bluuebunny/crossref_metadata_2025 huggingface dataset:
Experimental Study on Thermal-Hydraulics During Start-Up in the Natural Circulation Boiling Water Reactor Concept: Effects of System Pressure and Increasing Heat Flux on the Geysering and Density Wave Oscillation
Authors: Subki M. Hadid , Aritomi Masanori , Watanabe Noriyuki , Muncharoen Chaiwat
Published: March 2009

The feasibility study in thermal-hydraulics for the future light water reactor concept is carried out. One of the essential studies is the two-phase flow instability during start-up in the natural circulation boiling water reactor (BWR) concept. It is anticipated that the occurrence of the two-phase flow instabilities during start-up significantly affects the feasibility concept, since it would cause the complexity in raising and maneuvering the power output. The purpose of the current study is to experimentally investigate the driving mechanism of the geysering and density wave oscillation in the natural circulation loop, induced by a range of system operating pressure and increasing heat flux in vertical parallel channels. The pressure range of atmospheric up to about 4 bars, and the input heat flux range of 0 up to 577 kW/m2 are applied in these experiments. An experimental apparatus of twin boiling upflow channels to simulate natural circulation flow loop has been designed, constructed and operated. The natural circulation in the loop occurs due to the density difference between two-phase region in the channels and the single-phase liquid in the downcomer. The objective of the study is to propose a rational start-up procedure in which the geysering and density wave oscillation can be prevented during startup, according to its system pressure and heat flux. Previous studies have clarified that three (3) kinds of thermo-hydraulics instabilities may occur during start-up in the natural circulation BWR depending on its procedure and reactor configuration, which are (1) geysering induced by condensation, (2) natural circulation induced by hydrostatic head fluctuation in steam separator, and (3) density wave oscillation.
Numerical Visualization of Water-Vapor Flow Configurations in Fusion Reactors During Ingress of Coolant Events
Authors: Ose Yasuo , Takase Kazuyuki , Yoshida Hiroyuki , Akimoto Hajime
Published: March 2009

An integrated Ingress-of-Coolant Event (ICE) test facility was constructed to demonstrate that the International Thermonuclear Experimental Reactor (ITER) safety design approach and design parameters for the ICE are adequate. Major objectives of the integrated ICE test facility are to estimate the performance of an integrated pressure suppression system and obtain the validation data for safety analysis codes for fusion reactors. The integrated ICE test facility simulates the current ITER components with a scaling factor of 1/1600. The modified Transient Reactor Analysis Code (TRAC) is used to verify the integrated ICE test results and clarify quantitatively the two-phase flow behavior in ITER during the ICE. From the results of the present study the effectiveness of the ITER pressure suppression system was verified experimentally, and then, water-vapor flow configurations in ITER at the ICE were visualized numerically by the three-dimensional computations using the modified TRAC.
Direct Contact Heat Exchange Interfacial Phenomena for Liquid Metal Reactors: Part II — Void Fraction
Authors: Abdulla S. , Liu X. , Anderson M. H. , Bonazza R. , Corradini M. L. , Cho D. H. , Page R. J. , Hurtault D.
Published: March 2009

One concept being considered for steam generation in innovative nuclear reactor applications, involves water coming into direct contact with a circulating molten metal. The vigorous agitation of the two fluids, the direct liquid-liquid contact and the consequent large interfacial area can give rise to large heat transfer coefficients and rapid steam generation. For an optimum design of such direct contact heat exchange and vaporization systems, detailed knowledge is necessary of the various flow regimes, interfacial transport phenomena, heat transfer and operational stability. In order to investigate the interfacial transport phenomena, heat transfer and operational stability of direct liquid-liquid contact, a series of experiments are being performed in a 1-d test facility at Argonne National Laboratory and a 2-d experimental facility at UW-Madison. Each of the experimental facilities primarily consist of a liquid-metal melt chamber, heated test section (10cm diameter tube for 1-d facility and 10cm × 50cm rectangle for 2-d facility), water injection system and steam suppression tank. This paper is part II which, primarily addresses results and analysis of a set of preliminary experiments and void fraction measurements conducted in the 2-d facility at UW-Madison, part I deals with the heat transfer in the 1-d test facility at Argonne National Laboratory. A real-time high energy X-ray imaging system was developed and utilized to visualize the multiphase flow and measure line-average local void fractions, time-dependent void fraction distribution as well as estimates of the vapor bubble sizes and velocities. These measurements allowed us to determine the volumetric heat transfer coefficient and gain insight into the local heat transfer mechanisms. In this study, the images were captured at frame rates of 100 fps with spatial resolution of about 7mm with a full-field view of a 15cm square and five different positions along the test section height. The full-field average void fraction increases rapidly to about 15% in these preliminary tests, with the apparent boiling length of less than 20cm. The volumetric heat transfer coefficient between the liquid metal and water are compared to the CRIEPI data, the only prior data for direct contact heat exchange for these liquid metal/water systems.
Heat Transfer Crisis and Liquid Distribution Between the Flow Core and Film in Steam Generating Channels With Swirling
Authors: Boltenko E. A.
Published: March 2009

