bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of Technique for Controlling Natural Circulation Flow by Using Helium Gas
Authors: Hatori Hirofumi , Yanagawa Naoto , Takeda Tetsuaki , Funatani Shumpei
Published: November 2014

The purpose of this study is to investigate a control method of natural circulation flow of air by injection of helium gas. A depressurization accident by the primary pipe rupture is one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the double coaxial duct connecting between a reactor core and an intermediate heat exchanger (IHX) module breaks, air is expected to enter the reactor pressure vessel from the breach and oxidize in-core graphite structures. Then, it seems to be probable that the natural circulation flow of air in the reactor pressure vessel produce continuously. In such condition, injection of helium gas into the channel by a passive method can prevent occurrence of the natural circulation flow of air in the reactor pressure vessel. Therefore, it is thought that oxidation of in-core graphite structures by air ingress can be prevented by establishing this method.

The experiment has been carried out regarding the natural circulation flow using a circular tube consisting of a reverse U-shaped type. The vertical channel consists of one side heated tube and the other side cooled tube. The experimental procedure is as follows. Firstly, the apparatus is filled with air and one vertical tube is heated. Then, natural circulation of air will be produced in the channel. After the steady state is established, a small amount of helium gas is injected from the top of the channel. The velocity, mole fraction, temperature of gas, and temperature of the tube wall are measured during the experiment. The results were obtained as follows. When the temperature difference between the both vertical tubes was kept at about 60K, the velocity of the natural circulation flow of air was measured about 0.17m/s. During a steady state, a small amount of helium gas was injected into the channel. When the volume of injected helium gas is about 5.7% of the total volume of the channel, the velocity of the natural circulation flow of air became around zero. After 810 seconds elapsed, the natural circulation flow suddenly produced again. The natural circulation flow of air can be controlled by injecting of helium gas.
Estimation of Source Term Uncertainty in a Severe Accident With Correlated Variables
Authors: Zheng Xiaoyu , Itoh Hiroto , Tamaki Hitoshi , Maruyama Yu
Published: November 2014

The quantitative evaluation of the fission product release to the environment during a severe accident is of great importance. In the present analysis, integral severe accident code MELCOR 1.8.5 has been applied to estimating uncertainty of source term for the accident at Unit 2 of the Fukushima Daiichi nuclear power plant (NPP) as an example and to discussing important models or parameters influential to the source term. Forty-two parameters associated with models for the transportation of radioactive materials were chosen and narrowed down to 18 through a set of screening analysis. These 18 parameters in addition to 9 parameters relevant to in-vessel melt progression obtained by the preceding uncertainty study were input to the subsequent sensitivity analysis by Morris method. This one-factor-at-a-time approach can preliminarily identify inputs which have important effects on an output, and 17 important parameters were selected from the total of 27 parameters through this approach. The selected parameters have been integrated into uncertainty analysis by means of Latin Hypercube Sampling technique and Iman-Conover method, taking into account correlation between parameters. Cumulative distribution functions of representative source terms were obtained through the present uncertainty analysis assuming the failure of suppression chamber. Correlation coefficients between the outputs and uncertain input parameters have been calculated to identify parameters of great influences on source terms, which include parameters related to models on core components failure, models of aerosol dynamic process and pool scrubbing.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: (12) Bubble Motion Along the Flow in Structure Vibration
Authors: Kato Yuki , Arai Rie , Kaneko Akiko , Monji Hideaki , Abe Yutaka , Yoshida Hiroyuki , Takase Kazuyuki
Published: November 2014

In a nuclear power plant, one of the important issues is an evaluation of the safety of the reactor core and its pipes when an earthquake occurs. Many researchers have conducted studies on constructions of plants. Consequently, there is some knowledge about earthquake-resisting designs.

