bluuebunny/crossref_metadata_2025 huggingface dataset:
Assessment of Spatial Property Variability of the Subsurface in SSI Studies of the Armenian Nuclear Power Plant
Authors: Simos Nicholas , Zadoyan Paruyr , Labbe Pierre , Sollogoub Pierre , Touret J.-Pierre
Published: November 2008

This paper addresses issues surrounding soil property variability including uncertainties associated with “best estimate” values and searches for practical ways to assess the impact on the seismic response of a facility, such as a nuclear power plant, resting on it. Specifically, it attempts, using a parametric study, to formulate a probabilistic model that enables the enveloping of uncertainties associated with the soil-structure-interaction component of the seismic problem. The effects of most-likely sources of uncertainty, such as variability of “distinct” soil layer profile and variability of controlling soil properties, are to be addressed by generating a probabilistic profile in which randomization of key parameters that appear to have the most impact on the results of deterministic analyses is implemented. The use of stochastic finite elements and the introduction of correlation functions, in conjunction with finite element discretization of the foundation soil, are explored as means of achieving an enveloped structural response. The on-going evaluation of the Armenian nuclear plant site prompted this study. In order to stress the importance and relevance of the stated goal, the soil-structure-interaction of the nuclear power plant, subject to significant variation of the foundation soil, is examined. The conflicting results of two independent studies of the subsurface provide the basis for the variation range used in this study.
The Combined Thermohydraulics-Neutronics Code TRAB-SMABRE for 3D Plant Transient and Accident Analyses
Authors: Jaakko Miettinen , Timo Vanttola , Antti Daavittila , Hanna Ra¨ty
Published: November 2008

TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenised fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensables in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation.
An Overview of the Direct Energy Conversion Proof of Principle Power Production Program
Authors: King D. , Rochau G. , Oscar D. , Morrow C. , Tsvetkov P. , Hart R. , Gallix R.
Published: November 2008

The United States Department of Energy, Nuclear Energy Research Initiative (NERI) Direct Energy Conversion Proof of Principle (DECPOP) project has as its goal the development of a direct energy conversion process suitable for commercial development. We define direct energy conversion as any fission process that returns usable energy without an intermediate thermal process. A prior Direct Energy Conversion (DEC) project [1] has been completed and indicates that a viable direct energy device is possible if several technological issues can be overcome. The DECPOP program is focusing on two of the issues: charged particle steering and high voltage hold-off. This paper reports on the progress of the DECPOP project. Two prototype concepts are under development: a Fission Electric Cell using magnetic insulation and a Fission Fragment Magnetic Collimator using magnetic fields to direct fission fragments to collectors. Included in this paper are a short project description, an abbreviated summary of the work completed to date, a description of ongoing and future project activities, and a discussion of the potential for future research and development.
Fission Products Particular Peak Measurement for UO2-Gd2O3-MOX γ-Scanning Renormalization in 100% MOX ABWR Mock-Up Cores
Authors: Blaise P. , Cathalau S. , Thiollay N. , Fougeras P. , Laval V. , Philibert H. , Girard J. M. , Hudelot J. P.
Published: November 2008

This paper presents the principles of the peak check measurements by gamma spectrometry. One details the main equations used for the analysis of the raw data as the calculation of the different sources of uncertainties and their propagation on the result. The method is illustrated with actual examples from the French-Japanese BASALA ABWR 100% MOX experimental program. • For each individual fission rate, the systematical uncertainty due to the radioactive decay data is smaller than the statistical uncertainty due to the counting process. • The main part of the final uncertainty on the scaling factor is brought by the systematical uncertainty on the average fission yields. This study enables to propose some recommendations for fission products and nuclear data evaluation used. • Only the analysis of the 140La peak at 1596 keV leads to acceptable uncertainties on the fission rate maps renormalization, with a good consistency with the integral γ-scanning results. • The JEF-2.2 evaluation on the fission yields and their associated uncertainties seems more realistic than the ENDF/B-6 and is recommended for the scaling factor analysis.
A Balancing Method for Calculating a Component RAW Including CCF
Authors: Kim Kilyoo , Kang Dae Il , Yang Joon-Eon
Published: November 2008

In this paper, a method called the ‘Balancing Method’ to derive a component RAW with basic event RAWs including a CCF (Common Cause Failure) RAW is summarized, and compared with the NEI proposed method by mathematically checking the background on which the two methods are based. It is proved that the Balancing Method has a strong mathematically background.
PWSCC Experience of Pressurizer Dissimilar Metal Welds at Tsuruga Unit-2
Authors: Kobayashi Hiroyuki , Takeuchi Kimihito , Nakama Shohei , Mukai Masayuki , Ohta Takahiro
Published: November 2008

