bluuebunny/crossref_metadata_2025 huggingface dataset:
Self-Propelling Cooling Systems: Back-Fitting Passive Cooling Functions to Existing Nuclear Power Plants
Authors: von Lavante Dominik , Kuhn Dietmar , von Lavante Ernst
Published: October 2013

The present paper describes a back-fit solution proposed by RWE Technology GmbH for adding passive cooling functions to existing nuclear power plants. The Fukushima accidents have high-lighted the need for managing station black-out events and coping with the complete loss of the ultimate heat sink for long time durations, combined with the unavailability of adequate off-site supplies and adequate emergency personnel for days. In an ideal world, a nuclear power plant should be able to sustain its essential cooling functions, i.e. preventing degradation of core and spent fuel pool inventories, following a reactor trip in complete autarchy for a nearly indefinite amount of time. RWE Technology is currently investigating a back-fit solution involving “self-propelling” cooling systems that deliver exactly this long term autarchy. The cooling system utilizes the temperature difference between the hotter reactor core or spent fuel pond with the surrounding ultimate heat sink (ambient air) to drive its coolant like a classical heat machine. The cooling loop itself is the heat machine, but its sole purpose is to merely achieve sufficient thermal efficiency to drive itself and to establish convective cooling (∼2% thermal efficiency). This is realized by the use of a Joule/Brayton Cycle employing supercritical CO2. The special properties of supercritical CO2 are essential for this system to be practicable. Above a temperature of 30.97°C and a pressure of 73.7bar CO2 becomes a super dense gas with densities similar to that of a typical liquid (∼400kg/m3), viscosities similar tothat of a gas (∼3×105Pas) and gas like compressibility. This allows for an extremely compact cooling system that can drive itself on very small temperature differences. The presented parametric studies show that a back-fitable system for long-term spent fuel pool cooling is viable to deliver excess electrical power for emergency systems of approximately 100kW. In temperate climates with peak air temperatures of up to 35°C, the system can power itself and its air coolers at spent fuel pool temperatures of 85°C, although with little excess electrical power left. Different back-fit strategies for PWR and BWR reactor core decay heat removal are discussed and the size of piping, heat exchangers and turbo-machinery are briefly evaluated. It was found that depending on the strategy, a cooling system capable of removing all decay heat from a reactor core would employ piping diameters between 100–150mm and the investigated compact and sealed turbine-alternator-compressor unit would be sufficiently small to be integrated into the piping.
Investigation on Velocity Distribution Around the Wrapping Wire in an Inner Subchannel of Fuel Pin Bundle
Authors: Nishimura Masahiro , Sato Hiroyuki , Kamide Hideki , Ohshima Hiroyuki , Nagasawa Kazuyoshi , Imai Yasutomo
Published: October 2013

A sodium cooled fast reactor is designed to attain a high burn-up core in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the flow velocity distribution in a wire wrapped pin bundle. In this study, water experiments were carried out to investigate the detailed velocity distribution in inner subchannel of the pin bundle geometry. These basic data are not only useful for understanding of pin bundle thermal hydraulics but also code validation.

A wire-wrapped 3-pin bundle water model was applied to investigate the detailed velocity distribution in an inner subchannel surrounded by 3 pins with the wrapping wire. The test section consists of an irregular hexagonal acrylic duct tube and fluorinated resin pins which have nearly the same refractive index with that of water and a high light transmission rate. This refractive index matching enables to visualize the inner subchannel through the outer pins. The velocity distribution in the inner subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through two sides of the duct tube. Typical flow velocity conditions in the pin bundle were 1.6m/s (Re = 13,500) and 0.36m/s (Re = 2,700).

Feature of stream regime in the subchannel existing wrapping wire was visualized in vertical and horizontal plane. The time averaged velocity field in the horizontal plane was reconstructed from the two vertical plane data in different directions. A detailed simulation code based on FEM was applied to the experimental analysis. The calculated velocity distributions were consistent with the experimental data.
Jet Breakup Behavior With Surface Solidification
Authors: Iwasawa Yuzuru , Abe Yutaka , Kaneko Akiko , Kuroda Taihei , Matsuo Eiji , Nariai Hideki , Koyama Kazuya , Sakaba Hiroshi , Itoh Kazuhiro
Published: October 2013

When a hypothetical Core Disruptive Accident (CDA) occurs in Fast Breeder Reactor (FBR), it is strongly required that the molten core material can be solidified and cooled down by the sodium coolant in a reactor vessel. In order to estimate whether the molten material jet is completely solidified by sodium coolant, it is necessary to understand the interaction between the molten core material and the coolant.

