bluuebunny/crossref_metadata_2025 huggingface dataset:
The Need for Strategic Planning for Nuclear Power Plant I&C Upgrades
Authors: Hill J. Douglas , Moore Paul
Published: March 2009

Nuclear power plants rely on Instrumentation and Control (I&C) systems for control, monitoring and protection of the plant. The original, analog designs used in most nuclear plants have become or soon will be obsolete, forcing plants to turn to digital technology. Many factors affect the design of replacement equipment, including long-term and short-term economics, regulatory issues, and the way the plant operates on a day-to-day basis. The first step to all modernization projects should involve strategic planning, to ensure that the overall long and short-term goals of the plant are met. Strategic planning starts with a thorough evaluation of the existing plant control systems, the available options, and the benefits and consequences of these options.
Experimental Study on the Performance of IIST Passive Core Cooling System
Authors: Chang Chin-Jang , Lee Chien-Hsiung , Hong Wen-Tan , Wang Lance L. C.
Published: March 2009

The purpose of this study is to conduct the experiments at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility for evaluation of the performance of the passive core cooling system (PCCS) during the cold-leg small break loss-of-coolant accidents (SBLOCAs). Five experiments were performed with (1) three different break sizes, 2%, 0.5%, and 0.2% (approximately corresponding to 1 1/4”, 2”, and 4” breaks for Maanshan nuclear power plant), and (2) 0.2% and 0.5% without actuation of the first-stage and third-stage automatic depressurization valve (ADS-1 and ADS-3) to initiate PCCS for assessing its capacity in accident management. The detailed descriptions of general system response and the interactions of core makeup tanks (CMTs), accumulators (ACCs), automatic depressurization system (ADS), passive residual heat Removal (PRHR), and in-containment refueling water storage tank (IRWST) on the core heat removal are included. The results show: (1) core long term cooling can be maintained for all cases following the PCCS procedures, (2) the core can be covered for the cases of the 0.2% and 0.5% breaks without actuation of ADS-1 and ADS-3.
Preliminary Analysis of the BASALA-H Experimental Programme
Authors: Blaise Patrick , Fougeras Philippe , Philibert Herve´ , Laval Vale´rie , Perret Gre´gory
Published: March 2009

This paper is focused on the preliminary analysis of results obtained on the first cores of the first phase of the BASALA (Boiling water reactor Advanced core physics Study Aimed at mox fuel LAttice) programme, aimed at studying the neutronic parameters in ABWR core in hot conditions, currently under investigation in the French EOLE critical facility, within the framework of a cooperation between NUPEC, CEA and COGEMA. The first “on-line” analysis of the results has been made, using a new preliminary design and safety scheme based on the French APOLLO-2 code in its 2.4 qualified version and associated CEA-93 V4 (JEF-2.2) Library, that will enable the Experimental Physics Division (SPEx) to perform future core designs. It describes the scheme adopted and the results obtained in various cases, going to the critical size determination to the reactivity worth of the perturbed configurations (voided, overmoderated, and poisoned with Gd2O3-UO2 pins). A preliminary study on the experimental results on the MISTRAL-4 is resumed, and the comparison of APOLLO-2 versus MCNP-4C calculations on these cores is made. The results obtained show very good agreements between the two codes, and versus the experiment. This work opens the way to the future full analysis of the experimental results of the qualifying teams with completely validated schemes, based on the new 2.5 version of the APOLLO-2 code.
Fundamental Experiments on Local Hydrogen Behaviors Under Severe Accident Conditions in a Rectangular Subcompartment of Nuclear Power Plant
Authors: Lee Jung-Jae , Lee Un-Jang , Park Goon-Cherl
Published: March 2009

During severe accidents in nuclear power plant (NPP), the limit of local hydrogen concentration is regulated to ensure the integrity of containment. In this study hydrogen mixing experiments were conducted to investigate the hydrogen concentration distribution in a subcompartment of NPP. The mixing compartment is a vertical rectangular type with the dimension of 1×1×1.5 m3. The helium gas was used as a simulant of hydrogen. The goals of this study are to understand local hydrogen mixing phenomena and to examine the effects of the wall condensing, of the existence of obstacle and of the containment spray operation on local hydrogen concentration in NPP. Experimental results showed that the hydrogen might be locally accumulated in the subcompartment and the local hydrogen concentration could instantaneously rise during the spray operation.
Radio Active Waste Management: Underground Repository Method
Authors: Sandesh Kumar Rudrapati , Shrivastava Payal
Published: March 2009