During the 13th refuel outage in September 2003, the small boric acid deposit was observed at one dissimilar weld joint between Relief Valve Nozzle and Safe End of Pressurizer in Tsuruga unit 2. The other four indications were also observed at the weld joint by non-destructive examination. Similar indications were also observed at Safety Valve Nozzle. Both Nozzles were under the steam condition at 345°C. As the first root cause investigation, the visual observation after etching was conducted at these weld joint on site. The observation revealed the repair weld print. Both weld joints with cracks were taken to conduct some investigations such as metallographic examination of fracture surface at hot laboratory. Hot laboratory examination results revealed that all cracks were extended axially in the Ni base metal weld joint, and there was no evidence of any cracks extending into the low alloy steel Nozzle and stainless steel Safe end. All cracks were interdendritic and branched. Metallographic examination results revealed that the crack growth mainly was caused by interdendritic stress corrosion cracking (IDSCC). These Nozzles including the weld joint with indications were repaired by jointing the spool piece as new Safe ends. The repair process includes the weld buttering on the low alloy steel Nozzle and joint welding between Nozzle and the spool piece. These welding were conducted by GTAW using alloy 52. It took twenty days to complete the repair works.
Seismic Hazard Analysis for an NPP Site
Authors: Ghosh A. K. , Kushwaha H. S.
Published: November 2008

The various uncertainties and randomness associated with the occurrence of earthquakes and the consequences of their effects on the NPP components and structures call for a probabilistic seismic risk assessment (PSRA). However, traditionally, the seismic design basis ground motion has been specified by normalised response spectral shapes and peak ground acceleration (PGA). The mean recurrence interval (MRI) used to be computed for PGA only. The present work develops uniform hazard response spectra i.e. spectra having the same MRI at all frequencies for Kakrapar Atomic Power Station site. Sensitivity of the results to the changes in various parameters has also been presented. These results determine the seismic hazard at the given site and the associated uncertainties. The paper also presents some results of the seismic fragility for an existing containment structure. The various parameters that could affect the seismic structural response include material strength of concrete, structural damping available within the structure and the normalized ground motion response spectral shape. Based on this limited case study the seismic fragility of the structure is developed. The results are presented as families of conditional probability curves plotted against the peak ground acceleration (PGA). The procedure adopted incorporates the various randomness and uncertainty associated with the parameters under consideration.
Inspection and Repair Technology for BMI Penetrations
Authors: Glass S. W. , Cole R. F.
Published: November 2008

Historically, United States (US) nuclear power plant inspections of the reactor vessel bottom-mounted-instrument (BMI) penetrations have been limited to visual verification via a combination of walk-downs and pressure tests. However in France, more than 18 UT and weld-surface VT campaigns had been performed to inspect the BMI penetration nozzles and welds since 1992 with no observed failures through 2002. In April 2003, South Texas Project Unit 1 discovered apparent leakage from two nozzles during a bare-metal examination. Based on the French inspection experience, Framatome ANP was selected for inspection and repair services to address the leaking nozzles. Inspection activities included ultrasonic examination (UT) of the tube, enhanced visual test (VT) and eddy current testing (ECT) of the J-groove weld, bobbin ECT and profilometry of the tube ID, helium leak test, phased-array UT, borescopic VT, and boat-sample removal with destructive metallurgical analysis. Repair activities included sealing the top portion of the penetration, cutting out the old nozzles, and welding in new nozzles. This presentation discusses BMI inspection and repair technology focused particularly on the South Texas Project experience.
A Stability Investigation of Two-Dimensional Surface Waves on Evaporating, Isothermal or Condensing Liquid Films: Part III — Temporal Stability of Traveling Waves
Authors: Ye Xuemin , Li Chunxi , Yan Weiping
Published: November 2008

The temporal stability equation of the two-dimensional traveling waves of evaporating or condensing liquid films falling down an inclined wall is established based on the Prandtl boundary layer theory and complete boundary conditions. By investigating the flow temporal characteristics curves, including the stability curves and stability curves of the fastest wave, the effects on flow stability of evaporating, isothermal and condensing states, thermocapillarity, Reynolds number, fluid property and inclined angle are discussed, and are compared in different Reynolds numbers. The theoretical study indicated that evaporation process destabilizes the film flow and condensation process stabilizes the film flow, the thermocapillarity take a destabilizing effect in evaporation condition and an adverse effect in condensation condition. Present study indicates that the temporal growth rate increases with increase of the Reynolds number and inclination angle, and decreases with increase of Ka numbers. And the effects on flow stability of liquid properties and inclination angle are always significant.
Validation of Trio_U: Numerical Simulations of a ROCOM Buoyancy Driven Test Case
Authors: Ho¨hne T. , Bieder U. , Kliem S. , Prasser H.-M.
Published: November 2008