The objectives of the present study are to clarify the correlation of the jet breakup and fragmentation behavior and the dominant factors of both behaviors considering surface solidification. In order to investigate the influence of surface solidification on jet breakup and fragmentation behavior, experiments under surface solidification and liquid-liquid contact condition are conducted. Although the molten material jet is fragmented with each condition, jet breakup and fragmentation behaviors on each condition are different. In addition, when the surface solidification occurs, there is possibility that the material strength of solidified crust on the surface affects jet breakup and fragmentation behaviors. Then, numerical calculation based on hydrodynamics and material mechanics is conducted to evaluate the influence of the material strength on jet breakup and fragmentation behaviors. In comparison with the numerical estimation and mass median diameters obtained from experimental results, the effect of solidification on jet breakup and fragmentation behavior of molten material jet is discussed.
The Effect of Flow Field on the Mass Transfer Coefficient in the Orifice Downstream
Authors: Fujishiro Atsushi , Shan Feng , Kojo Retsu , Takeuchi Momoe , Tsuji Yoshiyuki
Published: October 2013

Flow accelerated corrosion (FAC) caused many accidents of nuclear energy or thermal-power-generation plants. The main cause of FAC is known as promotion of mass transfer from the wall by the turbulence accompanied by separation, reattachment, and rotating of the flow. However, the relation between mass transfer and flow field is not clarified. In this study, we investigate the relation between the flow field behind the orifice and the mass transfer rate by means of PIV measurement and electrochemical method. It is concluded that the mass transfer coefficient is scaled by the cross-sectional averaged velocity at the orifice UD and the parameter Res in an interaction with a swirling flow.
Simulation of the Natural Convection Test of Phenix Reactor Using Modified ATHLET
Authors: Zhou Chong , Huber Klaus , Cheng Xu
Published: October 2013

The sodium-cooled fast reactor (SFR) is gathering worldwide attention for its features of the fast-spectrum reactor and closed fuel recycle system. This paper presents the modification of the ATHLET code for application to SFRs. The thermal-hydraulic computer code ATHLET is developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) in Germany for light water reactors analysis. In this paper, a sodium properties package is implemented in to the ATHLET code, and heat transfer correlations for sodium are also added for heat transfer prediction. To evaluate the capability of the modified code, the Phenix reactor, a SFR operated by French Alternative Energies and Atomic Energy Commission (CEA) from 1973 to 2009, is modeled. The scenario of transient from forced to natural convection is simulated and analyzed. The results are compared with the experimental data of the natural convection ultimate test of the Phenix facility. Results achieved so far indicate good applicability of the modified ATHLET code.
Thermal Hydraulic Test of Advanced Fuel Bundle With Spectral Shift Rod (SSR) for BWR: Steady State and Transient Test Results and Analysis
Authors: Kondo Takao , Kitou Kazuaki , Chaki Masao , Ohga Yukiharu , Makigami Takeshi
Published: October 2013

Japanese national project of next generation light water reactor (LWR) development started in 2008. Under this project, spectral shift rod (SSR) is being developed.

SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR’s merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies the change of average core void fraction, resulting in the amplified spectral shift effect.

This paper presents the steady state test with varied SSR geometry parameters, the transient test, and the simulation analysis of these steady state and transient tests. The steady state test results showed that the basic functioning principle such as the controllability of SSR water level by flow rate was maintained in the possible range of geometry parameters. The transient test results showed that the change rate of SSR water level was slower than the initiating parameters. The simulation analysis of steady state and transient test showed that the analysis method can simulate the height of SSR water level and its change with a good agreement.

As a result, it is shown that the SSR design concept and its analysis method are feasible in both steady state and transient conditions.
Measurement of Void Fraction in Dispersed Bubbly Flow With Constant Electric Current Method
Authors: Uesawa Shin-ichiro , Kaneko Akiko , Abe Yutaka
Published: October 2013

Void fraction is one of important physical values for gas-liquid two-phase flow in nuclear power plants, and it is an essential parameter for designs and performance evaluations of devices, including a core of BWR and so on. Therefore void fraction measurement with real time, high temporal resolution and high spatial resolution has been needed. In several void fraction measurement methods, electric measurement methods of void fraction can realize real-time measurement. In previous studies, conductance methods capacitance methods, wire mesh methods and tomography techniques have been studied.

In the present study, we research on a constant electric current method. This method can measure void fraction with higher temporal resolution and simpler systems. In previous study, the constant electrical current method has been applied in annular flow mainly. However, the method cannot be applied to three dimensional dispersed bubbly flow. This is because void fraction is estimated by cross-section ratio of gas and liquid phases in this method. In the present study, Maxwell’s theory and polarization method are applied to calculation method of void fraction from voltage measured by constant electrical current method, and we try to measure void fraction in dispersed bubbly flow. Maxwell’s theory is a calculation of a mixture with two materials of different conductivity. The polarization method proposed in this study assumes bubbles to be dielectric bodies and void fraction is estimated by the polarization electric field of the bubbles.

