bluuebunny/crossref_metadata_2025_split huggingface dataset:
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: (13) Rising Bubble Motion Under Horizontal Vibration
Authors: Arai Rie , Kaneko Akiko , Monji Hideaki , Abe Yutaka , Yoshida Hiroyuki , Takase Kazuyuki 
Published: November 2014

An earthquake is one of the most serious phenomena for the safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors ware contracted with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void fraction is an important factor for the safety operation of the nuclear reactor. In the previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically, to investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In the case of the earthquake, the fluctuation is not only the flow rate, but also a body force on the two-phase flow and a shear force through a pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake.
Therefore, in this research project, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in the series of study. In this study, to investigate the effects of vibration on bubbly flow in the components and construct an experimental database for validation, we performed visualization experiments of vertical bubbly flow in a rectangular water tank on which a sine wave vibration was applied. In this paper, results of visualized experiment evaluated by the visualization techniques, including positions of bubbles, shapes of bubbles and liquid velocity distributions around bubbles, were shown. And liquid velocity distribution around bubbles by the PIV measurement was also shown. In the results, bubble behaviors were affected by oscillation. And the cycle of the bubble tilt angle was almost same as the cycle of oscillation table velocity.
Analytical Evaluation of Debris Cooling and Spreading Behaviors at Molten Core in Severe Accident
Authors: Kobayashi Akihiro , Miwa Shuichiro , Mori Michitsugu 
Published: November 2014

On March 11, 2011, severe accident occurred at Fukushima Daiichi Nuclear Power Plant, and Units 1 to 3 of the plant have led to core melt. That is to say, melted fuel rods and core internals fell to the bottom of the Reactor Pressure Vessel (RPV). It is also believed that molten core has leaked into the reactor containment vessel. In order to plan for a safe molten core removal from the reactor, it is important to estimate the conditions of molten core by conducting analysis. Particular importance of the analysis is to understand the mechanisms of molten core spreading-cooling processes. However, sufficient understanding of this process has not been obtained yet.
The main purpose of this study is to evaluate molten metal spreading-cooling phenomena and subsequently, estimate the conditions of the molten metal. In order to achieve the purpose, the Computational Fluid Dynamics (CFD) for thermal fluid analysis, STAR-CCM+ was utilized. In the simulation of the unsteady two-phase flow, the volume of fluid model was applied for the spreading and interfacial surface formation of molten metal with the surrounding air. The key parameter for the molten metal spreading is the temperature dependent viscosity of molten metal. To assess the validity of this model, the analysis of the VULCANO VE-U7, molten metal spreading experiment, has been compared with simulation results.
Analysis of Bypass Effect on a 250 MW High Temperature Gas-Cooled Reactor
Authors: Zheng Yanhua , Chen Fubing , Shi Lei 
Published: November 2014

Pebble bed modular high temperature gas-cooled reactors (HTR), due to their characteristics of low power density, slender structure, large thermal inertia of fuel elements and reactor component materials (graphite), have good inherent safety features. However, the reflectors consisting of large piles of graphite blocks will form huge numbers of certain bypass gaps in the radial, axial and circumferential directions, thus affecting the effective cooling flow into the reactor core, which is one of the concerned issues of HTRs.
According to the preliminary design of the Chinese high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the thermal-hydraulic calculation model is established in this paper. Based on this model, considering different bypass flow, that is to say, different core cooling flow, fuel element temperature, outlet helium temperature and the core pressure drop in the normal operation, as well as the maximal fuel temperature during the depressurized loss of forced cooling (DLOFC) accident are analyzed. This study on bypass effects on the steady-state and transient phases can further demonstrate the HTR safety features.
A New Paradigm: Cycling Operations at Nuclear Power Plants in the United States
Authors: Hilleman Douglas , Kumar Nikhil , Lefton Steven 
Published: November 2014