A generic investigation of the influence of density differences between the primary loop inventory and the ECC water on the mixing in the downcomer was made at the ROCOM Mixing Test Facility at Forschungszentrum Rossendorf (FZR)/Germany. ROCOM is designed for experimental coolant mixing studies over a wide variety of possible scenarios. It is equipped with advanced instrumentation, which delivers high-resolution information characterizing either temperature or boron concentration fields in the investigated pressurized water reactor. For the validation of the Trio_U code an experiment with 5% constant flow rate in one loop (magnitude of natural circulation) and 10% density difference between ECC and loop water was taken. Trio_U is a CFD code developed by the CEA France, aimed to supply an efficient computational tool to simulate transient thermal-hydraulic single-phase turbulent flows encountered in the nuclear systems as well as in the industrial processes. For this study a LES approach was used for mesh sizes according to between 300000–2 million control volumes. The results of the experiment as well as of the numerical calculations show, that a streak formation of the water with higher density is observed. At the upper sensor, the ECC water covers a small azimuthal sector. The density difference partly suppresses the propagation of the ECC water in circumferential direction. The ECC water falls down in an almost straight streamline and reaches the lower downcomer sensor position directly below the affected inlet nozzle. Only later, coolant containing ECC water appears at the opposite side of the downcomer. The study showed, that density effects play an important role during natural convection with ECC injection in pressurized water reactors. Furthermore it was important to point out, that Trio_U is able to cope the main flow and mixing phenomena.
Life Cycle Management Application at the STARS Plants
Authors: Gregor Frank E. , Hromyak Daniel E.
Published: November 2008

Six of the southwest and western nuclear plants have formed an alliance to jointly apply and share in new technologies to increase plant reliability. Life cycle management (LCM) is an EPRI developed process to evaluate and select the best long-term maintenance strategies for important plant systems, structures and component groups. Both, technical and economic assessments are made to arrive at a strategy that provides the highest reliability at the least cost. Among those SSCs chosen for LCM are the Station Transformers, the Main Generator, Turbine Controls, Underground Piping and the Circulating Water System (CWS). The application of the LCM process to the STARS plants is demonstrated for the CWS. The process steps from information gathering, establishing component maintenance and performance history, failure rates and current maintenance practices to aging and technical obsolescence assessment are reviewed. Information sources for generic failure rates, best industry practices, PM and PdM technologies, aging/degradation and performance trending are discussed. Lastly, methods for the modeling and analysis of the economic parameters are presented. The results of the CWS LCM plans for four of the STARS plants are compared and conclusions are summarized.
Feasibility Study on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles for Reduced-Moderation Water Reactors
Authors: Ohnuki Akira , Kureta Masatoshi , Liu Wei , Tamai Hidesada , Akimoto Hajime
Published: November 2008

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute (JAERI) in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R&D plan and describe the current status focused on an experimental study using large-scale (37-rod bundle) test facility. Steady-state critical power experiments are conducted with the test facility and the experimental data reveal the feasibility of RMWR.
Comparison of Realistic Large Break LOCA Analyses of a 3-Loop Westinghouse Plant Using Response Surface and Statistical Sampling Techniques
Authors: Muftuoglu K. , Ohkawa K. , Frepoli C. , Nissley M.
Published: November 2008

In 1988, the U.S. Nuclear Regulatory Commission (NRC) issued a revision to 10CFR50.46 which allows realistic calculations for loss-of-coolant accident/emergency core cooling system (ECCS) analysis. The revision also requires that uncertainties in the analysis method and inputs be considered, such that there is a high level of probability that the ECCS criteria would not be exceeded. Currently, two methodologies are approved by the U.S. Nuclear Regulatory Commission for such calculations. One technique relies on response surface generation and applying Monte-Carlo technique to the regression model to characterize the PCT distribution and quantify the uncertainty. In this method, uncertainty contributors are ranged individually to compute PCT response. Later, these results are combined to generate a response surface. There exists an additional uncertainty since PCT response to any one parameter or group of parameters is not independent from other contributors. The other technique uses non-parametric order statistics where uncertainty parameters are sampled simultaneously for each transient calculation. The required number of transient runs depends on the desired tolerance intervals and is independent of the number of contributing uncertainty parameters. Westinghouse has developed realistic LOCA methodologies based on both techniques. In this study a 3-loop Westinghouse plant is analyzed using each technique. Use of the same input model and same computational tool for thermal-hydraulic response of the system gives the direct comparison between the two uncertainty calculation techniques. For the example used in comparison, the calculated 95th percentile PCT with the response surface technique is 1171 °C (2140 °F) as opposed to 1045 °C (1913 °F) using the statistical sampling technique. The difference between the computed results is mainly associated with the elimination of superposition uncertainty. In addition, there are inherent differences between the two techniques, in their theory and as well as in their application. This study also compares their main features.
IRIS PRA Preliminary Results and Future Direction
Authors: Finnicum David J. , Kling Charles L. , Carelli Mario D.
Published: November 2008

