis model. On the other hand, the discharged flow in the diffuser was observed with a high speed camera.
According to the measurement of pressure, a significant pressure rise in the diffuser was measured as the back pressure increased. In addition, the value of the discharged pressure just before the SI becomes inoperative by the back pressure rise was defined as the maximum discharge pressure. It is suggested that this maximum discharge pressure is one of the indicators which predicts the operating limit of the SI. From visualization result of the diffuser, it was confirmed that there is propagation of white color. The void fraction in the discharge flow was estimated at about 2% from the velocity of the propagation.
From the experimental results, effect of two-phase flow behavior of discharged water jet on the operating limits of a SI is discussed.
Study on Flushing Phenomena by Microwave Heating
Authors: Yamaki Tatsunori , Kaneko Akiko , Abe Yutaka , Segawa Tomoomi , Kawaguchi Koichi , Yamada Yoshikazu , Suzuki Masahiro , Fujii Kan-ichi 
Published: February 2014

Recently, the number of nuclear power plants has been increased in many countries. In contrast, uranium fuels used in nuclear power plants are exhaustible resources. Therefore, it is required to exploit uranium resources effectively, and reprocessing of spent fuel is indispensable. To use recovered uranium and plutonium as raw material of nuclear fuel, reprocessing solution (uranium and plutonium mixed nitrate solution) of the spent nuclear fuel is converted to uranium and plutonium mixed oxide (MOX) powder. Microwave heating direct denitration method (MH method) is one of such methods to convert nitrate solution to MOX powder. The cylindrical denitration vessel can be expected to realize high-speed and high-capacity processing against traditional shallow vessel. However, flushing and overflow phenomena of solution have been confirmed in cylindrical vessel. Thus, the safety and the optimization of the vessel shape during microwave heating. In the present study, the purpose of this paper is to clarify generation conditions and generation mechanism of flushing phenomena that is not fully understood. In experiment, flushing phenomena was observed and the liquid temperature was measured using microwave heating device. The main parameters are the vessel diameter, initial water level and the magnetron power.
There was tendency of flushing in the case of short vessel diameter and high initial water level when magnetron power was constant. There was also tendency of flushing in the case of large magnetron power when vessel diameter and initial water level was constant. From visualization, it was clarified that generation of singular bubble triggers flushing. If flushing occurs, the liquid is blown up at a burst, and the vessel become almost empty. From temperature measurement results, it was clarified that the liquid had over 10 °C superheat just before flushing. Therefore, it was suggested that the liquid superheat affected flushing significantly. Generation conditions of flushing are different with the vessel diameter, initial water level and the magnetron power because it is considered that these characteristics have influence on the liquid superheat.
Experimental Study on Sodium Column Fire of Sodium-Cooled Fast Reactor
Authors: Peng Kang-wei , Zhang Zhi-gang , Guo Ming , Wang Chao , Sun Shu-bin 
Published: February 2014

In the operation of sodium-cooled fast reactor, accident caused by the leakage and combustion of liquid sodium is common, and it is a pivotal and difficult problem in research, construction and operation of sodium-cooled fast reactor internationally. In actual operation of sodium-cooled fast reactor, liquid sodium in sodium fire accident is difficult to form fog but mainly in columnar flow form due to low operation pressure and thermal insulation material wrapping the pipe, so experimental research about columnar fire is of much more practical significance.
This paper focuses on combustion property on sodium column fire in the sodium-cooled fast reactor. Liquid sodium with high temperature will be poured into the combustion room via pouring high pressure nitrogen into sodium storage tank when solid sodium has been heated to enactment temperature; liquid sodium with high temperature will burn at combustion room with air forming sodium column fire. Initial temperature of sodium jet is about 200°C in experiment and spurt pressure is 0.2MPa, spurt flow is about 0.4–0.6m3 each hour. Temperature fields in combustion tank space and the catch plate in the bottom are measured through dozens of thermocouple distributed in combustion tank. No atomization phenomenon would exist in the pure sodium columnar fire in the space, the atomized fire triggered by splashed sodium from sodium stream striking the admittance plat and generate more than the high temperature of 900°C.
The Investigation on Interactive Products for Propaganda of Nuclear Safety in Public
Authors: Hou Fei 
Published: February 2014

Depend on the development of technology, the interaction products have been gradually applied into public awareness in the part of Nuclear Safety. Therefore how to put the information about Nuclear Safety into entertainment produces become a crucial research direction in related industry. The study is to investigate current interaction products based on collection of related information in order to find valuable results for effective propaganda on Nuclear Safety.
After the nuclear accident happened in Japan, the popular distrust of Nuclear Energy is spreading all over the world. Due to the kind of negative voices from public in China, several Nuclear constructions have been suspended. Consequently public awareness on Nuclear Safety is supposed to gain more attention for developing Nuclear energy.
According to the investigation on current products on Nuclear Safety, the most of exhibited products are unattractive. How to integrate knowledge of Nuclear Safety into interactive technology becomes a key question. The interaction products need to be improved by designing, developing and manufacture. As a result of complex production process, one interaction product needs interdisciplinary cooperation, so that the future products could become attractively, entertainingly, effectively, pertinently from the conclusions of this research.
Generally, this study is aim to collect the information of interaction products on Nuclear Safety in the whole world, then investigate and analyze current status using appropriate research method. Finally valuable results might be presented for improving interaction products on Nuclear Safety from the whole production process.
Preliminary Investigation of Heat-Transfer Correlation for Upward Flow of CO2 at Supercritical Pressure
Authors: Saltanov Eugene , Pioro Igor , Harvel Glenn 
Published: February 2014

The development of Generation IV nuclear reactors is an ongoing worldwide interdisciplinary effort. Canada is involved with the development of the SuperCritical Water-cooled Reactor (SCWR). One of the numerous engineering challenges associated with this development is to ensure intensive and stable heat transfer to SuperCritical Water (SCW) in a core.
It is very important to develop sophisticated theoretical models to improving understanding of physical processes behind forced-convective heat transfer and its deterioration in near-critical region. However, not all turbulent models implemented in Computational Fluid Dynamics (CFD) codes are applicable to heat transfer at supercritical pressures. Additionally these codes should be first tuned on the basis of experimental data and, after that, used in similar calculations. Therefore, there is still a great reliance on 1D heat-transfer correlations for preliminary calculations.
Performing experiments in SCW is extremely expensive due to the high temperatures and pressures involved. Therefore, it is reasonable to study the general properties of SuperCritical Fluids (SCF) by running experiments with modelling fluids; for example, carbon dioxide, which is widely used as a modeling fluid. In view of this, there is a need to compile a large database that will include experimental data on forced-convection heat-transfer to SuperCritical (SC) CO2 within a wide range of operational parameters. To date, a part of this database has been produced. It contains the bulk-fluid and wall temperatures of CO2 flowing upward in a vertical bare tube. CO2 is at supercritical pressure and the bulk-fluid and wall temperatures are below or above the pseudocritical temperature at the inlet of the test-section.
Therefore, the objectives of this paper are:
1) to propose a correlation in a standard non-dimensional form, which will generalize data within ±25%;
2) to compare this correlation with the most recent, as well as previous 1D correlations for SC CO2 and SCW (appropriate scaling is performed).