The results of an investigation of the critical heat flux on the concave and convex heat transfer surfaces of a steam generating channel with coolant flow swirl have been presented. It is established that the presence of convex heat transfer surfaces results in CHF decrease and, consequently, in channel power rating decrease.
Flow Induced Vibration and Fretting Wear in PWR Fuel
Authors: King S. J. , Young M. Y. , Seel D. D. , Conner M. E. , Lu Roger Y. , Paramonov D. V.
Published: March 2009

Fuel rod fretting wear is due to complex combinations of factors. Excitation of the fuel rod and motion relative to its supports can be caused by coolant flow and mechanical forces. The amplitude of response for a given set of flow and mechanical forces is dependent on the fuel rod and its support system. The fuel rod mass and stiffness and location of supports along its length dictate the mode shapes and natural frequencies. The contact geometry between the support system and fuel rod is also an important factor. The two primary areas covered in this paper are 1) the types of flow conditions that may exist in a reactor core and how they would result in excitation of the fuel rod and 2) how the relative motion of the rod to its supports can result in fretting wear. Numerous testing techniques, ranging from microscopic, to single grid cell tests, to high temperature and flow full scale multiple fuel assembly tests, are discussed.
Possibility of Different Fuel Cycles Usage in GT-MHR
Authors: Kodochigov N. , Sukharev Yu. , Marova E. , Ponomarev-Stepnoy N. , Glushkov E. , Fomichenko P. , Ellis C.
Published: March 2009

The GT-MHR reactor core is characterized by flexibility of neutronic characteristics at the given average power density and fixed geometrical dimensions of reactor core. Such flexibility makes it possible to start the reactor operation with one fuel cycle, and then to turn to another type of core fuel load without changes of main reactor elements: fuel block design, core and reflector size, control rod number etc. Preliminary analysis reindicates the commercial viability of the GT-MHR, part of which is due to the ability to accommodate different fuel types and cycles. This paper presents the results of studies of the neutronic characteristics of reactor cores using different fuel (low- and high-enriched uranium, MOX fuel). Comparison of different fuel cycles is carried out for a three-batch refueling option with respect to following characteristics: discharged fuel burnup, reactivity change during one partial cycle of fuel burnup, consumption of fissile isotopes per unit of produced energy, power distribution, reactivity effects, control rods worth. It is shown, that the considered options of fuel loads provide the three-year fuel campaign (with accounting of capacity factor ∼ 0,8) without change of core design, number and design of control rods at transition from the one fuel type to another.
Risk-Informed Decisions Optimization in Inspection and Maintenance
Authors: Alzbutas Robertas
Published: March 2009