Westinghouse is currently conducting the pre-application licensing of the International Reactor Innovative and Secure (IRIS) on behalf of the IRIS Consortium. One of the key aspects of the IRIS design is the concept of safety-by-design. The PRA is being used as an integral part of the design process. As part of this effort, a PRA of the initial design was generated to address two key areas. First, the IRIS PRA supported the evaluation of IRIS design issues by providing a solid risk basis for design and analyses required for the pre-licensing evaluation of the IRIS design. The PRA provides the tool for quantifying the benefit of the safety-by-design approach. Second, the current PRA task is beginning the preparation of the more complete PRA analyses and documentation eventually required for Design Certification. One of the key risk-related goals for IRIS is to reduce the EPZ to within the exclusion area by demonstrating that the off-site doses are consistent with the US Protective Action Guidelines (PAGs) for initiation of emergency response so that the required protective actions would be limited to the exclusion area. The results of the preliminary PRA indicated a core damage frequency of 1.2E−08 for internal initiators. This is a very good result but much work is needed to meet the ambitious goal of no emergency response. The next phase of the PRA analyses will involve a two-fold expansion of the PRA. First, as the design and analyses approach a greater level of detail, the assumptions used for the initial PRA will be reviewed and the models will be revised as needed to reflect the improved knowledge of the system design and performance. Furthermore, as the full plant design advances, the PRA will be expanded to incorporate risk associated with external challenges such as seismic and fire, and to address low power and shutdowns modes of operation. As with the initial work, the PRA will serve as a tool to help improve, not just to evaluate the design.
Ratchetting Behavior of Primary Heat Transport (PHT) Piping Material SA-333 Carbon Steel Subjected to Cyclic Loads at Room Temperature
Authors: Kulkarni Sandeep , Desai Y. M. , Kant T. , Reddy G. R. , Gupta C. , Chakravarthy J. K.
Published: November 2008

Ratchetting behavior of SA-333 Gr. 6 carbon steel used as primary heat transport (PHT) piping material has been investigated with three constitutive models proposed by Armstrong-Frederick, Chaboche and Ohno-Wang involving different hardening rules. Performance of the above mentioned models have been evaluated for a broad set of uniaxial and biaxial loading histories. The uniaxial ratchetting simulations have been performed for a range of stress ratios (R) by imposing different stress amplitudes and mean stress conditions. Numerical simulations indicated significant ratchetting and opening of hysteresis loop for negative stress ratio with constant mean stress. Application of cyclic stress without mean stress (R = −1.0) has been observed to produce negligible ratchet-strain accumulation in the material. Simulation under the biaxial stress condition was based on modeling of an internally pressurized thin walled pipe subjected to cyclic bending load. Numerical results have been validated with the experiments as per simulation conditions. All three models have been found to predict the observed accumulation of circumferential strain with increasing number of cycles. However, the Armstrong Frederick (A-F) model was found to be inadequate in simulating the ratchetting response for both uniaxial as well as biaxial loading cases. The A-F model actually overpredicted the ratchetting strain in comparison with the experimental strain values. On the other hand, results obtained with the Chaboche and the Ohno-Wang models for both the uniaxial as well as biaxial loading histories have been observed to closely simulate the experimental results. The Ohno-Wang model resulted in better simulation for the presents sets of experimental results in comparison with the Chaboche model. It can be concluded that the Ohno-Wang model suited well compared to the Chaboche model for above sets of uniaxial and biaxial loading histories.
Transition Risk Method and Example
Authors: Schneider Raymond E. , Farkas Steven E.
Published: November 2008

