bluuebunny/crossref_metadata_2025 huggingface dataset:
R&D of the Next Generation Safety Analysis Methods for Fast Reactors With New Computational Science and Technology: 3 — Improvement of Basic Fluid Dynamics Models for the COMPASS Code
Authors: Zhang Shuai , Morita Koji , Shirakawa Noriyuki , Yamamoto Yuichi
Published: July 2009

A new next generation safety analysis code, COMPASS, is designed based on the moving particle semi-implicit (MPS) method to provide local information for various key phenomena in core disruptive accidents of sodium-cooled fast reactors. In FY2006, improvement of basic fluid dynamics models for the COMPASS code was carried out and verified with fundamental verification calculations. In order to improve the numerical stability of MPS simulations, a fully implicit pressure solution algorithm was introduced instead of the two-stage MAC algorithm originally applied by MPS. With a newly developed free surface model, numerical difficulty caused by poor pressure solutions is overcome by involving free surface particles in the pressure Poisson equation. An improved algorithm was also proposed for surface tension calculation with the continuous surface force (CSF) model applied to the moving particle method. This algorithm provides higher numerical precision with the CSF model by interpolation between moving particles and background mesh. Application of the fully Lagrangian MPS method to solid-fluid mixture flow problems is straightforward. In FY2006, applicability of the MPS method to interactions between fluid and multi-solid bodies was investigated in comparison with dam-break experiments with solid balls. It was found that a modified pressure solution algorithm makes simulation with the passively moving solid model stable numerically. Though characteristic behavior of solids was successfully reproduced by the present numerical simulations, the comparisons with the experimental results showed that interactions between solids and solid-wall should be modeled for more precise simulations. Therefore, the discrete element method will be considered in the next stage.
Development of Three-Dimensional Virtual Plant Vibration Simulator on Grid Computing Environment ITBL-IS/AEGIS
Authors: Suzuki Yoshio , Nakajima Norihiro , Araya Fumimasa , Hazama Osamu , Nishida Akemi , Kushida Noriyuki , Akutsu Taku , Teshima Naoya , Nakajima Kohei , Kondo Makoto , Hayashi Sachiko , Aoyagi Tetsuo
Published: July 2009

Center for computational science and e-systems of Japan Atomic Energy Agency is carrying out R&D in the area of extra large-scale simulation technologies for solving nuclear plant structures in its entirety. Specifically, we focus on establishing a virtual plant vibration simulator on interconnected supercomputers intended for seismic response analysis of a whole nuclear plant. The simulation of the whole plant is a very difficult task because an extremely large dataset must be processed. To overcome this difficulty, we have proposed and implemented a necessary simulation framework and computing platform. The simulation framework based on the computing platform has been applied to a linear elastic analysis of the reactor pressure vessel and cooling systems of a nuclear research facility, the HTTR. The simulation framework opens a possibility of new simulation technologies for building a whole virtual nuclear plant in computers for virtual experiments.
Westinghouse Advanced Loop Tester (WALT) Update
Authors: Wang G. , Byers W. A. , Young M. Y. , Karoutas Z. E.
Published: July 2009

In order to understand crud formation on the fuel rod cladding surfaces of pressurized water reactors (PWRs), a crud Thermal-Hydraulic test facility referred to as the Westinghouse Advanced Loop Tester (WALT) was built at the Westinghouse Science and Technology Department Laboratories in October 2005. Since then, a number of updates have been made and are described here. These updates include heater rod improvements, system pressure stabilization, and more effective protection systems. After these updates were made, the WALT system has been operated with higher stability and fewer failures. In this test loop, crud can be deposited on the heater rod surface and the character of the crud is similar to what has been observed in the PWRs. In addition, chemistry in the WALT loop can be varied to study its impact on crud morphology and associated parameters. The WALT loop has been successful in generating crud and measuring its thermal impact as a function of crud thickness. Currently, this test facility is supporting an Electric Power Research Institute (EPRI) program to assess the impact of zinc addition to PWR reactor coolant. Meanwhile, the WALT system is also being utilized by Westinghouse to perform dry-out and hot spot tests. These tests support the industry goal of 0 fuel failures by 2010 set by Institute of Nuclear Power Operations (INPO). Another major goal of the Westinghouse tests is to gain a better understanding of unexpected changes in core power distributions in operating reactors known as crud induced power shifts (CIPS) or axial offset anomalies (AOA).
Experimental and Numerical Modeling of Transition Matrix From Momentum to Buoyancy-Driven Flow in a Pressurized Water Reactor
Authors: Ho¨hne Thomas , Kliem So¨ren , Vaibar Roman
Published: July 2009

