bluuebunny/crossref_metadata_2025 huggingface dataset:
Wall Shear Prediction for Wavy and Annular Air-Water Two-Phase Flow
Authors: Schubring DuWayne , Ashwood Andrea C. , Shedd Timothy A.
Published: July 2009

Non-intrusive pressure drop, film thickness, and wave behavior measurements have been performed for horizontal air-water two-phase flow in round tubes. A databank of over 300 flow conditions in the wavy, wavy-annular, and annular regimes was compiled for three tube diameters (8.8 mm, 15.1 mm, and 26.3 mm). A correlation was developed based on the product of total mass flux, gas superficial velocity, and an empirical friction factor, selected from the lesser of a modified Froude number form and a Reynolds number form. This correlation performed to within 11% mean absolute error across the data range considered. A variety of correlations were also selected from the literature for comparison. Those based on film roughness (Wallis, Hurlburt-Newell), two-phase multipliers (Lockhart-Martinelli, Friedel, Chisholm, Olujic, and Gro¨nnerud), and empirical data fits (Mu¨ller-Steinhagen and Heck) were among those tested. As needed, correlations were rephrased in terms of wall shear to provide a consistent basis for comparison. A complex film roughness model (Hurlburt et al.) was also compared to the experimental data.
Two-Zone PIV Analysis of Wavy and Wavy-Annular Two-Phase Flow
Authors: Schubring DuWayne , Foster Rory E. , Shedd Timothy A.
Published: July 2009

Cross-sectional thin film micro-particle image velocimetry (cTFMPIV) analysis has been performed for air-water two-phase flow in a 19.05 mm horizontal duct. Images were captured by introducing a light sheet from a Nd:YAG laser into the flow, with the fluorescent emissions from three micron particles captured by a CCD camera behind a microscope objective and red filter. These images were grouped into pairs and digitally enhanced to isolate particle traces using commercial image processing software. Image pairs were collected for five air flow rates at each of four water flow rates and separated into base film and wave zones for PIV analysis. Using DaVis 7.1 (LaVision GmbH), vector maps were obtained for each pair. Less than 2% of the image pairs were removed due to clearly spurious vectors, while the appropriate zone for less than 8% of the image pairs could not be determined These pairs were grouped into a third batch. The vector maps were analyzed to calculate velocity profiles and turbulence (fluctuation) statistics in terms of wall coordinates. These statistics included intensities and Reynolds stresses. The base film shows laminar behavior beyond what would be expected from the universal velocity profile. The wave zone displays significantly different behavior and appears to reach an asymptote rapidly. Calculation of fluctuation intensities indicates a maximum within the wave zone along the gas-liquid interface, while the distribution of radial fluctuations in the base film and axial turbulence in all zones is approximately uniform. Calculation of the u′v′ shear Reynolds stress suggests that it has a negative value.
Development of an Innovative Plate-Type SG for Sodium Cooled Fast Reactor: Concept and Test Fabrication
Authors: Nishi Yoshihisa , Kinoshita Izumi
Published: July 2009

The concept of an innovative plate-type steam generator (SG) for a fast reactor, fabricated using hot isostatic pressing (HIP), was presented. The heat transfer plate, which is assembled with rectangular tubes and fabricated using HIP, is surrounded by a leak-detection layer. The optimum form for the detection layer was determined by crack extension analysis. The concept can be applied in both pool-type and loop-type liquid-metal cooled fast reactors (LMFR). In this report, the fabrication technique is described as applied to a loop-type LMFR. Optimum HIP conditions of [1423K (1150°C), 1200kgf/cm2 (117MPa), 3hours] for modified 9Cr-1Mo steel were determined from HIP tests, tensile tests, and structural inspection. Nickel-type solder (BNi-5) and gold-type solder (BAu-4) were examined as potential joining materials to laminate the heat transfer plates. Tensile test comparisons showed that BAu-4 was superior, so it was used. No problems were observed in the fabrication of a partial model of an SG (HIP of the rectangular tubes, brazing the plates, welding the header, etc.).
Study on Boiling Heat Transfer in Narrow and Flat Space
Authors: Oshikawa Takaaki , Sakurai Hisashi , Koizumi Yasuo , Ohtake Hiroyasu
Published: July 2009