Paragraph (a)(4) of the Maintenance Rule (re 10CFR§50.65) states that before performing maintenance activities, the licensees shall assess and manage the increase in risk that may result from the maintenance activities. The rule is explicitly applicable to all operating modes. Currently, most plants use a qualitative tool for assessing and controlling the risk of the various plant conditions during an outage. Fewer plants have any means of performing a quantitative or qualitative assessment of the associated risks of transitioning the plant in each configuration from power to “cold shutdown.” Typically, only the end-state of shutdown is considered. The transition-period includes short-duration configurations when the available set of equipment is not what it was during power operations, e.g., having only one main feedwater train in-service. Given the concern that the NRC may require quantitative risk assessments of plant transitions and plant configurations during shutdown operations, Omaha Public Power District (OPPD) pro-actively authorized Westinghouse Engineering Services to develop a method for assessing risk associated with a transition from all power to shutdown and back to full power. An outage schedule is highly plant specific, with plant-to-plant and outage-to-outage variations in system configurations, and maintenance practices. Accordingly, the duration of the transition largely depends on equipment maintenance activities driving the decision to shutdown and repair. The time spent in various parts of the transition is a strong determinant in the associated risk of the transition. A transition takes the plant through a series of Plant Operational States (POSs). The features important to the characterization of each of the POSs include decay-heat level, plant activities involved, available equipment, as well as RCS temperature and pressure. The risk of the entire transition comes from calculating a figure-of-merit of each POS which can be loosely thought of as a per-hour core-damage frequency (CDF). This number gets multiplied by the associated duration of the POS. The sum is the transition risk. The effective CDF associated with the transition comes from dividing the POS-specific CDF sum by the total transition time, and converting that result to an annual frequency. Our paper describes decomposing OPPD operating procedures into periods for which we quantified sequences. In particular, the method considers the dominant shutdown failure modes: loss of shutdown cooling, loss of inventory, and loss of offsite power (including both plant centered and grid-related events). The transition example presented herein covers a simple shutdown and restart stemming from an indeterminate-quality problem. That is, all equipment is functional and available to the plant operators.
Strategic Planning for RPV Head Nozzle PWSCC
Authors: White Glenn A. , Hunt E. Stephen , Nordmann Nicolas S.
Published: November 2008

Life cycle management of RPV heads is a significant issue for utilities with PWR plants. Experience has shown the potential for primary water stress corrosion cracking (PWSCC) of the Alloy 600 CRDM nozzles and/or Alloy 182 J-groove welds which can lead to leaks if allowed to propagate. Non-destructive examinations to detect the PWSCC at an early stage are expensive and potentially time consuming while failure to perform timely inspections can lead to leaks, boric acid corrosion, and the possible need to perform repairs during outages when no provisions have been made for inspections or repairs. Decisions regarding the optimum long term strategic plan can be made based on economic modeling of alternative strategies in combination with probabilistic risk modeling based on plant-specific and industry generic inspection results. This paper describes the key factors in developing a strategic plan that takes into account planned costs, potential repair costs, costs of lost production, and other consequential costs, while ensuring safety. Modeling tools discussed include welding residual stress analyses, fracture mechanics calculations, crack initiation and growth modeling, and probabilistic Monte Carlo simulation.
Analytical and CFD Investigation of Ex-Core Cooling of the Nuclear Fuel Rod Bundle in a Water Pool
Authors: Stevanovic Vladimir , Stosic Zoran , Stoll Uwe
Published: November 2008

The efficiency of ex-core cooling of nuclear fuel assemblies under decay heat generation is influenced by many conditions, among them being coolant flow rate, position of fuel assemblies in a water pool, and position of coolant inlets and outlets. Although the decay heat generation is lower than the nominal heat power of fuel bundles in operation, the much lower coolant flow rates and coolant inlets and outlets positions can lead to incidents conditions, with a violation of the fuel assembly integrity. Such a combination of unacceptable thermal-hydraulic conditions occurred at the Nuclear Power Plant Paks in Hungary in April 2003, during the process of nuclear fuel assembly chemical cleaning in a specially designed tank. The cooling of the nuclear fuel rod bundles in the tank was not efficient under low coolant flow rates through the cleaning tank, and after several hours the boiling of cooling water occurred with subsequent dry-out of nuclear fuel rod bundles. The thermal-hydraulic conditions in the cleaning tank that led to the unexpected event are analysed both analytically and with a CFD approach for idealised conditions of one nuclear fuel rod bundle with the bottom by-pass opening. The analytical analysis is based on a pressure balance of low Reynolds number upward water coolant flow through the bundle, downward water flow in the pool around the bundle, flow across the by-pass opening and outlet flow from the cleaning vessel. The transient CFD simulations are performed in order to demonstrate multidimensional effects of the event. The water density dependence on the temperature is taken into account in both analytical and CFD investigation, as the dominant effect that influences the buoyancy forces between the water flow streams inside and outside the vertically positioned bundle in the water pool. The influence of the bundle bottom by-pass area on the water pool thermal-hydraulic conditions and on the efficiency of the nuclear fuel rods cooling is analysed. Both analytical and CFD results show that the continuous cooling of the fuel rods can not be achieved for higher values of the bundle bottom by-pass areas. The averaged coolant temperature in the water pool outside the bundle becomes higher than the average temperature along the rod bundle, providing a “negative” buoyancy force that tends to stop the upward coolant flow through the bundle, and, hence, increases the coolant flow through the bundle by-pass at the bottom. The critical value of the by-pass area, above which the rod bundle cooling is deteriorated, is predicted.
HELIOS Code Simulation of Condensation Induced Hydrogen Accumulation in Slightly Inclined Non-Vented Pipe
Authors: Stevanovic Vladimir , Stosic Zoran , Stoll Uwe
Published: November 2008

