bluuebunny/crossref_metadata_2025_split huggingface dataset:
MCCI of a Metal and Oxide Melt With Reinforced Siliceous Concrete in MOCKA Experiments
Authors: Foit Jerzy Jan 
Published: November 2014

Within the framework of large-scale MOCKA (KIT, Germany) experiments, a series of experiments have been performed to study the interaction of a simulant oxide (Al2O3, ZrO2, CaO) and metal melt (Fe) in a stratified configuration. To allow for a longer-term interaction, additional heating was provided by alternating additions of thermite and Zr metal to the melt. Since the heat generated by the thermite reaction and the exothermal oxidation reaction of Zr is mainly deposited in the oxide phase, prototypic heating of both melt phases is achieved. This allows the investigation of concrete erosion by metal melt as well as by the oxide which was not possible in all former experiments.
Current tests in the MOCKA (KIT, Germany) program are focused on assessing the influence of concrete reinforcement (rebars) on the cavity erosion behaviour using a simulant oxide-iron melt in a stratified configuration. The experiments are performed in siliceous concrete crucibles with an inner diameter of 25 cm containing 12 wt.% reinforcement. In these experiments, the overall downward erosion by the metal melt was of the same order as the sideward one. In addition, the lateral erosion in the overlaid oxide melt region was about the same as in the metal melt region. The former experiments (BETA, COMET-L) and MOCKA tests on siliceous concrete without reinforcement have produced results with pronounced downward erosion by the metal phase. This pronounced downward erosion of the siliceous concrete without rebars seems to be inherent for melts containing a significant fraction of iron.
Burnup Credit in the Criticality Safety Analysis of Spent Fuel in Storage Systems
Authors: Soloviov Vladyslav 
Published: November 2014

In this paper accounting of spent nuclear fuel (SNF) burnup of RBMK-1000 with actinides and full isotopic composition has been performed.
The following characteristics were analyzed: initial fuel enrichment, burnup fraction, axial burnup profile in the fuel assembly (FA) and fuel weight.
As the results show, in the first 400 hours after stopping the reactor, there is an increase in the effective neutron multiplication factor (keff) due to beta decay of 239Np into 239Pu. Further, from 5 to 50 years, there is a decrease in keff due to beta decay of 241Pu into 241Am. Beyond 50 years there is a slight change in the criticality of the system.
Accounting for nuclear fuel burnup in the justification of nuclear safety of SNF systems will provide an opportunity to increase the volume of loaded fuel and thus significantly reduce technology costs of handling of SNF.
Dedicated Commercial Power Distribution Panels for Nuclear Safety Related Application
Authors: Hsu Yao-Tung 
Published: November 2014

Six nuclear power reactors in Taiwan have been operating over beyond thirty years. They are all operated by Taiwan Power Company (TPC) and expected to have 40-year lifetimes. The limited original suppliers and obsolete components are the challenge to comply with current licensing basis and maintaining a high level reliability. Therefore, the procurement of basic components from the second source is very important to the plant safety and operation.
This paper describes the dedication process applied to commercial power distribution panels in a harsh environment. The safety functions of power distribution panels provided backup power input connection for mobile diesel generators while station blackout (SBO). After Fukushima-Accident, the utility needs to setup diversity power to comply with regulatory requirements in Taiwan. The power distribution panels dedication activity include the function testing, aging, seismic qualifications (SQ), and environmental qualifications (EQ) based on EPRI NP 5652, IEEE Std. 323, and IEEE Std. 344 standards. Some subcomponents could not meet the acceptance criteria during testing and the anomalies were noticed to the customer and the utility. One of these anomalies reported to regulatory due to the subcomponent failure after accident radiation endurance test.
Commercial-Grade Item dedication is second source to obtain safety related components according to 10 CFR 21.3 definitions. In the past nineteen years, Institute of Nuclear Energy Research (INER) has actively performed the dedication service to help local nuclear power plants solve their procurement problems of nuclear grade items, due to reduced availability of qualified suppliers and/or obsolete issues of qualified components. Although the codes and standards for dedication in Taiwan refer to those in USA, the challenges may happen due to different regulators, utility, manufacture’s quality culture, and personal responsibility. This paper introduces the self-reliant experiences in dedication and economic benefit to local nuclear power plants.
A Remote Operated Quadruped Robot System for Investigation of Reactor Building
Authors: Suganuma Naotaka , Uehara Takuya , Matsuzaki Kenji , Ochiai Makoto , Terai Fujio , Tsuyuki Akira , Chida Itaru 
Published: November 2014