However the influence of an earthquake vibration on thermal fluid inside a nuclear reactor plant is not fully understood. Especially, there is little knowledge how coolant in a core response when large earthquake acceleration is added. Some studies about the response of fluid to the vibration were carried out. And it is supposed that the void fraction and/or the power of core are fluctuated with the oscillation by the experiments and numerical analysis. However the detailed mechanism about a kinetic response of gas and liquid phases is not enough investigated, therefore the aim of this study is to clarify the influence of vibration of construction on bubbly flow behavior. In order to investigate the influence of vibration of construction on bubbly flow behavior, we visualized bubbly flow in pipeline on which sine wave was applied. In a test section, bubbly flow was produced by injecting gas into liquid flow through a horizontal circular pipe. In order to vibrate the test section, an oscillating table was used. The frequency and acceleration of vibration added from the oscillating table was from 1.0 Hz to 10 Hz and . 0.4 G (1 G=9.8 m/s2) at each frequency. The test section and a high speed video camera were fixed on the oscillating table. Thus the relative velocity between the camera and the test section was ignored. PIV measurement was also conducted to investigate interaction between bubble motion and surround in flow structure. Liquid pressure was also measured at upstream and downstream of the test section. The effects of oscillation on bubbly flow were quantitatively evaluated by these pressure measurements and the velocity field. In the results, it was observed that the difference of bubble motion by changing oscillation frequency. Moreover it was suggested that the bubble deformation is correlated with the fluctuation of liquid velocity field around the bubble and the pressure gradient in the flow area. In addition, these experimental results were compared with numerical simulation by a detailed two-phase flow simulation code with an advanced interface tracking method, TPFIT. Numerical simulation was qualitatively agreed with experimental results.
Study on Criteria of Flushing Phenomena in Boiling Transition by Microwave Heating
Authors: Yamaki Tatsunori , Abe Yutaka , Kaneko Akiko , Segawa Tomoomi , Kawaguchi Koichi , Yamada Yoshikazu
Published: November 2014

In the Microwave Heating de-nitration (MH) method developed in Japan, a mixed solution of uranyl nitrate and plutonium nitrate (Pu/U mixed nitrate solution) recovered from the spent fuel in the reprocessing plant is converted directly to mixed oxide (MH-MOX) powder. This MH-MOX powder is utilized to fabricate MOX fuel with UO2 powder for FBR.

The MH method is accompanied with transient boiling phenomena such as overflow and flushing. Toward high-speed and high-capacity conversion by MH-method in the future, it is required to avoid overflow and flushing and to understand optimal conditions for vessel shape design and microwave output operation. The objective of this paper is to elucidate occurrence criteria of flushing phenomena.

At the first step for this objective, basic knowledge of transient boiling phenomena by the MH-method has been mainly acquired with using distilled water. From the results, it is observed that generation of singular bubble triggers flushing and distilled water just before flushing is superheated more than 10 °C in conditions that flushing is confirmed. Also, the water temperature reaches its peak, and it is almost unchanged in conditions that flushing is not confirmed. In no flushing conditions, it is found that the evaporation starts from the point where the water temperature reaches its peak, and water level is decreased gradually. Thus, the difference of thermal characteristics greatly affects whether flushing occurs.

The second step, by focusing on the process before flushing occurs, we investigate flow structure, and heat amount through the temperature distribution inside the jelly just after microwave heating. Potassium chloride (KCl) aqueous solution and water are used. KCl solution is electrolyte solution same as reprocessing solution. From the results, upward flow is observed near the center of the solution in the case of the water just after microwave heating. On the other hand, downward flow is observed in the case of KCl solution. As a result of temperature distribution, the water is heated its near center, and KCl solution is heated the around. Thus, it is considered that heat amount of microwave varies depending on solution characteristics, and the difference greatly affects the flow structure. The heat amount and flow structure are critical factors that can determine transient boiling phenomena. We need to investigate transient boiling phenomena of KCl solution in the future.
Relationship Between Flow Pattern and Pressure Distribution in Venturi Scrubber
Authors: Horiguchi Naoki , Yoshida Hiroyuki , Kaneko Akiko , Abe Yutaka
Published: November 2014

As revealed by Fukushima Daiichi nuclear disaster, countermeasures against severe accidents in nuclear power plants are an urgent need. In particular, from the viewpoint of protecting containment and suppressing diffusion of the radioactive materials, it is most important to install filtered venting devices to release high pressure contaminated gas to the atmosphere with elimination radioactive materials in the gas.

A Multi Venturi Scrubber System (MVSS) is one of the filtered venting devices, and used in European reactors [1, 2]. One of the main components of the MVSS is a Venturi Scrubber (VS). It is considered that a dispersed or dispersed annular flow is formed in the VS by a self-priming phenomena. In the self-priming phenomena, the liquid was suctioned from a surrounding region of the VS to the inside of the VS. And a part of the radioactive materials are eliminated through the gas-liquid interface of the dispersed or annular dispersed flow. Therefore, to consider the MVSS operation characteristics, it is important whether to occur the self-priming or not and the liquid flow rate of the self-priming of the VS.