The influence of density differences on the mixing of the primary loop inventory and the Emergency Core Cooling (ECC) water in the cold leg and downcomer of a Pressurised Water Reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. This paper presents a matrix of ROCOM experiments in which water with the same or higher density was injected into a cold leg of the reactor model with already established natural circulation conditions at different low mass flow rates. Wire-mesh sensors measuring the concentration of a tracer in the injected water were installed in the cold leg, upper and lower part of the downcomer. A transition matrix from momentum to buoyancy-driven flow experiments was selected for validation of the CFD software ANSYS CFX. A hybrid mesh with 4 million elements was used for the calculations. The turbulence models usually applied in such cases assume that turbulence is isotropic, whilst buoyancy actually induces anisotropy. Thus, in this paper, higher order turbulence models have been developed and implemented which take into account for that anisotropy. Buoyancy generated source and dissipation terms were proposed and introduced into the balance equations for the turbulent kinetic energy. The results of the experiments and of the numerical calculations show that mixing strongly depends on buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation with lower mass flow rates and/or higher density differences. The ECC water falls in an almost vertical path and reaches the lower downcomer sensor directly below the inlet nozzle. Therefore, density effects play an important role during natural convection with ECC injection in PWR and should be also considered in Pressurized Thermal Shock (PTS) scenarios. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.
R&D of the Next Generation Safety Analysis Methods for Fast Reactors With New Computational Science and Technology: 5 — Study of Eutectic Reaction Between Metals: Classical Molecular Dynamics Approach
Authors: Ito Takahiro
Published: July 2009

Molecular dynamics study on eutectic phenomena between two metal materials is performed for development of a new MPS-based code named ‘COMPASS’ [1] being developed for meso-scale analysis of core disruptive accidents in Sodium-cooled fast reactor. In this study, Cu-Ag system is simulated by classical molecular dynamics to provide knowledge on detailed mechanisms on the inception and extension of eutectic melting. Embedded Atom Method [2] which has been used in many studies associated with atomic behavior of metals is applied. First, melting temperature of solid solution of Cu-Ag binary system is investigated. The analytical result on the dependence of the melting temperature on the atomic fraction of Cu shows good agreement with the experiments. The melting behavior on the interface between two pure Cu and Ag slabs are then simulated. It is shown that the melting temperature at the interface is lowered depending on the local value of the atomic fraction there and is almost identical to that of solid solution with the corresponding atomic fraction.
Experimental Analysis of Containment Atmosphere Mixing and Stratification Induced by Vertical Fluid Release
Authors: Gupta Sanjeev , Paladino Domenico , Benz Pascal , Prasser Horst Michael
Published: July 2009

Continuing safety assessments are important for the present and future generations of nuclear reactors to ensure safe and efficient use of nuclear energy. For safety analysis purpose, lumped parameter codes as well as codes with 3D capabilities are used to predict and simulate the containment conditions following a hypothetical accident. The assessment of code capability to predict consequences of hypothetical accidents can be done by comparing code simulations with experimental data obtained in large scale facilities under prototypical thermal hydraulic conditions. The large scale PANDA test facility built at the Paul Scherrer Institute (PSI) in Switzerland is well suited to perform thermal hydraulic experiments for investigating integral containment system response during accidents and also to study multicompartmental 3D effects related to a Light Water Reactor (LWR). This paper presents the experimental results and the analysis of two tests performed in a containment that is represented by two large vessels (volume of each vessel is 90 m3) interconnected by a bended pipe of 1 m diameter. The tests address the possibility of the permanence of hydrogen stratification just beneath the containment dome under conditions determined by low velocity and a plume or jet of higher density which could be, for instance, due to steam produced by evaporation of the sump. Since practically no data are available on these specific phenomena in large scale test facilities, the capabilities of the codes could not be yet verified. During the experiments, a stratified helium-rich layer is created in upper part of one of the two vessels and thereafter, interaction of an ascending buoyant jet with the stratified helium layer is investigated. Helium is used to simulate hydrogen released by cladding metal water reaction. Two tests with initial densimetric Froude number equal to 22.6 (TestAir/He) and 104 (TestSteam/He) are investigated. The distance between the pipe exit for vertical injection of the fluid and the bottom front of the stratified helium rich layer is about 2m for both the investigated test cases. Results depicting the mixing of lower and upper vessel regions are presented in terms of helium concentration profiles and temperature contour maps. The results show that the maximum penetration height is controlled by the initial exit conditions (momentum flux and the buoyancy flux) and density stratification inside the test vessel. Additionally, for the given test conditions and geometry, most of the injected fluid remains trapped in upper part of one of the two test vessels in which helium layer is created.
Near-Term Deployment of Advanced Light Water Reactors
Authors: Hamel Jeffrey F.
Published: July 2009