The Risk-Informed Approach (RIA) used to support decisions related to inspection and maintenance program is considered. The use of risk-informed methods can help focus the adequate in-service inspections and control on the more important locations of complex dynamic systems. The focus is set on the highest risk measured as conditional core damage frequency, which is produced by the frequencies of degradation and final failure at different locations combined with the conditional failure consequence probability. The probabilities of different degradation states per year and consequences are estimated quantitatively. The investigation of inspection and maintenance process is presented as the combination of deterministic and probabilistic analysis based on general risk-informed model, which includes the inspection and maintenance program features. Such RIA allows an optimization of inspection program while maintaining probabilistic and fundamental deterministic safety requirements. The failure statistics analysis is used as well as the evaluation of reliability of inspections. The assumptions regarding the effectiveness of the inspection methods are based on a classification of the accessibility of the welds during the inspection and on the different techniques used for inspection. The probability of defect detection is assumed to depend on the parameters either through logarithmic or logit transformation. As example the modeling of the pipe systems inspection process is analyzed. The means to reduce a number of inspection sites and the cumulative radiation exposure to the NPP inspection personnel with a reduction of overall risk is presented together with used and developed software. The developed software can perform and administrate all the risk evaluations and ensure the possibilities to compare different options and perform sensitivity analysis. The approaches to define an acceptable level of risk are discussed. These approaches with appropriate software in partial case is used for the construction and research of the models related to inspections and maintenance planning of Ignalina Nuclear Power Plant (RBMK-1500) piping components. The discussed example is related to risk analysis and inspection program improvements for selected pipe systems. The new risk-informed inspection and maintenance program for selected pipe systems are compared with various alternatives. The usage of risk evaluations to optimize the selection of inspection locations, the inspection interval, and the changes in risk and cost due suggested modifications are demonstrated. The proposed integrated modeling methodic and general model of inspection process can be used as a base for other risk-informed models of inspection process control and risk monitors of complex dynamic systems.
CFD Analyses for Water-Air Flow With the Euler-Euler Two-Phase Model in the Fluent4 CFD Code
Authors: Miettinen Jaakko , Schmidt Holger
Published: March 2009

Framatome ANP develops a new boiling water reactor called SWR 1000. For the case of a hypothetical core melt accident it is designed in such a way that the core melt is retained in the Reactor Pressure Vessel (RPV) at low pressure owing to cooling of the RPV exterior and high reliable depressurization devices. Framatome ANP performs — in co-operation with VTT — tests to quantify the safety margins of the exterior cooling concept for the SWR 1000, for determining the limits to avoid the critical heat fluxes (CHFs). The three step procedure has been set up to investigate the phenomenon: 1. Water-air study for a 1:10 scaled global model, with the aim to investigate the global flow conditions. 2. Water-air study for a 1:10 scaled, 10% sector model, with the aim to find a flow sector with almost similar flow conditions as in the global model. 3. Final CHF experiments for a 1:1-scaled, 10% sector., the boarders of this model have been selected based on the first two steps. The instrumentation for the water/air experiments included velocity profiles, the vertically averaged average void fraction and void fraction profiles in selected positions. The experimental results from the air-water experiments have been analyzed at VTT using the Fluent-4.5.2 code with its Eulerian multiphase flow modeling capability. The aim of the calculations was to learn how to model complex two-phase flow conditions. The structural mesh required by Fluent-4 is a strong limitation in the complex geometry, but modeling of the 1/4 sector from the facility was possible, when the GAMBIT pre-processor was used for the mesh generation. The experiments were analyzed with the 150 × 150 × 18 grid for the geometry. In the analysis the fluid viscosity was the main dials for adjusting the vertical liquid velocity profiles and the bubble diameter for adjusting the phase separation. The viscosity ranged between 1 to 10000 times the molecular viscosity, and bubble diameter between 3 to 100 mm, when the calculation results were adjusted for a good agreement with the experimental data. The analysis results were very valuable for designing the final water/steam facility for final CHF tests. The validation against data from the air-water experiments proved that the present CFD codes approach to the state where they can be used for simulating such two-phase experiments, where the fraction of both phases is essential and the flow is strongly affected by the density differences. It is still too early to predict, if the CFD calculation of the 1:1 scale critical heat flux experiments is successful, could the result be used for formulating a new type of a critical heat flux correlation, where the effects of CRD’s on the flow patterns and gap dimensions are model parameters.
Experimental Study and Analytical Methods for Particle Bed Dryout With Heterogeneous Particles and Pressure Variation
Authors: Miettinen Jaakko , Sairanen Risto , Holmstrom Stefan , Lindholm Ilona
Published: March 2009