The radiolytic gases (hydrogen and oxygen) accumulation in a gas mixture with steam, due to the steam condensation in a nearly horizontal pipe, closed at one end and open at the other, is simulated and analyzed with the HELIOS code. The thermal-hydraulic and physico-chemical code HELIOS is based on three-dimensional two-fluid model of steam, hydrogen and oxygen gas mixture and water condensate flow, as well as on the radiolytic gases mass conservation equations. Transport processes at the gas mixture-water condensate interface are modelled with appropriate closure laws. Obtained results show that the occurrence of the radiolytic gases accumulation in the non-vented horizontal pipe depends on the complex influence of several parameters such as pipe diameter, thermo-physical properties of gas mixture and condensate and heat loss rate from the gas mixture to the ambient atmosphere. Performed parametric analyses show that under lower heat loss rates or higher pipe diameters, the hydrogen-oxygen accumulation does not occur due to the buoyancy driven circulation of the gas mixture, which expels the gas mixture stream with the higher radiolytic gases concentration out of the pipe. This behavior is opposite to the hydrogen and oxygen accumulation in the vertical pipe closed at the top and open at the bottom, where the front of the accumulated radiolytic gases always propagates from the pipe top.
A Hybrid Solution Method for the Two-Phase Fluid Flow Equations
Authors: Mousseau Vincent A.
Published: November 2008

This paper will present a hybrid solution algorithm for the two-phase flow equations coupled to wall heat conduction. The partial differential equations in the physical model are the same as in RELAP5. The hybrid solution algorithm couples two solution methods, the solution method currently employed by RELAP5 and an implicitly balanced solution method. The RELAP5 solution method provides a fast solution that is “close” to the correct solution. The implicitly balanced solution method provides an accurate solution that is very stable. The resulting hybrid solution method is both fast and accurate. Results will be presented that show the hybrid solution method is more accurate than the RELAP5 solution method for the same size time step. In addition, results will be presented that show when accuracy and CPU time are considered simultaneously that there are ranges when the hybrid solution algorithm is preferred over the RELAP5 solution method.
A Comparative Evaluation of Licensing Requirements for Dry Storage of Spent Nuclear Fuel in the United States, Germany, Canada and the Russian Federation
Authors: Wonder Edward , Duncan David S. , Howden Eric A.
Published: November 2008

Technical activities to support licensing of dry spent nuclear fuel storage facilities are complex, with policy and regulatory requirements often being influenced by politics. Moreover, the process is often convoluted, with numerous and diverse stakeholders making the licensing activity a difficult exercise in consensus-reaching. The objective of this evaluation is to present alternatives to assist the Republic of Kazakhstan (RK) in developing a licensing approach for a planned Dry Spent Fuel Storage Facility. Because the RK lacks experience in licensing a facility of this type, there is considerable interest in knowing more about the approval process in other countries so that an effective, non-redundant method of licensing can be established. This evaluation is limited to a comparison of approaches from the United States, Germany, Russia, and Canada. For each country considered, the following areas were addressed: siting; fuel handling and cask loading; dry fuel storage; and transportation of spent fuel. The regulatory requirements for each phase of the process are presented, and a licensing approach that would best serve the RK is recommended.
Tracking Aerosols in Large Volumes With the Help of CFD
Authors: Dehbi A.
Published: November 2008

