Sample records for reactors transients identification

  1. Nuclear reactors transients identification and classification system; Sistema de identificacao e classificacao de transientes em reatores nucleares

    Bianchi, Paulo Henrique


    This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)

  2. Transients in reactors for power systems compensation

    Abdul Hamid, Haziah

    This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the

  3. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri


    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  4. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    Jensen, C.; Wachs, D.; Carmack, J.; Woolstenhulme, N.


    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

  5. Adaptive Nodal Transport Methods for Reactor Transient Analysis

    Thomas Downar; E. Lewis


    Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.

  6. Transient two-phase performance of LOFT reactor coolant pumps

    Chen, T.H.; Modro, S.M.


    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed.

  7. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    Laureau, A., E-mail:; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.


    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  8. FRINK - A Code to Evaluate Space Reactor Transients

    Poston, David I.; Dixon, David D.; Marcille, Thomas F.; Amiri, Benjamin W.


    One of the biggest needs for space reactor design and development is detailed system modeling. Most proposed space fission systems are very different from previously operated fission power systems, and extensive testing and modeling will be required to demonstrate integrated system performance. There are also some aspects of space reactors that make them unique from most terrestrial application, and require different modeling approaches. The Fission Reactor Integrated Nuclear Kinetics (FRINK) code was developed to evaluate simplified space reactor transients (note: the term ``space reactor'' inherently includes planetary and lunar surface reactors). FRINK is an integrated point kinetic/thermal-hydraulic transient analysis FORTRAN code - ``integrated'' refers to the simultaneous solution of the thermal and neutronic equations. In its current state FRINK is a very simple system model, perhaps better referred to as a reactor model. The ``system'' only extends to the primary loop power removal boundary condition; however this allows the simulation of simplified transients (e.g. loss of primary heat sink, loss of flow, large reactivity insertion, etc.), which are most important in bounding early system conceptual design. FRINK could then be added to a complete system model later in the design and development process as system design matures.

  9. Need for space-time analyses of research reactor transients

    Jatuff, F.E. [INVAP S.E., de Bariloche (Argentina)


    The success of the point-kinetics approximation to represent the time behavior of research reactors relies on the fact that research reactor cores are small enough to be neutronically tightly coupled; the core is small when measured in diffusion lengths. This fact implies that a certain change in a part of the core is immediately observed by the whole system. The propagation of changes is so fast that the core exhibits a shape function that is practically unchanged during the transient; the amplitude function, the only unknown of the problem, represents the full knowledge of the core response. One is immediately warned to look for the truth of this assumption. How small should a research reactor core be to be sure that point kinetics is a valid assumption? This question is becoming increasingly important because the tendency is to increase the size of research reactor cores to make them capable of various simultaneous uses (multipurpose characteristics), with powers in the range of tens of thermal megawatts. One of the lines of investigation at the Department of Reactor Physics is related to scenarios of Materials Test Reactor (MTR)-type research reactor transients for which space-time kinetics would bring a more profound insight than point kinetics.

  10. Simulation of Reactor Transient and Design Criteria of Sodium-cooled Fast Reactors

    Gottfridsson, Filip


    The need for energy is growing in the world and the market of nuclear power is now once more expanding. Some issues of the current light-water reactors can be solved by the next generation of nuclear power, Generation IV, where sodium-cooled reactors are one of the candidates. Phénix was a French prototype sodium-cooled reactor, which is seen as a success. Although it did encounter an earlier unexperienced phenomenon, A.U.R.N., in which a negative reactivity transient followed by an oscillati...

  11. Transient behavior of a nuclear reactor coupled to an accelerator

    Sadineni, Suresh Babu

    Accelerator Driven Systems (ADS) present one of the most viable solutions for transmutation and effective utilization of nuclear fuel. Spent fuel from reactors will be partitioned to separate plutonium and other minor actinides to be transmuted in the ADS. Without the ADS, minor actinides must be stored at a geologic repository for long periods of time. One problem with ADS is understanding the control issues that arise when coupling an accelerator to a reactor. "ADSTRANS" was developed to predict the transient behavior of a nuclear reactor coupled to an accelerator. It was based on MCNPX, a radiation transport code developed at the LANL, and upon a numerical model of the neutron transport equation. MCNPX was used to generate the neutron "source" term that occurs when the accelerator is fired. ADSTRANS coupled MCNPX to a separate finite difference code that solved the transient neutron transport equation. A cylindrical axisymmetric reactor with steel shielding was considered for this analysis. Multiple neutron energy groups, neutron precursor groups and neutron poisons were considered. ENDF/B cross-section data obtained through MCNPX was also employed. The reactor was assumed to be isothermal and near zero power level. Unique features of this code are: (1) it predicts the neutron behavior of an ADS for different reactor geometry, material concentration, both electron and proton particle accelerators, and target material, (2) it develops input files for MCNPX to simulate neutron production, runs MCNPX, and retrieves information from the MCNPX output files. Neutron production predicted by MCNPX for a 20 MeV electron accelerator and lead target was compared with experimental data from the Idaho Accelerator Center and found to be in good agreement. The spatial neutron flux distribution and transient neutron flux in the reactor as predicted by the code were compared with analytical solutions and found to be in good agreement. Fuel burnup and poison buildup were also as

  12. The Dynamic Monte Carlo Method for Transient Analysis of Nuclear Reactors

    Sjenitzer, B.L.


    In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing co

  13. On closed loop transient response system identification

    Christer Dalen


    Full Text Available Some methods for transient closed loop step response system identification presented in the literature are reviewed. Interestingly some errors in a method published in the early 80's where propagated into a recently published method. These methods are reviewed and some improved methods are suggested and presented. The methods are compared against each other on some closed loop system examples, e.g. a well pipeline-riser severe-slugging flow regime example, using Monte Carlo simulations for comparison of the methods.

  14. Qualitative diagnosis for transients analysis on nuclear reactors

    Lorre, J.P.; Dorlet, E.; Evrard, J.M.


    One of the major aims of an intelligent monitoring system, is the supervision task which assist the operator in understanding what occurs on a process. Failures hypotheses must be located and the inferring process must be explained. This paper demonstrate a second generation expert system (SEXTANT) decided to the transients analysis on PWR nuclear reactors. This system detects failures by simulating the process with a numerical model. A diagnosis module uses an even graph built from a causal graph model of the plant to generate hypotheses, and a numerical model to validate these hypotheses. Hypotheses are stored into scenarios which are concurrent possible interpretations of the process evolution. The approach is illustrated by an application for the analysis of the house load operation on a pressurized water reactor. (authors). 9 refs., 10 figs.

  15. A Nuclear Reactor Transient Methodology Based on Discrete Ordinates Method

    Shun Zhang


    Full Text Available With the rapid development of nuclear power industry, simulating and analyzing the reactor transient are of great significance for the nuclear safety. The traditional diffusion theory is not suitable for small volume or strong absorption problem. In this paper, we have studied the application of discrete ordinates method in the numerical solution of space-time kinetics equation. The fully implicit time integration was applied and the precursor equations were solved by analytical method. In order to improve efficiency of the transport theory, we also adopted some advanced acceleration methods. Numerical results of the TWIGL benchmark problem presented demonstrate the accuracy and efficiency of this methodology.

  16. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Bradley K. Heath


    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  17. Classifier-ensemble incremental-learning procedure for nuclear transient identification at different operational conditions

    Baraldi, Piero, E-mail: piero.baraldi@polimi.i [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milano (Italy); Razavi-Far, Roozbeh [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milano (Italy); Zio, Enrico [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milano (Italy); Ecole Centrale Paris-Supelec, Paris (France)


    An important requirement for the practical implementation of empirical diagnostic systems is the capability of classifying transients in all plant operational conditions. The present paper proposes an approach based on an ensemble of classifiers for incrementally learning transients under different operational conditions. New classifiers are added to the ensemble where transients occurring in new operational conditions are not satisfactorily classified. The construction of the ensemble is made by bagging; the base classifier is a supervised Fuzzy C Means (FCM) classifier whose outcomes are combined by majority voting. The incremental learning procedure is applied to the identification of simulated transients in the feedwater system of a Boiling Water Reactor (BWR) under different reactor power levels.

  18. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)


    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  19. Nonlinear damping identification from transient data

    Smith, Clifford B.; Wereley, Norman M.


    To study new damping augmentation methods for helicopter rotor systems, accurate and reliable nonlinear damping identification techniques are needed. For example, current studies on applications of magnetorheological (MR) dampers for rotor stability augmentation suggest that a strong Coulomb damping characteristic will be manifested as the field applied to the MR fluid is maximized. Therefore, in this work, a single degree of freedom (SDOF) system having either nonlinear Coulomb or quadratic damping is considered. This paper evaluates three analyses for identifying damping from transient test data; an FFT-based moving block analysis, an analysis based on a periodic Fourier series decomposition, and a Hilbert transform based technique. Analytical studies are used to determine the effects of block length, noise, and error in identified modal frequency on the accuracy of the identified damping level. The FFT-based moving block has unacceptable performance for systems with nonlinear damping. These problems were remedied in the Fourier series based analysis and acceptable performance is obtained for nonlinear damping identification from both this technique and the Hilbert transform based method. To more closely simulate a helicopter rotor system test, these techniques were then applied to a signal composed of two closely spaced modes. This data was developed to simulate a response containing the first lag and 1/rev modes. The primary mode of interest (simulated lag mode) had either Coulomb or quadratic damping, and the close mode (1/rev) was either undamped or had a specified viscous damping level. A comprehensive evaluation of the effects of close mode amplitude, frequency, and damping level was performed. A classifier was also developed to identify the dominant damping mechanism in a signal of 'unknown' composition. This classifier is based on the LMS error of a fit of the analytical envelope expression to the experimentally identified envelope signal. In most

  20. Steam separator modeling for various nuclear reactor transients

    Paik, C Y; Mullen, G; Knoess, C; Griffith, P


    In a pressurized water reactor steam generator, a moisture separator is used to separate steam and liquid and to insure that essentially dry steam is supplied to the turbine. During a steam line break or combined steam line break plus tube rupture, a number of phenomena can occur in the separator which have no counterparts during steady-state operation. How the separator will perform under these circumstances is important for two reasons, it affects the carry-over of radioactive iodine and the water inventory in the secondary side. This study has as its goal the development of a simple separator model which can be applied to a variety of steam generator for off-design conditions. Experiments were performed using air and water on three different types of centrifugal separators: a cyclone as a generic separator, a Combustion Engineering type stationary swirl vane separator, and a Westinghouse type separator. The cyclone separator system has three stages of separation: first the cyclone, then a gravity separator, and finally a chevron plate separator. The other systems have only a centrifugal separator to isolate the effect of the primary separator. Experiments were also done in MIT blowdown rig, with and without a separator, using steam and water. The separators appear to perform well at flow rates well above the design values as long as the downcomer water level is not high. High downcomer water level rather than high flow rates appear to be the primary cause of degraded performance. Appreciable carry-over from the separator section of a steam generator occurs when the drain lines from three stages of separation are unable to carry off the liquid flow. Failure scenarios of the separator for extreme range of conditions from the quasi-steady state transient to the fast transients are presented. A general model structure and simple separator models are provided.