Boiling heat transfer experiments were performed by using ethanol as test fluid. A heat transfer surface was at the bottom wall of the flow channel. Two types of experiments were conducted; horizontal-rectangular-narrow flow channel experiments and horizontal-narrow-flat space experiments. Experiments were conducted at the pool condition and at 0.1 MPa. In the rectangular-narrow flow channel experiments, the width of the flow channel was varied in the range of 1.0 ∼ 2.0 mm and the height of the flow channel was varied in the range of 0.5 mm ∼ 8 mm. The length of the flow channel was 30 mm. The diameter of the heat transfer surface was the same as the width of the flow channel. Both ends of the flow channel were opened to wide space filled with liquid. In the experiments of the narrow-flat space experiments, the heat transfer surface was at the center of the flat space of 20 mm × 20 mm or 30 mm × 30 mm. The size of the heat transfer surface diameter was 3.0 mm. The space was varied in the range of 0.5 mm ∼ 8 mm. Pool-type experiments of no flat space were also conducted for comparison. The circumference of the narrow-flat space was opened to wide space filled with liquid. In the rectangular-narrow flow channel experiments, a large bubble periodically left from the heat transfer surface and moved to the both ends of the flow channel. As the heat flux was increased, the critical heat flux condition was finally reached. The critical heat flux decreased with a decrease in the flow channel height. When the flow channel height was large, the critical heat flux was close to the pool boiling value. In the experiments of the narrow-flat space experiments, when the space was wide, bubbles generated on the heat transfer surface left freely from the flat space. The heat transfer characteristics were close to those of the pool boiling. As the space became narrow, a large bubble sat on the heat transfer surface, which resulted in the critical heat flux condition. When the space was narrow, the initiation of the bubble generation on the heat transfer surface immediately resulted in the critical heat flux condition.
Validation of FLUENT for Prediction of Flow Distribution and Pressure Gradients in a Multi-Branch Header Under Low Flow Conditions
Authors: Muhana A. , Novog D. R.
Published: July 2009

Flow headers are commonly used in nuclear reactors, boilers and heat exchangers to distribute fluid to branches or to combine flow from the branches along the header. In CANDU reactors the main heat transport system divides the flow from the pumps into approximately 120 individual feeder pipes which then direct the flow into separate fuel channels. Historically, nuclear safety analysis has been performed using one-dimensional averaged system codes, and as such the headers are cross-sectionally averaged. In this paper, flow distribution and pressure gradients along a multi-branch header have been predicted using the three dimensional computational fluid dynamics software FLUENT and were compared to results obtained from experimental data obtained from literature for single phase conditions. In order to assess FLUENTs capabilities this study was performed by comparing the predictions against separate effects experiments conducted on a smaller sized header available in literature. For these experiments, water inlet flow rate was varied and flow rates in the header branches were measured. The aim of this work is to validate FLUENT software for predicting flow distribution and pressure gradients in single phase flow in such a multi-branch geometry. The effects of flow model, grid density, convergence criteria, flow inlet velocity and header size on the computational results were studied. Vortex formation and flow separation were also studied and compared to the experimentally observed flow behaviour.
Electrical Insulation System Degradation Sensors: Improving Reliability of Power Generation and Distribution
Authors: Watkins Kenneth S.
Published: July 2009

As insulation systems of power system components such as electrical motors, generators and transformers degrade, they become brittle, crack and, eventually, fail to perform their intended function. Failure of the insulation system of these components often leads to costly power interruptions that could be prevented if the actual condition of the insulation system is known. The degradation mechanisms of modern insulation systems are highly dependent on the actual environmental and operational conditions of the component. Current methods to measure insulation system condition are often complex, expensive and require specialized training to interpret. In contrast, conductive composite sensors made of the same polymeric components as the insulation system itself monitor the actual environmentally and operationally induced degradation of the component insulation and provide a quick, objective indication of the current condition and remaining design life of the insulation. This innovative technology utilizes low-cost, inert conductive particles compounded with a portion of the insulation polymer to provide a tiny degradation sensor embedded into the winding, core or stator of the component. Sensor output correlates with the degraded state of the insulation system relative to standard industry thermal endurance testing, giving advanced warning of a degraded condition of the insulation system before design conditions are exceeded. Maintenance personnel, utilizing a simple ohmmeter, can read sensor output quickly and reliably without specialized equipment or training. Alternately, threshold-warning devices connected to the sensor provide constant monitoring. Conductive composite degradation sensors provide advance warning of prematurely degraded insulation systems and reduce the need for complex, intrusive and sometimes destructive electrical testing. Because conductive composite degradation sensors require no electrical power during the aging process, they are ideally suited to wireless, passive radio frequency identification (RFID), and “smart label” technologies.
Comparison of Plant Response of Forsmark 3 Nuclear Power Plant to Forsmark 1 Type of Transient (25th July 2006), Using the HAMBO BWR Simulator: Thermal-Hydraulic Aspects
Authors: Pekkarinen Esko
Published: July 2009

Modernisation of control rooms of the nuclear power plants has been a major issue during the last few years. With this as a basis, the BWR plants in Sweden and Finland funded, in co-operation with the Halden Project, an experimental HAMBO BWR simulator project based on the Forsmark 3 plant in Sweden. VTT Energy in Finland developed the simulator models for HAMBO with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator and its performance have been described in other publications [1, 2]. On July 25th 2006 there was a short circuit at Forsmark 1 nuclear power plant when manoeuvring equipment in the 400kV-switch yard. Due to the short circuit, the plant suffered an electrical disturbance that led to scram and failure of two out of four diesel generators. The purpose of the study carried out at VTT in 2007 was to assess the capabilities of the HAMBO BWR simulator to handle Forsmark 1 type of events in different nuclear power plants (Forsmark 3 in this case). The Forsmark 1 incident showed (among other things) that the intention to protect certain components (in this case the UPS-system) can in certain situations affect negatively to the safety functions. It is concluded that most of this type of BWR transients may be simulated to a certain extent using the existing HAMBO- and APROS- models. A detailed modelling of the automation and electric systems is required sometimes if the complex interplay between these systems and the process is to be predicted reliably. The modelling should be plant specific and level of detail should be assessed case-by-case (i.e. what kind of transient is in question, what systems are available, what is the main purpose of the analyses etc.). The thermal-hydraulic models of the APROS-code seem to replicate well the real behaviour of thermal-hydraulic process provided that there is enough information about the transient in consideration.
Refurbishment of Secondary System and High AVT Water Treatment of Genkai #1 and #2
Authors: Manabe Jun , Shoda Yasuhiko , Yamamura Tatsushi , Kusumoto Yuuichirou
Published: July 2009