Finding a solution for nuclear waste is a key issue, not only for the protection of the environment but also for the future of the nuclear industry. Ten years from now, when the first decisions for the replacement of existing nuclear power plants will have to be made, The general public will require to know the solution for nuclear waste before accepting new nuclear plants. In other words, an acceptable solution for the management of nuclear waste is a prerequisite for a renewal of nuclear power. Most existing wastes are being stored in safe conditions waiting for permanent solution, with some exceptions in the former Eastern Bloc. Temporary surface or shallow storage is a well known technique widely used all over the world. A significant research effort has been made by the author of this paper in the direction of underground repository. The underground repository appears to be a good solution. Trying to transform dangerous long lived radionuclides into less harmful short lived or stable elements is a logical idea. It is indeed possible to incinerate or transmute heavy atoms of long lived elements in fast breeder reactors or even in pressurised or boiling water reactors. There are also new types of reactors which could be used, namely accelerator driven systems. High level and long lived wastes (spent fuel and vitrified waste) contain a mixture of high activity (heat producing) short lived nuclides and low activity long lived alpha emitting nuclides. To avoid any alteration due to temperature of the engineered or geological barrier surrounding the waste underground, it is necessary to store the packages on the surface for several decades (50 years or more) to allow a sufficient temperature decrease before disposing of them underground. In all cases, surface (or shallow) storage is needed as a temporary solution. This paper gives a detailed and comprehensive view of the Deep Geological Repository, providing a pragmatic picture of the means to make this method, a universally acceptable one.
Implementation of Erbia With ZrB2 in 17×17 Assemblies for a 4-Loop Core With a 24-Month Cycle
Authors: Platte James E. , Pitruzzella Ernesto , Shatilla Youssef , Johansen Baard
Published: March 2009

There are many types of burnable absorbers currently used in power reactors. They are used to provide reactivity and power peaking control. Westinghouse reactors most commonly use Zirconium Diboride Integral Fuel Burnable Absorbers (ZrB2) while Combustion Engineering reactors most commonly use Erbia Integral Fuel Burnable Absorbers (Erbia) in Combustion Engineering reactors. This paper documents the study to determine the effect of placing Erbia and ZrB2 within a Westinghouse 17×17 fuel assembly, and the effect of these ZrB2/Erbia assemblies on the physics characteristics of a representative Westinghouse 4-loop, 24 month cycle length design. The study consisted first of producing optimal within-assembly burnable absorber configurations where ∼25% of the ZrB2-bearing fuel rods within an assembly were replaced with Erbia-bearing fuel rods. This ratio was selected in order to provide an effective balance between potential peaking factor improvements and the known Erbia disadvantage of increased residual absorber penalty compared with ZrB2. The optimal patterns were selected as the ones that most reduced the assembly-wise cumulative peak-to-average rod power during the depletion compared with existing all-ZrB2 BA configurations with the same BA rod quantity loading. The second part of this study consisted of substituting various quantities of these ZrB2/Erbia feed fuel assemblies in a representative Westinghouse 4-loop, 24 month cycle core design to study the effect on power peaking factors, moderator temperature coefficient (MTC), and cycle length.
Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes
Authors: Holowach M. J. , Hochreiter L. E. , Cheung F.-B.
Published: March 2009

A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation.
Repair Welding of Irradiated Materials: Modeling of Helium Bubble Distributions for Determining Crack-Free Welding Procedures
Authors: Feng Zhili , Wilkowski Gery
Published: March 2009

In this paper, a computational simulation study is presented on the prediction of helium bubble evolution during repair welding of irradiated 304 stainless steel. Realistic spatial and temporal temperature and stress evolution during welding were obtained from simulation of the repair welding operation using the finite element model approach. The helium bubble evolution model by Kawano et al. was adopted as a user subroutine in the finite element model to predict the spatial distribution and temporal evolution of the helium bubble size and density in the heat-affected zone (HAZ) of partial penetration welds. Comparisons with experimental results available in open literature show that the predicted average helium bubble sizes were consistent with those observed experimentally under similar conditions. In addition, the computer simulation revealed strong spatial variation of helium bubble size due to the differences in combined thermal and stress conditions experienced in different locations in the HAZ. The predicted location of the maximum helium bubble agreed well with the observed helium-induced cracking site. The effect of welding heat input and welding speed was also investigated numerically. The modeling approach adopted in this study could be used as a cost-effective tool to quantitatively correlate the welding condition, radiation damage, and the likelihood of cracking, under the influence of welding-induced thermal and stress cycles. The model will also be useful in studying the degradation of properties from helium bubble formation of post-welded structures, even if a successful weld is made.
Steam Condensation Experiments by the MISTRA Facility for Field Containment Code Validation
Authors: Caron-Charles M. , Quillico J. J. , Brinster J.
Published: March 2009

The MISTRA facility is a large scale experiment, designed for the validation of thermal-hydraulics multi-D codes. A short description of the facility, the set up of the instrumentation and the test program are presented. Then, experimental results, studying axisymmetric flows of steam injected in the containment and condensed on temperature regulated walls are detailed. Experimental evaluation of the mass and heat transfer coefficients is proposed, then they are compared with the prediction of usual models for condensation, such as Tagami Uchida (0D codes) and the dimensionless correlations (multi-D codes).
Power Production in ADS With Natural Uranium as Nuclear Fuel
Authors: Bergelson Boris R. , Tikhomirov Georgy V. , Gerasimov Aleksander S.
Published: March 2009