  1. Bagged ensemble of Fuzzy C-Means classifiers for nuclear transient identification

    Baraldi, Piero; Razavi-Far, Roozbeh [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, Via Ponzio 34/3, 20133 Milano (Italy); Zio, Enrico, E-mail: [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, Via Ponzio 34/3, 20133 Milano (Italy); Ecole Centrale Paris-Supelec, Paris (France)


    Research highlights: > A bagged ensemble of classifiers is applied for nuclear transient identification. > Fuzzy C-Means classifiers are used as base classifiers of the ensemble. > Transients are simulated in the feedwater system of a boiling water reactor. > Ensemble is compared with a supervised, evolutionary-optimized FCM classifier. > Ensemble improves classification accuracy in cases of large or very small sizes data. - Abstract: This paper presents an ensemble-based scheme for nuclear transient identification. The approach adopted to construct the ensemble of classifiers is bagging; the novelty consists in using supervised fuzzy C-means (FCM) classifiers as base classifiers of the ensemble. The performance of the proposed classification scheme has been verified by comparison with a single supervised, evolutionary-optimized FCM classifier with respect of the task of classifying artificial datasets. The results obtained indicate that in the cases of datasets of large or very small sizes and/or complex decision boundaries, the bagging ensembles can improve classification accuracy. Then, the approach has been applied to the identification of simulated transients in the feedwater system of a boiling water reactor (BWR).

  2. Identification of Transient and Permanent Faults

    李幼仪; 董新洲; 孙元章


    A new algorithm was developed for arcing fault detection based on high-frequency current transients analyzed with wavelet transforms to avoid automatic reclosing on permanent faults. The characteristics of arc currents during transient faults were investigated. The current curves of transient and permanent faults are quite similar since current variation from the fault arc is much less than the voltage variation. However, the fault current details are quite different because of the arc extinguishing and reigniting. Dyadic wavelet transforms were used to identify the current variation since wavelet transform has time-frequency localization ability. Many electric magnetic transient program (EMTP) simulations have verified the feasibility of the algorithm.

  3. A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers

    Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)

  4. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  5. Design criteria of integrated reactors based on transients; Criterios de diseno de reactores integrados basados en transitorios

    Zanocco, P.; Gimenez, M.; Delmastro, D. [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche


    A new tendency in integrated reactors conceptual design is to include safety criteria through accident analysis. In this work, the effect of design parameters in a Loss of Heat Sink transient using design maps is analyzed. Particularly, geometry related parameters and reactivity coefficients are studied. Also the effect of primary relief/safety valve during the transient is evaluated. A design map for valve area vs. coolant density reactivity coefficient is obtained. A computer code (HUARPE) is developed in order to simulate these transients. Coolant, steam dome, pressure vessel structures and core models are implemented. This code is checked against TRAC with satisfactory results. (author) 5 refs., 13 figs.

  6. Simulation of Safety and Transient Analysis of a Pressurized Water Reactor using the Personal Computer Transient Analyzer

    Sunday J. IBRAHIM


    Full Text Available Safety and transient analyses of a pressurised water reactor (PWR using the Personal Computer Transient Analyzer (PCTRAN simulator was carried out. The analyses presented a synergistic integration of a numerical model; a full scope high fidelity simulation system which adopted point reactor neutron kinetics model and movable boundary two phase fluid models to simplify the calculation of the program, so it could achieve real-time simulation on a personal computer. Various scenarios of transients and accidents likely to occur at any nuclear power plant were simulated. The simulations investigated the change of signals and parameters vis a vis loss of coolant accident, scram, turbine trip, inadvertent control rod insertion and withdrawal, containment failure, fuel handling accident in auxiliary building and containment, moderator dilution as well as a combination of these parameters. Furthermore, statistical analyses of the PCTRAN results were carried out. PCTRAN results for the loss of coolant accident (LOCA caused a rapid drop in coolant pressure at the rate of 21.8KN/m2/sec triggering a shutdown of the reactor protection system (RPS, while the turbine trip accident showed a rapid drop in total plant power at the rate of 14.3 MWe/sec causing a downtime in the plant. Fuel handling accidents mimic results showed release of radioactive materials in unacceptable doses. This work shows the potential classes of nuclear accidents likely to occur during operation in proposed reactor sites. The simulations are very appropriate in the light of Nigeria’s plan to generate nuclear energy in the region of 1000 MWe from reactors by 2017.

  7. Transient modeling of the thermohydraulic behavior of high temperature heat pipes for space reactor applications

    Hall, Michael L.; Doster, Joseph M.


    Many proposed space reactor designs employ heat pipes as a means of conveying heat. Previous researchers have been concerned with steady state operation, but the transient operation is of interest in space reactor applications due to the necessity of remote startup and shutdown. A model is being developed to study the dynamic behavior of high temperature heat pipes during startup, shutdown and normal operation under space environments. Model development and preliminary results for a hypothetical design of the system are presented.

  8. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    Galvez, Cristhian


    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the pa...

  9. An earthquake transient method for pebble-bed reactors and a fuel temperature model for TRISO fueled reactors

    Ortensi, Javier

    This investigation is divided into two general topics: (1) a new method for analyzing the safe shutdown earthquake event in a pebble bed reactor core, and (2) the development of an explicit tristructural-isotropic fuel model for high temperature reactors. The safe shutdown earthquake event is one of the design basis accidents for the pebble bed reactor. The new method captures the dynamic geometric compaction of the pebble bed core. The neutronic and thermal-fluids grids are dynamically re-meshed to simulate the re-arrangement of the pebbles in the reactor during the earthquake. Results are shown for the PBMR-400 assuming it is subjected to the Idaho National Laboratory's design basis earthquake. The study concludes that the PBMR-400 can safely withstand the reactivity insertions induced by the slumping of the core and the resulting relative withdrawal of the control rods. This characteristic stems from the large negative Doppler feedback of the fuel. This Doppler feedback mechanism is a major contributor to the passive safety of gas-cooled, graphite-moderated, high-temperature reactors that use fuel based on TRISO particles. The correct prediction of the magnitude and time-dependence of this feedback effect is essential to the conduct of safety analyses for these reactors. An explicit TRISO fuel temperature model named THETRIS has been developed in this work and incorporated in the CYNOD-THERMIX-KONVEK suite of coupled codes. The new model yields similar results to those obtained with more complex methods, requiring multi-TRISO calculations within one control volume. The performance of the code during fast and moderately-slow transients is verified. These analyses show how explicit TRISO models improve the predictions of the fuel temperature, and consequently, of the power escalation. In addition, a brief study of the potential effects on the transient behavior of high-temperature reactors due to the presence of a gap inside the TRISO particles is included

  10. RTC-control of power transients in nuclear reactors

    Ratemi, Wajdi Mohamed [Alfateh University, PO Box 13040, Tripoli (Libyan Arab Jamahiriya)


    In this paper, the new Reactivity Trace Curve (RTC) method (Ratemi 1993,1994), which is based on the dynamic period studies (Bernard et al.,1984), has been studied for maneuvering of the nuclear reactor power without power shooting. The reactor is modeled with one group of delayed neutrons with temperature feedback effect, as well as, Xenon feedback effect. A precursors concentration model is used to provide for the effective dynamic decay constant (in one group case, it is a static one). The RTC-identifier which is given by a differential equation is then solved at each sampling time (for one group, it has an analytical solution). Its solution is what is called the Reactivity Trace Curve which keeps the power steady at the desired power. An inverse kinetic model which uses the on-line power data for reactivity calculation is used to provide initial condition (initial reactivity) for the RTC- power controller. Also feedback model are needed to evaluate both the temperature and Xenon reactivities which when subtracted from the RTC-value, one then can determine the reactivity required to keep the reactor power steady without power shooting. (authors)

  11. Identification of Chemical Reactor Plant’s Mathematical Model

    Pyakillya Boris


    Full Text Available This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

  12. Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

    Khedr Ahmed


    Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.


    Du Xuesong; Li Runfang; Lin Tengjiao


    Sealing analysis of sealing system in reactor pressure vessels is relevant with multiple nonlinear coupled-field effects, so even large-scale commercial finite element software cannot finish the complicated analysis. A fmite element method of 3D transient coupled thermo-elastic-plastic contact sealing analysis for reactor pressure vessels is presented, in which the surface nonlinearity,material nonlinearity, transient heat transfer nonlinearity and multiple coupled effect are taken into account and the sealing equation is coupling solved in iterative procedure. At the same time, a computational analysis program is developed, which is applied in the sealing analysis of experimental reactor pressure vessel, and the numerical results are in good coincidence with the experimental results. This program is also successful in analyzing the practical problem in engineering.

  14. Analysis of transients in advanced heavy water reactor using lumped parameter models

    Manmohan Pandey; Venkata Ramana Eaga; Sankar Sastry, P. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati (India); Gupta, S.K.; Lele, H.G.; Chatterjee, B. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)


    Full text of publication follows: Analysis of transients occurring in nuclear power plants, arising from the complex interplay between core neutronics and thermal-hydraulics, is important for their operation and safety. Numerical simulations of such transients can be carried out extensively at very low computational cost by using lumped parameter mathematical models. The Advanced Heavy Water Reactor (AHWR), being developed in India, is a vertical pressure tube type reactor cooled by boiling light water under natural circulation, using thorium as fuel and heavy water as moderator. In the present work, nonlinear and linear lumped parameter dynamic models for AHWR have been developed and validated with a distributed parameter model. The nonlinear lumped model is based on point reactor kinetics equations and one-dimensional homogeneous equilibrium model of two-phase flow. The distributed model is built with RELAP5/MOD3.2 code. Various types of transients have been simulated numerically, using the lumped model as well as RELAP5. The results have been compared and parameters tuned to make the lumped model match the distributed model (RELAP5) in terms of steady state as well as dynamic behaviour. The linear model has been derived by linearizing the nonlinear model for small perturbations about the steady state. Numerical simulations of transients using the linear model have been compared with results obtained from the nonlinear model. Thus, the range of validity of the linear model has been determined. Stability characteristics of AHWR have been investigated using the lumped parameter models. (authors)

  15. Transient analyses for a molten salt fast reactor with optimized core geometry

    Li, R., E-mail: [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)


    Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

  16. Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

    Gonzalez-Pintor, S.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain); Ginestar, D. [Departamento de Matematica Aplicada, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain)


    Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)

  17. The pressurization transient analysis for Lungmen advanced boiling water reactor using RETRAN-02

    Tsai, C.-W., E-mail: [Department of Engineering and System Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Shih Chunkuan [Department of Engineering and System Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Wang, J.-R.; Lin, H.-T. [Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Cheng, S.-C. [Department of Nuclear Engineering, Taiwan Power Company, No. 242, Sec. 3, Roosevelt Rd., Taipei City 10016, Taiwan (China)


    A RETRAN-02 model was devised and benchmarked against the preliminary safety analysis report (PSAR) for the Lungmen nuclear power plant roughly 10 years ago. During these years, the fuel design, some of the reactor vessel designs, and control systems have since been revised. The Lungmen RETRAN-02 model has also been modified with updated information when available. This study uses the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen RETRAN-02 plant model. Five transients, load rejection (LR), turbine trip (TT), main steam line isolation valves closure (MSIVC), loss of feedwater flow (LOFF), and one turbine control valve closure (OTCVC), were utilized to validate the Lungmen RETRAN-02 model. Moreover, due to the strong coupling effect between neutron dynamics and the thermal-hydraulic response during pressurization of transients, the one-dimensional kinetic model with the cross-section data library is used to simulate the coupling effect. The analytical results show good agreement in trends between the RETRAN-02 calculation and the Lungmen FSAR data. Based on the benchmark of these design-basis transients, the modified Lungmen RETRAN-02 model has been adjusted to a level of confidence for analysis of pressure increase transients. Analytical results indicate that the Lungmen advanced boiling water reactor (ABWR) design satisfied design criteria, i.e., vessel pressure and hot shutdown capability. However, a slight difference exists in the simulation of the water level for cases with changes in water levels. The Lungmen RETRAN-02 model tends to predict the change in water level at a slower rate than that in the Lungmen FSAR. There is also a slight difference in void reactivity response toward vessel pressure change in both simulations, which causes the calculated neutron flux before reactor shutdown to differ to some degree when the reactor experiences a rapid pressure increase. Further studies will be performed in the future using

  18. Transient Behaviour of Superconducting Magnet Systems of Fusion Reactor ITER during Safety Discharge

    A. M. Miri


    Full Text Available To investigate the transient behaviour of the toroidal and poloidal field coils magnet systems of the International Thermonuclear Experimental Reactor during safety discharge, network models with lumped elements are established. Frequency-dependant values of the network elements, that is, inductances and resistances are calculated with the finite element method. That way, overvoltages can be determined. According to these overvoltages, the insulation coordination of coils has to be selected.