In the experimental results, the void fractions in the three dimensional dispersed bubbly flow can be estimated with Maxwell’s theory and the polarization method. The void fractions estimated with these methods are more accurately than the previous method. Furthermore, it is experimentally clarified that the present proposed method can follow highly temporal void fluctuations of bubbly and froth flows. In addition, effects of intervals between electrodes and structures of electrodes are experimentally investigated.
Preliminary Study on Features of Key Components for HTGR Helium Turbine System
Authors: Hao Haoran , Yang Xiaoyong , Wang Jie
Published: October 2013

High temperature gas cooled reactor (HTGR) is featured by inherent safety. With high outlet temperature, HTGR is of potentially economical competitiveness. Helium turbine system can take advantage of high temperature helium to achieve high efficiency as well as compact structure, with development perspective. Key components of HTGR helium turbine system include rotary machinery such as turbo compressor and motor/generator, and heat transfer components (i.e. precooler, intercooler, recuperator and nuclear reactor). Due to the closed cycle, the features of key components are not independent. This paper preliminary explores the inherent relationships among features of key components, especially changing with reactor power. For heat transfer components, the general relationship between pressure drop and the effectiveness of heat transfer is analyzed. Furthermore, it is found that the pressure drop in recuperator, precooler and inter-cooler depends on the reactor power with engineering constraints. Besides, There exists the approximate squared relation between power and pressure drop in heat exchangers. For rotary machinery, it is found that the moment of inertia is the function of reactor power. For example, the moment of inertia of motor/generator is proportional to the reactor power. The relationship between moment of inertia of turbo-compressors and reactor power is more complex. This study is expected to be helpful to the design of key components for HTGR helium turbine system and can be applied to analyze the cycle’s dynamic characteristics.
Subcooled Flow Boiling in a Rectangular Channel With Added Turbulence and Longitudinal Vortices
Authors: Bloch Gregor , Jochum Christina , Schechtl Tobias , Sattelmayer Thomas
Published: October 2013

Experiments are conducted to analyze the influence of turbulence and secondary flows on heat transfer and CHF in sub-cooled flow boiling. Inserts creating turbulence and stationary vortices are placed below a vertical channel with a heated wall and upward flow direction with flow velocities up to 1.2 m/s. The boiling chamber is of square shape with inner dimensions of 40 × 40 mm2. Boiling regimes range from onset of nucleate boiling up to fully developed film boiling. Influence of the inserts is measured for varying flow velocities and subcooling from 4 K to 27 K. Flow parameters are measured with Particle Image Velocimetry (PIV). A decay of nearly isotropic turbulence within only few diameters is observed, while stationary swirls exhibit longer penetration depths. Boiling experiments are conducted with unsteady heating with a low boiling hydrocarbon (dodekafluoromethylpentanone) as working fluid. Results from boiling experiments show a positive influence of the inserts on the boiling process, increasing with higher subcooling and flow velocities.
CANDU 6 Severe Accident Prevention and Mitigation Features
Authors: Ford Steven , Lekakh Boris , Choy Ed , Verma Kamal , Ghelbereu Sorin
Published: October 2013

The CANDU 6 design includes features, both engineered and inherent, that act as barriers to prevent and mitigate severe accidents at progressive stages of a beyond design basis event such as that which occurred at Fukushima in March 2011.

CANDU 6 has ample design margins including multiple layers of defense. Large inventories of water slow down any accident progression to severe accident conditions, even when multiple failures are assumed; giving operations staff more time to manage the event.

Ongoing improvements to operating plants, and enhancements made to future evolutions of the CANDU design (including the Enhanced CANDU 6) improve upon these inherent features, further strengthening the CANDU 6 design to withstand severe core damage accidents.
Probability of the Loss of Offsite Power and Damage to Road Network due to a Rare Event
Authors: Cavaluzzi Jack , Gilmore Chase , Khan Bilal , Tran Minh Hong
Published: October 2013

The main goal of the project was to better understand the impact to a nuclear power plant due to the unavailability of critical infrastructure. We evaluated the use of rare event analysis to establish rare event occurrences in the vicinity of the plant. For the purpose of this report, rare events were considered extreme scenarios of natural disasters. The initial data was acquired by viewing STP FSARs and topographical maps of the region. The two specific factors that were extracted by analyzing the initial data of the plant design and location were the road networks and the electrical grid system. These two variables were then analyzed in light of the rare event analysis model that was used as a building block for this project. The rare event analysis also provided information regarding the types of data and distributions that were likely to be seen as a result. For the road networks, the road layout around the plant was mapped and the relevant data was used to consider the possible routes to the plants in case of a rare event. Data regarding LOOP Events were gathered mainly from NRC Documents and Institutional Sources. These data were used to analyze how many events occurred per year and the downtime associated with the rare event.
Failure of Modern Generators
Authors: Maughan Clyde V.
Published: October 2013

Making generators at lower cost has had a significant negative impact on the current fleet of generators, units less than about 30 years old. A number of factors have come into play, including the “reinventing old problems”, designs that push duties to higher and uncharted levels, and pressures to manufacture new machines more quickly and with less costly materials and processes.