New nuclear power plants incorporating advanced light water reactor (ALWR) technology must overcome a number of regulatory, economic, technical, and social barriers prior to licensing, construction and successful start-up. Many of these barriers can be addressed through technical products and targeted tools that minimize deployment risks. EPRI’s Advanced Nuclear Technology (ANT) Program has been initiated to complement — and help accelerate — industry activities aimed at enabling and building confidence in new nuclear plant deployment through coordinated work on cross cutting issues. The program is built around three core elements: transferring technology to new plant designs, developing effective tools for performance management, and providing robust planning and evaluation tools for new plant deployment. This paper discusses the projects EPRI is working on with industry to enable and build confidence in new nuclear plant deployment.
Assessment of DES Multiscale Turbulence Models for Prediction of Flow and Heat Transfer in an Axial-Channel Rod Configuration
Authors: Basu Debashis , Das Kaushik , Painter Scott L. , Howard Lane D. , Green Steven T.
Published: July 2009

This paper presents results of a computational study conducted to assess the multiscale resolution capabilities and limitations of different Detached Eddy Simulation (DES) multiscale turbulence models in unsteady flow predictions for internal axial flow in a single rod channel configuration. Two different DES models are compared in the present analysis. The DES models are based on the Spalart-Allmaras (S-A) one-equation model and the two-equation realizable k-ε model. A detailed assessment of the DES turbulence model coefficient for the S-A based DES model is presented. The predicted time-averaged mean velocity and turbulent stresses are compared with the available experimental results. Flow unsteadiness, which is important for determining heat, momentum, and mass transfer in the gap region, is presented through time histories and spectra of flow quantities. The unsteady spectra for the velocity fluctuations are also compared with the experimental observations. The results demonstrate that the DES turbulence model coefficient significantly influence the predicted solution. The realizable k-ε-model-based DES model is found to be numerically more stable than the one-equation S-A-based DES model. Predicted results demonstrate that the modifications need to be incorporated in the current DES model formulations for proper prediction of wall bounded internal turbulent flows.
Experimental Investigation of Saturated Flow Boiling Heat Transfer of Vertical Narrow Rectangular Channel
Authors: Chen De-Qi , Pan Liang-Ming , Yuan De-Wen
Published: July 2009

Narrow rectangular channel has been widely used at evaporator of cryogenic engineering and electronics cooling. At present paper, saturated flow boiling heat transfer at vertical narrow rectangular channel with gaps of 1.0, 1.5 and 2.5 was experimentally investigated, the heat transfer characteristics at various operating conditions were tested and discussed, and the correlation of saturated boiling heat transfer coefficient was regressed used Least Square Method according to experimental results. It shows that heat transfer coefficient decreases at wider gaps and higher inlet pressure, and it increases at higher flow rate, heat flux and increased vapor quality. Compared with traditional correlation for ordinary channel, the discrepancy is significant. Considering the size effects, a new correlation was proposed to estimate the heat transfer coefficient. The error of the regressed correlation is within ±12%.
Analysis and Strategy of Ageing and Ageing Management for Steam Generator in 300MWe NPP
Authors: Ning Dong , Zhang Kefeng , Yao Weida
Published: July 2009

The definition of ageing can be understood as a continuous and gradual process of degradation which is relative to environment, operating conditions, structure and material, etc. of components. It has characteristic of complexity that one or several factors in combination affect a component to degrade so as to result in losing function and structure failure. Much knowledge is involved in understanding and predicting degradation (ageing) mechanism as well as mitigating and managing degradation (ageing) phenomenon. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing errors) can jeopardize plant safety as well as plant life. At present, ageing in NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements.
Analysis of Aerosol Collection by Droplets: Application to Fission Products Removal in Case of Severe Accident
Authors: Porcheron Emmanuel , Lemaitre Pascal
Published: July 2009

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Suˆrete´ Nucle´aire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulate typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation. The present work is devoted to study the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between the spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.
Design of Wire-Wrapped Rod Bundle Matched Index-of-Refraction Experiments
Authors: McCreery Glenn E. , McIlroy Hugh M. , Hamman Kurt D. , Zhang Hongbin
Published: July 2009

Experiments will be conducted in the Idaho National Laboratory (INL) Matched Index-of-Refraction (MIR) Flow Facility [1] to characterize the three-dimensional velocity and turbulence fields in a wire-wrapped rod bundle typically employed in liquid-metal cooled fast reactors and to provide benchmark data for computer code validation. Sodium cooled fast reactors are under consideration for use in the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program. The experiment model will be constructed of quartz components and the working fluid will be mineral oil. Accurate temperature control (to within ± 0.05 °C) matches the index-of-refraction of mineral oil with that of quartz and renders the model transparent to the wavelength of laser light employed for optical measurements. The model will be a scaled 7-pin rod bundle enclosed in a hexagonal canister. Flow field measurements will be obtained with a LaVision 3-D particle image velocimeter (PIV) and complimented by near-wall velocity measurements obtained from a 2-D laser Doppler velocimeter (LDV). These measurements will be used as benchmark data for computational fluid dynamics (CFD) validation. The rod bundle model dimensions will be scaled up from the typical dimensions of a fast reactor fuel assembly to provide the maximum Reynolds number achievable in the MIR flow loop. A range of flows from laminar to fully-turbulent will be available with a maximum Reynolds number, based on bundle hydraulic diameter, of approximately 22,000. The fuel pins will be simulated by 85 mm diameter quartz tubes (closed on the inlet ends) and the wire-wrap will be simulated by 25 mm diameter quartz rods. The canister walls will be constructed from quartz plates. The model will be approximately 2.13 m in length. Bundle pressure losses will also be measured and the data recorded for code comparisons. The experiment design and preliminary CFD calculations, which will be used to provide qualitative hydrodynamic information, are presented in this paper.
Validation of the Coupled System Code ATHLET/BIPR-VVER on Local Core Measured Data
Authors: Nikonov S. , Lizorkin M. , Langenbuch S. , Velkov K.
Published: July 2009