  19. CFD simulation of the IAEA 10 MW generic MTR reactor under loss of flow transient

    Salama, Amgad, E-mail: [Konkuk University, Seoul 143-701 (Korea, Republic of); Atomic Energy Authority, Reactors Department, 13759 Cairo (Egypt); El-Morshedy, Salah El-Din, E-mail: [Atomic Energy Authority, Reactors Department, 13759 Cairo (Egypt)


    Three-dimensional simulation of the IAEA 10 MW generic reactor under loss of flow transient is introduced using the CFD code, Fluent. The IAEA reactor calculation is a safety-related benchmark problem for an idealized material testing reactor (MTR) pool type specified in order to compare calculational methods used in various research centers. The flow transients considered include fast loss of flow accidents (FLOFA) and slow loss of flow accidents (SLOFA) modeled with exponential flow decay and time constants of 1 and 25 s, respectively. The transients were initiated from a power of 12 MW with a flow trip point at 85% nominal flow and a 200 ms time delay. The simulation shows comparable results as those published by other research groups. However, interesting 3D patterns are shown that are usually lost based on the one-dimensional simulations that other research groups have introduced. In addition, information about the maximum clad surface temperature, the maximum fuel element temperature as well as the location of hot spots in fuel channel is also reported.

  20. Application of transient analysis using Hilbert spectra of electrochemical noise to the identification of corrosion inhibition

    Homborg, A.M.; Westing, E.P.M. van; Tinga, T.; Ferrari, G.M.; Zhang, X.; Wit, J.H.W. de; Mol, J.M.C.


    This study validates the ability of Hilbert spectra to investigate transients in an electrochemical noise signal for an aqueous corrosion inhibition process. The proposed analysis procedure involves the identification and analysis of transients in the electrochemical current noise signal. Their

  1. Nuclear reactor transient analysis via a quasi-static kinetics Monte Carlo method

    Jo, YuGwon; Cho, Bumhee; Cho, Nam Zin


    The predictor-corrector quasi-static (PCQS) method is applied to the Monte Carlo (MC) calculation for reactor transient analysis. To solve the transient fixed-source problem of the PCQS method, fission source iteration is used and a linear approximation of fission source distributions during a macro-time step is introduced to provide delayed neutron source. The conventional particle-tracking procedure is modified to solve the transient fixed-source problem via MC calculation. The PCQS method with MC calculation is compared with the direct time-dependent method of characteristics (MOC) on a TWIGL two-group problem for verification of the computer code. Then, the results on a continuous-energy problem are presented.

  2. Nuclear reactor transient analysis via a quasi-static kinetics Monte Carlo method

    Jo, YuGwon; Cho, Bumhee; Cho, Nam Zin, E-mail: [Korea Advanced Institute of Science and Technology 291 Daehak-ro, Yuseong-gu, Daejeon, Korea 305-701 (Korea, Republic of)


    The predictor-corrector quasi-static (PCQS) method is applied to the Monte Carlo (MC) calculation for reactor transient analysis. To solve the transient fixed-source problem of the PCQS method, fission source iteration is used and a linear approximation of fission source distributions during a macro-time step is introduced to provide delayed neutron source. The conventional particle-tracking procedure is modified to solve the transient fixed-source problem via MC calculation. The PCQS method with MC calculation is compared with the direct time-dependent method of characteristics (MOC) on a TWIGL two-group problem for verification of the computer code. Then, the results on a continuous-energy problem are presented.

  3. Multimodal Person Re-identification Using RGB-D Sensors and a Transient Identification Database

    Møgelmose, Andreas; Moeslund, Thomas B.; Nasrollahi, Kamal


    . Subjects are added to a transient database and re-identified based on the distance between recorded biometrics and the currently measured metrics. The system works on live video and requires no collaboration from the subjects. The system achieves a 68% re-identification rate with no wrong re...

  4. A jet-stirred reactor for transient studies of char gasification

    Calo, J.M.; Suuberg, E.M.; Hradil, G.; Perkins, M.T.; Lilly, W.D.


    For some time now in our laboratory we have been using well-mixed, ''gradientless'' reactors to study the dynamics of gas-carbon reactions. In the past we have used a mechanically-stirred, Berty-type reactor to examine the kinetics of carbon dioxide and steam gasification of chars. More recently, we have been involved in investigating the behavior of ''active sites'' in certain carbonaceous chars. As a result of the requirements of this work, we were led to the development of a jet-stirred, ''gradientless'' reactor. The current paper describes the design, development and application of this reactor to char gasification studies. The research project for which the reactor was developed is concerned with the study of the nature and behavior of ''active sites'' in the gasification of chars produced from synthesized model compounds, primarily those of the phenol-formaldehyde family of resins. One of the tasks defined for this project was to conduct transient kinetic studies on the gasification of these model compound chars, for both unlabelled and /sup 13/C-labelled resins. Due to the requirements of the expense of the labelled resins we were led to a design that minimized the relatively large sample requirements of the existing autoclave reactor. Once the necessity for developing a new reactor was realized, it was decided to employ jet-mixing in order to avoid some of the potential complexities of mechanical mixing under the particular operating conditions dictated by the nature of the reaction system. 6 refs., 8 figs., 1 tab.

  5. Inert matrix fuel neutronic, thermal-hydraulic, and transient behavior in a light water reactor

    Carmack, W. J.; Todosow, M.; Meyer, M. K.; Pasamehmetoglu, K. O.


    Currently, commercial power reactors in the United States operate on a once-through or open cycle, with the spent nuclear fuel eventually destined for long-term storage in a geologic repository. Since the fissile and transuranic (TRU) elements in the spent nuclear fuel present a proliferation risk, limit the repository capacity, and are the major contributors to the long-term toxicity and dose from the repository, methods and systems are needed to reduce the amount of TRU that will eventually require long-term storage. An option to achieve a reduction in the amount, and modify the isotopic composition of TRU requiring geological disposal is 'burning' the TRU in commercial light water reactors (LWRs) and/or fast reactors. Fuel forms under consideration for TRU destruction in light water reactors (LWRs) include mixed-oxide (MOX), advanced mixed-oxide, and inert matrix fuels. Fertile-free inert matrix fuel (IMF) has been proposed for use in many forms and studied by several researchers. IMF offers several advantages relative to MOX, principally it provides a means for reducing the TRU in the fuel cycle by burning the fissile isotopes and transmuting the minor actinides while producing no new TRU elements from fertile isotopes. This paper will present and discuss the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials in comparison with the results of similar analyses for reference UOX fuel bundles. The results of this work are to be used for screening purposes to identify the general feasibility of utilizing specific inert matrix fuel compositions in existing and future light water reactors. Compositions identified as feasible using the results of these analyses still require further detailed neutronic, thermal-hydraulic, and transient analysis study coupled with rigorous experimental testing and qualification.


    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart


    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  7. Dynamic modeling of primary and secondary systems of IRIS reactor for transient analysis using SIMULINK

    Magalhaes, Mardson Alencar de Sa; Lira, Carlos Alberto Brayner de Oliveira; Silva, Mario Augusto Bezerra da, E-mail: cabol@ufpe.b [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear; Lima, Fernando Roberto de Andrade, E-mail: [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN-PE), Recife, PE (Brazil)


    The IRIS project has significantly advanced in the last few years in response to a demand for a new generation reactor, that could fulfill the essential requirements for a future nuclear power plant: better economics, safety-by-design, low proliferation risk and environmental sustainability. IRIS reactor is a integral type PWR in which all primary components are arranged inside the pressure vessel. This configuration involves important changes in relation to a conventional PWR. These changes require several studies to comply with the safe operational limits for the reactor. In this paper, a study has been conducted to develop a dynamic model (named MODIRIS) for transient analysis, implemented in the MATLAB'S software SIMULINK, allowing the analysis of IRIS behavior by considering the neutron point kinetics for power production. The methodology is based on generating a set of differential equations of neutronic and thermal-hydraulic balances which describes the dynamics of the primary circuit, as well as a set of differential equations describing the dynamics of secondary circuit. The equations and initialization parameters at full power were into the SIMULINK and the code was validated by the confrontation with RELAP simulations for a transient of feedwater reduction in the steam generators. (author)

  8. MASTER-2.0: Multi-purpose analyzer for static and transient effects of reactors

    Cho, Byung Oh; Song, Jae Seung; Joo, Han Gyu [Korea Atomic Energy Research Institute, Taejon (Korea)


    MASTER-2.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the two group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM(Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with AFEN/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. Master-2.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P model can be used selectively. In addition, MASTER-2.0 is designed to cover various PWRs including SMART as well as WH-and CE-type reactors, providing all data required in their design procedures. (author). 39 refs., 12 figs., 4 tabs.

  9. Simulation and analysis of a WWER-1000 reactor under normal and transient conditions

    Baghban Ghonche


    Full Text Available An accurate analysis of the flow transient is very important in safety evaluation of a nuclear power plant. In this study, analysis of a WWER-1000 reactor is investigated. In order to perform this analysis, a model is developed to simulate the coupled kinetics and thermal-hydraulics of the reactor with a simple and accurate numerical algorithm. For thermal-hydraulic calculations, the four-equation drift-flux model is applied. Based on a multi-channel approach, core is divided into some regions. Each region has different characteristics as represented in a single fuel pin with its associated coolant channel. To obtain the core power distribution, point kinetic equations with different feedback effects are utilized. The appropriate initial and boundary conditions are considered and two situations of decreasing the coolant flow rate in a protected and unprotected core are analyzed. In addition to analysis of normal operation condition, a full range of thermal-hydraulic parameters is obtained for transients too. Finally, the data obtained from the model are compared with the calculations conducted using RELAP5/MOD3 code and Bushehr nuclear power plant data. It is shown that the model can provide accurate predictions for both steady-state and transient conditions.

  10. Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method

    He, Mingtao; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Zheng, Youqi, E-mail: [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China); Li, Xunzhao; Zhou, Shengcheng [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China)


    Highlights: • A transport-based kinetics code for Accelerator Driven Subcritical Reactors is developed. • The performance of different kinetics methods adapted to the ADSR is investigated. • The impacts of neutronic parameters deteriorating with fuel depletion are investigated. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is almost external source dominated since there is no additional reactivity control mechanism in most designs. This paper focuses on beam-induced transients with an in-house developed dynamic analysis code. The performance of different kinetics methods adapted to the ADSR is investigated, including the point kinetics approximation and space–time kinetics methods. Then, the transient responds of beam trip and beam overpower are calculated and analyzed for an ADSR design dedicated for minor actinides transmutation. The impacts of some safety-related neutronics parameters deteriorating with fuel depletion are also investigated. The results show that the power distribution varying with burnup leads to large differences in temperature responds during transients, while the impacts of kinetic parameters and feedback coefficients are not very obvious. Classification: Core physic.