Couple these competitive market realities with reduction in number of engineers in OEM organizations and the loss of institutional knowledge as elders have retired, and it is not surprising that some machines are failing much sooner than historically expected. This paper will present examples of some of the more critical missteps for various 3600 (3000) and 1800 (1500) RPM generators and propose practical maintenance approaches for power plant engineers.
Power Distribution and Fuel Centerline Temperature in a Pressure-Tube Supercritical Water-Cooled Reactor (PT SCWR)
Authors: Peiman W. , Saltanov Eu. , Grande L. , Pioro I. , Rouben B. , Gabriel K.
Published: October 2013

SuperCritical Water-cooled nuclear Reactor (SCWR) designs are one of six nuclear-reactor concepts being developed under the Generation IV International Forum (GIF) initiative. A generic pressure-tube SCWR consists of distributed fuel channels with coolant inlet and outlet temperatures of 350 and 625°C at 25 MPa, respectively. Such reactor coolant outlet conditions allow for high thermal efficiencies of SCW Nuclear Power Plant (NPP) of about 45–50%. In addition to high thermal efficiencies, SCWR designs provide the means for co-generation of hydrogen through thermochemical processes such as the Cu–Cl cycle.

The main objective of this paper is to determine the power distribution inside the core of an SCWR by using a lattice code - DRAGON and a diffusion code - DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermal-hydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature for UO2 and UC nuclear fuels. Results of an analysis showed that the fuel centerline temperature of UC was significantly lower than that of UO2. This paper also studies effects of energy groups on multi-group diffusion calculations and proposes nine energy groups for further neutronic studies related to SCWRs.
The Benefits of Using a Risk-Informed Approach to Resolving GSI-191
Authors: Sande Timothy D. , Zigler Gilbert L. , Kee Ernie J. , Letellier Bruce C. , Grantom C. Rick , Mohaghegh Zahra
Published: October 2013

The emergency core cooling system (ECCS) and containment spray system (CSS) in a pressurized water reactor (PWR) are designed to safely shutdown the plant following a loss of coolant accident (LOCA). The assurance of long term core cooling in PWRs following a LOCA has a long history dating back to the NRC studies of the mid 1980s associated with Unresolved Safety Issue (USI) A-43. Results of the NRC research on boiling water reactor (BWR) ECCS suction strainer blockage of the early 1990s identified new phenomena and failure modes that were not considered in the resolution of USI A-43. As a result of these concerns, Generic Safety Issue (GSI) 191 was identified in September 1996 related to debris clogging of the ECCS sump suction strainers at PWRs. Although plants have taken steps to prevent strainer clogging (by increasing the screen area, for example), satisfactory closure of this issue has proved elusive due to long term cooling issues and the effect of chemical precipitates on head loss. Previous investigators have identified bounding scenarios using conservative inputs, methods, and acceptance criteria. The acceptance criteria are applied in a “pass/fail” fashion that ignores risk. That is, if the results are acceptable, the issue has been resolved. Otherwise, it is necessary to either redo the analysis with partial relaxation of analytical conservatisms or perform additional plant modifications to ensure that the acceptance criteria are met. This article describes a new approach to close out the GSI-191 issue by evaluating the risk associated with ECCS performance on post-LOCA core cooling as a basis to change the plant license. The approach includes an assessment of LOCA frequencies as a function of break size at locations along the reactor coolant system, as well as a full quantification of the uncertainties associated with LOCA frequencies and the generation, transport, accumulation, and subsequent impact of debris on ECCS performance. The overall frameworks for the deterministic and risk-informed approaches are summarized with emphasis on the risk-informed method. The differences between the deterministic approach taken in the past and the new risk-informed approach are described. Advantages and disadvantages between the two methods are described and contrasted for the GSI-191 issue. The South Texas Project (STP) GSI-191 risk-informed closure efforts are presented.
Experimental Investigation of Nucleate Boiling on Capillary Tubes Under PWR Specific Subcooling and Flow Parameters
Authors: Schneider C. , Hampel R. , Traichel A. , Hurtado A. , Meissner S. , Koch E.
Published: October 2013

During full power operation of Pressurized Water Reactors (PWR), heat transfer phenomena of subcooled nucleate boiling may occur on the surface of the fuel rods. Despite high subcooling, this behavior results from the high heat flux up to 100 W/cm2 where vapor bubbles condensate when they are detached from the rod surface. In case of an accident with disturbance of cooling during transition from bubble to film boiling the critical heat flux (CHF) can be reached.

This paper outlines the experimental investigation of heat transfer during subcooled flow boiling on a capillary tube. To investigate the heat transfer processes under these boiling conditions, a test facility for flow boiling with access for optical measuring methods was constructed. The temperature is measured with a thermocouple inside the tube while boiling bubbles are generated on the outside. For different subcooling and flow velocity the heat flux is increased in a range from zero up to approximately 115 W/cm2. The major aims of these investigations are to generate a database for modeling of these dependencies in computational fluid dynamic (CFD) codes and enhance the knowledge of phenomenological effects of subcooled flow boiling. This provides a contribution for the prediction of the critical heat flux with simulation codes.
Use of Risk-Management Methods in NPP Safety Regulation
Authors: Taranovski A. V.
Published: October 2013

The object of research is risk management method in nuclear power plant (NPP) safety regulation.