The objective of this paper is to understand the self-priming phenomena of the VS for the filtered venting. And theoretical analysis and experiment were conducted. By comparing these results, we discussed about the mechanism of the self-priming phenomena. As results, the self-priming phenomena in the VS was confirmed and, at a high gas flow rate, the suspension of the self-priming is confirmed experimentally and theoretically.
Study on Thermalhydraulic Effects of Nuclear Steam Superheat for a Generic 1200-MWel Pressure-Channel Reactor
Authors: Sidawi Khalil , Vincze Andrei , Abdullah Rand , Baldock Matthew , Peiman Wargha , Pioro Igor
Published: November 2014

Current generation water-cooled Nuclear Power Plants (NPPs) have significantly lower thermal efficiencies than their thermal counterparts; due, partially, to their lower turbine-inlet steam temperature. Nuclear steam superheat can be implemented in a generic pressure-channel nuclear reactor to increase the temperature of the steam at the inlet of the turbine, and thus increase the thermal efficiency of a NPP.

A heat flux is computed specifically for a stable SuperHeated Steam (SHS) and Pressurized Water (PW) 520 pressure-channel reactor core configuration, from which a unique temperature profile for each coolant (as a bulk fluid) is calculated. Using the coolant temperature profile of each coolant, the sheath temperature distribution is calculated, using Fourier’s law, and the fuel pellets’ axial and radial temperature profiles are determined using an analytical solution to the temperature distribution in a solid with uniform heat generation.

Properties of the coolant, sheath, and fuel were calculated based on the temperature (and pressure, in the case of coolant) along the heated length of a channel. The effects on the flow rates and the differences in the required channel powers, due to the addition of the SHS channels, were also considered. To ensure safe operating parameters, the maximum sheath and fuel centerline temperatures were shown to be much lower than the operating limits.

The implementation of steam superheat in a generic 1200-MWel pressure-channel nuclear reactor allows for an increase in the temperature of steam at the inlet of a turbine from ∼319°C to ∼550°C, and ultimately an increase in the thermal efficiency of the NPP by about 5–7%.
Monte Carlo Neutron Transport on GPUs
Authors: Bergmann Ryan M. , Vujić Jasmina L.
Published: November 2014

GPUs have gradually increased in computational power from the small, job-specific boards of the early 90s to the programmable powerhouses of today. Compared to CPUs, they have a higher aggregate memory bandwidth, much higher floating-point operations per second (FLOPS), and lower energy consumption per FLOP. Because one of the main obstacles in exascale computing is power consumption, many new supercomputing platforms are gaining much of their computational capacity by incorporating GPUs into their compute nodes. Since CPU optimized parallel algorithms are not directly portable to GPU architectures (or at least without losing substantial performance gain), transport codes need to be rewritten in order to execute efficiently on GPUs. Unless this is done, we cannot take full advantage of these new supercomputers for reactor simulations. In this work, we attempt to efficiently map the Monte Carlo transport algorithm on the GPU while preserving its benefits, namely, very few physical and geometrical simplifications. Regularizing memory access and introducing parallel-efficient search and sorting algorithms are the main factors in completing the task.
Development of a Multi-Dimensional Measurement Sensor of Void Fraction and Phasic Velocity for Boiling Two-Phase Flow in a 5×5 Heated Rod Bundle
Authors: Arai Takahiro , Furuya Masahiro , Kanai Taizo , Shirakara Kenetsu , Nishi Yoshihisa
Published: November 2014