  11. Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

    Ishii, Mamoru [Purdue Univ., West Lafayette, IN (United State


    The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup

  12. Thermal-hydraulic characteristics in a tokamak vacuum vessel of fusion reactor after transient events occurred

    Takase, Kazuyuki; Kunugi, Tomoaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Seki, Yasushi


    The thermal-hydraulic characteristics in a vacuum vessel (VV) of fusion reactor under the ingress-of-coolant-event (ICE) or loss-of-vacuum-event (LOVA) condition were carried out to investigate experimentally the thermofluid safety in the International Thermonuclear Experimental Reactor (ITER) under transient events. In the ICE experiments, the pressure rise and wall temperatures in the VV were measured and the performance of a suppression tank was confirmed. In the LOVA experiments, the exchange time inside the VV from the vacuum to be the atmospheric pressure was measured for various breach size and the exchange flow rates through the breaches of the VV under the atmospheric pressure conditions were clarified. (author)

  13. Laser pulse heating of nuclear fuels for simulation of reactor power transients

    C S Viswanadham; K C Sahoo; T R G Kutty; K B Khan; V P Jathar; S Anantharaman; Arun Kumar; G K Dey


    It is important to study the behaviour of nuclear fuels under transient heating conditions from the point of view of nuclear safety. To simulate the transient heating conditions occurring in the known reactor accidents like loss of coolant accident (LOCA) and reactivity initiated accident (RIA), a laser pulse heating system is under development at BARC, Mumbai. As a prelude to work on irradiated nuclear fuel specimens, pilot studies on unirradiated UO2 fuel specimens were carried out. A laser pulse was used to heat specimens of UO2 held inside a chamber with an optically transparent glass window. Later, these specimens were analysed by metallography and X-ray diffraction. This paper describes the results of these studies.

  14. Investigation of natural circulation instability and transients in passively safe novel modular reactor

    Shi, Shanbin

    The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal

  15. Optimization of a heterogeneous fast breeder reactor core with improved behavior during unprotected transients

    Poumerouly, S.; Schmitt, D.; Massara, S.; Maliverney, B. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)


    Innovative Sodium-cooled Fast Reactors (SFRs) are currently being investigated by CEA, AREVA and EDF in the framework of a joint French collaboration, and the construction of a GEN IV prototype, ASTRID (Advanced Sodium Technical Reactor for Industrial Demonstration), is scheduled in the years 2020. Significant improvements are expected so as to improve the reactor safety: the goal is to achieve a robust safety demonstration of the mastering of the consequences of a Core Disruptive Accident (CDA), whether by means of prevention or mitigation features. In this framework, an innovative design was proposed by CEA in 2010. It aims at strongly reducing the sodium void effect, thereby improving the core behavior during unprotected loss of coolant transients. This design is strongly heterogeneous and includes, amongst others, a fertile plate, a sodium plenum associated with a B{sub 4}C upper blanket and a stepwise modulation of the fissile height of the core (onwards referred to as the 'diabolo shape'). In this paper, studies which were entirely carried out at EDF are presented: the full potential of this heterogeneous concept is thoroughly investigated using the SDDS methodology. (authors)

  16. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Yan Wang


    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

  17. Effect of Transient Nicotine Load Shock on the Performance of Pseudomonas sp. HF-1 Bioaugmented Sequencing Batch Reactors

    Dong-sheng Shen


    Full Text Available Bioaugmentation with degrading bacteria can improve the treatment of nicotine-containing tobacco industrial wastewater effectively. However, the transient and extremely high feeding of pollutants may compromise the effectiveness of the bioaugmented reactors. The effect of transient nicotine shock loads on the performance of Pseudomonas sp. HF-1 bioaugmented SBRs were studied. The results showed that, under 500–2500 mg/L of transient nicotine shocks, all the reactors still could realize 100% of nicotine degradation in 4 days of recovery, while the key nicotine degradation enzyme HSP hydroxylase increased in expression. Though the dramatic increase of activities of ROS, MDA, SOD, and CAT suggested that transient nicotine shock loads could induce oxidative stress on microorganisms in activated sludge, a decrease to control level demonstrated that most of the microorganisms could resist 500–1500 mg/L of transient nicotine shock under the protection from strain HF-1. After 8 cycles of recovery, high ROS level and low TOC removal in high transient shock reactors implied that 2000–2500 mg/L of transient nicotine shock was out of its recovery of strain HF-1 bioaugmented system. This study enriched our understanding on highly efficient nicotine-degrading strain bioaugmented system, which would be beneficial to tobacco waste or wastewater treatment in engineering.

  18. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)


    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  19. Simulation of a channel blockage transient in the Angra 2 Nuclear Reactor using a RELAP5-3D model

    Gonzalez-Mantecon, Javier; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Pereira, Claubia; Reis, Patricia A.L.; Scari, Maria E., E-mail:, E-mail:, E-mail:, E-mail:, E-mail:, E-mail: [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear


    The Angra 2 Nuclear Power Plant (NPP) is a Pressurized Water Reactor (PWR) type with electrical output of about 1350 MW. The RELAP5-3D code was used to develop a detailed thermal hydraulic model of such reactor using reference data from the Angra 2 Final Safety Analysis Report (FSAR). In this work, a blockage transient has been investigated at full power operation. The transient herein considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  20. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    Ruggles, A. E. [Oak Ridge National Lab., TN (United States); Tennessee Univ., Knoxville, TN (United States); Cheng, L. Y. [Brookhaven National Lab., Upton, NY (United States); Dimenna, R. A. [Westinghouse Savannah River Co., Aiken, SC (United States); Griffith, P. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Wilson, G. E. [Idaho National Engineering Lab., Idaho Falls, ID (United States)


    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis.

  1. Reactors

    International Electrotechnical Commission. Geneva


    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  2. CFD investigation of flow inversion in typical MTR research reactor undergoing thermal-hydraulic transients

    Salama, Amgad, E-mail: [Atomic Energy Authority, Reactors Department, 13759 Cairo (Egypt)


    Highlights: > The 3D, CFD simulation of FLOFA accident in the generic IAEA 10 MW research reactor is carried out. > The different flow and heat transfer mechanisms involved in this process were elucidated. > The transition between these mechanisms during the course of FLOFA is discussed and investigated. > The interesting inversion process upon the transition from downward flow to upward flow is shown. > The temperature field and the friction coefficient during the whole transient process were shown. - Abstract: Three dimensional CFD full simulations of the fast loss of flow accident (FLOFA) of the IAEA 10 MW generic MTR research reactor are conducted. In this system the flow is initially downward. The transient scenario starts when the pump coasts down exponentially with a time constant of 1 s. As a result the temperatures of the heating element, the clad, and the coolant rise. When the flow reaches 85% of its nominal value the control rod system scrams and the power drops sharply resulting in the temperatures of the different components to drop. As the coolant flow continues to drop, the decay heat causes the temperatures to increase at a slower rate in the beginning. When the flow becomes laminar, the rate of temperature increase becomes larger and when the pumps completely stop a flow inversion occurs because of natural convection. The temperature will continue to rise at even higher rates until natural convection is established, that is when the temperatures settle off. The interesting 3D patterns of the flow during the inversion process are shown and investigated. The temperature history is also reported and is compared with those estimated by one-dimensional codes. Generally, very good agreement is achieved which provides confidence in the modeling approach.

  3. System identification during a transient via wavelet multiresolution analysis followed by spectral techniques

    Antonopoulos-Domis, M.; Tambouratzis, T. [NCSR, Athens (Greece). Institute of Nuclear Technology-Radiation Protection


    A method is proposed for system identification during a transient, employing wavelet multiresolution analysis and denoising followed by classical (FFT) special analysis. The method has been tested on numerical experiments. (Author).

  4. System identification during a transient via wavelet multiresolution analysis followed by spectral techniques

    Antonopoulos-Domis, M.; Tambouratzis, T


    A method is proposed for system identification during a transient, employing wavelet multiresolution analysis and denoising followed by classical (FFT) special analysis. The method has been tested on numerical experiments.

  5. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.


    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

  6. Application of transient analysis using Hilbert spectra of electrochemical noise to the identification of corrosion inhibition

    Homborg, A.M.; Westing, E.P.M. van; Tinga, T.; Ferrari, G.M.; Zhang, X.; Wit, J.H.W. de; Mol, J.M.C.


    This study validates the ability of Hilbert spectra to investigate transients in an electrochemical noise signal for an aqueous corrosion inhibition process. The proposed analysis procedure involves the identification and analysis of transients in the electrochemical current noise signal. Their deco

  7. Application of transient analysis using Hilbert spectra of electrochemical noise to the identification of corrosion inhibition

    Homborg, A.M.; Westing, van E.P.M.; Tinga, T.; Ferrari, G.M.; Zhang, X.; Wit, de J.H.W.; Mol, J.M.C.


    This study validates the ability of Hilbert spectra to investigate transients in an electrochemical noise signal for an aqueous corrosion inhibition process. The proposed analysis procedure involves the identification and analysis of transients in the electrochemical current noise signal. Their deco

  8. Inverse transient heat conduction problems and identification of thermal parameters

    Atchonouglo, K.; Banna, M.; Vallée, C.; Dupré, J.-C.


    This work deals with the estimation of polymers properties. An inverse analysis based on finite element method is applied to identify simultaneously the constants thermal conductivity and heat capacity per unit volume. The inverse method algorithm constructed is validated from simulated transient temperature recording taken at several locations on the surface of the solid. Transient temperature measures taped with infrared camera on polymers were used for identifying the thermal properties. The results show an excellent agreement between manufacturer and identified values.

  9. Transient recovery voltage analysis for various current breaking mathematical models: shunt reactor and capacitor bank de-energization study

    Oramus Piotr


    Full Text Available Electric arc is a complex phenomenon occurring during the current interruption process in the power system. Therefore performing digital simulations is often necessary to analyse transient conditions in power system during switching operations. This paper deals with the electric arc modelling and its implementation in simulation software for transient analyses during switching conditions in power system. Cassie, Cassie-Mayr as well as Schwarz-Avdonin equations describing the behaviour of the electric arc during the current interruption process have been implemented in EMTP-ATP simulation software and presented in this paper. The models developed have been used for transient simulations to analyse impact of the particular model and its parameters on Transient Recovery Voltage in different switching scenarios: during shunt reactor switching-off as well as during capacitor bank current switching-off. The selected simulation cases represent typical practical scenarios for inductive and capacitive currents breaking, respectively.

  10. Application programming interface document for the modernized Transient Reactor Analysis Code (TRAC-M)

    Mahaffy, J. [Pennsylvania State Univ., University Park, PA (United States); Boyack, B.E.; Steinke, R.G. [Los Alamos National Lab., NM (United States)


    The objective of this document is to ease the task of adding new system components to the Transient Reactor Analysis Code (TRAC) or altering old ones. Sufficient information is provided to permit replacement or modification of physical models and correlations. Within TRAC, information is passed at two levels. At the upper level, information is passed by system-wide and component-specific data modules at and above the level of component subroutines. At the lower level, information is passed through a combination of module-based data structures and argument lists. This document describes the basic mechanics involved in the flow of information within the code. The discussion of interfaces in the body of this document has been kept to a general level to highlight key considerations. The appendices cover instructions for obtaining a detailed list of variables used to communicate in each subprogram, definitions and locations of key variables, and proposed improvements to intercomponent interfaces that are not available in the first level of code modernization.