A remote operated quadruped robot has been developed for disaster site which can move on stairs, slopes, and uneven floor under the radiation-polluted environment, such as TEPCO Fukushima Daiichi nuclear power plants [1][2].
In particular, the control method for stable walking and the remote operation system have been developed to move on stairs in the reactor building.
We applied this robot to investigation of suspicious water leakage points in reactor building at Fukushima Daiichi nuclear power plants unit2[3]. In this investigation, a small vehicle equipped with camera and a manipulator which is connected the vehicle with cable were mounted on the robot and were carried to near the target by the quadruped robot and the investigation was carried out with the small vehicle.
Dynamic Correcting Dispersion Parameters of Lagrangian Puff Model in Atmospheric Tracer Experiments
Authors: Ma Yuanwei , Wang Dezhong , Ji Zhilong , Qian Nan 
Published: November 2014

In atmospheric dispersion models of nuclear accident, the empirical dispersion coefficients were obtained under certain experiment conditions, which is different from actual conditions. This deviation brought in the great model errors. A better estimation of the radioactive nuclide’s distribution could be done by correcting coefficients with real-time observed value. This reverse problem is nonlinear and sensitive to initial value. Genetic Algorithm (GA) is an appropriate method for this correction procedure. Fitness function is a particular type of objective function to achieving the set goals.
To analysis the fitness functions’ influence on the correction procedure and the dispersion model’s forecast ability, four fitness functions were designed and tested by a numerical simulation. In the numerical simulation, GA, coupled with Lagrange dispersion model, try to estimate the coefficients with model errors taken into consideration.
Result shows that the fitness functions, in which station is weighted by observed value and by distance far from release point, perform better when it exists significant model error. After performing the correcting procedure on the Kincaid experiment data, a significant boost was seen in the dispersion model’s forecast ability.
Study on the Effect of One-Dimensional Natural Circulation on the Mixing Process of Two Component Gases
Authors: Funatani Shumpei , Takeda Tetsuaki 
Published: November 2014

This study is to investigate the effect of one-dimensional natural circulation on the mixing process of two component gases by evaluating the onset time of natural circulation through the apparatus under the stable density stratified fluid layer.
The experimental apparatus consists of a reverse U-shaped vertical slot and a storage tank. The left side vertical slot consists of the heated wall and the cooled wall. The right side vertical slot consists of the two cooled walls. Temperature difference between the vertical walls was set to 50, 70, and 100 K. In this study, the combination of the two component gases is He/Ar and density ratio of each component is 1.4/10.
The heavy gas was filled with the storage tank and light gas was filled with the reverse U-shaped vertical slot. Before the experiment starts, the localized natural convection was generated in the heated side vertical slot. After the experiment starts, the heavy gas will be transported to the slot by the molecular diffusion and natural convection. And then, natural circulation occurs abruptly through the reverse U-shaped passage. The mixing process of two component gases and the onset time of natural circulation in the vertical fluid layer were affected not only by the localized natural convection but also by the molecular diffusion.
The wall and gas temperatures were measured by thermocouples and the velocity of natural convection was measured to evaluate the characteristics of the mixing process and the natural convection.
These experimental results show that generation time of natural circulation was affected by molecular diffusion and localized natural convection. When the two components of gases have large density ratios and large Gr numbers, the mixing process of two components of gases was affected by more intensively molecular diffusion than localized natural convection when temperature difference was 50K. The mixing process of two component gas was affected by more intensively localized natural convection than molecular diffusion when temperature difference was 70 to 100K. However, two component gases were affected by more intensively molecular diffusion than localized natural convection at small density ratios and small Gr numbers.
Application of Ultrasonic Technique for Precipitate Detection and Viscosity Measurement of Molten Glass at Elevated Temperature
Authors: Uchiyama Midori , Yanagihara Arisa , Koikegami Hajime , Hatanaka Hiroaki , Maruyama Tadashi 
Published: November 2014