Kyushu electric co. Genkai #1 and #2 are twin 500 Mw class first generation PWR power stations starting their commercial operation in 1975–1981. The units were recently altered their secondary water treatment from AVT to HAVT (High All Volatile Treatment) operation aiming to suppress erosion in piping and equipment, resulting in feed water iron concentration reduction to around 1 ppb as indication of the effects. The units had been successfully operated from the start of their commercial operation except for scale adhesion to SG and others, degradation of copper alloy material tubes in auxiliary heat exchangers and lower condenser vacuum derived from protective ferrous sulfate coating. Life cycle management program was implemented resulting in the alteration of water treatment to HAVT adopting the SG blow down demineralizing and the replacement of copper alloy tube heat exchangers to stainless steel and titan tubes. Further more the examination results were introduced, of the scale adhesion mechanism in the high temperature region of the secondary system based on actual plants iron characterization data and field examination results of HAVT of Genkai units, expecting HAVT application would be effective for the scale adhesion reduction.
Computational Methodology for the Thermal-Hydraulic Behaviour of a Fuel Rod Surrounded by a Coolant Under PWR Conditions
Authors: Laszlo Jean-Louis , Bignalet Julien , Gounand Stephane , Gallo-Lepage Danielle
Published: July 2009

The present study deals with the thermal-hydraulic modelization of the heating part of a test loop containing a fuel rod cooled by forced convection under PWR conditions, while the pressure tube is cooled naturally by the water from an OSIRIS reactor open pool. A renovation policy of calculation schemes leads us to carry out these studies with a 3D CEA computer F.E. code, named CAST3M, coupling 3D thermal and 1D hydraulic phenomena. In addition to the liquid single phase character of the flow cooling the fuel rod, safety concerns are taken into account, such as the possibility of nucleate boiling regime, without affecting the main safety criteria. A few numerical results show the heat balance for different values of the main inlet relevant parameters.
3D Visualization in Industrial Monolayer CT Inspection System
Authors: Zhang Xi , Liu Jinhui
Published: July 2009

3D visualization plays an important role in industrial CT inspection system. Though at present it is not widely used in CT inspection systems, 3D images betray more visual pictures, and can make easier observation. Taking 3D-imaging methods in other fields, such as medical or geological field, as references, this paper introduces three feasible imaging algorithms, contour connecting method, marching cubes, and an improved marching cubes method basing on distance function. These three methods are all suited to small quantity of slice images, which is accord with most industrial applications. Contour connection has the most directive approach of rending, but has some unsatisfied defects on edge extraction and edge connection. Marching Cubes is a classic method of 3D visualization; it is feasible and commonly used in many realms. However, applied in industrial CT inspection systems, there are still some discontented effects, such as fracture, in certain condition. Accordingly, an improved marching cube method that based on edge extraction is brought forward to conquer these defects. Compare of these methods is given in this paper, and implemental instances are provided as well. It is of importance for the development of 3D visualization in air container and luggage CT inspection systems.
Radiation Induced Surface Activation on SUS304 Rod Surface Coated With Thin Film Layer Formed by Vacuum Evaporation Method
Authors: Zhang Jian , Mishima Kaichiro , Sugitani Naoki , Hino Masahiro , Takamasa Tomoji
Published: July 2009

In this study, the effect of film layer deposited on a metal surface by vacuum evaporation method on the RISA phenomenon was investigated from the view points of surface wettability and quenching velocity. Test rods were made of SUS304 with the outer diameter of 24mm and the length of 150mm. Four kinds of materials, i.e. titanium, zircaloy-II, germanium and silicon, were deposited separately on the half-circumference surface of a SUS304 rod with thickness of 200 nm. It was concluded that the RISA effect was observed on the film-coated surfaces after gamma-ray irradiation, which is similar to that on an oxide layer of SUS304 surface. It was also found that the quenching velocity on film-coated surfaces was much faster than that on an un-coated surface.
Remotely Operated Equipments for In-Vessel Inspection in BWRs
Authors: Yuguchi Yasuhiro , Naruse Katsuhiko , Shimamura Mitsuaki , Sato Takao , Maehara Takeshi , Iimori Hiroshi
Published: July 2009