The interest to study the dryout heat flux in particle beds is related to interest of quantify the debris coolability margins during a hypothetical severe reactor accident. When the molten core has relocated to the containment floor, one accident management concept is based on the cooling of the corium by the water injection on top. Earlier experimental and analytical work has concentrated on homogeneous particle beds at atmospheric pressures. For plant safety assessment in Finland, there is a need to consider heterogeneous particle mixtures, layered particle bed setups and varied pressures. A facility has been constructed at VTT to measure dryout heat flux in a heterogeneous particle bed. The bed dimensions are 0.3 m in diameter and 0.6 m in height, with a mixture of 0.1 to 10 mm particles. The facility has a pressure range from atmospheric to 6 bar (overpressure). The bed is heated by spirals of a resistance band. The preliminary experiments have been carried out, but a more systematic set of data is expected to be available in the spring 2002. To support the experiments analytical models have been developed for qualification of the experimental results. The first comparison is done against various critical heat flux correlations developed in 1980’ies and 1990’ies for homogeneous bed conditions. The second comparison is done against 1-D and 0-D models developed by Lipinski. The most detailed analysis of the transient process conditions and dryout predictions are done by using the two-dimensional, drift-flux based thermohydraulic solution for the particle bed immersed into the water. The code is called PILEXP. Already the first validation results against the preliminary tests indicate that the transient process conditions and the mechanisms related to the dryout can be best explained and understood by using a multidimensional, transient code, where all details of the process control can be modeled as well. The heterogeneous bed and stratified bed can not be well considered by single critical heat flux correlations.
Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model
Authors: Miettinen Jaakko , Ha¨ma¨la¨inen Anitta , Pekkarinen Esko
Published: March 2009

Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and anticipated transients without scram. Experiences with the GENFLO code have pointed out that the efforts needed for the code development and its validation can be minimized with this kind of the generalized solution.
The Renewal of HTR Development in Europe
Authors: Hittner Dominique
Published: March 2009

The European HTR-Technology Network (HTR-TN), created in 2000, presently groups 20 organisations from European nuclear research and industry for developing the technologies of direct-cycle modular HTRs, which presently raise a large world-wide interest, because of their high potential for economic competitiveness, natural resource sparing, safety and minimisation of the waste impacts, in line with the goals of sustainable development of Generation IV. All aspects of HTR technologies are addressed by HTR-TN, from the reactor physics to the development of materials, fuel and components. Most of this activity is supported by the European Commission in the frame of its 5th Euratom Framework Programme. The first results of HTR-TN programme are given: the analysis of the reactor physics international benchmark on the commissioning tests of HTTR (Japan), the long term behaviour of spent HTR fuel in geologic disposal conditions, the preparation of a very high burnup fuel irradiation and the development of fabrication processes for producing high performance coated particles, etc.
Development of Correlations for Pressure Loss/Drop Coefficients Obtained From Flow Testing of Fuel Assemblies in Framatome ANP’s PHTF
Authors: Madni Imtiaz K. , Stephens Lance G. , Turner Dave M.
Published: March 2009

Thermal-hydraulic analyses of pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies are generally performed for either assembly thermal-hydraulic design, thermal-hydraulic compatibility evaluation, or cycle licensing thermal-hydraulic characterization. A key issue in all cases is the hydraulic resistance characterization of the assembly in which the assembly, its components and support plates, etc., are represented by their respective pressure loss and pressure drop coefficients. These hydraulic coefficients can be determined by single-phase flow testing in an experimental facility such as the Framatome ANP Portable Hydraulic Test Facility (PHTF) located at Richland Test Facilities (RTF) in Richland, WA. The goal of this paper is to present a uniform and consistent methodology for the development of coefficient correlations from data obtained from single phase pressure drop testing of PWR and BWR fuel assemblies and their components performed in the PHTF. This methodology reflects the years of accumulated experience from an existing facility with an ongoing test program.
Liquid Ejection Through Orifices by Sparging Gas: The PERCOLA Program
Authors: Tourniaire B. , Seiler J. M. , Bonnet J. M. , Amblard M.
Published: March 2009

Corium coolability after a severe PWR accident involving core meltdown and RPV failure is one of the main items in nuclear safety. The case considered here is a situation in which the corium is supposed to spread over a concrete floor and is flooded by water. In this frame, many researches are performed to study the physical phenomena which may enhance the heat transfer between the corium and the water pool. Among them, the melt entrainment above the corium crust by the sparging gas released by the concrete ablation appears as a potentially efficient cooling mechanism. The main target of the experimental program PERCOLA is to provide qualitative and quantitative information on this entrainment phenomenon. The first part of this paper is devoted to a general description of the experimental program and to the presentation of the main results. In a second part, the attention is focused on the modelling of the liquid entrainment phenomenon and to the comparison between the experimental data and the calculation results of two different entrainment models.
Evaluation of the Risk of Damages in Mixing Zones: EDF R&D Programme
Authors: Stephan J. M. , Curtit F. , Vindeirinho C. , Taheri S. , Akamatsu M. , Peniguel C.
Published: March 2009