The modern coupled thermal-hydraulics neutron–kinetics code systems require validation by comparison with measured local core data. Measured assembly coolant temperatures at core outlet are applied for this purpose in this study. Experimental data have been collected during the commissioning phase of the NPP Kalinin-3. One of the experimental data sets is selected to be used for international benchmark purposes (1). The new modelling capabilities of the ATHLET/BIPR-VVER code as well as the validation on plant measurements are discussed in the paper. The detailed modelling features of the reactor core allow to predict the coolant temperature at the assembly outlet within a rather high accuracy even though ATHLET code is based on one dimensional thermal-hydraulic pipe models.
Transient Performance of Medium Small LWR “DMS-400” Evaluated Using TRACG Code
Authors: Ikegawa Tomohiko , Ishii Yoshihiko , Matsuura Masayoshi , Hoshi Takashi
Published: July 2009

A new concept of a small and medium sized light water reactor, named the DMS (double MS: modular simplified & medium small reactor) has been developed. The DMS features significantly simplified plant systems realized by adoption of a natural circulation system of coolant and a free surface separation (FSS) system that is based on the gravitational separation of steam and water. With these systems, reactor internal pumps and steam separators are not needed, reducing plant cost. In this study, the DMS plant performance under transient conditions has been evaluated using TRACG code. TRACG code, which can treat multi-dimensional hydrodynamic calculations in a reactor pressure vessel (RPV), is well suited for evaluating DMS reactor transient performance because it can evaluate the void fraction in the chimney and therefore evaluate the natural circulation flow. As critical transient cases, generator load rejection with total turbine bypass failure (LRNBP) and loss of feedwater heating (LFWH) were chosen to evaluate. LRNBP and LFWH are the most severely recognized events as a pressure increase event and a thermal power increase event, respectively. In case of LRNBP, heat flux increased to about 110% of rated power, and the natural circulation flow barely changed, resulting in a lower ΔMCPR than that of LFWH case. The reason that heat flux only increased to 110% was because the RPV of the DMS has a large steam region volume in the chimney region compared to the thermal power. As a result, the change in the void fraction with a pressure increase in the core was small. In case of LFWH, the maximum heat flux, calculated using the neutron flux, was 121% of rated power when a scram occurred, and ΔMCPR was 0.14, almost the same as for current ABWRs. Since the analysis conditions were set conservatively, these results show that the DMS performs as well for transient events as conventional BWRs.
IAEA CRP Benchmark of Kalinin VVER-1000 NPP: An Analysis Using EXCEL-TRIHEX-FA Code System
Authors: Khan Suhail Ahmad , Jagannathan V. , Jain R. P.
Published: July 2009

Two units of VVER-1000Mwe reactors are in an advanced stage of construction at Kudankulam, Tamilnadu, India. With a view to assess the capability of analyzing the physics characteristics of VVER cores, the IAEA CRP benchmark problem of Kalinin VVER-1000 MWe NPP [1] is studied using the indigenous code system EXCEL-TRIHEX-FA [2,3]. The lattice burnup code EXCEL is based on a combination of 1-D multigroup transport theory and 2-D few group diffusion theory. Nuclear data in 172 group WIMS-D format based on JEFF-3.1 [4] has been used in the present analysis. The core level calculations are performed using the code TRIHEX-FA which solves the 3-D multigroup diffusion equation using the finite difference method with fine triangular meshes. Power dependent feedback models for xenon, Doppler, coolant temperature and density values have been incorporated in TRIHEX-FA. Keff for the critical soluble boron concentration, assembly power distribution and axial power distribution are calculated as a function of fuel cycle burnup. In the present paper, lattice level results are compared with the results of other participants reported in Ref. [1]. The results of core level calculations have been compared with the experimental data provided [1].
Mixing Conditions in the Lower Plenum and Core Inlet of a BWR
Authors: Tinoco Hernan , Ahlinder Stefan
Published: July 2009