Regulations, which are functioning in Ukraine, are based on the deterministic approaches and criteria. They proved themselves in the period of lack of experience and knowledge about mechanics of material aging. Despite the fact that such regulations are providing necessary safety levels, sometimes deterministic approaches are leading to redundant conservatism of specifications. This is driving to safety-related negative consequences, which are equipment-accelerated deterioration and increased radioactive influence on operative and maintenance staff. Now there is a situation, when new requirements, based on the increasing experience, are arising. So the costs are rising for getting compliance of those requirements.

From this point of view, an urgent problem is what tools and methods of safety analysis we need to use regulating safety levels of nuclear facilities. This work is aimed to improve safety, reliability, and efficiency of NPP operation. It is supposed, that the goal will be achieved by developing and adjusting risk-management methods for the NPPs in Ukraine.
Comparison of Transient Responses for SCWRs With Different Flow Path Designs Under Loss of Flow Accident
Authors: Zhu Dahuan , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: October 2013

Many various supercritical water-cooled reactor (SCWR) conceptual designs have been proposed by different researchers in last few years. However, the research on comparison of these concepts to judge the advantage and shortcoming of each conceptual design has little been carried out. This study is devoted to comparing transient flow characteristics and the consequent safety performance of SCWRs with different flow path design. Three different SCWR concepts such as SCWR with mixed spectrum core (SCWR-M), American SCWR design and Simplified Supercritical Water-Cooled Reactor (SSCWR) are respectively selected as the representations of multi-pass core, one-pass core with water rods and one-pass core without water rods. The comparative analysis of loss of flow accident is performed for the three SCWR concepts by new developed code TACOS. The flow rate and cladding temperature behaviors following the same transient assumptions are compared to each other. The results show that the SCWR-M and the American SCWR concepts have more complicated flow characteristics than the SSCWR. Inverse flow occurs both in the SCWR-M and the American SCWR when pump coastdown ends. The effect evaluation of Auxiliary Feedwater System (AFS) flow rate indicates that the flow path design of the SCWR-M would require larger injection flow rate than the other two concepts. The comparison indicates that the flow path design of SSCWR could adapt better for the transient conditions than the other two concepts.
Characteristics Analysis of SCWR During Partial Loss of Reactor Coolant Flow Transient
Authors: Chen Juan , Zhou Tao , Hou Zhousen , Sun Canhui
Published: October 2013

Partial loss of reactor coolant flow is one of the most important transients for safety analysis of supercritical water-cooled reactor (SCWR). Taking the super LWR concept provided by Japan as research object, transient analysis of partial loss of coolant flow rate is given by coupled neutronics and thermal hydraulics calculation method. The results show that, when 5% partial loss of coolant flow is happening, maximum cladding temperature would increase firstly with the decreasing of fuel channel inlet coolant flow. Then followed with the neutronic feedback and control operation, maximum cladding temperature decreases and finally return to normal. When 50% partial loss of coolant flow is happening, a scram signal will be given to ensure system safety, but the maximum cladding temperature still shows a significant increase early. On this basis, sensitivity analysis is performed considering the influence of core power and main coolant flow. It is found that maximum peaking value increases significantly following the initial flow rate decreasing, but shows a very little increase caused by core power increasing.
A Compressible Three-Dimensional Inverse Design Method Based on the Streamline Curvature Approach and Clebsch Formulation for Radial and Mixed Flow Turbomachines
Authors: Tian Xiao Pei , Shan Peng
Published: October 2013

The through-flow inverse design method based on the streamline curvature approach is nowadays a widely used quasi-3-dimensional blades design method for radial and mixed flow turbomachines. The main limitation of this method is using the flow field on the mean stream surface S2,m to approximate the actual 3-dimensional flow field. Without an effective description of the periodic flow, it is impossible for this method to realize exactly the prescribed circumferentially averaged swirl rVθ. Is there any way to develop this classical through-flow inverse method to a 3-dimensional one conveniently? The answer is yes. A new compressible 3-dimensional inverse design method for radial and mixed flow turbomachines is presented in this paper. This new 3-dimensional inverse method provides a convenient and effective way to obtain the periodic flow field for the streamline curvature through-flow inverse method. Meanwhile, compared with another type of similar 3-dimensional inverse method firstly described by Tan etc. based on Stokes stream functions and Monge potential functions from the Clebsch formulation to calculate the circumferentially averaged flow and the periodic flow respectively, this new method has its own advantages. In order to assess the usefulness of the new method, four centrifugal impellers are designed under the same design specifications by four different inverse methods respectively. They are two quasi-3-dimensional streamline curvature through-flow inverse methods without and with a slip factor model, a 3-dimensional approximated inverse approach based on stream functions and Monge potential functions and the 3-dimensional inverse method presented here. The performances of the four impellers yielding from a RANS commercial solver are compared. The capabilities of the four methods to realize the target circumferentially averaged swirl are also studied.
Analysis of the Onset of Dryout Conditions in Diabatic Annular Two-Phase Flows
Authors: Anglart Henryk
Published: October 2013