ADS operating with solid fuel for power production is discussed. The components of the facility are accelerator, two targets und two blankets. Accumulation of fissile isotopes occurs in first blanket fed by natural or depleted uranium. Power generates in the second blanket fed by fuel elements containing fissile isotopes extracted from the first blanket. In contrast to the natural fuel reactors, the deep fuel burnup can be achieved in ADS by converting of the part of produced power into neutrons.
Cyclic Mode of Transmutation of Minor Actinides in Heavy-Water Reactor
Authors: Gerasimov Aleksander S. , Kiselev Gennady V. , Myrtsymova Lidia A. , Zaritskaya Tamara S.
Published: March 2009

Characteristics of process of transmutation of americium and curium from spent nuclear fuel in heavy-water reactor during first 10 lifetimes and at transition to equilibrium mode are calculated. During transmutation, dangerous nuclides, first of all, 244Cm and 238Pu are accumulated. They cause an increase of radiotoxicity. At first 10 cycles of a transmutation, the radiotoxicity is increased by 11 times in comparison with initial load of transmuted actinides. Heavy-water reactor with thermal power of 1000 MW can transmute americium and curium extracted from 7–8 VVER-1000 type reactors. It means that the required power of transmutation reactor makes about 4% of thermal power of VVER-1000 type reactors.
Comparison of Decay Heat Power of Uranium, Plutonium, and Thorium Spent Fuel at Long-Term Storage
Authors: Gerasimov Aleksander S. , Bergelson Boris R. , Zaritskaya Tamara S. , Tikhomirov Georgy V.
Published: March 2009

Accumulation and subsequent storage of actinides from uranium, plutonium, and thorium spent fuel of PWR type reactor is discussed in the paper. During period of accumulation 100 or 300 years, actinides are periodically introduced in the storage facility. Subsequent period of 300 000 years is long-term storage without introduction of new portions of spent fuel. This time includes both period of controllable storage and ultimate geological storage. Decay heat power of actinides was calculated. It affects heat removal system of the storage facility. Maximal decay heat power of actinides from plutonium spent fuel corresponding to the end of the period of accumulation of 100 years is 2.5 times higher than that of uranium spent fuel. Maximal decay heat power of actinides from uranium spent fuel is 1.2 times higher than that of thorium spent fuel.
Code of a Tokamak Fusion Energy Facility ITER
Authors: Asada Yasuhide , Miya Kenzo , Hada Kazuhiko , Tada Eisuke
Published: March 2009

The technical structural code for ITER (International Thermonuclear Experimental Fusion Reactor) and, as more generic applications, for D-T burning fusion power facilities (hereafter, Fusion Code) should be innovative because of their quite different features of safety and mechanical components from nuclear fission reactors, and the necessity of introducing several new fabrication and examination technologies. Introduction of such newly developed technologies as inspection-free automatic welding into the Fusion Code is rationalized by a pilot application of a new code concept of “system-based code for integrity”. The code concept means an integration of element technical items necessary for construction, operation and maintenance of mechanical components of fusion power facilities into a single system to attain an optimization of the total margin of these components. Unique and innovative items of the Fusion Code are typically as follows; • Use of non-metals; • Cryogenic application; • New design margins on allowable stresses, and other new design rules; • Use of inspection-free automatic welding, and other newly developed fabrication technologies; • Graded approach of quality assurance standard to cover radiological safety-system components as well as non-safety-system components; • Consideration on replacement components.
Thermophysical Analysis of the ADS Systems
Authors: Krepel Jiri
Published: March 2009

For the Thermophysical analysis of the Accelerator Driven Systems (ADS) an simplified model of ADS core was used. The neutron flux was calculated by means of diffusion theory for subcritical homogeneous reactor with an external neutron source. From this result the distribution of the heat sources was found and next the distribution of temperatures was calculated.
Improving the Efficiency of the Nodal Integral Method With the Portable, Extensible Toolkit for Scientific Computation
Authors: Toreja Allen J. , Rizwan-Uddin
Published: March 2009

An existing implementation of the nodal integral method for the time-dependent convection-diffusion equation is modified to incorporate various PETSc (Portable, Extensible Toolkit for Scientific Computation) solver and preconditioner routines. In the modified implementation, the default iterative Gauss-Seidel solver is replaced with one of the following PETSc iterative linear solver routines: Generalized Minimal Residuals, Stabilized Biconjugate Gradients, or Transpose-Free Quasi-Minimal Residuals. For each solver, a Jacobi or a Successive Over-Relaxation preconditioner is used. Two sample problems, one with a low Peclet number and one with a high Peclet number, are solved using the new implementation. In all the cases tested, the new implementation with the PETSc solver routines outperforms the original Gauss-Seidel implementation. Moreover, the PETSc Stabilized Biconjugate Gradients routine performs the best on the two sample problems leading to CPU times that are less than half the CPU times of the original implementation.
Assessment of the Hydraulic Flows in the MARA8 Test-Facility Simulating a HCDA
Authors: Robbe M. F. , Lepareux M.
Published: March 2009