This paper describes the results of ultrasonic examination of molten glass in the glass melter for the vitrification process of high-level radioactive liquid waste (HLW). The propagation and the attenuation of ultrasound in the molten glass were measured to study feasibility of detecting noble metal precipitates and the viscosity measurement of glass melter.
The ultrasonic propagation experiment at 2.25MHz were carried out in a simulant fluid (glycerin) at room temperature and the molten glass at 1000°C using ultrasonic wave guide rods. The wave guide rod materials used in this research are a nickel-based heat resistant alloy 690 and a high density isotropic graphite T-6 (Ibiden Co., Ltd.) with a corrosion resistant coating. Although the nickel-based alloy 690 has a favorable property such as superior structural integrity or corrosion resistant against molten glass at high temperatures, the ultrasonic attenuation was quite large thereby not suited for the wave guide rod. On the other hand, the wave guide rod made of high density isotropic graphite showed good capabilities of ultrasonic propagation both in glycerin and the molten glass at 1000°C.
The changes in ultrasonic intensity of reflecting echoes vs. distance were measured both in glycerin and the molten glass to obtain attenuation coefficient of ultrasound. Then, the viscosity of glycerin and molten glass was calculated from the result of ultrasonic attenuation measurement. The viscosity of glycerin and molten glass thus obtained showed good agreement with those obtained by ordinary viscous test.
Design Improvement Studies for Future SFR
Authors: Babu V. Rajan , Balasubramaniyan V. , Sundararajan Raghupathy , Puthiyavinayagam P. , Perumal Chellapandi 
Published: November 2014

Prototype Fast Breeder Reactor (PFBR), a 500 MWe, (U-Pu)O2 fuelled, sodium cooled, pool type fast reactor, is in advanced stage of construction at Kalpakkam, India. Based on the experience gained during the design, manufacture and erection of various reactor components of PFBR, it is planned to construct Sodium cooled Fast Reactors (SFR) by adopting twin unit (2×500 MWe reactors) concept. The future Fast Breeder Reactor (FBR) – 1 & 2 have three main heat transport circuits, namely primary sodium, secondary sodium and steam-water systems. All the reactor internals including core and primary heat transport circuit systems are contained in a single vessel called main vessel and it is closed with top shield. Reactor assembly forms the heart of the Nuclear Steam Supply System. A detailed and exhaustive design / optimisation exercise was initiated towards improving the economic competitiveness and enhancing the safety of future FBRs. It is observed that the overall dimensions of the reactor assembly contribute immensely to the capital cost. In this context, detailed studies were carried out towards optimizing the overall dimensions of the reactor assembly. Further, the reactor assembly design in particular has been engineered to favour manufacture of integrated assembly and erection of the same, as a single unit, in reactor vault to reduce construction time. Various activities undertaken towards technology development of critical components have enhanced the confidence level in the improved design concepts and reducing time for manufacture and erection.
In addition to the reactor assembly, specific improvements have been made in decay heat removal systems and sodium purification system. The layout incorporates a twin unit concept in which the ex-vessel fuel handling system and fuel storage building are shared. This paper discusses the basis for undertaking the review exercise and experience gained during construction of PFBR and highlights the design studies and technology development carried out for future SFR.
Prediction of Gas-Holdup in Filtered Containment Venting System
Authors: Strömgren Tobias , Perzon Sven , Karlsson Martin , Lillberg Eric 
Published: November 2014