An analysis is presented that predicts the conditions which allow for a formation of a stable dry patch in diabatic annular two phase flows. The analysis employs a force balance formulated for the leading edge of the liquid film. In addition to stagnation, thermo-capillary and vapor thrust forces, the analysis includes effects of the pressure gradient and the interfacial shear stress. It is shown that the equilibrium conditions of a dry patch are dominated by the stagnation force, the surface tension force, the capillary force and the skin drag force. For high heat flux conditions only the first three forces are important.
Underground Siting of Nuclear Power Plants: Insights From Fukushima
Authors: Kunze Jay F. , Mahar James M. , Giraud Kellen M. , Myers C. W.
Published: October 2013

Siting of nuclear power plants in an underground nuclear park has been proposed by the authors in many previous publications, first focusing on how the present 1200 to 1600 MW-electric light water reactors could be sited underground, then including reprocessing and fuel manufacturing facilities, as well as high level permanent waste storage. Recently the focus has been on siting multiple small modular reactor systems. The recent incident at the Fukushima Daiichi site has prompted the authors to consider what the effects of a natural disaster such as the Japan earthquake and subsequent tsunami would have had if these reactors had been located underground. This paper addresses how the reactors might have remained operable — assuming the designs we previously proposed — and what lessons from the Fukushima incident can be learned for underground nuclear power plant designs.
Neutron Generation Based Method for Monte Carlo Three-Dimensional Reactor Time-Dependent Simulation
Authors: Xu Qi , Yu Ganglin , Wang Kan
Published: October 2013

A novel Monte Carlo time-dependent simulation method, named neutron generation based method (NGBM), is proposed for three-dimensional reactor dynamic analysis. Different from the traditional direct simulation method (DSM) based on neutron history, the new method, originating from the process of Monte Carlo criticality calculation, is based on neutron generation. In order to turn the original static calculation into a dynamic one, the time mark and time-dependent flux tally are added, the weight of neutron is adjusted while accumulating the flux estimator and the criteria for ending simulation is set. This new method is of higher computing efficiency than the direct simulation method for super-critical time-dependent situation, because it is able to take the advantage of Monte Carlo criticality calculation to keep the number of neutrons per generation approximately constant while the direct simulation method cannot stop the exponential increase of neutron population. The new method was integrated into RMC (Reactor Monte Carlo code developed by Tsinghua University). A numerical experiment was performed. The results demonstrate the feasibility and accuracy of the neutron generation based method for reactor dynamic analysis. The relative deviation of the time-dependent neutron flux tells that the accuracy of the neutron generation based method is enough for routine reactor safety analysis. And the experiment also shows the high efficiency of this method for super-critical reactor systems, since in the experiment, RMC runs nearly 7 times faster than MCNP which uses the direct simulation method.
The Screening Methodologies and/or Achievement Evaluation in Japanese FR Cycle Development Program With the Changing Needs for Evaluation
Authors: Shiotani Hiroki , Uto Nariaki , Kawaguchi Koichi , Shinoda Yoshihiko , Ono Kiyoshi , Namba Takashi
Published: October 2013

This paper argues the characteristics evaluation of Fast Reactor and fuel cycle concepts in the FS “Feasibility Study on commercialized fast reactor cycle systems” and the achievement of the performance evaluation conducted in FaCT (Fast Reactor Cycle System Technology Development) project in Japan.

At the beginning of FS phase-I (JFY1999), a combinatorial number of candidate concepts with innovative technologies were screened. After the FS phase-I evaluation, the several promising FR cycle concepts were selected to achieve consistency between FR system and fuel cycle system.

Analytical evaluation methodologies were developed to compare candidate FR cycle concepts in the FS phase-II (JFY2001 to JFY2005). Finally, the main concepts were decided mainly based on the technical summary (a kind of qualitative evaluation) of FR cycle concepts besides quantitative evaluation in FS phase-II. Sodium cooled FR combined with simplified pelletizing fuel fabrication, and advanced aqueous reprocessing was selected as “main concept” after Japanese government review.

In the FaCT phase-I (JFY2006 to JFY2010), the achievement toward the performance criteria / design requirements for the conceptual design of the commercialized facility developed in the FaCT phase-I was also evaluated. However, the objectives of the evaluation were to confirm the direction and the problems of R&D plans, development targets and design requirements. No aggregation methodologies were used in the FaCT final evaluation. Since the design requirements were set as challenges to achieve higher performance, although some items of respective facilities showed an insufficient achievement, a comprehensive evaluation determined that the performance criteria set by the JAEC were achieved in general.