To ensure a high safety level to the French nuclear power plants, safety authorities impose to the plant designers to take into consideration several cases of hypothetical severe accidents to which the plant must be able to resist. The aim is to guarantee that even a core melt accident would not require any stringent countermeasures such as evacuation or relocation of the population beyond the immediate vicinity of the plant.
Comparison of a HCDA Numerical Simulation With the Results of the MARA 8 Test
Authors: Robbe M. F. , Cariou Y. , Lepareux M. , Treille E.
Published: March 2009

To ensure a high safety level to the French nuclear power plants safety authorities impose to the plant designers to take into consideration several cases of hypothetical severe accidents to which the plant must be able to resist. The aim is to guarantee that even a core melt accident would not require any stringent countermeasures such as evacuation or relocation of the population beyond the immediate vicinity of the plant.
Rod Control Assemblies Wear Mechanisms
Authors: Kaczorowski Damien , Georges Jean-Mary , Bec Sandrine , Vannes Andre´-Bernard , Tonck Andre´ , Vernot Jean-Philippe
Published: March 2009

In nuclear power plants, slender tubular components are subjected to vibrations in a PHTW environment. As a result, the two contacting surfaces, tubes and their guides undergo impact at low contact pressures [1]. The components are usually made of stainless steel and it was found that the influence of the PHTW, combined with other actions (such as corrosion, erosion, squeeze film effect, third body effect and cavitation) leads to a particular wear of the material [2] [3]. Therefore, this paper aims to show that the colloidal oxides, formed on the steel surfaces in PHTW, play a principal role in the wear of the surfaces. Actually, due to the specific kinematic conditions of the contact, the flow of compacted oxides abrades the surfaces.
Simeco Tests in a Melt Stratified Pool
Authors: Gubaidullin A. A. , Sehgal B. R.
Published: March 2009

In the last phase of the core degradation, an oxidic melt pool of mainly UO2, ZrO2, and unoxidized Zircaloy and stain-less steel will form in the lower head of the RPV (Theofanous et al., 1996). A molten metal layer (composed mainly of Fe and Zr) will rest on the top of the crust of the oxidic pool. A thin oxidic crust layer of frozen core material is formed on the vessel’s inside wall. In this bounding configuration, thermal loads to the RPV walls are determined by natural convection heat transfer driven by internal heat sources. Decay heat from fission products is assumed to be generated uniformly within the oxidic pool and generally no heat generation is considered in the upper metallic layer. For example, in a hypothetical severe accident scenario for an AP600-like reactor, the following values can be expected: volumetric heat generation Qv ∼ 1 MW/m3, volume of the oxidic pool V ∼ 10 m3, radius R = 2 m, temperatures in the oxidic pool T ∼ 2700°C, temperatures in the metal layer T ∼ 2000°C, maximum depth ratio of the metal layer to the oxidic pool L12 ∼ 0:3, properties of the oxidic pool, depending on melt composition, as characterized by the Prandtl number, Pr ∼ 0:6, properties of the metallic layer Pr < 0:1, the intensity of convective motion, as characterized by the Rayleigh number, Ra ∼ 1015–1016 (Theofanous et al., 1996). The time scale of core melt pool formation is estimated as 1/2 to 1 hour (Sehgal, 1999). Indeed, these estimates could vary, depending very much on the accident scenario and the type of reactor.
Next Generation CANDU Performance Assurance
Authors: Wren David J. , Allsop P. J. , Hopwood J. M.
Published: March 2009

AECL is developing a next generation CANDU design to meet market requirements for low cost, reliable energy supplies. The primary product development objective is to achieve a capital cost substantially lower than the current nuclear plant costs, such that the next generation plant will be competitive with alternative options for large-scale base-load electricity supply. However, other customer requirements, including safety, low-operating costs and reliable performance, are being addressed as equally important design requirements. The main focus of this paper is to address the development directions that will provide performance assurance. The next generation CANDU is an evolutionary extension of the proven CANDU 6 design. There are eight CANDU 6 units in operation in four countries around the world and further three units are under construction. These units provide a sound basis for projecting highly reliable performance for the next generation CANDU. In addition, the next generation CANDU program includes development and qualification activities that will address the new features and design extensions in the advanced plant. To limit product development risk and to enhance performance assurance, the next generation CANDU design features and performance parameters have been carefully reviewed during the concept development phase and have been deliberately selected so as to be well founded on the existing CANDU knowledge base. Planned research and development activities are required only to provide confirmation of the projected performance within a modest extension of the established database. Necessary qualification tests will be carried out within the timeframe of the development program, to establish a proven design prior to the start of a construction project. This development support work coupled with ongoing AECL programs to support and enhance the performance and reliability of the existing CANDU plants will provide sound assurance that the next generation CANDU plants will meet customer expectations.
Integrated Multi Path Model to Calculate Radionuclide Release From a Repository Using Wavelet Galerkin Method
Authors: Nasif Hesham R. , Neyama Atsushi
Published: March 2009