A thermal mixing analysis of the Downcomer, Main Recirculation Pumps (MRPs) and Lower plenum of Forsmark’s Unit 3 has been carried out with three separate CFD models. Several difficulties with the boundary conditions have been encountered, particularly with the MRP model. The results obtained predict stable temperature differences of around 8 K at the core inlet. Such large temperature differences have never been observed at Forsmark NPP. Temperature measurements at four positions above the Reactor Pressure Vessel (RPV) bottom give the mean value used as the core inlet temperature for core analyses with codes such as POLCA. The temperature transmitters used are rather slow and inaccurate. Still, they should be able to detect large stable temperature differences such as those predicted by the aforementioned computations. Indirect indication of the incongruity of these predictions is the possibility of fuel damage caused by such large temperature differences. Fuel damage other than the one caused by debris fretting (thread-like particles) through mechanical influence has not been reported at Forsmark NPP since the implementation of liner cladding in fuel design. Also, the aforementioned difficulties with the connection of the models throw some doubt upon the accuracy of these predictions. A completely connected model of the same RPV volume covered by the separate models predicts temperature differences at core inlet that are almost a fourth of those mentioned above, i. e. approximately 2.5 K. Most of the mixing occur downstream of the MRP diffusers, at the Lower Plenum “inlet”. The reason for this prediction divergence is an impossibility of a correct transfer of complete three-dimensional flow field properties by means of boundary conditions defined at a two-dimensional inlet section.
Hydraulic Autofrettage Technology: A Review
Authors: Malik M. Afzaal , Khushnood Shahab , Khan Muddasar , Rashid Badar , Khan Musarrat
Published: July 2009

Thick-walled cylinders play a unique role in military, automobiles, aircraft and oil fields. The autofrettage method is used to increase the fatigue life of thick-walled components like gun barrel, titanium alloy tubes, nuclear reactor vessels, high pressure oil field pipes and LDPE. There are three types of autofrettage processes and hydraulic is the one applied mostly. The autofrettage process uses a pressure that is high enough to plastically deform the bore of the part but not so high that it bursts the component, is applied to the inside of tube. The result is that after the pressure is removed, the elastic recovery of the outer wall put the inner wall into compression, providing a residual compressive stress. In this research paper a study has been carried out to investigate the progress made in different fields by the hydraulic technique and proposals are sought for future work.
Effect of Bolt Spacing on the Circumferential Distribution of Gasket Contact Stress in Bolted Flange Jonts
Authors: Do Tan Dan , Bouzid Abdel-Hakim , Dao Thien-My
Published: July 2009

Bolted flange joints are part of pressure vessel and piping components and are used extensively in the chemical, petrochemical and nuclear power industries. They are simple structures and offer the possibility of disassembly which make them attractive to connect pressurized equipments and piping. In addition of being prone to leakage, they often require maintenance while in operation in which case the bolts are either retightened as in hot torquing or untightened to be replaced. Although costly shutdown are avoided, such a maintenance work exposes the operator to a potential risk because the bolt load alteration can produce a gasket load unbalance which results in a local gasket contact stress to drop below some critical value causing major leak and hence jeopardizing the life of the operator. This paper addresses the issue of the contact stress level unbalance around the flange when the bolts are subjected to bolt-up condition. This study is developed for the purpose of helping limit the degree of load increase in hot torquing or the maximum number of bolts to be replaced at a time and identify those flanges the bolt of which cannot be replaced in service.
Sensitivity of Response Spectra to ARMA Model Process
Authors: Brahimi Malek , Berri Sidi
Published: July 2009

The sensitivity analysis in this study focuses on the sensitivity of response spectra to the number of peaks in the envelope function and the order of the Autoregressive Moving Average of order p, q (ARMA( p,q)) process used to represent an earthquake time series. Results consist of two sets of response spectra: one set corresponds to the one peak for envelope function and ARMA(2,1), the second set involves a change in the number of peaks assumed and the order p, q of fitted ARMA(2,1) models. For comparison purposes, response data were normalized by dividing all spectral ordinates by the corresponding ordinates for the one peak, ARMA(2,1) models for the same earthquake. It was conclude that, models of order (2,1) are sufficient to estimate the damage effect of an earthquake with relatively few parameters.
Multiple Models Predictive Control of Once-Through Steam Generator
Authors: Tai Yun , Hou Su-Xia , Zhao Fu-Yu
Published: July 2009

Because of multiphase flow during the heat transfer, OTSG (Once-Through Steam Generator) is a complex nonlinear MIMO (Multiple Input and Multiple Output) system. This article sets the mathematics model of OTSG with internal screw and double tubes, uses the FOROTSG program to simulate OTSG, and identifies the system models of different power levels. Based on this, Multiple Model Predictive Control (MMPC) strategy is proposed, which designs the Model Predictive Controller of each model identified, and then integrates the multiple models by Membership Function Law, to achieve smoothly switch of the multiple models at last. The simulating result indicates the MMPC has the good control effects, and it is an available strategy to solve the problems of the nonlinear system control.
The Importance of Establishing an Operational Approach for the Selection of Materials in the Reactors of the Future
Authors: Horowitz Emmanuel
Published: July 2009