The appropriate methodologies to evaluate the total FR and related fuel cycle system varies at different periods, according to our experience in Japanese FR cycle development program. This paper argues the way of methodologies with the changing needs and objectives of evaluation.
Study of Droplets Behaviors in the Steam-Water Separator With Monte Carlo Method
Authors: Zhang Huang , Bo HanLiang
Published: October 2013

The Monte Carlo method is used to study the behaviors of droplets in the steam-water separator(SWS). Different from previous researches, this paper presents the significance of droplets collision in the SWS. Our major objectives focus on the outcomes of binary-droplet collision. Three regimes of binary-droplet collision including coalescence, reflexive separation and stretching separation are considered. Then we prove that the relative velocity and initial positions of the binary-droplet determine the outcomes of collision at the same ambient atmosphere and properties of droplets(e.g. viscosity and coefficient of surface tension). The probability distribution of the collision outcomes of the binary-droplet associated with average radius in the condition of both cold and actual operating state in the SWS is performed by the Monte Carlo method.
Evaluation of Debris Transport Augmentation by Turbulence During the Recirculation Cooling Phase of a LOCA
Authors: Park Jong-Pil , Choi Kyung Sik , Jeong Ji Hwan , Jeong Jae Jun
Published: October 2013

The determination of the debris transport fraction during a LOCA of a PWR is very important in the sizing of the sump screen area. In this study, the debris transport fraction during the recirculation cooling phase is evaluated with and without consideration of turbulence effect. To do this, first experiments involving tumbling velocities measurements and supplementary CFD analyses are performed to verify the turbulence effect on debris transport. From these findings, the turbulence effect on the degree of debris tumbling augmentation was found to be represented by the algebraic sum of the mean velocity and the fluctuating velocity. Then, a CFD analysis of the flooding containment floor during the recirculation cooling phase of the OPR1000 plant is performed. Based on these studies, the debris transport fraction is evaluated for NUKON. The result shows a considerable increase in the debris transport fraction when a turbulence effect is implemented compared to when it is not. Increases of 5.55 and 2.06 times are observed for large NUKON and small/fine NUKON, respectively. This result implies that the turbulence effect should be considered in the debris transport quantification for conservatism.
An Experimental Study on Fracture Toughness of a Fine-Grained Isotropic Graphite
Authors: Wang Hongtao , Sun Libin , Wang Haitao , Shi Li , Zhang Zhensheng
Published: October 2013

Graphite is a brittle, polygranular material, which brings difficulties to the research of its fracture toughness. In the present paper, an experimental work on measuring the fracture toughness of graphite IG-11 is described. Three-point bending tests were employed. The effects of graphite specimen size, thickness, and the relative crack length a/W are studied in the tests. The deformation and strain near the crack tips are measured by both the strain gauge method and the digital speckle correlation method (DSCM). The results indicate that the fracture toughness of graphite IG-11 is in a range of 0.82∼1.27 MPa.m1/2, with an averaged value of 1.038 MPa.m1/2.
Tank Diaphragm Kinematic Behavior and Conditions for Safe Tank Withdrawal
Authors: Haber Ludwig C. , Humphrey-Facendola Amie
Published: October 2013

Power plants, including nuclear power plants regularly employ tanks whose contents need to be kept isolated from atmospheric conditions. One way to satisfy these requirements is to provide a liner for the tank which completely fits the interior shape of the tank but floats on top of the tank contents when the tank contains fluid. As the volume in the tank changes, the liner or diaphragm accommodates the changes in volume by sliding along the tank walls. To allow free movement of the diaphragm, management of the gas volume above the fluid and behind the diaphragm is of prime importance. The work described in this paper elaborates on the conditions required to prevent the tank diaphragm from becoming damaged. To develop potential failure modes, the kinematics of the diaphragm and the interaction with the gas volume between the diaphragm and the tank fluid are considered in detail. The developed model is applied to the case of a condensate storage tank at Comanche Peak Nuclear Power Plant (CPNPP). Two physical scale models of the tank were constructed and tested to validate the model and allow the safe operation principles to be quantified for use in the operation of the condensate storage tank at CPNPP. The work allowed CPNPP to design appropriate periodic checks and maintenance activities to ensure the diaphragm will not be damaged due to tank volume changes while still ensuring the required water chemistry criteria for the tank contents can be met.
Reliability Analysis for Passive Containment Cooling System in Seismic Hazard
Authors: Yu Yu , Wang Shengfei , Lv Xuefeng , Niu Fenglei
Published: October 2013