This work represents a WIRS code developed using wavelet Galerkin method to solve radionuclide transport model in near field and far field of a repository for high-level radioactive waste. After overpack failure, radionuclides diffuse through the bentonite buffer material to the water bearing fracture around the repository transport horizontally through this geosphere then transport vertically through the major water conducting fault (MWCF) reach the biosphere. The radionuclides transport barriers considered in this model are engineered barrier system (EBS), geosphere, and MWCF. Hydraulic conductivity of the bentonite is more than three orders of magnitude smaller than that of the surrounding host rock, so the only transport mechanism through EBS is diffusion. In the host rock, the problem is of advection-diffusion type with highly varying parameters from one medium to other due to the variability in length, transmissivity and other transport-relevant properties of the transport paths. Daubechies’ wavelet is used as a basis function to solve the nonlinear partial differential equations arising from the model formulation of the radionuclides transport. Since the scaling functions are compactly supported, only a finite number of the connection coefficients are nonzero. The resultant matrix has a block diagonal structure, which can be inverted easily. We tested our WGM algorithm with several problems to verify the model. The solutions are very accurate with a proper selection of Daubechies’ order and dilation order. The solution is very accurate at the interfaces where the radionuclide concentration exhibits very steep gradients.
New Methods of Studying Nuclear Reactor Thermohydraulics
Authors: Su Guanghui , Fukuda K.
Published: March 2009

New methods of studying reactor thermohydraulic problematics such as CHF, DNBR, boiling and instability etc. were introduced briefly in this paper. Some of these methods have been successfully applied in the reactor thermohydraulic analysis. And the others have not yet been applied in this field. The further study work must be done to show that these new methods can be successfully applied in this field with high accuracy because they are interdisciplinary fields.
Inspection and Repair Techniques and Strategies for Alloy 600 PWSCC in Reactor Vessel Head CRD Nozzles and Welds
Authors: Glass S. W. , Schlader D. M.
Published: March 2009

As a result of the Alloy 600 PWSCC CRD nozzle leaks discovered in the fall of 2000 and spring of 2001 in several US plants, the NRC has recommended a more pro-active effort by U.S. utilities to inspect similarly susceptible nozzles in all US plants. The primary safety concern is circumferential cracks that can permit the nozzle to separate from the head at high velocity and produce a large-break leak in the reactor vessel. A secondary concern is head leakage from any through-wall cracks in the nozzle or J-groove weld area. Although the fundamental weld and seal design are similar for all US PWR plants, the various surrounding geometry and repair probability considerations require multiple inspection and repair alternatives. Geometry issues include the head insulation design that influences the ability to perform visual examinations from above the head, and the presence or absence of thermal sleeves and funnels governing the type of NDE probes than can be used. Repair probability considerations primarily include the likelihood for repair of a small or large number of nozzles and the length of time the repair must last before a head replacement. This paper discusses the various inspection and repair alternatives offered by one service vendor and discusses a decision process for planning the inspection and repair effort.
Scaling Methodology for the Direct ECC Bypass During LBLOCA Reflood Phase With Direct Vessel Injection System: Its Development and Validation
Authors: Cho H. K. , Yun B. J. , Kwon T. S. , Song C.-H. , Park G. C.
Published: March 2009

From the two dimensional two-fluid model a new scaling methodology, named the “modified linear scaling”, is suggested for the scientific design of a scaled-down experimental facility and data analysis of the direct ECC bypass under LBLOCA reflood phase. The characteristics of the scaling law are its velocity is scaled by a Wallis-type parameter and the aspect ratio of experimental facility is preserved with that of prototype. For the experimental validation of the proposed scaling law, the air-water tests for direct ECC bypass were performed in the 1/4.0 and 1/7.3 scaled UPTF downcomer test section. The obtained data are compared with those of UPTF Test21-D. It is found that the modified linear scaling methodology is appropriate for the preservation of multi-dimensional flow phenomena in downcomer annulus, such as direct ECC bypass.
Radiation Induced Surface Activity Phenomenon: 2nd Report — Radiation Induced Boiling Enhancement
Authors: Koga Tatsuya , Imai Yasuyuki , Takamasa Tomoji , Okamoto Koji , Mishima Kaichiro
Published: March 2009