Nuclear scientists and engineers should consider adopting a more operational approach for the purpose of selecting their future materials. For each type of nuclear power generating reactor, for each coolant (water, helium or liquid metal), the next generation of specialists and decision-makers will need to choose and optimise the iron or nickel alloys, steels, ODS (oxide dispersed strengthened steels) and ceramics that are going to be used. It may well be considered that either each reactor type has its own, specific materials, or, in a complementary manner, that the efforts for improvements should be shared. At high temperatures, as found on fuel-cladding liners, heat exchangers or even tubes or tube liners, different types of steels and alloys may be envisaged. It is considered that austenitic steels provide a better creep resistance at high temperature but they must be stabilized by nickel, thereby becoming more expensive. Ferrite steels could be better as far as swelling, mechanical strength and thermal behaviour are concerned. To withstand corrosion, chromium or aluminium, ODS steels could turn out to be good solutions, if they can comply with stringent criteria. Concerning heat exchangers, choices must be made between iron and nickel alloys, according to proposed operating conditions. In the case of sodium-cooled rapid neutron reactors (RNRs), ferritic-martensitic alloys with 9%–12% chromium or chromium ODS steels could prove suitable, especially if we judge by their specific mechanical behaviour, up to at least 700°C. Nevertheless, behaviour of these steels — with respect to ageing, anisotropy, radiation induced segregation, radiation induced precipitation, reduction of activation products and welding — needs be better understood and qualified. Sodium heat exchanger materials should be carefully chosen since they have to withstand corrosion arising from the primary flow and also from the secondary or tertiary flow (either sodium or molten salts, gas or water); therefore, experimental loops are necessary to gain improved understanding and assessment of the designs envisioned. One way to improve alloys is through thermal, mechanical treatments or by surface treatments. A better way could, however, be to improve the nanostructure and mesostructure of the materials chosen at the drawing-board stage, for instance by nano-size cluster dispersion and grain size controls; experimental tests, microscope and spectroscope observations, multi-scale modelling and thermodynamics computing could also help calibrate and implement these improvements. Large, experimental databases and codes will be the keystone to defining more operational knowledge bases that will then allow us to determine terms of reference for the new materials. Failing this, time will be running out — within the next twenty years — to design and develop nuclear prototypes consistent with the criteria laid down for “Generation IV” reactors.
Dynamic Sensitivity Analysis of Thermal-Mechanical Deformation of a CANDU Fuel Channel
Authors: Luxat J. C.
Published: July 2009

In a limiting critical break loss of coolant accident in a CANDU reactor significant degradation of heat transfer from the fuel can occur. As a result of the subsequent increase in fuel temperature it is possible that the pressure tube undergoes heat up at intermediate pressure during blowdown. This can result in ballooning deformation of the pressure tube into contact with its calandria tube. It is required that fuel channels not fail as a consequence of the thermal mechanical deformation of the pressure tube and calandria tube in such events. Dynamic sensitivity functions are derived as analytical partial differential equations derived from the equations used to model the time-dependent behavior of physical systems. The dynamic sensitivity functions can be used to propagate uncertainties using a time-dependent perturbation approach in which the variations in a set of output variables, with respect to perturbations of the input parameters, are evaluated about reference response trajectories of the input parameters and associated output variables. The dynamic sensitivity method is described in this paper and results are presented for the pressure tube heatup phase of a LOCA. These results show the importance of all key parameters with respect to specified safety evaluation criteria. The dynamic sensitivity method is applied in a probabilistic uncertainty analysis to evaluate the probability of a pressure tube experiencing creep strain deformation to contact its calandria tube during the early stages of a LOCA.
Effect of Experimental Conditions on Gas Core Length and Downward Velocity of Free Surface Vortex in Cylindrical Vessel
Authors: Monji Hideaki , Shinozaki Tatsuya , Kamide Hideki , Sakai Takaaki
Published: July 2009

This paper deals with characteristics of surface vortex in a cylindrical vessel. One of the characteristics is a gas core length which is important to estimate the onset condition of the gas entrainment but influenced easily by the experimental condition. In the experiment using water, the effects of the water temperature, water level and the surface tension on the gas core length were investigated. The onset condition of the gas entrainment is sometimes estimated by using the Burgers vortex model but the real flow in the vessel is different from the model. The velocity fields were measured by PIV and the velocity gradient of the downward flow was discussed. The proper flow conditions for the Burgers vortex model are a high water level and a high flow rate.
Research on the Radioactivity of Thorium Fuel in Different Reactors
Authors: Yu Ganglin , Wang Kan
Published: July 2009