This paper describes development and application of remotely operated equipments for reactor internals ultrasonic testing (UT) and visual testing (VT) in boiling water reactors (BWRs) to enhance the availability of operating nuclear power plants. Stress Corrosion Cracking (SCC) in reactor internal components causes relatively low operating rate of BWR as one of the most important issues since the 1990s, and demand to inspect the reactor internals become greater. The remotely operated equipments have been developed for in-vessel inspection of core shrouds, shroud supports, jet pumps and so on. The equipments are evaluated in the point of the positioning precision as an application of a phased array UT technique and visual length measurement system. And as results of application to the actual plants and improvement, a Shroud vehicle, a Crawler vehicle and a Visual position measurement system are enhanced in an aspect of the accessible range and/or positioning precision. Their configurations and functional testing results are shown in this paper. As the other application, micro replication devices are developed for CRD Stub tube and shroud in BWR. This paper indicates the functional testing result using the prototype for the CRD stub tube with observing material microstructure.
Pre and Post Test Analysis of LBLOCA Late Reflood Phase in ATLAS Using RELAP5/MOD3.3
Authors: Lee Seok Ho , Kim Han Gon
Published: July 2009

The code simulation capability for the KAERI (Korea Atomic Energy Research Institute) integral effect test facility, ATLAS (Advanced Thermal hydraulic Test Loop for Accidents Simulaion), has been assessed for LBLOCA (Large Break Loss Of Coolant Accidents) reflood test. IETs (Integral Effect Tests) using ATLAS facility for directing simulation of the prototype plant APR1400 LBLOCA reflood with the same control logics and accident scenarios have been performed in 2007. The experimental results show that Maximum temperature of heater rods is 722°C and the reflooding rate and quenching speed is ∼6.2cm/sec and ∼0.3cm/sec, respectively. Also, quenching time is about 650 sec from the beginning of reflood. The best-estimate code, RELAP5/MOD3.3 is chosen to examine the code prediction capability. In basecase analysis, initial safety injection water is not nearly injected into core through downcomer. As the result, the maximum rod temperature, 1400°C is nearly twice as high as the experimental data. And the quenching time is similar to the experimental data. We adjusted a hydraulic behavior in core and downcemer through the improvement of the adequate nodalization. However, the thermal behavior for maximum heater rod temperature in core still included the limitation. To adjust the thermal behavior of core heater, we have performed the several sensitivity analyses for heat transfer region in core. The results of sensitivity analyses showed that the dominant factor is the wall heat transfer model in film boiling region. Therefore, we developed and assessed the sensitivity analysis model to examine the effect of void fraction for thermal behaviors. As the result, void fraction within film coefficient correlation had a considerable effect on the temperature behavior of heater rods.
Evaluation of the Influence of Hydrides to the Stress Intensity Factor of Zirconium–2.5% Niobium Alloy
Authors: Janulionis Remigijus , Dundulis Gintautas , Karalevicˇius Renatas , Grybenas Albertas
Published: July 2009

The NPP containing RBMK-1500 type reactor is in operation in Lithuania. The RBMK-1500 is graphite moderated, boiling water, multi-channel reactor. Zr-2.5% Nb alloys are used as a constructional material for manufacturing claddings of both fuel assemblies and fuel channels (FC). Fuel channels of RBMK-1500 reactors are the major structural elements of the reactor core that have to meet strict requirements in terms of operational reliability. Therefore it is necessary to evaluate the influence of ageing mechanism on mechanical properties of zirconium alloys during operation of the reactor. Hydrogen absorption by zirconium alloy during corrosion process is the ageing mechanism of Zr-2.5% Nb fuel channel. When hydrogen concentration in fuel channel exceeds solubility limit, formation of hydrides under certain conditions can reduce resistance to brittle fracture and cause initiation and development of hydride cracks. Therefore the evaluation of the resistance to brittle fracture of zirconium alloy is important. The objective of this paper is modelling and assessment of the influence of hydrogen to the stress intensity factor of Zr - 2.5% Nb alloy using finite element method. The stress intensity factor is a parameter, which is used to estimate a material resistance to brittle fracture. Modelling of the stress intensity factor was performed using finite element method. The influence of the hydrogen concentration to stress intensity factor was evaluated. The prognosis results and experimental data are in close agreement. It was demonstrated that applied methodology for modelling of the influence of hydrogen on the stress intensity factor can be used.
External Events Importance for Safety of the Ignalina Nuclear Power Plant
Authors: Kupciuniene Kristina , Alzbutas Robertas
Published: July 2009