  11. Experimental and analytical study of stability characteristics of natural circulation boiling water reactors during startup transient

    Woo, Kyoungsuk

    Two-phase natural circulation loops are unstable at low pressure operating conditions. New reactor design relying on natural circulation for both normal and abnormal core cooling is susceptible to different types of flow instabilities. In contrast to forced circulation boiling water reactor (BWR), natural circulation BWR is started up without recirculation pumps. The tall chimney placed on the top of the core makes the system susceptible to flashing during low pressure start-up. In addition, the considerable saturation temperature variation may induce complicated dynamic behavior driven by thermal non-equilibrium between the liquid and steam. The thermal-hydraulic problems in two-phase natural circulation systems at low pressure and low power conditions are investigated through experimental methods. Fuel heat conduction, neutron kinetics, flow kinematics, energetics and dynamics that govern the flow behavior at low pressure, are formulated. A dimensionless analysis is introduced to obtain governing dimensionless groups which are groundwork of the system scaling. Based on the robust scaling method and start-up procedures of a typical natural circulation BWR, the simulation strategies for the transient with and without void reactivity feedback is developed. Three different heat-up rates are applied to the transient simulations to study characteristics of the stability during the start-up. Reducing heat-up rate leads to increase in the period of flashing-induced density wave oscillation and decrease in the system pressurization rate. However, reducing the heat-up rate is unable to completely prevent flashing-induced oscillations. Five characteristic regions of stability are discovered at low pressure conditions. They are stable single-phase, flashing near the separator, intermittent oscillation, sinusoidal oscillation and low subcooling stable regions. Stability maps were acquired for system pressures ranging 100 kPa to 400 kPa. According to experimental investigation


    Goldstein, D. A.; Nugent, P. E. [Department of Astronomy, University of California, Berkeley, 501 Campbell Hall #3411, Berkeley, CA 94720 (United States); D’Andrea, C. B.; Nichol, R. C.; Papadopoulos, A. [Institute of Cosmology and Gravitation, University of Portsmouth, Dennis Sciama Building, Burnaby Road, Portsmouth, PO1 3FX (United Kingdom); Fischer, J. A.; Sako, M.; Wolf, R. C. [Department of Physics and Astronomy, University of Pennsylvania, Philadelphia, PA 19104 (United States); Foley, R. J. [Astronomy Department, University of Illinois at Urbana-Champaign, 1002 West Green Street, Urbana, IL 61801 (United States); Gupta, R. R. [Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States); Kessler, R. [Kavli Institute for Cosmological Physics, University of Chicago, Chicago, IL 60637 (United States); Kim, A. G.; Thomas, R. C. [Lawrence Berkeley National Laboratory, 1 Cyclotron Road, Berkeley, CA 94720 (United States); Smith, M.; Sullivan, M. [School of Physics and Astronomy, University of Southampton, Highfield, Southampton, SO17 1BJ (United Kingdom); Wester, W. [Fermi National Accelerator Laboratory, P.O. Box 500, Batavia, IL 60510 (United States); Abdalla, F. B.; Benoit-Lévy, A. [Department of Physics and Astronomy, University College London, Gower Street, London, WC1E 6BT (United Kingdom); Banerji, M. [Kavli Institute for Cosmology, University of Cambridge, Madingley Road, Cambridge CB3 0HA (United Kingdom); Bertin, E. [Institut d’Astrophysique de Paris, Univ. Pierre et Marie Curie and CNRS UMR7095, F-75014 Paris (France); and others


    We describe an algorithm for identifying point-source transients and moving objects on reference-subtracted optical images containing artifacts of processing and instrumentation. The algorithm makes use of the supervised machine learning technique known as Random Forest. We present results from its use in the Dark Energy Survey Supernova program (DES-SN), where it was trained using a sample of 898,963 signal and background events generated by the transient detection pipeline. After reprocessing the data collected during the first DES-SN observing season (2013 September through 2014 February) using the algorithm, the number of transient candidates eligible for human scanning decreased by a factor of 13.4, while only 1.0% of the artificial Type Ia supernovae (SNe) injected into search images to monitor survey efficiency were lost, most of which were very faint events. Here we characterize the algorithm’s performance in detail, and we discuss how it can inform pipeline design decisions for future time-domain imaging surveys, such as the Large Synoptic Survey Telescope and the Zwicky Transient Facility. An implementation of the algorithm and the training data used in this paper are available at at

  13. Automated transient identification in the Dark Energy Survey

    Goldstein, D. A. [Univ. of California, Berkeley, CA (United States); Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). et al.


    We describe an algorithm for identifying point-source transients and moving objects on reference-subtracted optical images containing artifacts of processing and instrumentation. The algorithm makes use of the supervised machine learning technique known as Random Forest. We present results from its use in the Dark Energy Survey Supernova program (DES-SN), where it was trained using a sample of 898,963 signal and background events generated by the transient detection pipeline. After reprocessing the data collected during the first DES-SN observing season (2013 September through 2014 February) using the algorithm, the number of transient candidates eligible for human scanning decreased by a factor of 13.4, while only 1.0 percent of the artificial Type Ia supernovae (SNe) injected into search images to monitor survey efficiency were lost, most of which were very faint events. Here we characterize the algorithm's performance in detail, and we discuss how it can inform pipeline design decisions for future time-domain imaging surveys, such as the Large Synoptic Survey Telescope and the Zwicky Transient Facility.

  14. Automated Transient Identification in the Dark Energy Survey

    Goldstein, D A; Fischer, J A; Foley, R J; Gupta, R R; Kessler, R; Kim, A G; Nichol, R C; Nugent, P; Papadopoulos, A; Sako, M; Smith, M; Sullivan, M; Thomas, R C; Wester, W; Wolf, R C; Abdalla, F B; Banerji, M; Benoit-Lévy, A; Bertin, E; Brooks, D; Rosell, A Carnero; Castander, F J; da Costa, L N; Covarrubias, R; DePoy, D L; Desai, S; Diehl, H T; Doel, P; Eifler, T F; Neto, A Fausti; Finley, D A; Flaugher, B; Fosalba, P; Frieman, J; Gerdes, D; Gruen, D; Gruendl, R A; James, D; Kuehn, K; Kuropatkin, N; Lahav, O; Li, T S; Maia, M A G; Makler, M; March, M; Marshall, J L; Martini, P; Merritt, K W; Miquel, R; Nord, B; Ogando, R; Plazas, A A; Romer, A K; Roodman, A; Sanchez, E; Scarpine, V; Schubnell, M; Sevilla-Noarbe, I; Smith, R C; Soares-Santos, M; Sobreira, F; Suchyta, E; Swanson, M E C; Tarle, G; Thaler, J; Walker, A R


    We describe an algorithm for identifying point-source transients and moving objects on reference-subtracted optical images containing artifacts of processing and instrumentation. The algorithm makes use of the supervised machine learning technique known as Random Forest. We present results from its use in the Dark Energy Survey Supernova program (DES-SN), where it was trained using a sample of 898,963 signal and background events generated by the transient detection pipeline. After reprocessing the data collected during the first DES-SN observing season (Sep. 2013 through Feb. 2014) using the algorithm, the number of transient candidates eligible for human scanning decreased by a factor of 13.4, while only 1 percent of the artificial Type Ia supernovae (SNe) injected into search images to monitor survey efficiency were lost, most of which were very faint events. Here we characterize the algorithm's performance in detail, and we discuss how it can inform pipeline design decisions for future time-domain imaging...

  15. A three-dimensional transient neutronics routine for the TRAC-PF1 reactor thermal hydraulic computer code

    Bandini, B.R. [Los Alamos National Lab., NM (United States)


    No present light water reactor accident analysis code employs both high state of the art neutronics and thermal-hydraulics computational algorithms. Adding a modern three-dimensional neutron kinetics model to the present TRAC-PFI/MOD2 code would create a fully up to date pressurized water reactor accident evaluation code. After reviewing several options, it was decided that the Nodal Expansion Method would best provide the basis for this multidimensional transient neutronic analysis capability. Steady-state and transient versions of the Nodal Expansion Method were coded in both three-dimensional Cartesian and cylindrical geometries. In stand-alone form this method of solving the few group neutron diffusion equations was shown to yield efficient and accurate results for a variety of steady-state and transient benchmark problems. The Nodal Expansion Method was then incorporated into TRAC-PFl/MOD2. The combined NEM/TRAC code results agreed well with the EPRI-ARROTTA core-only transient analysis code when modelling a severe PWR control rod ejection accident.

  16. Transient plasma estimation: a noise cancelling/identification approach

    Candy, J.V.; Casper, T.; Kane, R.


    The application of a noise cancelling technique to extract energy storage information from sensors occurring during fusion reactor experiments on the Tandem Mirror Experiment-Upgrade (TMX-U) at the Lawrence Livermore National Laboratory (LLNL) is examined. We show how this technique can be used to decrease the uncertainty in the corresponding sensor measurements used for diagnostics in both real-time and post-experimental environments. We analyze the performance of algorithm on the sensor data and discuss the various tradeoffs. The algorithm suggested is designed using SIG, an interactive signal processing package developed at LLNL.

  17. Cynod: A Neutronics Code for Pebble Bed Modular Reactor Coupled Transient Analysis

    Hikaru Hiruta; Abderrafi M. Ougouag; Hans D. Gougar; Javier Ortensi


    The Pebble Bed Reactor (PBR) is one of the two concepts currently considered for development into the Next Generation Nuclear Plant (NGNP). This interest is due, in particular, to the concept’s inherent safety characteristics. In order to verify and confirm the design safety characteristics of the PBR computational tools must be developed that treat the range of phenomena that are expected to be important for this type of reactors. This paper presents a recently developed 2D R-Z cylindrical nodal kinetics code and shows some of its capabilities by applying it to a set of known and relevant benchmarks. The new code has been coupled to the thermal hydraulics code THERMIX/KONVEK[1] for application to the simulation of very fast transients in PBRs. The new code, CYNOD, has been written starting with a fixed source solver extracted from the nodal cylindrical geometry solver contained within the PEBBED code. The fixed source solver was then incorporated into a kinetic solver.. The new code inherits the spatial solver characteristics of the nodal solver within PEBBED. Thus, the time-dependent neutron diffusion equation expressed analytically in each node of the R-Z cylindrical geometry sub-domain (or node) is transformed into one-dimensional equations by means of the usual transverse integration procedure. The one-dimensional diffusion equations in each of the directions are then solved using the analytic Green’s function method. The resulting equations for the entire domain are then re-cast in the form of the Direct Coarse Mesh Finite Difference (D-CMFD) for convenience of solution. The implicit Euler method is used for the time variable discretization. In order to correctly treat the cusping effect for nodes that contain a partially inserted control rod a method is used that takes advantage of the Green’s function solution available in the intrinsic method. In this corrected treatment, the nodes are re-homogenized using axial flux shapes reconstructed based on the

  18. Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

    Kalcheva, S.; Ponsard, B.; Koonen, E.


    The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

  19. Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

    Jeong Soon Park


    Full Text Available The failure probabilities of the reactor pressure vessel (RPV for low temperature over-pressurization (LTOP and cool-down transients are calculated in this study. For the cool-down transient, a pressure–temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT. The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

  20. Development of a preliminary PIRT (Phenomena Identification and Ranking Table) of thermal-hydraulic phenomena for 330MWt SMART integral reactor

    Chung, B. D.; Lee, W. J.; Sim, S. K.; Song, J. H.; Kim, H. C.


    The work reported in this document identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in a 330 MWt SMART integral reactor which is under development at KAERI. The result of this efforts is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary Phenomena Identification and Ranking Table (PIRT) has been developed based on the experts` knowledge and experience. The preliminary PIRT has been developed by the consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this report is intended for use to identify and integrate development areas of further experimental tests needed and thermal-hydraulic models and correlations and code improvements for the safety analysis of the SMART integral reactor. (author). 7 refs., 21 tabs., 22 figs.