It’s very important to estimate the mass and radiotoxicity of isotopes in spent fuel of thorium fuel cycle, which will benefit the application of the thorium fuel. Much research work has been done on the spent fuel and radioactivity of thorium-based fuel before, yet the difference in usage is always ignored. This paper studies the raise of isotopes in spent fuel of thorium-based fuel cycle in pressurized water reactors and fast neutron reactors, focus on the radioactivity level of actinides and fission products, the important nuclides which have long term radiological impact or give the highest contribution to the total dose on short term after different decay periods. The paper discuss the important nuclides in the measurement of thorium fuel burnup, the conclusion will benefit the actual application of thorium fuel.
A Comparison of Fast Thorium Breeder Reactor Designs With Oxide and Metallic Fuels
Authors: Jagannathan V. , Pal Usha , Karthikeyan R. , Raj Devesh
Published: July 2009

Loading of seedless thoria rods in internal blanket regions and using them later as part of seeded fuel assemblies is the central theme of the thorium breeder reactor (ATBR) concept [1]. The fast reactors presently consider seedless blanket region surrounding the seeded core region. This results in slower fissile production rate in comparison to fissile depletion rate per unit volume. The overall breeding is achieved mainly by employing blanket core with more than double the volume of seeded core. The blanket fuel is discharged with fissile content of ∼30g/kg, which is much less than the asymptotic maximum possible fissile content of 100g/kg. This is due to smaller coolant flow provided for in the blanket regions. In a newly proposed fast thorium breeder reactor (FTBR) [2], the blanket region is brought in and distributed through out the core. By this the fissile depletion and production rates per unit volume become comparable. The core considered simultaneous breeding from both fertile thoria and depleted uranium and hence the concept can be called as fast twin breeder reactor as well. Sodium is used as coolant. The blanket fuel rods achieve nearly 80% of the seed fuel rod burnup and also contain nearly the maximum possible fissile content at the time of discharge. In this paper a comparison of FTBR core characteristics with oxide and metallic fuel are compared.
Determination of Oxidation and Metallic Cations Release Kinetics on Nickel Base Alloys in PWR: Description of the On-Line Measurement Techniques
Authors: Proust Antoine , Guillodo Michael , Pijolat Miche`le , Wolski Krzysztof
Published: July 2009

A common issue in Nuclear Power Plants (NPP) is the release of corrosion products from the steam generator tubes in primary circuit which gives rise, after their neutronic activation, to a residual activity very problematic during chemistry transients. The objectives of this work are not to solve directly this problem but rather to improve the understanding of this phenomenon by determining experimentally the kinetics of oxidation and metallic cations release of nickel base alloys. These results will contribute to progress in the building of a physical mechanism. The specificity of the test facility is (i) the on-line measurements of cation release rate by a highly sensitive analytical device and (ii) the in-situ measurement of oxidation rate by Diffuse Reflectance Spectroscopy (DRS). The primary water chemistry is controlled by means of a circulating micro-loop in titanium to avoid any interference with the test samples. A specific test vena is connected between the titanium micro-loop and the analytical device for on-line measurements. Metallic cation release from SG tube-samples is deduced from the measurement of nickel and iron concentrations in upstream the test vena by use of differential polarography and ICP-AES techniques. The precision on the determination of Ni and Fe concentrations is assumed to be a few tens of ppt. Oxidation or corrosion measurements are performed off-line by optical diffuse reflectance measurements specifically developed in this framework. The corrosion sample is mounted in a High Temperature and Pressure (HT-HP) spectroscopic flow-through cell connected to the test vena and fed with the primary chemistry from the titanium micro-loop. Analysis of reflection spectra provides the thickness of oxide layer. A set of parameters such as (i) temperature, (ii) hydrogen content and (iii) metallurgical surface state in nominal primary chemistry for test durations of 1500–2000 hours is planned to be investigated in the frame of metallic cation release and oxidation kinetics measurements of Ni base alloys. The aim of this paper is focused on the presentation of the online technique specifically developed in the French Technical Centre of AREVA NP to determine the metallic cation release and oxidation kinetic of Ni base alloys in primary water chemistry. First results of cation release and oxidation rate are given for shorter duration test.
Experimental Study on Natural Circulation and Air-Injection Enhanced Circulation With Different Fluids
Authors: Ambrosini W. , Forgione N. , Oriolo F. , Semeraro E. , Tarantino M.
Published: July 2009

The paper reports about an experimental investigation on natural circulation and air-injection enhanced circulation performed adopting different fluids. The work is aimed at providing information on the basic mechanisms proposed in the design of future reactors relying on such circulation mechanisms for core cooling. Though the final objective of the research is the study of heavy metal cooling, the work is here limited to non-metallic fluids. The working fluid adopted in past analyses was water. Further experimental campaigns were recently performed using the Novec™ HFE-7100 fluid, providing additional information on basic phenomena and the related scaling laws. The new fluid has a greater density and a greater thermal expansion coefficient with respect to water. Air was adopted for gas injection. Both natural circulation and gas-injection enhanced circulation are addressed in this work, drawing quantitative conclusions about the observed parametric trends. A systematic comparison is performed with the results obtained in previous experimental activities using water.
Current Issues in Operated WWER Type RPVs
Authors: Brumovsky´ Milan , Kytka Milos
Published: July 2009