Probabilistic risk analysis of external events impact on the safety important systems of Ignalina nuclear power plant (NPP) is performed. At first, the methodology was established for screening out external events, which impact on Ignalina NPP safety is not significant. In order to estimate probabilities of external events occurrences the statistical data were collected, mathematical models were constructed and probabilities of these events occurrences were determined. For risk estimation, the following external events were studied: forest fires, external floods, aircraft crash, seismic events, extreme winds, fall of frequency in electricity network and loss of an external electricity supply. The aircraft crash event was modeled and analyzed in more detail. The crash probability estimation model is improved considering uncertain data. The aircraft crash probability on the territory of the Ignalina NPP with the radius r is expressed by the derived analytical formula. This formula is used to calculate the aircraft crash probability applying the most recent statistical data. The aircraft crash dependencies upon the amount of flights and crashes in the different radius territories were analyzed. Relationship of non-flying zone around Ignalina NPP and aircraft crash frequency was investigated. As a part of the initial conditions and parameters of aircraft crash model are not well-known or have different values for various types of aircrafts, the sensitivity analysis was performed for the model. This analysis determines the parameters that have the greatest influence on the model results. Conclusions about the importance of the parameters and sensitivity of the results are obtained using a sensitivity measures of the model under analysis.
Generic Issues Program Overview and Update
Authors: Foster Jack W. , Kauffman John V.
Published: July 2009

The United States Nuclear Regulatory Commission (NRC) has a Generic Issues Program (GIP) to address Generic Issues (GI). A GI is defined as “a regulatory matter involving the design, construction, operation, or decommissioning of several, or a class of, NRC licensees or certificate holders that is not sufficiently addressed by existing rules, guidance, or programs.” This rather legalistic definition has several practical corollaries: First, a GI must involve safety. Second, the issue must involve at least two plants, or it would be a plant-specific issue rather than a GI. Third, the potential safety question must not be covered by existing regulations and guidance (compliance). Thus, the effect of a GI is to potentially change the body of regulations and associated guidance (e.g., regulatory guides). The GIP was started in 1976, thus it is a relatively mature program. There have been approximately 850 issues processed by the program to date. More importantly, even after 30 years, new GIs continue to be proposed. The entire set of Generic Issues (GIs) is updated annually in NUREG-0933, “A Prioritization of Generic Safety Issues.” GIs tend to involve complex questions of safety and regulation. The efficient and effective means of addressing these issues is very important for regulatory effectiveness. If an issue proves to pose a genuine, significant safety question, then swift, effective, enforceable, and cost-effective action needs to be taken. Conversely, if an issue is of little safety significance, the issue should be dismissed in an expeditious manner, avoiding unnecessary expenditure of resources and regulatory burden or uncertainty. This paper provides an overview of the 5-stage program, from identification through the regulatory assessment stage. The paper also includes a discussion of the program’s seven criteria, sources of proposed GIs, recent improvements, publicly available information, historical performance, and status of current GIs.
Assessment of Dynamic Instability in Once Through Steam Generator
Authors: Prakash V. , Thirumalai M. , Murugesan P. , Vinod V. , Sureshkumar V. A. , Noushad I. B. , Rajan K. K. , Kalyanasundaram P. , Vaidyanathan G.
Published: July 2009

Hydrodynamic flow instability in Once Through Steam Generators (OTSG) is one of the important problems in the design and operation of Liquid Metal Fast Breeder Reactors (LMFBRs). Under certain operating conditions, water flow in OTSG is susceptible to instability due to the close coupling between the thermal and hydraulic processes. Sustained flow oscillations due to instability are undesirable since they result in flow mal-distribution among the tubes resulting in thermal stress, mechanical vibrations and system control problems. It is therefore, necessary to assess the operating conditions, under which instability occurs so that the system may be designed to operate always under stable conditions. The cause of the main type of instability, important for the design of SGs is the propagation of density waves. This type of low frequency instability is referred to in literature as parallel-channel, density wave, time delay or mass flow-void feedback instability. Dynamic instability (density wave oscillation DWO) occurs because of the phase mismatch between the primary perturbation (water flow) and the response to this perturbation (pressure drop). As many tubes are operating under essentially constant pressure heads, this mismatch can lead to sustained/diverging oscillations. Water flow oscillation in tubes manifests as oscillations in the steam temperature at the tube outlet/pressure fluctuations. However it is difficult to instrument individual tubes in SG for such measurement in an operating plant. If the flow oscillation in the tube manifests itself in the overall module flow, then fluctuation in the overall flow/flow noise could be utilized for on-line stability measurements. Towards this, experiments were conducted in the sodium heated once through steam generator in an OTSG model. To confirm the extent of oscillation in the steam temperature and in inlet water flow, 3 tubes out of 19, were monitored besides overall module flow. Main objective of the present study was to assess the occurrence of dynamic instability in SG through module inlet flow perturbations, measured by ΔP measurements across the orifice at entry to the tubes and steam temperature fluctuation measurement at the outlet of tube by bare thermocouples. This paper discusses the experiments carried out in the Steam Generator model of Prototype Fast Breeder Reactor to investigate the instability phenomenon, the instrumentation details, the results and its discussion.
Review of Mathematical Models for Pebble Dynamics
Authors: Hu Wenping , Jiang Shengyao , Yang Xingtuan
Published: July 2009