A subchannel void sensor (SCVS) was developed to measure the cross-sectional distribution of a void fraction in a 5×5 heated rod bundle with o.d. 10 mm and heated length 2000 mm, and applied in a boiling two-phase flow experiment under the atmospheric conditions assumed in an accident and spent fuel pool. The SCVS comprises 6-wire by 6-wire and 5-rod by 5-rod electrodes. Wire electrodes 0.2 mm in diameter are arranged in latticed patterns between the rod bundle, while a conductance value in a region near one wire and another gives a local void fraction in the central-subchannel region. 32 points (= 6×6−4) of the local void fraction can be obtained as a cross-sectional distribution. In addition, a local void fraction near the rod surface can be estimated by a conductance value in a region near one wire and one rod using the simulated fuel rods as rod electrodes, which allows 100 additional points (=4×25) of the local void fraction to be acquired. The devised sensors are installed at five height levels to acquire two-phase flow dynamics in an axial direction. A pair of SCVS is mounted at each level and placed 30 mm apart to estimate the one-dimensional phasic velocity distribution based on the cross-correlation analysis of both layers. The time resolution of void measurement exceeds 800 frames (cross-sections) per second. The heated rod bundle has an axially and radially uniform power profile, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The boiling two-phase flow experiment, which simulated a boil-off process, was conducted with the devised SCVS and experimental data was acquired under various experimental conditions, such as inlet-flow velocity, rod-bundle power and inlet subcooling. The experimental results exhibited axial and radial distribution of two-phase flow structures, i.e. void-fraction and phasic-velocity distributions quantitatively.
CFD Study of the Condensation of Water Vapor on the Particles Flow in Rectangular Narrow Channel
Authors: Lin Daping , Zhou Tao , Yang Xu , Fang Xiaolu
Published: November 2014

This paper presents a CFD study of the flow and deposition of particles in vapor in a 1000mm by 20mm narrow rectangular channel. The temperature of the wall is set below the condensation point, so the vapor will change into the liquid phase near the wall. The flow and deposition of the particles in the condition when the phase transition of water occurs is simulated by ANSYS Fluent code. The result shows that the condensation of the water vapor will change the temperature field and the velocity of the flow in the channel. The particles will be forced to the wall due to the thermophoretic force. The condensation developing process will enhance the deposition of the particles while the developed liquid layer will stop the particles flow to the wall.
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.
New Exploration on TMSR: The Lattice Optimization
Authors: Si Shengyi , Chen Qichang , Bei Hua , Zhao Jinkun
Published: November 2014

TMSR fuel lattice was reevaluated by varying the P/D ratio (Ratio of Lattice Pitch to Channel Diameter) and Lattice Size (Pitch); an in-house developed computer code named SONG/TANG-MSR was used to calculate the key TMSR parameters such as core reactivity, conversion or breeding ratio and power reactivity coefficient. The optimized lattice was recommended with trade off of the above key parameters, which can satisfy the requirements of Sustainability and Safety.
Research of Liquid Entrainment Through ADS-4 in AP1000
Authors: Meng Zhaoming , Xiao-liang Fu , Ding Lei , Tian Wenxi , Dong Bo , Su Guanghui
Published: November 2014

Experimental and theoretical studies have been performed for entrainment at T-junction consisting of hot leg and ADS-4 pipe. Research of basic influence factors on onset entrainment and stable entrainment is conducted and the effects of gas chamber height and gas Froude number on entrainment are discussed on the basis of experimental data. Based on visualization research, the flow regime in the main pipe is recorded by high-speed camera. Three kinds of entrainment phenomena including stratified entrainment, annular entrainment and intermittent entrainment can be clearly found, and feature of each phenomenon is summarized. Also, Onset entrainment model is established and accords with experimental data well.
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.
Experimental Investigation of Liquid Entrainment in ADS-4 Branch Line
Authors: Sun Ducheng , Liu Jianchang , Tian Wenxi , Qiu Suizheng , Su Guanghui , Zhang Peng , Ma Yingying
Published: November 2014

In order to gain more insights into the system depressurization and entrainment behavior after actuation of the fourth-stage (ADS-4) valves during a loss-of-coolant-accident, the ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) scaled to AP1000 was constructed to simulate the accident scenario with air-water and steam-water. A brief scaling analysis with emphasis on related thermal hydraulic processes was presented. Entrainment phenomena at vertical up tee branch were observed and analyzed. Preliminary test data of onset of entrainment and entrainment rate were collected with air-water tests and relevant conclusions were obtained.
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.
Long Term SBO With Selected Mitigative Measures: MELCOR Parametric Calculations for a 2-Loop PWR
Authors: Rydl Adolf , Jaeckel Bernd , Birchley Jonathan C. , Lind Terttaliisa
Published: November 2014