Nuclear power plants are no longer immune to cycling operation. While certain nuclear power plants in Europe have been performing load following operation, this type of operation has largely been avoided in the United States. Due to increasing contribution of nuclear generation in the mix, European operators were forced to make modifications to increase the maneuverability of their nuclear generation assets. However, in the United States, nuclear generation is still a relatively smaller contributor (19%). Still, with rapid increase in renewable generation, some nuclear plants are being asked to operate at reduced power and cycle to lower power levels. With most future renewable integration studies advocating for increased flexibility on the grid, nuclear generation maneuverability will allow system operators with another resource to mitigate system costs.
This paper presents the results of a detailed study of a 1,150 MW boiling water reactor nuclear plant when cycled to low loads. The authors present the relative damage of cycling to various reduced power levels 80% to 15% power levels compared to a cold startup and shutdown of a nuclear plant. An assessment was made of the systems that had fatigue damage and costs. We also discuss some of the limitations of cycling that a nuclear plant has and present and discuss recommendations to reduce damage and costs.
Adsorption Properties of Sr(II) on Zeolite Type Adsorbents and Their Irradiation Stabilities
Authors: Mimura Hitoshi , Susa Shunsuke , Ito Yoshiyuki , Saito Yasuo , Matsukura Minoru 
Published: November 2014

Radioactive waste treatment is planned in LWTF (Low-level radioactive Waste Treatment Facility, JAEA) for LLW generated from the Tokai-reprocessing facility. The target LLW consists of highly concentrated sodium nitrate (5 M NaNO3) containing low-level 90Sr. In this study, selective adsorption properties of Sr2+ for highly functional A type zeolites (A51-JHP, A51-J (Union Showa) and A-4, X type zeolite (F-9) and Titanic acid-PAN (polyacrylamide) were clarified by batch and column adsorption methods. The irradiation stabilities of these adsorbents were also evaluated.
The distribution properties of Sr2+ on different adsorbents were compared in simulated waste solution (5 M NaNO3, 0.1 ppm Sr2+, 85Sr as tracer). The order of distribution coefficients (Kd,Sr) was Titanic acid-PAN > A51-JHP > A51-J > A-4 > F-9. The largest value of Kd,Sr for titanic acid-PAN was estimated to be 218 cm3/g, while the saturated capacity (Qmax) was very small. Titanic acid-PAN had also the largest uptake rate of Sr2+ ions and the uptake attained equilibrium within 8 h. On the other hand, A51-JHP had a relatively large Kd,Sr value above 100 cm3/g and a Qmax value of 0.65 mmol/g.
The breakthrough properties of Sr2+ were examined by varying cations present (single and mixed solutions) and flow rate (0.08 and 0.17 cm3/min). The components for the single solution were 400 g/L NaNO3, 100 ppm Sr2+, 85Sr as tracer, and the mixed solution contains 200 ppm Cs+, 100 ppm Ca2+, 50 ppm Mg2+, 50 ppm RuNO3+ in addition to the single solution components. The breakthrough curve for Titanic acid-PAN column using single solution had an S-shaped profile, while the “concentration phenomenon” exceeding C/C0 (breakthrough ratio) = 1 was observed in the case of mixed solution. As for the A51-JHP column, the breakthrough curve for single solution was similar to that for mixed solution and the 5% breakpoint was enhanced by decreasing the flow rate.
The A51-JHP was stable under 60Co-irradiation up to 2.54 MGy; Kd,Sr and Qmax values were almost constant. In contrast, Titanic acid-PAN was affected above 0.28 MGy, due to the radiolysis of PAN matrix, and this surface alteration led to the release of active component of titanic acid.
The novel A type zeolite (A51-JHP) is thus expected for the selective removal of Sr2+ in LWTF. The optimization of particle size and flow rate should be examined before practical use.
Analysis of Some Accidents With Deborated Slug Passing Through the WWER Core
Authors: Ponomarenko Grigory L. 
Published: November 2014