Following the incident (through-wall crack and crazing zones) of May 1998 on the principal mixing zone of the residual heat removal system at CIVAUX nuclear plant, EDF has initiated a R&D programme to understand the incident and assess the risks of damage on other nuclear plant mixing zones. The programme includes different sectors of development: assessment of temperature fluctuations in mixing zones, study of high-cycle thermal fatigue behaviour of austenitic steel and development of mechanical methodologies for damage assessment and propagation of crazing zones (thermal striping). A series of tests is being performed in support of the programme. The paper develops the description of each sector of the programme: tests performed, on-site measurements and numerical interpretations. The first results are presented.
Passive and Active Vibration Control With Piezoelectric Fiber Composites
Authors: Vigier Yves , Agbossou Amen , Richard Claude
Published: March 2009

The possibility of dissipating mechanical energy with piezoelectric fiber composites (PFC) is investigated. The techniques for manufacturing an active beam with integrated (PFC) are presented and applied to a cantilevered beam experiment. We evaluated experimentally the performances of the active beam in passive energy dissipation. Three vibration cases were analysed: electrodes of the PFCs are (i) in open circuit, (ii) short circuit and (iii) shunted with electrical impedance designed to dissipate the electrical energy, which has been converted from the beam mechanical energy by the PFCs. Then we presented numerical models to analyze the vibration of active beams connect to electrical impedance. The proposed models point out with an accurate order of magnitude the change in vibration amplitude of the analysed beam. Hence we validate experimentally and numerically the concept of vibration control with PFCs and point out some new contributions of PFCs in active or passive damping.
Generation-IV Multi-Application Small Light Water Reactor (MASLWR)
Authors: Modro S. Michael , Fisher James , Weaver Kevan , Babka Pierre , Reyes Jose , Groome John , Wilson Gary
Published: March 2009

The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) have developed a Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, safe and economic natural circulation pressurized light water reactor. MASLWR reactor module consists of an integral reactor/steam generator located in a steel cylindrical containment. The entire module is to be entirely shop fabricated and transported to site on most railways or roads. Two or more modules are located in a reactor building, each being submersed in a common, below grade cavity filled with water. For the most severe postulated accident, the volume of water in the cavity provides a passive ultimate heat sink for 3 or more days allowing the restoration of lost normal active heat removal systems. MASLWR thermal power of a single module is 150 MWt, primary system pressure 10.5 MPa, steam pressure 1.52 MPa and the net electrical output is 35–50 MWe.
Heat and Fission Product Transport in a Molten U-Zr-O Pool With Crust
Authors: Yun J. I. , Suh K. Y. , Kang C. S.
Published: March 2009

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry and the change of pool geometry during the numerical calculation was neglected. Results of the numerical calculation revealed that the peak temperature of the molten pool significantly decreased and most of the volatile fission products were released from the molten pool during the accident.
Thoria-Based Cermet Nuclear Fuel: Sintered Microsphere Fabrication by Spray Drying
Authors: Solomon Alvin A. , McDeavitt Sean M. , Chandramouli V. , Anthonysamy S. , Kuchibhotla S. , Downar T. J.
Published: March 2009

Spray drying is a physical process of granulating fine powders that is used widely in the chemical, pharmaceutical, ceramic, and food industries. It is generally used to produce flowable fine powders for mechanized processing. Occasionally it is used to fabricate sintered bodies like cemented carbides, and has been used to produce sintered fuel and actinide microspheres [1]. As a physical process, it can be adapted to many powder types and mixtures and thus, has appeal for dispersion nuclear fuels, and waste forms of various compositions. It also permits easy recycling of unused powders, and generates minimal chemical waste streams that can arise in chemical sol/gel processing. On the other hand, the containment of the radioactive powders, present safety challenges that need to be addressed [2]. Detailed formal procedures and methods for characterizing and processing UO2/ThO2 mixtures have been established and approved by the Purdue Radiological Control Committee for (1) ball-milling, (2) viscosity and rheology measurements on slurries, (3) sintering, (4) co-precipitation, (5) particle size analysis using laser scattering, (6) surface area analysis using the BET technique, (7) X-ray diffraction, (8) stoichiometry measurement, (9) zeta potential measurements and (10) ceramographic preparation. The spray drying procedures represented a particular challenge since they deal with the handling of loose powders. Studies were carried out to formulate suitable stable, dense and homogeneous aqueous slurries of urania and thoria powders for the production of urania-thoria microspheres by the spray drying method. The studies included (a) particle size distribution after ball-milling, (b) viscosity, (c) zeta potential, (d) slurry flowability, stability and cleanability, (e) microsphere green strength, and f) effects of organic dispersants on the above properties. After formulating the slurry, U,ThO2 microspheres were produced using a commercial, laboratory-scale spray dryer modified for handling these radioactive materials. The microspheres thus obtained were dried at 473 K for 4 hours, presintered at 1173 K for 2 hours and sintered at 1923 K for 10 hours.
Horizontal Steam Generator Thermal-Hydraulics at Various Steady-State Power Levels
Authors: Stevanovic Vladimir D. , Stosic Zoran V. , Kiera Michael , Stoll Uwe
Published: March 2009