Analyses of three long term PWR Station Black-Out (SBO) scenarios with and without mitigation are performed using the US NRC source term code MELCOR. Refilling of the secondary side of the steam generator (SG) with fire water pumps was previously identified as a potentially important accident management measure. To assess parametrically the effect of the restored availability to refill of SG secondary, SBO sequences are analyzed both without any mitigation and with mitigation at different times into the accident representing different stages of the accident progression. The scenarios studied were (i) base-line SBO without any thermally-induced Reactor Coolant System (RCS) breach, and then sequences with assumption of (ii) Surge Line failure and (iii) thermally-induced SG tube rupture (SGTR). A detailed model was used for the description of flows in the hot legs and in SGs to enable us to simulate the counter-current natural circulation which is inherent in these types of scenarios, with the cold leg plugged by water in the loop seal. These were long-term simulations, some of them up to 5 days of the transient, with the analyses of the containment response and fission product (FP) releases to environment. For the relevant cases, the molten core-concrete interactions (MCCI) are modeled in detail in the complex lower containment geometry with several distinct volumes (cavity and other rooms with concrete walls) subject to ablation by the molten corium; the challenges to containment barrier by the late overpressurization and by concrete ablation are evaluated. With the Passive Autocatalytic Recombiners, PARs, installed the hydrogen is found to be of less threat to containment integrity. For the thermally-induced SGTR the impact of the chosen mitigation strategy on potential bypass FP release to environment is also assessed for different times of the refill. The results indicate that mitigation by the SG secondary refill can be very effective for the base-line SBO sequences. For the thermally-induced SGTR it is effective only when the refill is achieved early after the SG tube failure. Mitigation by the pressurizer-loop SG refill was much less successful in the case of the Surge Line failure sequences.
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.
Influence of Inlet Velocity Condition on Unsteady Flow Characteristics in Piping With a Short-Elbow at High Reynolds Number Condition
Authors: Ono Ayako , Tanaka Masaaki , Kobayashi Jun , Kamide Hideki
Published: November 2014

In design of the Japan Sodium-cooled Fast Reactor (JSFR), mean velocity of the coolant is approximately 9 m/s in the primary hot leg (H/L) piping which diameter is 1.27 m. The Reynolds number in the H/L piping reaches 4.2×107. Moreover, a short-elbow which has Rc/D = 1.0 (Rc: Curvature radius, D: Pipe diameter) is used in the hot leg piping in order to achieve compact plant layout and reduce plant construction cost. In the H/L piping, flow-induced vibration (FIV) is concerned due to excitation force which is caused by pressure fluctuation on the wall closely related with the velocity fluctuation in the short-elbow. In the previous study, relation between the flow separation and the pressure fluctuations in the short-elbow was revealed under the specific inlet condition with flat distribution of time-averaged axial velocity and relatively weak velocity fluctuation intensity in the pipe. However, the inlet velocity condition of the H/L in a reactor may have ununiformed profile with highly turbulent due to the complex geometry in reactor vessel (R/V). In this study, the influence of the inlet velocity condition on unsteady characteristics of velocity in the short-elbow was studied. Although the flow around the inlet of the H/L in R/V could not simulate completely, inlet velocity conditions were controlled by installing the perforated plate with plugging the flow-holes appropriately. Then expected flow patterns were made at 2D upstream position from the elbow inlet in the experiments. It was revealed that the inlet velocity profiles affected circumferential secondary flow and the secondary flows affected an area of flow separation at the elbow, by local velocity measurement by the PIV (particle image velocimetry). And it was found that the low frequent turbulence in the upstream piping remained downstream of the elbow though their intensity was attenuated.
Effect of Two-Phase Flow Structure in Decontamination Factor of Filtered Containment Venting System
Authors: Kanai Taizo , Furuya Masahiro , Arai Takahiro , Tanaka Nobuyuki , Nishi Yoshihisa , Shirakawa Kenetsu , Nishimura Satoshi , Satake Masaaki
Published: November 2014