Passive containment cooling system is an important safety-related system in AP1000 nuclear power plant, by which heat produced in reactor is transferred to the heat sink–atmosphere based on natural circulation. So it is important to the plant safety whether the system can work well or not in the seismic hazard. Since the system operation is independent of human interfere or the operation of outside equipments, the reliability of the system is improved, however, physical process failure become one of the important contributors to the system operation failure because natural circulation may not keep when the system configuration is different from the design. So it is necessary to analyze the system reliability in seismic situation. The equipment failure probability under earthquake is a function of the peak ground acceleration which is stochastic, and the fault tree method used in traditional probability safety assessment (PSA) for system reliability analysis is not power enough to deal with conditional probability. In this paper, a new analysis method for system reliability evaluate at seismic situation based on Monte Carlo (MC) simulation is put forward, and annual failure probability of passive containment cooling system in AP1000 in seismic hazard is calculated, the result is according with the AP1000 seismic margin evaluation.
Power Coefficient of Reactivity Determination for HTPBR and its Application for Reactivity Initiated Transients
Authors: DinkarBadgujar Kushal , Singh Om Pal , Singh Suneet , Rewankar Shripad T.
Published: October 2013

Safety evaluation is one of the most important aspects of the nuclear reactors. Here some safety studies have been carried out on High Temperature Gas cooled Reactors (HTGR) with fuel in the form of pebbles. This is one of the six types of GEN IV reactors that are being developed internationally with emphasis on features like inherent safety, nuclear proliferation resistance and high thermal to electric conversion efficiency. This work pertains to the dynamic simulation of the reactor with, as well as without reactivity feedback. It involves the solution of point kinetics equations with reactivity feedback arising from the power rise. First, the reactivity feedback arising from the power rise is calculated in the form of power coefficient of reactivity. The steady state temperatures are calculated at two different steady state power levels of the reactor using the heat balance between the fission power produced in the reactor and the heat removed by the coolant. A simplified model of the reactor and lumped model of heat transfer is developed and used. The temperature rise in going from one power level to other power level is calculated. This is multiplied by typical value of the temperature coefficient of reactivity available in the literature. The reactivity changes divided by the power rise provides the power coefficient of reactivity. For performing dynamic analysis of the reactor, the kinetics equations along with feedback are solved. The methodology is verified against the analytical expressions available. Simulations are carried out for the case of raising the power of the reactor from low power to high power and transients at full power and shutting down of the reactor by reactor SCRAM. It is observed that for raising the power of the reactor, the reactivity addition beyond 20pcm/second may not be acceptable as the higher reactivity addition rates result in lower instantaneous reactor period resulting in SCRAM of the reactor. This information is useful in designing the short term and long term decay heat power removal system. Moreover, it provides insight on the determination of the maximum permissible reactivity addition rates in the reactor and optimization of the reactivity feedback due to rise in power.
Study on Saturated Flow Boiling Heat Transfer Under Vibration Conditions
Authors: Fujiyama Yoshitaro , Ohtake Hiroyasu
Published: October 2013

The ability to predict void formation, void fraction and critical heat flux —CHF— in flow boiling under oscillatory flow and vibration conditions is important to the safety technology of nuclear reactor during earthquake. In the present study, the onset of nucleate boiling —ONB— and CHF on saturated flow boiling under vibration conditions were investigated experimentally.

Steady state experiments were conducted using a copper thin-film and saturated and subcooled water at 0.1 MPa. The liquid velocity was 0.25, 1.38, 3.20 and 4.07 m/s, respectively; the liquid subcooling was 0 K and 20 K. A heater was made of a printed circuit board. A test section was a rectangular flow channel of 10 mm width and 10 mm height. The test heater was heated by Joule heating of d.c. current from a low-voltage high-current stabilizer. The heating rate of the heater was determined from supplied current and voltage. The temperature of the heater was obtained by referring to the measured electric resistance. The test section was arranged for horizontal position facing upward and for vertical position, respectively. For the vibration condition, the test section was set on a vibration table. The ONB was decided as an occurrence of the first boiling bubble. The critical heat flux was determined as that immediately before the heating surface physically burned-out.

The CHF on saturated flow boiling under vibration conditions were investigated experimentally.
The Effects of Cycling on Generator Rotors
Authors: Moore Bill
Published: October 2013

Cycling can cause accelerated deterioration in generator rotors. EPRI (Electric Power Research Institute) states that “Cyclic operation can result in an increase in forced outage rates, higher operation and maintenance (O&M) costs, and further wear and tear on components…” Most industry references do not differentiate between speed cycling and load cycling. Speed cycling is the worst kind of cycling, leading to cracking and fatigue failure of many generator rotor components.

The outline of the paper will include the following major topics:

• Definition of Cycling

• Focus on Speed Cycling versus Load Cycling/Load Following

• Low Cycle versus High Cycle Fatigue -

• Fatigue progresses to cracking; cracking to failure, failure to forced outages, schedule delays, additional costs

• Key Generator Rotor Components Affected by Low Cycle Fatigue

• Rotor Forging Shaft including the rotor bore, tooth tops, snap ring groove cracking

• Rotor Field Winding including pole crossovers, J-straps/terminal studs, bore copper, conductor end turns

• Support Components including slot wedges, retaining rings

• Preventative Action by Investigation and NDE/DE

• Conclusions and Recommendations.