Three-dimensional computer simulation and analyses of the horizontal steam generator thermal-hydraulics of the WWER 1000 nuclear power plant have been performed for 50% and 75% partial loads, 100% nominal load and 110% over-load. Presented results show water and steam mass flow rate vectors, steam void fraction spatial distribution, recirculation zones, swell level position, water mass inventory on the shell side, and other important thermal-hydraulic parameters. The simulations have been performed with the computer code 3D ANA, based on the “two-fluid” model approach. Steam-water interface transport processes, as well as tube bundle flow resistance, energy transfer, and steam generation within tube bundles are modelled with “closure laws”. Applied approach implies non-equilibrium thermal and flow conditions. The model is solved by the control volume procedure, which has been extended in order to take into account the 3D flow of liquid and gas phase. The methodology is validated by comparing numerical and experimental results of real steam generator operational conditions at various power levels of the WWER Novovoronezh, Unit 5. One-dimensional model of the horizontal steam generator has been built with the RELAP 5 standard code on the basis of the multidimensional two-phase flow structure obtained with the 3D ANA code. RELAP 5 and 3D ANA code results are compared, showing acceptable agreement.
Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage
Authors: Einziger R. E. , Tsai H. C. , Billone M. C. , Hilton B. A.
Published: March 2009

Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150°C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission gas release from the fuel pellets occurred during the thermal benchmark tests or storage. Measurements of the cladding outer-diameter, oxide thickness and wall thickness are in the expected range for cladding of the Surry exposure. The measured hydrogen content is consistent with the oxide thickness. The volume of hydrides varies azimuthally around the cladding, but there is little variation across the thickness, of the cladding. It is most significant that all of the hydrides appear to have retained the circumferential orientation typical of prestorage PWR fuel rods.
Performance Evaluation of Passive Safety Injection Flow Controllers for the APR1400 Reactor
Authors: Chu I. C. , Chung H. J. , Park W. M. , Song C. H. , Park J. K.
Published: March 2009

This paper provides experimental results obtained in the full-scale tests for the performance evaluation of the Fluidic Device which is expected to be adopted in a Korean next generation reactor (APR1400). The tests have been carried out using the full-scale test facility called VAPER (Valve Performance Evaluation test Rig). Following characteristics of the fluidic device are investigated in the present work: peak discharge flow rate, duration of high flow, flow rate switching function, water level depression in stand pipe, etc. Experimental results show that the present fluidic device meets the primary performance requirements of APR1400 plant design.
Users’ Requirements for Environmental Effects From Innovative Nuclear Energy Systems and Their Fuel Cycles
Authors: Carreter M. , Gray M. , Falck E. , Bonne A. , Bell M.
Published: March 2009

The objective of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) is to support the safe, sustainable, economic and proliferation resistant use of nuclear technology to meet the needs of the 21st century. The first part of the project focusses on the development of an understanding of the requirements of possible users of innovative concepts for reactors and fuel cycle applications. This paper reports progress made on the identification of user requirements as they relate to the environment and environmental protection. The user requirements being formulated are intended to limit adverse environmental effects from the different facilities involved in the nuclear fuel cycles to be well below maximum acceptable levels. To determine if the user requirements are met, it is necessary to identify those factors that are relevant to assessment of the environmental performance of innovative nuclear systems. To this effect, Environmental Impact Assessment (EIA) and the Material Flow accounting (MFA) methodologies are being appraised for the suitability for application. This paper develops and provides the rationale for the “users’ requirements” as they are currently defined. Existing Environmental Impact Assessment and Materials Flow Accounting methodologies that can be applied to determine whether or not innovative technologies conform to the User Requirements are briefly described. It is concluded that after establishing fundamental principles, it is possible to formulate sets of general and specific users’ requirements against which, the potential adverse environmental effects to be expected from innovative nuclear energy systems (INES) can be assessed. The application of these users’ requirements should keep the adverse environmental effects from INES’s within acceptable limits.
Renovation of Chemical Processing Facility for Development of Advanced Fast Reactor Fuel Cycle System in JNC
Authors: Aoshima Atsushi , Miyachi Shigehiko , Suganuma Takashi , Nemoto Shinichi
Published: March 2009