In order to gain the best use of filtered containment venting systems (FCVSs), the decomtamination factor of FCVSs is to be investigated as a function of system parameter including steam flow rate, pressure, temperature, water level, and operating time. A full-height test facilities were designed and constructed in Central Research Institute of Electric Power Industry (CRIEPI), Japan to evaluate the decontamination factor (DF) in FCVSs. The target types are the orifice and the venturi FCVSs. The height and the internal diameter of the cylindrical test vessel is 8 m and 0.5 m. Bubbly flows were visualized through the view window up to 0.8 MPa and 170 °C. Steam bubbles in 0.2 wt% sodium thiosulfate and 0.5 wt% sodium hydroxide were found to be much smaller than those in water. The DF were evaluated for the aerosol, elemental iodine and organic iodine. The installed aerosol optical spectrometer measures the number density and the diameter of aerosols. The concentrations of elemental iodine were quantified with an inductively-coupled plasma with mass spectrometry (ICP-MS). The concentration of organic iodine was quantified with a gas chromatography with mass spectrometry (GC-MS). In order to investigate two-phase flow dynamics in the vessel, separate effect tests were conducted with air-water test facility. The height of cylindrical test vessel is 8 m. Visual observation was conducted for two internal diameter levels: 0.05 and 0.5 m. High speed video frames were recorded through the transparent (acrylic) vessel wall. Wire-Mesh Sensors (WMS) were installed to acquire a cross-sectional void fraction to compare with DF in the facility. On the basis of the obtained database, we develop the FCVSs performance evaluation technique and propose an optimal FCVSs operation method for a further safety improvements of the nuclear power plant.
Comparison Between PARCS and MCNP6 Codes on VVER1000/V320 Core
Authors: Mazzini Guido , Miglierini Bruno , Ruščák Marek
Published: November 2014

Research Centre Rez solves several safety related projects dealing with safety of Czech NPPs, some of which require fully functioning Three Dimensional (3D) model of the reactor core. While in a number of safety analysis of various accident scenarios it is sufficient to use one point reactor kinetics, there are selected types of accidents in which it is useful to model the space (3D) neutron kinetics, in particular control rod ejections, boron dilution scenarios, including transitions from design basis to beyond design basis accidents. This paper is focused to analyze the present model of the core of VVER1000/V320 reactor. Which is applicable for 3D modeling of neutron kinetics in selected design and beyond design basis accidents. The model is based on a cross-sections library created by SCALE 6.1.2/TRITON simulations. PARCS 3.2 code uses homogenized cross-sections libraries to calculate neutronic and other core parameters of the PWR reactors. Similar model is prepared with MCNP6 for comparison between deterministic (Pn spherical-harmonics method used in PARCS) and the stochastic (Monte Carlo) approach (used in MCNP6). Such comparison will serve as a demonstration of the capability of the PARCS code for VVER1000/V320 analyses.
The Effects of Rolling Motion Upon the Coefficient of Flow Boiling Heat Transfer
Authors: Chen Chong , Gao Pu-zhen
Published: November 2014

Experimental investigations on boiling heat transfer coefficients of boiling flows in rectangular narrow channel under rolling motion condition are performed. The cross section of the testing rectangular narrow channel is 2×40 mm, and the mechanical rolling thermal-hydraulic experimental facility is used in the experimental research of boiling heat transfer characteristics. Deionized water is used as the working fluid. The results show that the amplitude of boiling heat transfer coefficients of rectangular narrow channel increases with increasing rolling amplitude and rolling period of the rolling platform, the time average boiling heat transfer coefficients of test section in rolling motion are equal to the coefficients of the test section at equilibrium position, and with the increase of rolling amplitude and rolling period the time average boiling heat transfer coefficient almost unchanged. The amplitude of boiling heat transfer coefficients increases with increasing heat flux and flow rate, while decreases with the increase of system pressure. The curve of boiling heat transfer coefficient fluctuations of rectangular narrow channel is close to sine or cosine curve when the rolling period less than 15 seconds.
Effect of Regulation Valves Installation in High Pressure Injection System Pipelines on the Formation of Reactor Vessel Thermal Stress Conditions
Authors: Mazurok Aleksander , Vyshemirskyi Maksym
Published: November 2014

Effect of regulation valves (RV) installation in high pressure injection system (HPIS) pipelines on the formation of reactor pressure vessel (RPV) thermal stress conditions was analyzed. Modernization is implemented at South-Ukrainian nuclear power plant (SUNPP) Unit 1 within the framework of life extension, which finished by the end of 2013. The main goal of the modernization is to expand the HPIS functionality for small leak accident and protection against the cold overpressurization due to flow rate and primary pressure effectively regulation. The thermal hydraulic model for RELAP5/mod3.2 code with detailed downcomer (DC) model and changes in accordance with modernization was used for calculations. Detailed (realistic) modeling of piping and equipment was performed. Also, an algorithm for the RVs was developed. Applying of cooling water flow rate regulation avoids excessive primary cooling and, consequently, helps to preserve the RPV integrity and to prevent reaching through crack formation, which can lead to a severe accident.
Atmosphere Isolation During the Maintenance of HTR-10 Helium Circulator
Authors: Liu Ling , Wei Liqiang , Chen Xiaoming , Wang Yucheng
Published: November 2014