Pebble-bed nuclear reactor technology, with a reactor core typically composed of spherical pebbles draining very slowly in a continuous refueling process, is currently being revived around the world. But the dense slow pebble flow in the reactor, which has an important impact on reactor physics, is still poorly understood. Under such circumstance, this article studies mathematical models which are potential to research the pebbles motion in the pebble-bed reactor, including void model, spot model and DEM model. The fundamental principles of these models are introduced, the success and deficiency of each model is briefly analyzed. Theoretically, it’s expected that spot model and DEM model may be more practical to apply on studying the pebble dynamics. Though, spot model still needs to be refined based on further experimentation, and more research is necessary to solve the problem of huge computational time in order to make the DEM model simulation technique a really practical notion.
Axial Offset Anomaly Modeling Studies on EDF Cores
Authors: De´cossin Etienne , Bremnes O̸ystein
Published: July 2009

The Axial Offset Anomaly phenomenon, commonly called AOA, is one of the possible consequences of the undesirable presence of deposits on the nuclear fuel. AOA appears in PWR cores as abnormal distribution of the power, as compared to the design reference values. If the amplitude of the phenomenon becomes significant, it may lead to additional constraints in operating the reactor. Several factors contribute to the root cause of AOA. The state-of-the art knowledge relies on the fact that deviation should appear when the conditions of temperature, boron and corrosion product concentrations are appropriate to form crud on the fuel surface, that is thick enough to allow precipitation of lithium borates. Then, the neutron capture is locally enhanced by the additional presence of boron, leading to local flux depression and redistribution of the power all over the core. The experience feedback on AOA is rather consequent, mainly in the United States. On the EDF nuclear fleet, excepted two notable cases in the mid 90’s, the phenomenon remained limited to a few, low amplitude observations on the 1300 MW type cores. To perform AOA studies, EDF has developed its own numerical tool based on: • the neutron kinetics COCCINELLE code for power distribution computations, • the thermal-hydraulics THYC code, • the dedicated BOA code to evaluate, at the core scale, crud deposition and boron loading. The EDF software COCCINELLE and THYC are commonly used for core design and safety analysis. For AOA studies, they provide a 3D, best-estimate representation of the clean core. Then the thermal-hydraulic data are used by the BOA code as boundary conditions to determine, in both single-phase flow and sub-cooled nucleate boiling, how and where deposits form on the fuel surface. In the standard approach, the total mass of trapped boron is compared to a threshold defining the AOA risk limit. In a prospective approach, an advanced COCCINELLE, THYC and BOA coupling is proposed to account for the power distribution changes all along the cycle, instead of using the clean core data. Numerical simulations of an AOA cycle of a 900 MW core show that this feedback effect has a visible effect on the final axial-offset prediction.
Research on Two-Phase Flow Instability in Parallel Multi-Channel Under Rolling Motion Condition
Authors: Zhang Y. J. , Su G. H. , Qiu S. Z. , Yang X. B.
Published: July 2009

Two-phase flow instability of the parallel multi-channel system has been studied under rolling motion condition in this paper. Based on the homogeneous flow model with considering the rolling motion condition, the parallel multi-channel model is established by using the control volume integrating method. Gear method is used to solve the system equations. The influences of the inlet and upward sections and the heating power on the flow instability under rolling motion condition have been analyzed. The marginal stability boundary (MSB) under rolling motion condition is obtained and the unstable regions occur in both low and high equilibrium quality regions. The region with low inlet subcooling is also instable. In high equilibrium quality region, the multiplied period phenomenon is found and the chaotic phenomenon appears at the MSB. The oscillation part of mass flow rate (amplitude) may be averaged into other channels so that the influence of rolling motion is weakened. But the stability of multi-channel system is independent of the channel number and the increase of the channel number could only make the amplitude more uniformity in channels.
Study on Transitional Reactivity Variation of Natural and Forced Circulation
Authors: He Ke-Yu , Wang Zhang-Li , Han Wei-Shi
Published: July 2009

In the transient state of natural and forced circulation, the soaring flux of coolant will lead to the variation of core temperature. Reactivity will also be changed because of the negative temperature coefficient. To study the disciplinarian of reactivity variation, a calculation model was established for reactor core and primary coolant system based on Matlab. The point-reactor kinetics equations were solved by third-order Hermite interpolation approach and fourth-order Runge-Kutta method was used to solve the heat transfer equations. In addition, the variation rule of coolant flux was calculated by one-dimensional flow model of the primary circuit. The results indicate that the coolant flux changes in conformity to hyperbolic tangent curve in the transient state. The coefficient k and flux rising multiple, W∞/W0, determine the flux grads and time for flux to steady. It can effectively improve the minimum reactor period to choose appropriate value of coefficient k in reactor design. And trying the best to reduce the flux grads of transient state can decrease the peak value of reactivity and increase the security of reactor.
The Technical Development for Reuse of Radioactive Concrete Waste Generated by Dismantling of Nuclear Facilities
Authors: Min Byung Youn , Choi Wang Kyu , Park Jung Woo , Jung Chong Hun , Oh Won Zin
Published: July 2009