A filter system to be used in order to reduce the radioactive pollution when reducing the pressure in the containment in case of a severe accident is developed. The radioactive pollution will be captured in the vessel that acts as a bubble column. To be able to maximize the water content in order to ensure that the radioactive pollution is captured it is of importance to predict the swelled water level in the vessel. In this study is three different methods to predict the gas holdup is presented and compared to experimental results; empirical correlations, thermal-hydraulic codes and Computational Fluid Dynamics (CFD). The advantages and disadvantages with the different methods are discussed.
Both Computational Fluid Dynamics and the thermal hydraulic code (Gothic) show good predictions of the water level. Using empirical models it is crucial to have knowledge of the two-phase flow region in the bubble column.
The Gamma Radiation Emitted by a PWR Core Under Severe Accident Conditions
Authors: Brachem Carsten , Konheiser Jörg , Hampel Uwe 
Published: November 2014

The gamma radiation emitted by a nuclear reactor core might contain information about the reactor state. This information may be used in a monitoring system for severe accidents. The Technische Universität Dresden and the Zittau/Görlitz University of Applied Sciences are currently carrying out feasibility studies for the development of such a system in a collaborative effort. As one part of such feasibility studies we performed Monte Carlo simulations on a simplified model of a generic pressurized water reactor. For a set of states which represent scenarios of a coolant level decrease and core melt, the gamma radiation distribution outside the reactor pressure vessel has been computed. The results are presented in this paper.
They indicate that different coolant levels yield different gamma radiation distributions, and that an accumulation of corium inside the lower head is detectable from the outside.
Characterization of Incandescent Particles Emitted by a Cut-Off Grinder During Decommissioning Operations for Evaluating Filters Degradation
Authors: Marchal Pierre , Porcheron Emmanuel , Grehan Gérard , Lafanechere Luc , Walter Jonathan , Gelain Thomas 
Published: November 2014

The use of mechanical or thermal cutting tools in decommissioning of nuclear facilities generates a lot of incandescent particles. Those particles may represent a deterioration risk of the containment barriers associated with a potential risk of fire starting.
The aim of this study is to characterize the incandescent particles emitted by a wheel grinder (in terms of temperature, diameter and velocity) and to follow those parameters all along their path from their emission point to their impact on the air filter. The characteristics of particles correlated with a possible loss of filter efficiency should highlight the destructing particles for the filter. All the measurement techniques used to characterize experimentally the incandescent particles are presented in this article. Particles are characterized in terms of diameter by microscope visualizations. The particle velocity is measured with a high speed camera using Particle Tracking Velocimetry (PTV) technique. An adaptation of a commercial monochromatic pyrometer is achieved to measure the in-flight particles temperature in our specific configuration. All of these techniques have been implemented on an experimental facility which reproduces representative conditions of the cutting processes realized during dismantling operations. Concerning the investigation of the filter, a global and a local approaches about filter degradation are used. The decontamination factor of High Efficiency Particle Air (HEPA) filter is measured, and detailed visualizations of the filter fibers deteriorations are obtained using Scanning Electrons Microscope (SEM).
Development of Inherently Safe Technologies for Large Scale BWRs: (2) Passive Water-Cooling System
Authors: Ishida Naoyuki , Watahiki Naohisa , Fujimoto Kiyoshi , Hosoi Hideaki , Kitou Kazuaki 
Published: November 2014

Working from the lessons of the Fukushima Daiichi nuclear accident, we have been developing the following various safe technologies for boiling water reactors (BWRs), a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system for reactor accidents. The objective of the study reported here was development of the passive water-cooling system. The above technologies are referred to as ‘Inherently Safe Technology’.
The passive water-cooling system works without electricity for the first 10 days after an event to remove a relatively large mount of decay heat from the core. The system consists of a condenser and a steam turbine-driven pump for transferring water from a suppression pool to the reactor. Steam from the reactor pressure vessel is condensed in the condensation tubes of the condenser, and the condensate flows out into the suppression pool in the primary containment vessel (PCV). The water temperature at the condensation tube outlet is lowered to less than the saturated temperature at the partial steam pressure of the maximum PCV design pressure to prevent the PCV failure. The condenser is located at a lower level, e.g., underground, for easier access and for supplying cooling water to a condenser pool without electricity during an event. The lower level condenser pool has an advantage that it can be seismically designed.
To evaluate our concept of the water-cooling system, heat transfer tests were conducted using full-scale U-shaped single tubes with three diameter sizes under a wide range of pressure and inlet steam velocity conditions. The heat transfer data were obtained at system pressures of 0.2 to 3.0 MPa (absolute) and inlet steam velocities of 5 to 56 m/s. The heat transfer data with this wide range of pressure and inlet velocity conditions include thermal hydraulics conditions for a passive containment cooling system (PCCS) and some of the data can be extrapolated to isolation condenser (IC) conditions. We also confirmed thermal hydraulics conditions to determine the practicality of our new concept.
Multi-Scale 2D Sensitivity Analysis of the TALL-3D Experiment
Authors: Geffray Clotaire , Macián-Juan Rafael 
Published: November 2014