The CPF had 4 laboratories (operation room A, laboratory A, laboratory C and analysis laboratory) in connection with reprocessing technology. The main laboratory, operation room A, has 5 hot cells. Since equipments in the main cell had been designed for small-scale verification of existing reprocessing steps, it was hardly able to respond flexibly to experimental studies on advanced technology. It was decided to remodel the cell according to the design that was newly laid out in order to ensure the function and space to conduct various basic tests. The other laboratories had no glove boxes for conducting basic experiments of important elements in the advanced reprocessing, such as actinides except U and Pu, lanthanides and so on. In order to meet various requirements of innovative technologies on advanced fuel cycle development, one laboratory is established more for study on dry reprocessing, and glove boxes, hoods and analytical equipments such as NMR, FT-IR, TI-MS are newly installed in the other laboratories in this renovation. After the renovation, hot tests in the CPF will be resumed from April 2002.
Experimental Investigation of Minimum Film Boiling Temperature for Vertical Cylinders at Elevated Pressure
Authors: Peterson L. J. , Bajorek S. M.
Published: March 2009

The minimum film boiling temperature, Tmin, is of particular interest to nuclear safety. Many thermal-hydraulics codes depend on an accurate estimation of Tmin to determine the appropriate heat transfer regime. Heat transfer coefficients increase considerably when the surface temperature decreases below Tmin. This paper presents the results of an experimental investigation of Tmin for vertical cylinders at elevated pressure. While several previous studies have investigated Tmin under various conditions, relatively few have generated experimental data at pressures above atmospheric. As a result, correlations for Tmin often predict unrealistic values at high pressure. The present investigation has generated an experimental database examining the effects of pressure, material, subcooling, surface roughness, and oxidation on Tmin. The experimental apparatus and test specimens are described in addition to the evaluation procedure. Tmin results are reported in this paper for Type 316 stainless steel, Type 1018 carbon steel, and Zircaloy-4. Pressure was varied from 0.101 MPa to 3.0 MPa, and liquid subcooling was ranged up to 30 °C. Zircaloy-4 samples with a thick oxide layer were also tested. The results show that Tmin increases with pressure, liquid subcooling, surface roughness, and surface oxidation. Tmin is a strong function of pressure at low pressure, but asymptotically approaches a constant value as pressure increases. Tmin is also a function of surface material properties, with Zircaloy-4 being found to have the highest minimum film boiling temperature compared to carbon steel and stainless steel. For Zircaloy-4, an oxide layer was found to significantly increase Tmin.
Performance of Key Features of EBR-II and the Implications for Next Generation Systems
Authors: King Ronald W. , Porter Douglas L.
Published: March 2009

The Experimental Breeder Reactor No. 2 (EBR-II) began operation in August 1964 as an engineering test of the operation of a sodium-cooled fast reactor power plant system. Its primary mission was to demonstrate and evaluate the performance of a sodium-cooled reactor system using recycled fuel from an integral fuel cycle facility, and as an electrical power generator tied to the utility grid. It accomplished this mission in the early years of operation. Following the early successes, the mission evolved to include more extensive testing of fuels, materials, components, and safety features of a sodium-cooled fast reactor. The use of sodium as the coolant, use of metal fuel, and a piped-pool primary system configuration, were key contributors to its notable long-term performance. Extensive evaluation and examination of these features have provided a solid basis for and understanding of the technology. Recent interest in future designs for nuclear generating stations has generated renewed interest in liquid-metal-cooled fast reactors in the United States and elsewhere. The successful operation of EBR-II for thirty years and the demonstration of characteristics significant to the development of next-generation reactors, has prompted a re-examination of key features of EBR-II and a review of its performance record. This paper discusses selected key features and their contribution to the performance of EBR-II, evaluates the overall performance of the reactor, and discusses the implications for the development of next generation reactor concepts.
A Proactive Aging/Asset Management Model to Optimize Equipment Maintenance Resources Over Plant Lifetime
Authors: Meyer Theodore A. , Perdue Robert K. , Woodcock Joel , Elder G. Gary
Published: March 2009