To delineate the effect of Radiation Induced Surface Activity (RISA) on boiling phenomenon, surface wettability in high-temperature environment or Leidenfrost condition and critical heat flux (CHF) of oxide metals irradiated by gamma rays were investigated. When the temperature of the heating surface reaches the wetting limit temperature, water-solid contact vanishes because of a stable vapor film between the droplet and the metal surface, i.e., a Leidenfrost condition. The wetting limit temperature increased with integrated irradiation dose. The CHF of oxidized titanium was improved up to 100% after 800 kGy 60Co gamma ray irradiated. Radiation Induced Boiling Enhancement (RIBE) phenomenon was firstly confirmed through the experiments.
Characterization of Individual Fission Products in Terms of Their Production and Transmutation
Authors: Sagara Hiroshi , Yoshida Tadashi , Saito Masaki
Published: March 2009

In order to establish a simple and common basis which can be referred for the specific studies on the fission products (FP) transmutation and its strategies, the general characteristics of the dominant individual FPs in terms of their production and transmutation in the fast reactor have been studied with an ideal model in the present paper. The potential hazard of each nuclide in long-term utilization of nuclear energy in human society has been quantitatively evaluated. During utilization of fission energy, two short half-life FPs, 90Sr and 137Cs, almost determine the total toxicity of FP nuclides in spite of the effort of transmuting them in fast reactors. The innovative transmuters (such as High Flux Reactor, Accelerator-driven system and Fusion Neutron Source) are needed to reduce these 2 toxicities. In case of 1000-year fission energy utilization with transmutation in fast reactors, the total toxicity of FP nuclides except 126Sn go down below the level of 238U toxicity consumed. The more detailed study of transmutation of 126Sn is a very important in future. However, in the case of 10,000-year fission energy utilization with transmutation in fast reactor, total toxicities of all FP nuclides go down below the level of 238U toxicity consumed.
Study of a Station Blackout Event in the PWR Plant
Authors: Wu Ching-Hui , Lin Tsu-Jen , Kao Tsu-Mu
Published: March 2009

On March 18, 2001, a PWR nuclear power plant located in the Southern Taiwan occurred a Station BlackOut (SBO) event. Monsoon seawater mist caused the instability of offsite power grids. High salt-contained mist caused offsite power supply to the nuclear power plant very unstable, and forced the plant to be shutdown. Around 24 hours later, when both units in the plant were shutdown, several inadequate high cycles of bus transfer between 345 kV and 161 kV startup transformers degraded the emergency 4.16 kV switchgears. Then, in the Train-A switchgear room of Unit 1 occurred a fire explosion, when the degraded switchgear was hot shorted at the in-coming 345 kV breaker. Inadequate configuration arrangement of the offsite power supply to the emergency 4.16 kV switchgears led to loss of offsite power (LOOP) events to both units in the plant. Both emergency diesel generators (EDG) of Unit 1 could not be in service in time, but those of Unit 2 were running well. The SBO event of Unit 1 lasted for about two hours till the fifth EDG (DG-5) was lined-up to the Train-B switchgear. This study investigated the scenario of the SBO event and evaluated a risk profile for the SBO period. Guidelines in the SBO event, suggested by probabilistic risk assessment (PRA) procedures were also reviewed. Many related topics such as the re-configuration of offsite power supply, the addition of isolation breakers of the emergency 4.16 kV switchgears, the betterment of DG-5 lineup design, and enhancement of the reliability of offsite power supply to the PWR plant, etc., will be in further studies.
Measurements of Bubble Rise Characteristics in Various Temperatures and Liquid Velocities
Authors: Torimoto Kazuhiro , Nishiura Masanori , Tanaka Tomoe , Okawa Tomio , Kataoka Isao
Published: March 2009

Single bubble rise characteristics in turbulent flow in a vertical circular pipe were experimentally observed. As a result, it was found that bubble rise velocity relative to the time averaged local liquid velocity decreases with the increase of liquid flowrate and that it also decreases with the decrease of the distance between a bubble and wall. Investigating the mechanisms of the reductions of relative velocity, new correlations were developed for accurately evaluating the bubble rise velocities in turbulent flow. Since the predicted bubble rise velocities reasonably agreed with the measured data, the new correlations could contribute to the improvement of bubbly flow calculations in nuclear reactor safety analysis.
High Cycle Thermal Fatigue in French PWR
Authors: Blondet Eric , Faidy Claude
Published: March 2009