  1. Contraction of information and its inverse problem in reactor system identification and stochastic diagnosis

    Kishida, K. [Gifu Univ. (Japan)


    Research concerning power reactor noise analysis makes rapid progress in the areas of the system identification, prediction and diagnosis. Keywords in these studies are artificial intelligence, neural network, fuzzy, and chaos. Nonlinear, nonstationary, or non-Gaussian processes as well as linear and steady processes are also studied in fluctuation analysis. However, we have not enough time to study a fundamental theory, since we are urged to obtain results or applications in power reactor fluctuations. Furthermore, we have no systematic approach to handle observed time series data in the linear process, since power reactor noise phenomena are complicated. Hence, it is important to study it from the fundamental viewpoint. It is a main aim of the present review paper to describe a unified formalism for reactor system identification and stochastic diagnosis.

  2. Neutronics, Steady-State, and Transient Analyses for the Kazakhstan VVR-K Reactor with LEU Fuel

    Hanan, N.A.; Garner, P.L.


    Calculations have been performed for steady state and postulated transients in the VVRK reactor at the Institute of Nuclear Physics (INP) in Alatau, Kazakhstan. These calculations have been performed at the request of staff of the INP who have performed similar calculations. Calculations were performed for the fresh low-enriched uranium (LEU) core and for four subsequent cores as beryllium is added as a radial reflector to maintain criticality during the first 15 cycles of operation. The calculations include neutronics parameters, steady-state power and temperature distributions, and response to transients. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process.

  3. Adaptive identification and interpretation of pressure transient tests of horizontal wells: challenges and perspectives

    Sergeev, V. L.; Van Hoang, Dong


    The paper deals with a topical issue of defining oil reservoir properties during transient tests of horizontal wells equipped with information-measuring systems and reducing well downtime. The aim is to consider challenges and perspectives of developing models and algorithms for adaptive identification and interpretation of transient tests in horizontal wells with pressure buildup curve analysis. The models and algorithms should allow analyzing flow behavior, defining oil reservoir properties and determining well test completion time, as well as reducing well downtime. The present paper is based on the previous theoretical and practical findings in the spheres of transient well testing, systems analysis, system identification, function optimization and linear algebra. Field data and results of transient well tests with pressure buildup curve analysis have also been considered. The suggested models and algorithms for adaptive interpretation of transient tests conducted in horizontal wells with resulting pressure buildup curve make it possible to analyze flow behavior, as well as define the reservoir properties and determine well test completion time. The algorithms for adaptive interpretation are based on the integrated system of radial flow PBC models with time- dependent variables, account of additional a priori information and estimates of radial flow permeability. Optimization problems are solved with the case study of PBC interpretation for five horizontal wells of the Verkhnechonsk field.

  4. Non-Intrusive Demand Monitoring and Load Identification for Energy Management Systems Based on Transient Feature Analyses

    Hsueh-Hsien Chang


    Full Text Available Energy management systems strive to use energy resources efficiently, save energy, and reduce carbon output. This study proposes transient feature analyses of the transient response time and transient energy on the power signatures of non-intrusive demand monitoring and load identification to detect the power demand and load operation. This study uses the wavelet transform (WT of the time-frequency domain to analyze and detect the transient physical behavior of loads during the load identification. The experimental results show the transient response time and transient energy are better than the steady-state features to improve the recognition accuracy and reduces computation requirements in non-intrusive load monitoring (NILM systems. The discrete wavelet transform (DWT is more suitable than short-time Fourier transform (STFT for transient load analyses.

  5. Analysis of unprotected transients with control and safety rod drive mechanism expansion feedback in a medium sized oxide fuelled fast breeder reactor

    Sathiyasheela, T., E-mail:; Natesan, K.; Srinivasan, G.S.; Devan, K.; Puthiyavinayagam, P.


    Highlights: • Possibilities of enhancing safety under ULOF and UTOP accidents. • CSRDM expansion feedbacks under unprotected transients. • CSRDM expansion feedback enhances the safety of fast reactors. • CSRDM expansion feedbacks ensuring enough time for initiating safety actions. - Abstract: Possibilities of enhancing core safety under unprotected loss of flow (ULOF) and unprotected transient over power (UTOP) accidents with control and safety rod drive mechanism (CSRDM) expansion feedbacks are explored in a medium sized oxide fuelled fast breeder reactor. This feedback is expected to take the reactor to a safe shutdown under ULOF and to an another steady state under UTOP where there is no significant fuel melting. Under ULOF, with CSRDM feedback net reactivity was maintained negative throughout the transient (up to 2000 s) and the power dropped to a level of heat removal capacity of decay heat removal system based on natural circulation. Similarly, under UTOP with the above feedback reactor power goes to a lower peak value. The fuel temperature is just touching the melting temperature and the melt fraction does not cross 5%. With CSRDM expansion feedbacks both ULOF and UTOP transients prolong beyond 2000 s. It ensures, availability of time for initiating any safety actions against the transients, and thus it helps to preclude core disruptive accidents (CDA) in a medium sized oxide fuelled reactors.Classification: L. safety and risk analysis.

  6. Identification of the reduced order models of a BWR reactor; Identificacion de modelos de orden reducido de un reactor BWR

    Hernandez S, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail:


    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  7. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    Massoud, M


    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  8. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert


    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  9. Initial verification and validation of RAZORBACK - A research reactor transient analysis code

    Talley, Darren G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)


    This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.

  10. Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

    Roberto, Thiago D., E-mail: [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)


    The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

  11. Performance Test of System Identification Methods for a Nuclear Reactor

    Yu, Keuk Jong; Kim, Han Gon [KHNP, Daejeon (Korea, Republic of)


    An automatic controller that uses the model predictive control (MPC) method is being developed for automatic load follow operation. As described in Ref. a system identification method is important in the MPC method because MPC is based on a system model produced by system identification. There are many models and methods of system identification. In this study, AutoRegressive eXogenous (ARX) model was selected from among them, and the recursive least square (RLS) method and least square (LS) method associated with this model are used in a comparative performance analysis

  12. Equalisation of Transient Temperature Profile Within the Fuel Pin of a Miniature Neutron Source Reactor (MNSR During Total Loss of Coolant

    Christian Amevi Adjei


    Full Text Available Transient temperature distributions in cylindrical fuel element of Ghana Research Reactor-1 (GHARR-1 Miniature Neutron Source Reactor (MNSR following sudden total loss of cooling have been investigated. The loss of cooling in the reactor core resulting from a blockage of the inner orifice of coolant flow channels was assumed to occur during normal operations and led to sudden shut dow n of the reactor. The objective was to analyse the transient behaviour by solving analytically the heat transfer equation using Bessel functions and also develop from first principle the transient temperature equations for the fuel element. Results obtained during a sudden total lost of cooling showed a high transient temperature distribution at the centre of the fuel element, with the surface of the fuel clad recording the least temperature. The transient temperature distribution decreased from the centre of the fuel element to the surface of the fuel clad and followed a parabolic decay pattern which after increase in tim e follow ed an equalisation pattern. During sudden shut down, since there w as no heat generated and decay heat , the rate at which the fuel elem ent was cooled w as directly proportional to time.

  13. Vital area identification for U.S. Nuclear Regulatory Commission nuclear power reactor licensees and new reactor applicants.

    Whitehead, Donnie Wayne; Varnado, G. Bruce


    U.S. Nuclear Regulatory Commission nuclear power plant licensees and new reactor applicants are required to provide protection of their plants against radiological sabotage, including the placement of vital equipment in vital areas. This document describes a systematic process for the identification of the minimum set of areas that must be designated as vital areas in order to ensure that all radiological sabotage scenarios are prevented. Vital area identification involves the use of logic models to systematically identify all of the malicious acts or combinations of malicious acts that could lead to radiological sabotage. The models available in the plant probabilistic risk assessment and other safety analyses provide a great deal of the information and basic model structure needed for the sabotage logic model. Once the sabotage logic model is developed, the events (or malicious acts) in the model are replaced with the areas in which the events can be accomplished. This sabotage area logic model is then analyzed to identify the target sets (combinations of areas the adversary must visit to cause radiological sabotage) and the candidate vital area sets (combinations of areas that must be protected against adversary access to prevent radiological sabotage). Any one of the candidate vital area sets can be selected for protection. Appropriate selection criteria will allow the licensee or new reactor applicant to minimize the impacts of vital area protection measures on plant safety, cost, operations, or other factors of concern.

  14. Diagnosis of class using swarm intelligence applied to problem of identification of nuclear transient

    Villas Boas Junior, Manoel; Strauss, Edilberto, E-mail: junior@lmp.ufrj.b [Instituto Federal de Educacao, Ciencia e Tecnologia do Ceara/ Universidade do Estado do Ceara, Itaperi, CE (Brazil). Mestrado Integrado em Computacao Aplicada; Nicolau, Andressa dos Santos; Schirru, Roberto, E-mail: andressa@lmp.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Mello, Flavio Luis de [Universidade Federal do Rio de Janeiro (POLI/UFRJ), RJ (Brazil). Escola Politecnica. Dept. de Engenharia Eletronica e Computacao


    This article presents a computational model of the diagnostic system of transient. The model makes use of segmentation techniques applied to support decision making, based on identification of classes and optimized by Particle Swarm Optimization algorithm (PSO). The method proposed aims to classify an anomalous event in the signatures of three classes of the design basis transients postulated for the Angra 2 nuclear plant, where the PSO algorithm is used as a method of separation of classes, being responsible for finding the best centroid prototype vector of each accident/transient, ie equivalent to Voronoi vector that maximizes the number of correct classifications. To make the calculation of similarity between the set of the variables anomalous event in a given time t, and the prototype vector of variables of accident/transients, the metrics of Manhattan, Euclidean and Minkowski were used. The results obtained by the method proposed were compatible with others methods reported in the literature, allowing a solution that approximates the ideal solution, ie the Voronoi vectors. (author)

  15. Classification of Transient Events of Nuclear Reactor Using Hidden Markov Model

    P. Bečvář


    Full Text Available This article describes a part of on-line system for a residual life of a pressure vessel shell. In this system there appears a need for determining of a real history of a pressure vessel described as a sequence of so called transient events. Each event (there are 23 events now is given by its template. It is theoretically necessary to compare data measured in a real history with all possible sequences of transient events. This solution in intractable and that is why a more sophisticated solution had to be used. Because this task is very similar to task of an automatic speech recognition, models and algorithms used to solve speech recognition tasks can be efficiently used to solve our task. Of course there are some different circumstances caused by the fact that the transient events take much longer than words and their number is much smaller than the number of words in speech recognition system's vocabulary. But the inspiration from speech recognition was very useful and the known algorithms rapidly decreased the design time.