Following an accident, the gas speeds induced by natural convection inside a nuclear power plant containment can be of the order of several tens of cm/s, i.e. two orders of magnitude greater than the settling velocities of the generated aerosols. Accordingly, the airborne fission product particles will most of the time follow the streamlines rather than uniformly settle under the lone influence of gravity, as is traditionally modeled in lumped-parameter containment codes. As a result, the controlling phenomena for aerosol removal need to be assessed using a more rigorous treatment of the forces acting on the particle. The CFD code CFX4.4 is used for to compute the flow field and particle tracks in the Phebus containment during steady state conditions. It is found that the calculated half-lives of the aerosols are within 25% of the experimental values. In addition, 80 to 90% of the particles deposit on the bottom section of the containment, which is also consistent with experiments findings. In stark contrast to the classical treatment, it is found that the aerosol decay constant is a much weaker function of particle size.
Realistic Model of a Double-Walled Containment Vessel for the Assessment of the Areas to Be Repaired
Authors: Jean Cyril , Chantron Joe¨lle , Heinfling Gre´gory , Bisch Philippe
Published: November 2008

The leak tightness of French latest pressurized water reactors relies on the pre-stressing of the inner containment concrete wall. To improve the leak tightness of this wall, a procedure has been developed by EDF, including the application of a composite liner onto the intrados over a surface that encompasses the leaking areas identified during the pressure test as well as the “tensile stress area” derived from finite element computations. The method used for finite element computations (named ‘conventional’ method) is based on a number of conservative assumptions. With a view to optimizing the extent of the areas to be repaired, a realistic method (i.e. a less conservative method) has been developed. The paper presents the realistic model, its application, the main results and elements of validation and concludes with an assessment of the gain in realism.
ACR™ Constructability: Increasing Confidence in Project Objectives
Authors: Ricciuti Rick , Fairclough Neville , Hiebert Len , Elgohary Medhat , Yu Stephen K. W.
Published: November 2008

The ACR-700 has been designed with safety, cost and schedule as major drivers. Early in the concept stage it was demonstrated that short construction schedules can be achieved by paralleling activities, using such techniques as extensive modularization, combined with open top construction of the reactor building, as well as using the latest construction techniques available such as climbing formwork, prefabricated rebar, automatic welding, etc. This paper looks at the continued evolution of detailed construction sequences as a planning tool tied to the Primavera schedule. It also looks at examples where cost and schedule become important factors in making engineering decisions associated with the construction method and sequence. Experience on the modularization benefits achieved from the recently completed Qinshan CANDU project is presented. Finally it draws conclusions on the confidence in achieving construction schedules for both the first and nth units of a series.
Dynamic PIV Measurement of High-Speed Mist Flow
Authors: Ishikawa Masaaki , Hong Seong-Dae , Okamoto Koji , Madarame Haruki
Published: November 2008

Recently PIV (Particle Image Velocimetry) system which consists of high-speed cameras with 1k × 1k pixels and 2 kHz in frame rate and high-repetition rate and high-power double-pulse lasers has been developed. Previous PIV system is difficult to visualize the sequential instantaneous images in high-speed flow because of low frame rate as 30Hz and was used to apply to periodic flow and steady flow. According to this latest PIV system, it is possible to acquire high-speed flow of 1 msec consecutively. This camera system is called “Dynamic PIV system” or “High-Speed PIV system”. Also in PIV it is possible to capture two-consecutive images for about a micro-second to use the frame straddling technique. But in latest high-speed camera it has limitation because dead time which is to flush the charged electron perfectly is longer and it has a several micro-seconds. This purpose of study is to develop the new Dynamic PIV system which is possible to capture the two-consecutive images for less than dead time. Furthermore it is applied to the high-speed mist jet flow with 50m/s in order to elucidate this system.
Water Confinement Effects in Response of Fuel Assembly to Faulted Condition Loads
Authors: Shah S. J. , Brenneman B. , Williams G. T. , Strumpell J. H.
Published: November 2008

It has been established by other authors [1] that the accelerations of the water confined by the reactor core baffle plates has a significant effect on the responses of all the fuel assemblies during LOCA or seismic transients. This particular effect is a consequence of the water being essentially incompressible, and thus experiencing the same horizontal accelerations as the imposed baffle plate motions. These horizontal accelerations of the fluid induce lateral pressure gradients that cause horizontal buoyancy forces on any submerged structures. These forces are in the same direction as the baffle accelerations and, for certain frequencies at least, tend to reduce the relative displacements between the fuel and baffle plates. But there is another confinement effect — the imposed baffle plate velocities must also be transmitted to the water. If the fuel assembly grid strips are treated as simple hydro-foils, these horizontal velocity components change the fluid angle of attack on each strip, and thus may induce large horizontal lift forces on each grid in the same direction as the baffle plate velocity. There is a similar horizontal lift due to inclined flow over the rods when axial flow is present. These combined forces appear to always reduce the relative displacements between the fuel and baffle plates for any significant axial flow velocity. Modeling this effect is very simple. It was shown in previous papers [2,3] that the mechanism for the large fuel assembly damping due to axial flow may be the hydrodynamic forces on the grid strips, and that this is very well represented by discrete viscous dampers at each grid elevation. To include the imposed horizontal water velocity effects, on both the grids and rods, these dampers are simply attached to the baffle plate rather than “ground”. The large flow-induced damping really acts in a relative reference frame rather than an absolute or inertial reference frame, and thus it becomes a flow-induced coupling between the fuel and baffles. This has a significant effect on the fuel assembly motions and tends to reduce the relative displacements and impact forces between fuel assemblies and baffle walls.
Large-Scale Biomolecular Dynamics Using SMP Clusters
Authors: Suzuki Masaaki , Okuda Hiroshi , Yagawa Genki
Published: November 2008