Deborated slug may be formed by the different ways. Stagnant areas may be arise in circulation loops when Reactor Coolant Pumps (RCP) do not work and decay heat removing system works. First RCP start-up in the loop with formed deborated slug is the most dangerous situation. Accelerated slug enters into the core and quickly (during 1 s) may bring reactor from deep subcriticality into above-critical state and generate dangerous neutron splash. Discharged energy may damage or distract fuel rods. However in reality, at the core enter there is not a distillate with zero concentration of boric acid (CB=0 g/kg). There is a coolant with lowered 0 < CB < 16 as compared to shut-down value (16 g/kg), as a result of coolant mixing and washing-out of deborated slug.
Analysis of this beyond design basis accident was performed with use of certified system coupled code KORSAR/GP [1–3] with spatial neutron kinetics and account of incomplete coolant mixing in reactor.
Mixing process was modeled by KORSAR/GP on the base of experimental results obtained on the experimental facility scaled WWER-1000 as 1:5. As a result of the carried out work the technique of the analysis is developed. Essential features of mode with deborated slug are revealed. Necessary provisions are determined, and recommendations for safety maintenance and improve are formulated. Safe volumes of deborated slugs and the time intervals of inadvertent deboration were determined. These not dangerous volumes and corresponding time intervals appeared big enough to provide high probability not to exceed them during start-up of modernized WWER-1000 power units. Influence of increase in CR quantity from 61 up to 121 pcs. on safety parameters in the given mode is evaluated inter alia.
Development of Glass Melter Technology for HLLW Vitrification in Japan
Authors: Sakai Akira , Koikegami Hajime , Miura Nobuyuki , Ochi Eiji 
Published: November 2014

This paper describes the development of glass melter technology, primarily the liquid fed joule-heated ceramic melter process (LFCM) for the vitrificaton of high-level radioactive liquid waste (HLLW) since 1977 in Japan. In 2013 the active test at the vitrification facility (K-facility) in Rokkasho commercial reprocessing plant was successfully completed for the final acceptance test.
During this period many activities on LFCM process development have been carried out in the engineering scale or the full-scale inactive cold tests including the radioactive laboratory scale hot tests. In particular, the design of melter bottom structure and the operating method should be optimized in order to avoid the operational problems caused by accumulation of noble metals (Ru, Rh, Pd), electro-conducive deposits on the melter bottom.
Through the operation of inactive and active test facilities in Tokai, the design basis for the Tokai Vitrification Facility (TVF) has been provided. The hot operation of the TVF was started in 1995 to demonstrate the LFCM process including the performance of the melter off-gas clean-up system etc. The TVF has provided the basis of the process design and the operation method for the K-facility melter in Rokkasho.
In case of commercial scale vitrification, the glass production rate of the melter should be several times larger than that of the TVF. The K-facility full-scale inactive mock-up melter (KMOC) has been planned to confirm the influence of scale-up factors and the difference between Tokai and Rokkasho wastes. Through the testing operation of the KMOC, which was initially started in 2000, it has been found that the stable formation of a cold cap on a molten glass surface is fundamentally important to avoid the excessive precipitation of noble metals and the yellow phase formation.
The active test of the K-facility has been proceeding under the same conditions as the KMOC, and was successfully completed in May, 2013.
The advanced glass melter development programs have also commenced from 2009 to ensure a more robust and noble metals are compatible with the LFCM system and also to provide a higher processing rate. The second K-facility full-scale inactive mock-up melter (K2MOC) has been installed in the vitrification technology development facility (X-14) at Rokkasho. Its testing operation has commenced from November, 2013.
Electromagnetic Pump Flow Simulation by 3D MHD Code
Authors: Asada Takatoshi , Hirata Yosuke , Aizawa Rie , Suzuki Tetsu , Fujishima Yasushi 
Published: November 2014