As part of a fundamental study for the reuse and recycling of contaminated concrete wastes generated by dismantling a research reactor, KAERI has been developing separation technology which recovers for reuse and recycling aggregates from the dismantled concrete. The separation is based on a thermal treatment followed by mechanical processes such as a crushing, milling and sieving. The separation test of concrete waste (i.e., light concrete and heavy concrete) was performed using radioisotope (60Co). The distribution of 60Co has been investigated for the effects of the heating temperature, washing and aggregate size such as gravel, sand and paste using a mechanical and thermal unit process. The experimental results showed that most of the 60Co nuclide could be removed from the gravel, sand aggregate and concentrated into a paste. Especially, we found that the heating temperature played an important role in separating the 60Co nuclide from the concrete waste. The optimum heating temperature for the removal of the 60Co was about 500°C. According to an increase of the heating temperature, the amount of transferred 60Co to the paste linearly increased because the bond between the aggregate and paste matrix is reduced. Contamination of concrete is mainly concentrated in the porous paste and not in the dense aggregate such as the gravel and sand. Concrete can be separated into contaminate and clean parts by means of a unit process based on a crushing, heating, milling and sieving at over 1mm. Through experiments, the clean aggregate can be reused and recycled by up to 70%. By thermally and mechanically separating this concrete waste from the clean dense aggregate particles, a considerable volume reduction can be reacted.
Visualization Study on Complicated Flow Through Lower Plenum of BWR
Authors: Sano Yuta , Abe Yutaka , Fujiwara Akiko , Goto Shoji , Watanabe Fumitoshi , Mori Michitsugu
Published: July 2009

The increase of power outputs enables us to decrease a generation cost such as over thirty nuclear power plants which have adapted an up-rating method in the United States. To success it, it is necessary to evaluate detail thermal hydraulics behavior with high accuracies due to the severe use of internal core structures. The evaluation of coolant flow at a lower plenum of an advanced boiling water reactor (ABWR) is very important because there are a lot of structures such as control rod guide tubes (CRGT) and the core support beams on the fuel assemblies. The coolant flow direction changes from downward to upward with three-dimensional complicated flow in the lower plenum. The simulation results by a CFD (Computational Fluid Dynamics) code can predict such complicated flow in the lower plenum. It is necessary to compare the simulation results with the actual flow in wide range of high Reynolds numbers. And it is required to establish the database of flow structure in lower plenum of ABWR experimentally for the benchmark of CFD code. In the constructed model of the lower plenum of ABWR, we measured velocity profiles by LDV (Laser Doppler Velocimetry) and PIV (Particle Image Velocimetry) techniques with a high speed video camera. The turbulent flow structure of lower plenum of ABWR was evaluated experimentally. In the range of Reynolds number from 103 to 104, the velocity at the center of the test section was faster than the velocity near the wall. The intensity of turbulent increased when the Reynolds number was higher. The velocity profiles in downstream showed the tendency to be flat in the core support beam.
The Influence of the Grain Structure Size on Microstructurally Small Cracks
Authors: Simonovski I. , Cizelj L.
Published: July 2009

Material ageing and fatigue effects play a significant role in safe operation of nuclear power plants. Particular issues are the initialization and propagation of microstructurally small cracks which can represent a significant proportion of a component’s life time. These cracks are still not well understood, one of the reason being that a number of microstructural features have a significant influence on such cracks: crystallographic orientations of grains, grain boundaries, inclusions, voids, material phases, etc. Microstructural features away from the crack have smaller effect on the crack tip parameters. Crack length also plays a role. For a long crack the influence of the surrounding microstructural features will be smaller than for a small crack. A crack of sufficient length can be modeled using classical fracture mechanics methods. The question is when do we reach this point? To try to answer this question we create a model containing a large number of randomly sized, shaped and oriented grains with a crack inserted into a surface grain. Random grain structure is modelled using a given Voronoi tessellation. We then extend the crack up to 7 grains in length and estimate the standard deviation of crack tip opening displacements (CTOD) due to the random crystallographic orientations of the grains surrounding the crack. To account for the deformation mechanism at the grain size scale anisotropic elasticity and crystal plasticity constitutive models are employed. We show that even at crack length equivalent to 7 grains the standard deviation of the crack tip opening displacements due to the surrounding crystallographic orientations is still 7.3% and that from engineering point of view a crack with length of less than 10 average grain sizes can not be taken as independent of the surrounding microstructural features.
Effect of Inlet Configuration on Moderator Velocity and Temperature Distribution Inside the Calandria of a Heavy Water Reactor
Authors: Vaidya Abhijeet Mohan , Maheshwari Naresh Kumar , Vijayan Pallippattu Krishnan , Saha Dilip , Sinha Ratan Kumar
Published: July 2009

Computational study of the moderator flow in calandria vessel of a heavy water reactor is carried out for three different inlet nozzle configurations. For the computations, PHOENICS CFD code is used. The flow and temperature distribution for all the configurations are determined. The impact of moderator inlet jets on adjacent calandria tubes is studied. Based on these studies, it is found that the inlet nozzles can be designed in such a way that it can keep the impact velocity on calandria tubes within limit while keeping maximum moderator temperature well below its boiling limit.
Study on Chemical Reactivity Control of Liquid Sodium: Development of Nano-Fluid and Its Property and Applicability to FBR Plant
Authors: Saito Jun-Ichi , Ara Kuniaki , Sugiyama Ken-Ichiro , Kitagawa Hiroshi , Nakano Haruyuki , Ogata Kan , Yoshioka Naoki
Published: July 2009