The authors have applied Message Passing Interface (MPI) / OpenMP hybrid parallel programming model to molecular dynamics (MD) method for simulating a protein structure on a symmetric multiprocessor (SMP) cluster architecture. In that architecture, it can be expected that the hybrid parallel programming model, which uses the message passing library such as MPI for inter-SMP node communication and the loop directives such as OpenMP for intra-SMP node parallelization, is the most effective one. In this study, the parallel performance of the hybrid style has been compared with that of conventional flat parallel programming style, which uses only MPI, both in case that the fast multipole method (FMM) is employed for computing long-distance interactions and that is not employed. The computer environments used here are Hitachi SR8000/MPP placed at the University of Tokyo. The results of calculation are as follows: Without using FMM, the parallel efficiency using 16 SMP nodes (128 PEs) is: - 90% with the hybrid style, - 75% with the flat-MPI style, for MD simulation with 33,402 atoms. With FMM, the parallel efficiency using 16 SMP nodes (128 PEs) is: - 60% with the hybrid style, - 48% with the flat-MPI style, for MD simulation with 117,649 atoms.
Materials Interaction Tests to Identify Base and Coating Materials for an Enhanced In-Vessel Core Catcher Design
Authors: Rempe J. L. , Knudson D. L. , Condie K. G. , Swank W. D. , Suh K. Y. , Cheung F.-B. , Kim S.-B.
Published: November 2008

An enhanced in-vessel core catcher is being designed and evaluated as part of a joint United States (U.S.)–Korean International Nuclear Engineering Research Initiative (INERI) investigating methods to insure In-Vessel Retention (IVR) of core materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulating oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. Initial evaluations suggest that a thermally-sprayed oxide material is the most promising candidate insulator coating for a core catcher. As part of the effort to develop an in-vessel core catcher design, a series of high temperature materials interaction tests were conducted for thermal sprayed coatings and base materials with properties deemed most promising. This paper reports results from these materials interactions tests and efforts to optimize parameters for applying the thermal spray coatings.
Spectral Dynamics in a Channel Flow With Microbubbles
Authors: Jimenez Bernal Jose Alfredo , Hassan Yassin A. , Estrada Perez Carlos E. , Gutierrez-Torres Claudia del Carmen
Published: November 2008

Measurements were carried out near the upper wall of a turbulent water channel flow at Reynolds number equal to 10256 (considering total height of the channel and the bulk velocity). Particle Image Velocimetry (PIV) measurement technique was used to record instantaneous velocity fields in the x-y plane. The evaluation of the one dimensional spectra for the fluctuating streamwise velocity was performed in the streamwise wavenumber direction (k1) with and without microbubbles in the boundary layer. A redistribution of the energy from larger wavenumber to lower wavenumber was observed for the spectra close to the wall. However, the opposite trend was obtained in the outer region. The energy content with microbubbles was less than the energy without microbubbles in the near wall region of the boundary layer.
Heat Transfer Characteristic and Critical Heat Flux in Mist Dispersed Flow
Authors: Tanaki Tomoyasu , Nemoto Ken , Ohtake Hiroyasu , Koizumi Yasuo
Published: November 2008

The heat transfer in mist cooling for low droplet density, focusing on the heat transfer characteristics, behaviors of liquid on a heating surface and measurements of liquid droplets by LDA was investigated experimentally. Steady heat transfer experiments of a copper block were conducted for mist flow of air and water in a range of air flow rate from 40 to 120 1N/min. Water flow rate was 0.3, 0.9, 1.8, 4.0 and 8.0 1/hr, respectively. Mist flow of water and air forming in a fully conical nozzle with a mixture camber was supplied on the heating surface arranged for horizontal-upward position. The critical heat flux increased with an increasing liquid flow rate. The critical heat flux decreased as the air flow rate increased. Three correlations of the mist cooling rate for non-boiling, evaporation of droplets and evaporation of the liquid film were developed with microscopic parameters of two-phase flow, respectively.