Electromagnetic (EM) pumps, which drive liquid metal by electromagnetic force in Fast Reactors, have been developed [1] [2]. Annular Linear Induction Pump (ALIP) is one of EM pumps. There have been some reports in which developed pressure’s drop and fluctuation were described near the top of the relation curve between developed pressure and flow rate (P-Q curve). These phenomena were reported by Gailitis, Kirillov and Araseki. Araseki simulated this phenomenon by two-dimensional (2D) code and reported that it is characterized by vortexes, which were initiated by azimuthal non-uniformity of sodium flow velocity and/or magnetic field, in the sodium flow.
We calculated the developed pressure of an EM pump by 2D magnetohydrodynamics (MHD) code and designed the EM pump. The code simulated the developed pressure with high accuracy in normal operations. When the flow rate was lower than the one at the top of P-Q curve, the developed pressure’s drop and fluctuation occurred. The fluctuation could disturb the stable operation of the pump. For fear of this phenomenon, the design of EM pump is sometimes too conservative. To simulate this phenomenon and examine its occurring conditions, we have developed a new three-dimensional (3D) MHD code. Clarification of these conditions for the occurrence of this phenomenon will enable to design a new structure or determine its operation conditions reasonably.
This paper describes the simulation results focusing on the occurrence of developed pressure’s drop and fluctuation. We use initial conditions which have azimuthal non-uniformity of sodium velocity to simulate the initiation of vortexes in sodium. The 3D MHD code simulates the developed pressure’s drop and fluctuation by vortexes and provides more clear illustration of their occurrence.
Thermal Performance Monitoring and Optimization in Dukovany NPP
Authors: Pliska Jiri , Machat Zdenek , Veznik Libor , Smisek Jiri 
Published: November 2014

Competition in the electricity market forces producers to achieve — in compliance with safety — efficiency of production as high as possible. This efficiency and heat rate is an important indicator of both the condition of the power plant equipment and the quality of power plant operation. To cope with these challenges, powerful methods are process data reconciliation, statistical data processing of large data sets and process simulation. These functions and methods can be used to obtain useful information about process quality and equipment and sensor health.
The paper discusses practical experience from six years of using a thermal performance monitoring and optimization system in the Dukovany nuclear power plant. The system is integrated into the overall nuclear power plant process information system and data warehouse. The system provides information in near real time.
The major benefit of the system lies in a deep view into equipment behaviour and process which ensures timely detection and identification of functionality degradation of process and equipment or sensor faults. The system also helps to find and use margins of equipment operation and the overall thermal cycle.
Selected practical examples are used to demonstrate specific benefits of the system for operation and maintenance of the Dukovany nuclear power plant. There are examples of equipment fault detection and sensor degradation detection. The optimization function is explained with an example of cooling circuit optimization aimed to increase the delivery of electrical power into the grid. A detailed description of behaviour of the main components can be used for their performance evaluation and their repair planning. The benefit of more accurate determination of parameter values is reflected in more accurate determination of reactor thermal output.
The conclusion of the paper provides an overall evaluation of system benefits for operation and maintenance of a nuclear power plant.
The Role of CHECWORKS™ in an Effective FAC Program
Authors: Smith David , Horowitz Jeffrey 
Published: November 2014

Flow-accelerated corrosion (FAC) is a degradation mechanism that attacks carbon steels under conditions often found in nuclear and fossil power plants. FAC has been responsible for a number of significant accidents in nuclear power plants. The most recent noteworthy accident was at the Mihama Unit 3 (Japan) where the catastrophic failure of a pipe in the condensate system resulted in five fatalities.
FAC can affect virtually all of the carbon steel piping and components in the power cycle of nuclear reactors. The presence of wall thinning caused by FAC is determined through the use of non destructive examination (NDE) techniques. For large-bore piping components the most commonly used approach is ultrasonic technique (UT) using an inspection grid applied to the piping components.
As FAC is a reasonably well-understood degradation mechanism, a number of computer programs have been developed to help utility engineers determine the inspection locations and managing the data. CHECWORKS™ — the most commonly used computer program for this purpose — is the program discussed in this paper.
The use of CHECWORKS™ in utilities’ FAC programs will be described. Particular emphasis will be placed on: inspection planning, handling power uprates and other changes to operating conditions.
Inspection planning is one of the most common uses of the CHECWORKS™ software. As there are typically around 5,000 FAC-susceptible components in a reactor system, utility engineers must select the components with the highest FAC-rates for inspection. CHECWORKS™ uses its predictions combined with plant inspection data to provide a best estimate of FAC rates. From these FAC rate predictions and knowledge of the piping schedule and allowable wall thicknesses, inspection locations are determined.
Both the water chemistry used and the local operating conditions strongly influence the rate of FAC. CHECWORKS™ is used to study potential changes to water chemistry, system operation or power level to determine the impact of such changes on FAC rates and hence inspection locations. Of particular interest is the use of CHECWORKS™ to determine the impacts of power uprates. Because of the complicated parametric behavior of FAC rates, changing the power level will likely increase the FAC rates in some areas of the plant while other areas will likely see a decrease in FAC rates. This fact requires a pre-uprate analysis to determine how an inspection program will need to be modified.
This paper provides a description of how CHECWORKS™ is used in the above applications as well as showing typical examples of its usefulness in these analyses.
Water Chemistry Technology of Methanol Addition in PWR Primary Systems: Radiolysis of Methanol Solution at 320 °C
Authors: Domae Masafumi , Hojo Kosho , Sugino Wataru 
Published: November 2014