RPVs of WWER type reactors are manufactured from other type of steels (15Kh2MFA of Cr-Mo-V type for WWER-440 and 15Kh2NMFA of Ni-Cr-Mo-V type for WWER-1000) and according to other Codes and standards than PWR ones, thus some specific problems are currently more important for WWER. The principal problem lies in relatively small number of manufactured and operated WWER type NPPs. Even though a high level of unification in RPVs exists — practically only two designs of RPVs exists (WWER-440 and WWER-1000) — total number is still small. All WWER-440 RPV are practically identical, either they were manufactured for V-230 or V-213 model: the only difference is in the purity of used materials and existence/non-existence of the surveillance programmes. (Fact that some V-230 type vessels were not covered by austenitic cladding is not important from irradiation effects point of view.) Regarding WWER-440/V-230 types, it is necessary to take into account, that even though most of them were successfully annealed, only some of them are still in operation but most of them will be closed in near future. Similar situation is with WWER-1000 RPVs, either they were manufactured for V-320 (most frequent), or V-338 or the newest V-428 — differences are practically only in the content of nickel in critical weldments and/or in design of surveillance specimens capsules. But, Large advantage of all WWER surveillance programmes is in loading static fracture toughness specimens in all programmes. The papers tries to summarize and analyze all current issues connected with radiation embrittlement of operated reactor pressure vessels of WWER type.
A Computational Model of the Mark-IV Electrorefiner: Phase I — Fuel Basket/Salt Interface
Authors: Hoover Robert , Phongikaroon Supathorn
Published: July 2009

Spent driver fuel from the Experimental Breeder Reactor-II (EBR-II) is currently being treated in the Mk-IV electrorefiner (ER) in the Fuel Conditioning Facility at Idaho National Laboratory. The new modeling approach is developed to help understand the dynamics of different parameters in the system. The first phase of this new modeling approach is focused on the fuel basket/salt interface and involves the transport of various species in the driver fuels (e.g. uranium and zirconium). This new approach minimizes the guessed parameters from three down to one, the exchange current density (i0). This new method should allow a rapid, accurate understanding of the ER process permitting predictions of different system designs. U3+ and Zr4+ were the only species used for the current study. The result reveals that potential for each species is the fixed variable, it is a function of weight percent (wt%) of uranium in molten salts and is independent of i0. The dimensionless approach shows that the total potential is a strong function of i0 and a weak function of wt% of uranium in the salt system.
Bubbles and Droplets Measurement via Optical Fiber Probe Processed by Femtosecond Pulse Laser
Authors: Saito Takayuki , Ozawa Yusuke , Matsuda Keisuke , Aoshima Shin-Ichiro
Published: July 2009

An optical fiber has interesting and useful characteristics, which are able to be applied to scientific measurement. Its phase detection characteristic based on refraction difference between two phases is applied to bubbles/droplets measurement. Recently, demands for measurement of tiny bubbles/droplets increase in research fields of steam injectors, sprays, automotive engines, fine chemical reactors, and so on. Meanwhile, laser science and engineering, particularly femtosecond pulse laser, has made remarkable advances lately. Their unique properties of the interaction between the femtosecond pulses and materials can be utilized for microfabrication. The optical fiber probe methods have been repeatedly improved in order to measure bubbles/droplets efficiently and reliably in gas-liquid two-phase flows. However, one has been taking it for granted that simultaneous measurement of their diameters and velocities needs at least two optical fiber probes. To break through this situation, we newly developed a Single-Tip Optical Fiber Probe (F-STOP) microfabricated by femtosecond pulse laser (fs-pulses) which realizes simultaneous measurement of diameters and velocities of tiny bubbles/droplets. In the F-STOP measurement, the following properties are used to realize the simultaneous measurement of diameters and velocities of bubbles/droplets: the relation between the reflected-light intensity at the wedge-shaped probe tip and the tip-surface area covered with a phase; reflected-light intensity at the groove microfabricated by fs-pulses. The first aim of the present study is to provide data for evaluation of the influences of surface tension and wettability on the bubble measurement in order to develop precise and reliable F-STOP method. The second aim is to describe the process to make F-STOP via fs-pulses. The third aim is demonstration of the newly developed probe in real measurement of bubbles/droplets. On the basis of these, the performance of the new probe is discussed.
Uncertainty Analysis Tool for the Performance Assessment of the High Level Radioactive Waste Disposal System
Authors: Matsumoto Mitsuyo , Neyama Atsushi
Published: July 2009

The uncertainty analysis is one approach to indicate long-term safety of a HLW disposal system. Usually it is very difficult to construct the computer code for uncertainty analysis by using individual process existing computer codes without revising source program in a short time. Because detailed modification for an uncertainty analytical function and a post-processing procedure is needed in each computer code. We have developed the uncertainty analysis tool, which can apply to any computer codes. This uncertainty analysis tool is very effective to calculate the efficiency and quality for linkage of existing individual process computer codes.