  16. 1 GW固态燃料熔盐堆运行瞬态分析%Transient analysis with 1-GW solid fuel molten salt reactor

    张洁; 李明海; 何龙; 杨洋; 戴叶; 蔡翔舟


    Background:As a new type of reactor, thorium-based molten salt reactor (TMSR) has unique safety and operation characteristics. Its thermal-hydraulic features are enormously different from other reactors, thus worth doing transient analysis. Purpose:This study aims at disturbed transient analysis of TMSR for fundamental comprehension of its safety and operation characteristics. Methods:The structure and design scheme of the core of 1-GW solid fuel thorium-based molten salt reactor (TMSR-SF) have been presented. Structural emulation platform for transient analysis is proposed with the thermo-hydraulic model being developed on the basis of RELAP5, and the control system model being constructed by using the MATLAB/Simulink software. Results:The results of emulation test for operational transient, such as rapid power reduction, step load reducing, linear load reducing and temperature of secondary loop inlet reducing show that the reactor control system is effective to bring the reactor into a safe and steady state without actuation of reactor protection system. Conclusion: Analysis results show that the design satisfied the requirements of 1-GW TMSR-SF operational transient. It also indicates that the platform can perfectly simulate the variable power transient conditions.%钍基熔盐堆(Thorium-based Molten Salt Reactor, TMSR)作为一种新的堆型,具有独特的安全与运行特性。研究其热工水力特性,对其进行瞬态分析,将有助于深刻理解该反应堆。本文介绍了1 GW固态熔盐堆的堆芯设计方案,并描述了用于瞬态分析的详细程序结构。其中,利用 RELAP5对其热工水力模型进行模拟;利用 Simulink 对其控制系统模型进行模拟。通过预期运行瞬态,例如功率降低、堆芯反应性引入、二回路温度变化等工况显示了其运行特性,并验证了控制系统可以使反应堆达到安全稳定状态,而不触发保护系统动作。

  17. Crack growth behaviour of low alloy steels for pressure boundary components under transient light water reactor operating conditions (CASTOC)

    Foehl, J.; Weissenberg, T. [Materialpruefungsanstalt, Univ. Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT) (Spain); Ernestova, M.; Zamboch, M. [Nuclear Research Inst. (NRI) (Czech Republic); Seifert, H.P.; Ritter, S. [Paul Scherrer Inst. (PSI) (Switzerland); Roth, A.; Devrient, B. [Framatome ANP GmbH (F ANP) (Germany); Ehrnsten, U. [Technical Research Centre of Finland (VTT) (Finland)


    The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the transferability of the results to components in service. The obtained crack growth rate data were reflected on literature data and on commonly applied prediction curves as presented in the appropriate Code. At relevant stress intensity factors it could be shown that immediate cessation of growing cracks occurs after changing from cyclic to static load in high purity oxygenated BWR water and oxygen-free VVER water corresponding to steady state operation conditions. Susceptibility to environmentally assisted cracking under static load was observed for a heat affected zone material in oxygenated high purity water and also in base materials during a chloride transient representing BWR water condition below Action Level 1 of the EPRI Water Chemistry Guidelines according to the lectrical conductivity of the water but in the range of Action Level 2 according to the content of chlorides. Time based crack growth was also observed in one Russian type base material in oxygenated VVER water and in one Western type base material in oxygenated high purity BWR

  18. Study for identification of control rod drops in PWR reactors at any burnup step

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail:, E-mail:, E-mail:, E-mail: [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)


    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  19. Neutronic, steady-state, and transient analyses for the Kazakhstan VVR-K reactor with LEU fuel: ANL independent verification results

    Hanan, Nelson A. [Argonne National Lab. (ANL), Argonne, IL (United States); Garner, Patrick L. [Argonne National Lab. (ANL), Argonne, IL (United States)


    Calculations have been performed for steady state and postulated transients in the VVR-K reactor at the Institute of Nuclear Physics (INP), Kazakhstan. (The reactor designation in Cyrillic is BBP-K; transliterating characters to English gives VVR-K but translating words gives WWR-K.) These calculations have been performed at the request of staff of the INP who are performing similar calculations. The selection of the transients considered started during working meetings and email correspondence between Argonne National Laboratory (ANL) and INP staff. In the end the transient were defined by the INP staff. Calculations were performed for the fresh low-enriched uranium (LEU) core and for four subsequent cores as beryllium is added to maintain critically during the first 15 cycles. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process.

  20. Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

    Giacomino Bandini


    Full Text Available The development of a conceptual design of an industrial-scale transmutation facility (EFIT of several 100 MW thermal power based on accelerator-driven system (ADS is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS. In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios.

  1. Neural-net based unstable machine identification using individual energy functions. [Transient disturbances in power systems

    Djukanovic, M. (Institut Nikola Tesla, Belgrade (Yugoslavia)); Sobajic, D.J.; Yohhan Pao (Case Western Reserve Univ., Cleveland, OH (United States))


    The identification of the mode of instability plays an essential role in generating principal energy boundary hypersurfaces. We present a new method for unstable machine identification based on the use of supervised learning neural-net technology, and the adaptive pattern recognition concept. It is shown that using individual energy functions as pattern features, appropriately trained neural-nets can retrieve the reliable characterization of the transient process including critical clearing time parameter, mode of instability and energy margins. Generalization capabilities of the neural-net processing allow for these assessments to be made independently of load levels. The results obtained from computer simulations are presented using the New England power system, as an example. (author).

  2. Probabilistic margin evaluation on accidental transients for the ASTRID reactor project

    Marquès, Michel


    ASTRID is a technological demonstrator of Sodium cooled Fast Reactor (SFR) under development. The conceptual design studies are being conducted in accordance with the Generation IV reactor objectives, particularly in terms of improving safety. For the hypothetical events, belonging to the accidental category "severe accident prevention situations" having a very low frequency of occurrence, the safety demonstration is no more based on a deterministic demonstration with conservative assumptions on models and parameters but on a "Best-Estimate Plus Uncertainty" (BEPU) approach. This BEPU approach ispresented in this paper for an Unprotected Loss-of-Flow (ULOF) event. The Best-Estimate (BE) analysis of this ULOFt ransient is performed with the CATHARE2 code, which is the French reference system code for SFR applications. The objective of the BEPU analysis is twofold: first evaluate the safety margin to sodium boiling in taking into account the uncertainties on the input parameters of the CATHARE2 code (twenty-two uncertain input parameters have been identified, which can be classified into five groups: reactor power, accident management, pumps characteristics, reactivity coefficients, thermal parameters and head losses); secondly quantify the contribution of each input uncertainty to the overall uncertainty of the safety margins, in order to refocusing R&D efforts on the most influential factors. This paper focuses on the methodological aspects of the evaluation of the safety margin. At least for the preliminary phase of the project (conceptual design), a probabilistic criterion has been fixed in the context of this BEPU analysis; this criterion is the value of the margin to sodium boiling, which has a probability 95% to be exceeded, obtained with a confidence level of 95% (i.e. the M5,95percentile of the margin distribution). This paper presents two methods used to assess this percentile: the Wilks method and the Bootstrap method ; the effectiveness of the two methods

  3. Improved Recovery and Identification of Membrane Proteins from Rat Hepatic Cells using a Centrifugal Proteomic Reactor*

    Zhou, Hu; Wang, Fangjun; Wang, Yuwei; Ning, Zhibin; Hou, Weimin; Wright, Theodore G.; Sundaram, Meenakshi; Zhong, Shumei; Yao, Zemin; Figeys, Daniel


    Despite their importance in many biological processes, membrane proteins are underrepresented in proteomic analysis because of their poor solubility (hydrophobicity) and often low abundance. We describe a novel approach for the identification of plasma membrane proteins and intracellular microsomal proteins that combines membrane fractionation, a centrifugal proteomic reactor for streamlined protein extraction, protein digestion and fractionation by centrifugation, and high performance liquid chromatography-electrospray ionization-tandem MS. The performance of this approach was illustrated for the study of the proteome of ER and Golgi microsomal membranes in rat hepatic cells. The centrifugal proteomic reactor identified 945 plasma membrane proteins and 955 microsomal membrane proteins, of which 63 and 47% were predicted as bona fide membrane proteins, respectively. Among these proteins, >800 proteins were undetectable by the conventional in-gel digestion approach. The majority of the membrane proteins only identified by the centrifugal proteomic reactor were proteins with ≥2 transmembrane segments or proteins with high molecular mass (e.g. >150 kDa) and hydrophobicity. The improved proteomic reactor allowed the detection of a group of endocytic and/or signaling receptor proteins on the plasma membrane, as well as apolipoproteins and glycerolipid synthesis enzymes that play a role in the assembly and secretion of apolipoprotein B100-containing very low density lipoproteins. Thus, the centrifugal proteomic reactor offers a new analytical tool for structure and function studies of membrane proteins involved in lipid and lipoprotein metabolism. PMID:21749988

  4. Improved recovery and identification of membrane proteins from rat hepatic cells using a centrifugal proteomic reactor.

    Zhou, Hu; Wang, Fangjun; Wang, Yuwei; Ning, Zhibin; Hou, Weimin; Wright, Theodore G; Sundaram, Meenakshi; Zhong, Shumei; Yao, Zemin; Figeys, Daniel


    Despite their importance in many biological processes, membrane proteins are underrepresented in proteomic analysis because of their poor solubility (hydrophobicity) and often low abundance. We describe a novel approach for the identification of plasma membrane proteins and intracellular microsomal proteins that combines membrane fractionation, a centrifugal proteomic reactor for streamlined protein extraction, protein digestion and fractionation by centrifugation, and high performance liquid chromatography-electrospray ionization-tandem MS. The performance of this approach was illustrated for the study of the proteome of ER and Golgi microsomal membranes in rat hepatic cells. The centrifugal proteomic reactor identified 945 plasma membrane proteins and 955 microsomal membrane proteins, of which 63 and 47% were predicted as bona fide membrane proteins, respectively. Among these proteins, >800 proteins were undetectable by the conventional in-gel digestion approach. The majority of the membrane proteins only identified by the centrifugal proteomic reactor were proteins with ≥ 2 transmembrane segments or proteins with high molecular mass (e.g. >150 kDa) and hydrophobicity. The improved proteomic reactor allowed the detection of a group of endocytic and/or signaling receptor proteins on the plasma membrane, as well as apolipoproteins and glycerolipid synthesis enzymes that play a role in the assembly and secretion of apolipoprotein B100-containing very low density lipoproteins. Thus, the centrifugal proteomic reactor offers a new analytical tool for structure and function studies of membrane proteins involved in lipid and lipoprotein metabolism.

  5. Time-varying singular value decomposition for periodic transient identification in bearing fault diagnosis

    Zhang, Shangbin; Lu, Siliang; He, Qingbo; Kong, Fanrang


    For rotating machines, the defective faults of bearings generally are represented as periodic transient impulses in acquired signals. The extraction of transient features from signals has been a key issue for fault diagnosis. However, the background noise reduces identification performance of periodic faults in practice. This paper proposes a time-varying singular value decomposition (TSVD) method to enhance the identification of periodic faults. The proposed method is inspired by the sliding window method. By applying singular value decomposition (SVD) to the signal under a sliding window, we can obtain a time-varying singular value matrix (TSVM). Each column in the TSVM is occupied by the singular values of the corresponding sliding window, and each row represents the intrinsic structure of the raw signal, namely time-singular-value-sequence (TSVS). Theoretical and experimental analyses show that the frequency of TSVS is exactly twice that of the corresponding intrinsic structure. Moreover, the signal-to-noise ratio (SNR) of TSVS is improved significantly in comparison with the raw signal. The proposed method takes advantages of the TSVS in noise suppression and feature extraction to enhance fault frequency for diagnosis. The effectiveness of the TSVD is verified by means of simulation studies and applications to diagnosis of bearing faults. Results indicate that the proposed method is superior to traditional methods for bearing fault diagnosis.

  6. Development of a digital card to simulate period transients in research reactors; Desenvolvimento de um cartao digital para simulacao da variacao do periodo em reatores de pesquisa

    Masotti, Paulo Henrique Ferraz


    This work presents the development of a card to be used in a 'slot' of a micro-computer for evaluation of a nuclear channel used to monitor the start up of nuclear reactors. The results of the bench tests showed good linearity and 2% error deviation in the entire range of operation. Fields tests, performed with the start up channel of IEA-R1 research reactor showed that the card is an excellent device to verify the performance of the channel during steady state, and transient conditions. (author)


    C. Sayer


    Full Text Available Dynamic mathematical models are developed to simulate styrene emulsion polymerization reactions carried out in pulsed tubular reactors. Two different modeling approaches, the tanks-in-series model and the axial dispersion model, are compared. The models developed were validated with experimental data from the literature and used to study the dynamics during transient periods, e.g., the start-up of the reactor and the response to disturbances. The effect of the Peclet number on process variables such as conversion and particle concentration was also verified.