Different fatigue-related incidents which occurred in the world on the auxiliary lines of the reactor coolant system (SIS, RHR, CVC) have led EDF to search solutions in order to avoid or to limit consequences of thermodynamic phenomenal (Farley-Tihange, free convection loop and stratification, independent thermal cycling). Studies are performed on mock-up and compared with instrumentation on nuclear power stations. At the present time, studies allow EDF to carry out pipe modifications and to prepare specifications and recommendations for next generation of nuclear power plants. In 1998, a new phenomenal appeared on RHR system in Civaux. A crack was discovered in an area where hot and cold fluids (temperature difference of 140°C) were mixed. Metallurgic studies concluded that this crack was caused by high cycle thermal fatigue. Since 1998, EDF is making an inventory of all mixing areas in French PWR on basis of criteria. For all identified areas, a method was developed to improve the first classifying and to keep back only potential damage pipes. Presently, studies are performing on the charging line nozzle connected to the reactor pressure vessel. In order to evaluate the load history, a mock-up has been developed and mechanical calculations are realised on this nozzle. The paper will make an overview of EDF conclusions on these different points: • dead legs and vortex in a no flow connected line; • stratification; • mixing tees with high ΔT.
Non-Linear Fluid-Structure Interaction in Cylindrical Shells
Authors: Toorani M. H. , Lakis A. A.
Published: March 2009

Nuclear plant reliability depends directly on its component performance. The higher heat transfer performance of nuclear plant components often requires higher flow velocities through the shell and tube heat exchangers. So, these cylindrical structures are subjected to either axial or cross flow, while the excessive flow-induced vibrations, (which are a major cause of machinery downtime; fatigue failure and high noise), limit the performance of these structures. On the other hand, these shell components often experience large amplitude vibrations that are greater than the shell thickness. Therefore, the evaluation of complex vibrational behavior of these structures is highly desirable in the nuclear industry. A semi-analytical approach has been developed in the present theory to predict the geometrical non-linearity influence on the natural frequencies of anisotropic cylindrical shells conveying axial flow. Particular important in this study is to obtain the natural frequencies of the coupled system of the fluid-structure, taking into account the geometrical non-linearity of the structure, and also estimating the critical flow velocity at which the structure loses its stability. The displacement functions, mass and stiffness matrices, linear and non-linear ones, of the structure are obtained by exact analytical integration over a hybrid element developed in this work. Linear potential flow theory is applied to describe the fluid effect that leads to the inertial, centrifugal and Coriolis forces. Numerical results are given and compared with those of experiment and other theories to demonstrate the practical application of the present method.
Robotics for Long-Term Monitoring
Authors: Shahin Sarkis , Duran Celso
Published: March 2009

While long-term monitoring and stewardship means many things to many people, DOE has defined it as: “The physical controls, institutions, information, and other mechanisms needed to ensure protection of people and the environment at sites where DOE has completed or plans to complete cleanup (e.g., landfill closures, remedial actions, and facility stabilization).” Across the United States, there are thousands of contaminated sites with multiple contaminants released from multiple sources where contaminants have transported and commingled. The U.S. government and U.S. industry are responsible for most of the contamination and are landowners of many of these contaminated properties. These sites must be surveyed periodically for various criteria including structural deterioration, water intrusion, integrity of storage containers, atmospheric conditions, and hazardous substance release. The surveys, however, are intrusive, time-consuming, and expensive and expose survey personnel to radioactive contamination. In long-term monitoring, there’s a need for an automated system that will gather and report data from sensors without costly human labor. In most cases, a SCADA (Supervisory Control and Data Acquisition) unit is used to collect and report data from a remote location. A SCADA unit consists of an embedded computer with data acquisition capabilities. The unit can be configured with various sensors placed in different areas of the site to be monitored. A system of this type is static, i.e., the sensors, once placed, cannot be moved to other locations within the site. For those applications where the number of sampling locations would require too many sensors, or where exact location of future problems is unknown, a mobile sensing platform is an ideal solution. In many facilities that undergo regular inspections, the number of video cameras and air monitors required to eliminate the need for human inspections is very large and far too costly. HCET’s remote harsh-environment surveyor (RHES) is a robotic platform with SCADA capabilities equipped with a sonar-imaging scanner, a high-resolution color CCD camera, and various combinations of sensors. The RHES is controlled remotely via a PC. This paper will discuss the development and application of this system.
KEDO’s LWR Project: Unique Challenges in a Nuclear Power Project by Any Standard
Authors: Choi HanKwon , Crom Thomas D. , Mulligan John B.
Published: March 2009

In 1994 the United States of America and the Democratic People’s Republic of Korea (DPRK) signed the Agreed Framework in which the DPRK committed to shut down their graphite-moderated reactors and related facilities and to remain a party to the Treaty on the Non-Proliferation of Nuclear Weapons. In return the U.S. agreed, among other things, to deliver two light water reactor (LWR) units to the DPRK which should meet international safety standards. For the implementation of this Agreed Framework, the Korean Peninsula Energy Development Organization (KEDO) was founded. KEDO decided to build two units of the Korea Standard Nuclear Power Plant (KSNP) model, which is a proven design of U.S. origin. This paper describes the status of this project (the LWR project) and the unique challenges that KEDO must overcome to implement the construction and commissioning of this project and have the DPRK ready for plant takeover.
Risk-Informing: The SSC Scope of U.S. NRC Special Treatment Requirements – Treatment of Low Safety-Significant SSCs
Authors: Heymer Adrian P.
Published: March 2009