Experience has shown that proactive aging/asset management can best be defined as an ongoing process. Station goals directly supported by such a process include reducing Unplanned Capability Loss Factor and gaining the optimum value from maintenance and aging management budgets. An effective aging/asset management process must meet evolving and sometimes conflicting requirements for efficient and reliable nuclear power plant operation. The process should identify most likely contributors before they fail, and develop cost-effective contingencies. Current trends indicate the need for focused tools that give quantitative input to decision-making. Opposing goals, such as increasing availability while optimizing aging management budgets, must be balanced. Recognizing the importance of experience in reducing the uncertainty inherent in predicting equipment degradation rates, nuclear industry demographics suggest the need to capture existing expert knowledge in a usable form. The Proactive Aging/Asset Management Process has been developed to address these needs. The proactive approach is a process supported by tools. The process identifies goals and develops criteria — including safety, costs, and power production — that are used to prioritize systems and equipment across the plant. The process then draws upon tools to most effectively meet the plant’s goals. The Proactive Aging/Asset Management Model™ is one software-enabled tool designed for mathematical optimization. Results assist a plant in developing a plant-wide plan of aging management activities. This paper describes the proactive aging/asset management process and provides an overview of the methodology that has been incorporated in a model to perform a plant-wide optimization of aging management activities.
Development of a Thermal Equilibrium Prediction Algorithm
Authors: Aviles-Ramos Cuauhtemoc
Published: March 2009

A thermal equilibrium prediction algorithm is developed and tested using a heat conduction model and data sets from calorimetric measurements. The physical model used in this study is the exact solution of a system of two partial differential equations that govern the heat conduction in the calorimeter. A multi-parameter estimation technique is developed and implemented to estimate the effective volumetric heat generation and thermal diffusivity in the calorimeter measurement chamber, and the effective thermal diffusivity of the heat flux sensor. These effective properties and the exact solution are used to predict the heat flux sensor voltage readings at thermal equilibrium. Thermal equilibrium predictions are carried out considering only 20% of the total measurement time required for thermal equilibrium. A comparison of the predicted and experimental thermal equilibrium voltages shows that the average percentage error from 330 data sets is only 0.1%. The data sets used in this study come from calorimeters of different sizes that use different kinds of heat flux sensors. Furthermore, different nuclear material matrices were assayed in the process of generating these data sets. This study shows that the integration of this algorithm into the calorimeter data acquisition software will result in an 80% reduction of measurement time. This reduction results in a significant cutback in operational costs for the calorimetric assay of nuclear materials.
Conceptual Analysis of Criticality Aspects of Fission Electric Cell Reactors
Authors: Cash Jamie , Rochau Gary , King Donald , Morrow Charles
Published: March 2009

The U.S. Department of Energy’s Nuclear Energy Research Initiative Direct Energy Conversion project has a goal of developing direct energy conversion (DEC) processes suitable for commercial development. DEC is any fission process that returns usable energy with no intermediate thermal process. This project includes the study of the fission electric cell (FEC). In the FEC, fission fragments exit the fuel element cathode and are collected by the cell anode. Previous work [1] has shown the potential of FECs with theoretical efficiencies up to 60%. Inspection of this work indicates the need for additional system criticality studies prior to any conclusions regarding the final FEC reactor configuration. This paper outlines the development of models to facilitate reactor criticality design decisions. The models address criticality, design life, and reactor configuration. In addition, this paper proposes future work to complete the criticality model. The direct energy conversion concept converts nuclear energy to electrical energy without the use of a Carnot cycle based system. Kinetic energy of the highly charged fission fragments is converted directly into electrical potential using strong magnetic fields to separate the positive fission fragments from the electrons that are also produced during the fission process. A parametric analysis is performed using Monte Carlo N-Particle (MCNP) [2] simulations to calculate criticality for exact geometric models. The effect on criticality of changing enrichment, number of cells, size of cells, fuel thickness, and reactor size is determined. Heavy water, helium, and beryllium are each considered for a reflector design.