Liquid sodium is used as the coolant of the fast breeder reactor (FBR). A weak point of sodium is a high chemical reactivity with water or oxygen. So an idea of chemical reactivity suppression of liquid sodium itself is proposed. The idea is that nano-meter size particles (hereafter called nano-particles) are dispersed in liquid sodium, and an atomic interaction which is generated between nano-particle and sodium atoms is applied to suppress the chemical reactivity. We call sodium that has dispersed the nano-particle a Nano-fluid. Three key technologies which are the trial manufacture of Nano-fluid, the reaction property of the Nano-fluid and applicability of Nano-fluid to FBR Plant have been carried out to develop the Nano-fluid.
RANS and LES Investigations of Vertical Flows in the Fuel Passages of Gas-Cooled Nuclear Reactors
Authors: Keshmiri Amir , Cotton Mark A. , Addad Yacine , Rolfo Stefano , Billard Flavien
Published: July 2009

Coolant flows in the cores of current gas-cooled nuclear reactors consist of ascending vertical flows in a large number of parallel passages. Under post-trip conditions such heated turbulent flows may be significantly modified from the forced convection condition by the action of buoyancy, and the thermal-hydraulic regime is no longer one of pure forced convection. These modifications are primarily associated with changes to the turbulence structure, and indeed flow laminarization may occur. In the laminarization situation heat transfer rates may be as low as 40% of those in the corresponding forced convection case. The heat transfer performance of such ‘mixed’ convection flows is investigated here using a range of refined Reynolds-Averaged-Navier-Stokes (RANS) turbulence models. While all belong to the broad class of Eddy Viscosity Models (EVMs), the various RANS closures have different physical parameterizations and might therefore be expected to show different responses to externally-imposed conditions. Comparison is made against experimental and Direct Numerical Simulation (DNS) data. In addition, Large Eddy Simulation (LES) results have been generated as part of the study. Three different CFD codes have been employed in the work: ‘CONVERT’, ‘STAR-CD’, and ‘Code_Saturne’, which are respectively in-house, commercial, and industrial packages. It is found that the early EVM scheme of Launder and Sharma [1] is in the closest agreement with consistently-normalized DNS results for the ratio of mixed-to-forced convection Nusselt number (Nu/Nu0). However, in relation to DNS and experimental data for forced convection Nusselt number, other models perform better than the Launder-Sharma scheme. The present investigation has revealed discrepancies between direct-simulation, experimental, and the current LES studies.
Development of a J-Estimation Scheme for Surface Cracks in Piping Welds
Authors: Le Delliou Patrick , Marie Ste´phane , Kayser Yann , Barthelet Bruno
Published: July 2009

The RSE-M Code provides rules and requirements for in-service inspection of French Pressurized Water Reactor power plants. The Code gives non mandatory guidance for analytical evaluation of flaws. Flaw assessment procedures rely on fracture mechanics analyses based on simplified methods (i.e. analytical). Analytical methods were developed under a cooperative program between EDF, CEA and AREVA NP to calculate the J integral in various cracked piping components (straight pipe, tapered transition, elbow and pipe-to-elbow junction). These methods are available for mechanical loading (in-plane bending moment, pressure, torsion moment), thermal loading as well as for combined loading. Moreover, they can be used either for materials with Ramberg-Osgood stress-strain curves or for real materials (stainless steels and carbon manganese steels, including those with yield plateaus). However, for the analysis of cracks in welds, they use the tensile properties of the weakest material between the base material and the weld material. This induces some conservatism on the estimated J values. A cooperative program was launched in 2004 to develop a J estimation scheme which takes into account the strength mismatch effects. The scheme relies on the definition of an ‘equivalent’ stress-plastic strain curve, as proposed in the R6 rule (section III.8: allowance for strength mismatch effects). This curve is then used with the analytical methods for homogeneous cracked components. In a first step, the method is developed for circumferential surface cracks in straight buttwelded pipes submitted to mechanical loading. It takes into account the geometry of the weld joint (V-shaped), as well as the location of the crack within the weld. This paper presents the current state of development of this J estimation-scheme.
Numerical Investigation of On-Power Refueling in a Natural Circulation Boiling Water Reactor
Authors: Durga Prasad G. V. , Pandey Manmohan , Pradhan S. K. , Gupta S. K.
Published: July 2009

Most two-phase heat exchangers consist of multiple parallel boiling channels, and studies concerning multi-channel instabilities in boiling systems are of significant interest. The natural circulation boiling water reactor (NCBWR) considered in the present work is a pressure tube type boiling light water cooled and heavy water moderated reactor. The primary heat transport loop, comprising a number of parallel channels, is modeled by RELAP5/MOD3.4. The effects of on-power refueling on the flow rate and stability of the system are investigated. The conditions under which a sustainable flow rate can be maintained during on-power refueling are explored. It is observed that during on-power refueling, a near-stagnation condition or low-velocity reverse flow can occur, the possibility of reverse flow being higher at low pressures and low powers.