This paper provides a general description and rationale for licensee-established industrial controls for structures, systems and components (SSCs) that have been categorized as low safety-significant through a process consistent with 10 CFR 50.69 requirements.
Promoting Intellectual Discovery: Patents Versus Markets
Authors: Meloso Debrah , Copic Jernej , Bossaerts Peter
Published: March 2009

Because they provide exclusive property rights, patents are generally considered to be an effective way to promote intellectual discovery. Here, we propose a different compensation scheme, in which everyone holds shares in the components of potential discoveries and can trade those shares in an anonymous market. In it, incentives to invent are indirect, through changes in share prices. In a series of experiments, we used the knapsack problem (in which participants have to determine the most valuable subset of objects that can fit in a knapsack of fixed volume) as a typical representation of intellectual discovery problems. We found that our “markets system” performed better than the patent system.
Extremely High Mutation Rate of a Hammerhead Viroid
Authors: Gago Selma , Elena Santiago F. , Flores Ricardo , Sanjuán Rafael
Published: March 2009

The mutation rates of viroids, plant pathogens with minimal non-protein-coding RNA genomes, are unknown. Their replication is mediated by host RNA polymerases and, in some cases, by hammerhead ribozymes, small self-cleaving motifs embedded in the viroid. By using the principle that the population frequency of nonviable genotypes equals the mutation rate, we screened for changes that inactivated the hammerheads of
Chrysanthemum chlorotic mottle viroid
. We obtained a mutation rate of 1/400 per site, the highest reported for any biological entity. Such error-prone replication can only be tolerated by extremely simple genomes such as those of viroids and, presumably, the primitive replicons of the RNA world. Our results suggest that the emergence of replication fidelity was critical for the evolution of complexity in the early history of life.
Geometric Cue for Protein Localization in a Bacterium
Authors: Ramamurthi Kumaran S. , Lecuyer Sigolene , Stone Howard A. , Losick Richard
Published: March 2009

Proteins in bacteria often deploy to particular places within the cell, but the cues for localization are frequently mysterious. We found that the peripheral membrane protein SpoVM (VM) recognizes a geometric cue when localizing to a particular membrane during sporulation in
Bacillus subtilis
. Sporulation involves an inner cell maturing into a spore and an outer cell nurturing the developing spore. VM is produced in the outer cell, where it embeds in the membrane that surrounds the inner cell but not in the cytoplasmic membrane of the outer cell. We found that VM localized by discriminating between the positive curvature of the membrane surrounding the inner cell and the negative curvature of the cytoplasmic membrane. Membrane curvature could be a general cue for protein localization in bacteria.
Leucine-Rich Repeat Protein Complex Activates Mosquito Complement in Defense Against
Plasmodium
Parasites
Authors: Povelones Michael , Waterhouse Robert M. , Kafatos Fotis C. , Christophides George K.
Published: March 2009

Leucine-rich repeat–containing proteins are central to host defense in plants and animals. We show that in the mosquito
Anopheles gambiae
, two such proteins that antagonize malaria parasite infections, LRIM1 and APL1C, circulate in the hemolymph as a high-molecular-weight complex held together by disulfide bridges. The complex interacts with the complement C3-like protein, TEP1, promoting its cleavage or stabilization and its subsequent localization on the surface of midgut-invading
Plasmodium berghei
parasites, targeting them for destruction. LRIM1 and APL1C are members of a protein family with orthologs in other disease vector mosquitoes and appear to be important effectors in innate mosquito defenses against human pathogens.
Mutation of Cysteine 171 of Pestivirus ErnsRNase Prevents Homodimer Formation and Leads to Attenuation of Classical Swine Fever Virus
Authors: Tews Birke Andrea , Schürmann Eva-Maria , Meyers Gregor
Published: March 2009

Pestiviruses represent important pathogens of farm animals that have evolved unique strategies and functions to stay within their host populations. Erns, a structural glycoprotein of pestiviruses, exhibits RNase activity and represents a virulence factor of the viruses. Ernsforms disulfide linked homodimers that are found in virions and virus-infected cells. Mutation or deletion of cysteine 171, the residue engaged in intermolecular disulfide bond formation, results in loss of dimerization as tested in coprecipitation and native protein gel electrophoresis analyses. Nevertheless, stable virus mutants with changes affecting cysteine codon 171 could be recovered in tissue culture. These mutants grew almost as well as the parental viruses and exhibited an RNase-positive phenotype. Ernsdimerization-negative mutants of classical swine fever virus were found to be attenuated in pigs even though the virus clearly replicated and induced a significant neutralizing antibody response in the animals.