Atmosphere isolation is one of the pivotal operations during the maintenance of HTR-10 helium circulator, which main objects are to keep the high helium purity within the primary loop of HTR-10 and avoid radioactive contamination and diffusion of graphite dust. Several skillful and comprehensive processes are applied to achieve the object successfully.
Hydrodynamic Performance Study on Flow Field in Rod Bundles With Spacer Grids
Authors: Song Li-feng , Tian Rui-feng , Sun Lan-xin , Shi Cong , Li Peng
Published: November 2014

With the dramatic progress in the computational fluid dynamics (CFD) methodology, this technology can be used in researching the knowledge of thermal–hydraulic characteristics in the rod bundles, particularly with the spacer grids. These characteristics, including fluid flow, turbulence, and heat transfer and so on, all of this information can be applied in design and the improvement of rod bundles. This paper calculated the single-phase flow field in rod bundles with different spacer grids respectively by numerical method, which was based on the experiment performed by Korea Atomic Energy Research Institute (KAERI). Two types of grid designs are used, including the Split-type and Swirl-type, respectively. By analyzing the calculated results, discussing the feasibility of computational fluid dynamics (CFD) methodology in thermal-hydraulic analysis of nuclear reactor, and obtaining the hydrodynamic performance of the two different types of spacer grids. The result showed that both of the two spacer grids led to intense cross-flow in the channel, and the cross-flow intensity decayed with the increasing of downstream distance after the spacer grids. Axial velocities were distributed uniformly in the channel.
Source Term Estimation in Nuclear Accidents Using 4D Variational Data Assimilation for Heterologous Atmospheric Condition
Authors: Liu Yun , Fang Sheng , Li Hong , Qu Jingyuan , Fang Dong
Published: November 2014

The source information of the radionuclide release in nuclear accidents is a key issue of the nuclear emergency response. One way to estimate the source information is by inversing the radionuclide transportation process based on environment radiation monitoring data. The advantage of this method is that the required data are easy to obtain in accident. But it is vulnerable to large uncertainties in both data and transport model. To solve the problem, a source term estimation method based on four-dimensional variational (4DVAR) data assimilation technique was proposed for source term estimation in this study. The proposed method couples 4DVAR with the RIMPUFF air dispersion model. It formulates the inverse modelling of source term estimation as an optimization problem that is trying to find an optimal balance between real observation data and the background field. The advantage of this method is that the radionuclide transport in every time step is included in data assimilation and the result is global optimum in the whole assimilation period. The gradient for cost function is calculated by the backward integration of the adjoint model. Practical imperfectness of measurement were considered and integrated into the cost function. The proposed method was verified using numerical simulation for both homogeneous and heterologous atmospheric condition. The performance of the source term estimation method was also investigated with respect to different release profile, wind speed and atmospheric stability class. The simulation results demonstrate that the estimate matches the true release status well for both homogeneous and heterologous wind field. Also, the experimental results show that the proposed method has strong robustness to wind speed and atmospheric stability.
Effect of Interfacial Pressure Term on Numerical Stability of a Two-Fluid Model
Authors: Okawa Tomio , Kudo Yoshiro
Published: November 2014

Mathematical ill-posedness of the governing equations is one the main causes of numerical instability encountered in numerical simulation of two-phase flow using a two-fluid model. It is known that the ill-posedness can be mitigated if the difference between the average pressures of gas and liquid phases is taken into consideration appropriately. In the present work, it was investigated how the numerical stability of the one-dimensional, two-fluid model is influenced by the interfacial pressure terms that express the pressure difference between bubbles and continuous liquid phase in bubbly two-phase flow. Analyses were carried out for adiabatic air-water two-phase flow and subcooled flow boiling. It was confirmed that the interfacial pressure terms are effective to mitigate the numerical instability induced by the mathematical ill-posedness of the two-fluid model. However, the standard interfacial pressure terms deteriorated the numerical stability in some cases. It was found that the simplified model in which the spatial gradients of relative velocity and fluid density are eliminated is effective for the mitigation of numerical instability in wider analytical conditions.