Primary Water Stress Corrosion Cracking (PWSCC) is one of important ageing issues in PWR (Pressurized Water Reactor) primary systems. It has been pointed out that high concentration dissolved hydrogen may lead to occurrence of PWSCC. The authors have proposed to substitute hydrogen by methanol as a fundamental countermeasure of PWSCC. So far corrosion tests of stainless steels and Zircaloy-4 in methanol solutions at 320 °C were conducted under γ-ray irradiation and without irradiation. The test results show that methanol is promising.
In the present paper, γ-ray irradiation experiments of methanol solution at 320 °C were done up to 100 kGy. A study on the radiolysis of methanol solution is important from two aspects. One concerns corrosion of structural materials. The radiolysis of methanol may result in formation of harmful compounds to the structural materials, such as carboxylic acids. It is necessary to know the yields of such compounds. The other concern is possible polymerization of methanol and formation of organic polymer deposit on fuel claddings. Large amount of the deposit on fuel claddings should be avoided to keep integrity of fuel claddings. Therefore, it should be clarified whether gaseous species are major products and whether polymerized species of methanol such as ethylene glycol is formed.
After the γ-ray irradiation of methanol solution, following species were analyzed: CO2 and H2, methanol, formaldehyde, formate and acetate, and ethylene glycol and glycerin.
Without γ-ray irradiation, the major process of the thermal decomposition of methanol at 320 °C is oxidation of methanol by water and generation of one CO2 molecule and three H2 molecules. Under γ-ray irradiation, the decomposition of methanol is accelerated; little methanol remains after 10 kGy irradiation. The major product is CO2, and polymerization of methanol unlikely occurs. After methanol is completely decomposed, the hydrogen yield still increases. The reducing environment is maintained. Probably, transient organic species play important roles. The addition of low concentration methanol may be sufficient to maintain reducing environment of the PWR primary systems.
Genome Sequence of Corynebacterium pseudotuberculosis MB20 bv. equi Isolated from a Pectoral Abscess of an Oldenburg Horse in California
Authors: Baraúna Rafael A. , Guimarães Luís C. , Veras Adonney A. O. , de Sá Pablo H. C. G. , Graças Diego A. , Pinheiro Kenny C. , Silva Andreia S. S. , Folador Edson L. , Benevides Leandro J. , Viana Marcus V. C. , Carneiro Adriana R. , Schneider Maria P. C. , Spier Sharon J. , Edman Judy M. , Ramos Rommel T. J. , Azevedo Vasco , Silva Artur 
Published: November 2014

ABSTRACT

The genome of
Corynebacterium pseudotuberculosis
MB20 bv. equi was sequenced using the Ion Personal Genome Machine (PGM) platform, and showed a size of 2,363,089 bp, with 2,365 coding sequences and a GC content of 52.1%. These results will serve as a basis for further studies on the pathogenicity of
C. pseudotuberculosis
bv. equi.
Genome Sequence of “Thalassospira australica” NP3b2
T
Isolated from St. Kilda Beach, Tasman Sea
Authors: López-Pérez Mario , Rodriguez-Valera Francisco , Webb Hayden K. , Crawford Russell J. , Ivanova Elena P. 
Published: November 2014

ABSTRACT

Here, we present the draft genome of “
Thalassospira australica
” NP3b2
T
, a potential poly(ethylene terephthalate) (PET) plastic biodegrader. This genomic information will enhance information on the genetic basis of metabolic pathways for the degradation of PET plastic.