The TALL-3D facility built by KTH in the scope of the European THINS project aims at investigating the behavior of lead-bismuth eutectic under forced and natural circulation. The design additionally presents complex 3D flowing patterns in a pool.
GRS developed a multi-scale thermal-hydraulics computational framework based on a coupling between a CFD code and a system code. The coupled program enables the simulation of the entire primary side of the TALL-3D facility at affordable cost but still provides 3D information where complex flow phenomena need to be accounted for.
TUM performs Uncertainty and Sensitivity Analysis of the TALL-3D experiment using GRS’ computational framework. In the present paper a Sensitivity Analysis is performed. The influence of modeling uncertainty on the model output is investigated. This analysis is brought to a detailed level with the monitoring of local effects within the pool.
Variants of the Perspective Closed Fuel Cycle, Based on REMIX-Technology
Authors: Bobrov Evgeniy , Teplov Pavel , Alekseev Pavel , Chibinyaev Alexander , Dudnikox Anatoliy 
Published: November 2014

In the traditional closed fuel cycle, based on REMIX-technology (REgenerated MIXture of U and Pu oxides) the fuel composition is produced on the basis of a uranium and plutonium mixture from spent Light Water Reactor (LWR) fuel and additional natural uranium. In this case, there is some saving in the amount of natural uranium used.
The basic features of the WWER-1000 fuel loadings with a new variant REMIX-fuel during multiple recycle in the closed nuclear fuel cycle are described in this paper. Such fuel compositions are produced on a basis of a uranium and plutonium mixture allocated at processing the spent fuel after irradiation in the WWER-1000 core, depleted uranium and fission material such as: 235U as a part of high-enriched uranium from the warheads superfluous for defense.
Also here variants are considered of the perspective closed fuel cycle in which fissile feed materials for fuel manufacture is produced in the blankets of fast breeder reactors. The fissile material is 233U or Pu. The raw material is depleted uranium from the stocks of enrichment factories, or thorium. Natural uranium is not used in this case. The minimum feed material required for the REMIX technology in a closed fuel cycle was determined through calculations of different types of fissile and raw materials, with different cycle lengths and fuel-water ratios.
Influence of Silicone Carbide on the Reactivity of Nuclear Fuels Using Cerium Dioxide as a Surrogate Material
Authors: Foral Štěpán , Salamon David , Katovský Karel , Varmuža Jan , Roleček Jakub 
Published: November 2014

The most researched material in nuclear power industry is uranium dioxide however due to strict safety and sanitary restrictions this material can be researched only in specialized research institutes and universities which have sufficient technological background.
For this reason it can be suitable to find material which would show physical properties similar to UO2 but would not suffer by the strict limitations in storage and handling. In this case much more workplaces could be incorporated in the material research and the list of investigated problems could be significantly enlarged.
One of the possible substitutional materials is the cerium dioxide (CeO2) which shows similar chemical and physical properties like UO2 and in some cases shows also similar neutronic properties. The laboratory research was focused on comparing of the basic neutronic properties. For the comparison of nuclear properties the JANIS software [1] was used as it contains cross section libraries of both materials.
It will be shown that similarity of both materials is significant and in several cases application of cerium dioxide as alternative material is possible with sufficient accuracy. As an example of the use of CeO2, research of influence of the SiC content on the reactivity of nuclear fuels is presented in this work.