  8. Domain decomposition parallel computing for transient two-phase flow of nuclear reactors

    Lee, Jae Ryong; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of); Choi, Hyoung Gwon [Seoul National University, Seoul (Korea, Republic of)


    KAERI (Korea Atomic Energy Research Institute) has been developing a multi-dimensional two-phase flow code named CUPID for multi-physics and multi-scale thermal hydraulics analysis of Light water reactors (LWRs). The CUPID code has been validated against a set of conceptual problems and experimental data. In this work, the CUPID code has been parallelized based on the domain decomposition method with Message passing interface (MPI) library. For domain decomposition, the CUPID code provides both manual and automatic methods with METIS library. For the effective memory management, the Compressed sparse row (CSR) format is adopted, which is one of the methods to represent the sparse asymmetric matrix. CSR format saves only non-zero value and its position (row and column). By performing the verification for the fundamental problem set, the parallelization of the CUPID has been successfully confirmed. Since the scalability of a parallel simulation is generally known to be better for fine mesh system, three different scales of mesh system are considered: 40000 meshes for coarse mesh system, 320000 meshes for mid-size mesh system, and 2560000 meshes for fine mesh system. In the given geometry, both single- and two-phase calculations were conducted. In addition, two types of preconditioners for a matrix solver were compared: Diagonal and incomplete LU preconditioner. In terms of enhancement of the parallel performance, the OpenMP and MPI hybrid parallel computing for a pressure solver was examined. It is revealed that the scalability of hybrid calculation was enhanced for the multi-core parallel computation.

  9. Adaptation and implementation of the TRACE code for transient analysis on designs of cooled lead fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Lazaro, A.; Ammirabile, L.; Martorell, S.


    The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)

  10. Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

    Kontogeorgakos, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Derstine, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Bauer, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)


    The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO2 particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.

  11. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report; WTZ mit Russland. Transientenanalysen fuer wassergekuehlte Kernreaktoren. Abschlussbericht

    Rohde, Ulrich [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Kozmenkov, Yaroslav [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Institute of Physics and Power Engineering, Obninsk (Russian Federation); Pivovarov, Valeri; Matveev, Yurij [Institute of Physics and Power Engineering, Obninsk (Russian Federation)


    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

  12. Transient loads identification for a standoff metallic thermal protection system panel.

    Hundhausen, R. J. (Roy Jason); Adams, Douglas E.; Derriso, Mark; Kukuchek, Paul; Alloway, Richard


    Standoff thermal protection system (TPS) panels are critical structural components in future aerospace vehicles because they protect the vehicle from the hostile environment encountered during space launch and reentry. Consequently, the panels are exposed to a variety of loads including high temperature thermal stresses, thermal shock, acoustic pressure, and foreign object impacts. Transient impacts are especially detrimental because they can cause immediate and severe degradation of the panel in the form of, for example, debonding and buckling of the face sheet, cracking of the fasteners, or deformation of the standoffs. Loads identification methods for determining the magnitude and location of impact loads provide an indication of TPS components that may be more susceptible to failure. Furthermore, a historical database of impact loads encountered can be retained for use in the development of statistical models that relate impact loading to panel life. In this work, simulated inservice transient loads are identified experimentally using two methods: a physics-based approach and an inverse Frequency Response Function (FRF) approach. It is shown that by applying the inverse FRF method, the location and magnitude of these simulated impacts can be identified with a high degree of accuracy. The identified force levels vary significantly with impact location due to the differences in panel deformation at the impact site indicating that resultant damage due to impacts would vary with location as well.

  13. A Study on Development of Variable High Pressurizer Pressure Trip Function to Mitigate System Peak Pressure during Transients for Pressurized Water Reactors

    Kim, Ung Soo; Park, Min Soo; Huh, Jae Young; Lee, Gyu Cheon [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)


    According to intensified regulation environment such as separate safety analysis for the reactor coolant system (RCS) and the main steam system peak pressure, strict consideration of a control system malfunction as a single failure for the safety analysis and so on, the safety margin with respect to system pressure of pressurized water reactors (PWRs) has been decreased. Also, the possibility for that the main steam system pressure may violate the acceptance criteria during the LOCV event has been raised and relevant design modifications for the main steam safety valve (MSSV) have ever been performed as a solution. In order to overcome this problem, in this work, the variable high pressurizer pressure trip (VHPPT) function has been developed and a feasibility study on the application of this trip function has been performed. The VHPPT function has been devised to trip the reactor beforehand when a sharply pressurizing transient such as the LOCV occurs and to cutoff system pressure increase, resulting in reducing the system peak pressure. In this work, the VHPPT function has been suggested and developed to trip the reactor beforehand and to cutoff system pressure increase mitigating the system peak pressure of PWRs when a sharply pressurizing transient like the LOCV occurs. The VHPPT function uses the rate-limited variable setpoint and includes the existing HPPT function.

  14. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)


    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  15. Transient analysis for damping identification in rotating composite beams with integral damping layers

    Smith, Clifford B.; Wereley, Norman M.


    The first objective of this paper is to evaluate the performance of damping identification algorithms. The second objective is to determine the feasibility of damping augmentation in rotating composite beams via passive constrained layer damping (PCLD). Damping identification schemes were applied to four rectangular cross-section laminated composite beams with cocured integral damping layers over the span of the beam. The cocured beam consisted of a twenty-ply balanced and symmetric cross-ply Gr/Ep composite host structure, a top and bottom damping layer of viscoelastic material (VEM), and a 2-ply Gr/Ep constraining layer sandwiching the viscoelastic material to the host structure. Four VEM thicknesses were considered: 0, 5, 10, and 15 mils. The cantilevered beams were tested at rotational speeds ranging from 0 to 900 RPM in a vacuum chamber. Excitation in bending was provided using piezo actuators, and the bending response was measured using full strain gauge bridges. Transient data were analysed using logarithmic decrement, a Hilbert transform technique, and an FFT- based moving block analysis. When compared to the beam with no VEM, a 19.2% volume fraction (15 mil layer) of viscoelastic in the beam produced a 400% increase in damping ratio in the non-rotating case, while at 900 RPM, the damping ratio increased only 360%. Overall structural damping was reduced as a function of RPM, due to centrifugal stiffening.

  16. Identification of the autotrophic denitrifying community in nitrate removal reactors by DNA-stable isotope probing.

    Xing, Wei; Li, Jinlong; Cong, Yuan; Gao, Wei; Jia, Zhongjun; Li, Desheng


    Autotrophic denitrification has attracted increasing attention for wastewater with insufficient organic carbon sources. Nevertheless, in situ identification of autotrophic denitrifying communities in reactors remains challenging. Here, a process combining micro-electrolysis and autotrophic denitrification with high nitrate removal efficiency was presented. Two batch reactors were fed organic-free nitrate influent, with H(13)CO3(-) and H(12)CO3(-) as inorganic carbon sources. DNA-based stable-isotope probing (DNA-SIP) was used to obtain molecular evidence for autotrophic denitrifying communities. The results showed that the nirS gene was strongly labeled by H(13)CO3(-), demonstrating that the inorganic carbon source was assimilated by autotrophic denitrifiers. High-throughput sequencing and clone library analysis identified Thiobacillus-like bacteria as the most dominant autotrophic denitrifiers. However, 88% of nirS genes cloned from the (13)C-labeled "heavy" DNA fraction showed low similarity with all culturable denitrifiers. These findings provided functional and taxonomical identification of autotrophic denitrifying communities, facilitating application of autotrophic denitrification process for wastewater treatment.

  17. Real-Time Gas Identification by Analyzing the Transient Response of Capillary-Attached Conductive Gas Sensor

    Behzad Bahraminejad


    Full Text Available In this study, the ability of the Capillary-attached conductive gas sensor (CGS in real-time gas identification was investigated. The structure of the prototype fabricated CGS is presented. Portions were selected from the beginning of the CGS transient response including the first 11 samples to the first 100 samples. Different feature extraction and classification methods were applied on the selected portions. Validation of methods was evaluated to study the ability of an early portion of the CGS transient response in target gas (TG identification. Experimental results proved that applying extracted features from an early part of the CGS transient response along with a classifier can distinguish short-chain alcohols from each other perfectly. Decreasing time of exposition in the interaction between target gas and sensing element improved the reliability of the sensor. Classification rate was also improved and time of identification was decreased. Moreover, the results indicated the optimum interval of the early transient response of the CGS for selecting portions to achieve the best classification rates.

  18. Reactor

    Evans, Robert M.


    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  19. 基于模糊熵的核电站瞬态识别方法%Fuzzy Entropy Based Transient Identification in Nuclear Power Plant

    常远; 郝轶; 黄晓津; 李春文; 梁记兴; 刘景源


    For safe and economical operation of nuclear power plants (NPPs) ,the occur-ring anomalies should be promptly and correctly identified .In this paper ,the transients were identified by processing time series of critical variables .First ,based on its ability to measure the complexity of time series ,the fuzzy entropy (FuzzyEn) was used to determine whether the system was in normal state . Then cross fuzzy entropy was employed for classifying the occurring transients , using its ability to characterize the similarity between two time series .The feasibility and effectiveness were verified by simulator data of pebble-bed modular high temperature gas-cooled reactor nuclear power plant (HTR-PM ) .It is demonstrated that the proposed method is effective for transient identification and it dispenses with a complex training phase .%为保障核电站安全经济运行,需及时准确地识别核电站出现的异常。本文通过处理关键变量的时间序列数据,对瞬态过程进行识别:利用模糊熵度量时间序列复杂度的能力,判断系统是否处于正常状态;进而利用互模糊熵度量两时间序列相似度的能力,对出现的瞬态进行类型识别。利用模块式高温气冷堆核电站仿真机的数据验证了本方法的可行性和有效性,结果表明本文方法可有效进行瞬态识别,且不需复杂的训练过程。

  20. Assessment of SFR reactor safety issues: Part II: Analysis results of ULOF transients imposed on a variety of different innovative core designs with SAS-SFR

    Kruessmann, R., E-mail: [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Ponomarev, A.; Pfrang, W.; Struwe, D. [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Champigny, J.; Carluec, B. [AREVA, 10, rue J. Récamier, 69456 Lyon Cedex 06 (France); Schmitt, D.; Verwaerde, D. [EDF R& D, 1 avenue du général de Gaulle, 92140 Clamart (France)


    Highlights: • Comparison of different core designs for a sodium-cooled fast reactor. • Safety assessment with the code system SAS-SFR. • Unprotected Loss of Flow (ULOF) scenario. • Sodium boiling and core melting cannot be avoided. • A net negative Na void effect provides more grace time prior to local SA destruction. - Abstract: In the framework of cooperation agreements between KIT-INR and AREVA SAS NP as well as between KIT-INR and EDF R&D in the years 2008–2013, the evaluation of severe transient behavior in sodium-cooled fast reactors (SFRs) was investigated. In Part I of this contribution, the efficiency of newly conceived prevention and mitigation measures was investigated for unprotected loss-of-flow (ULOF), unprotected loss-of-heat-sink (ULOHS) and the unprotected transient-overpower (UTOP) transients. In this second part, consequence analyses were performed for the initiation phase of different unprotected loss-of-flow (ULOF) scenarios imposed on a variety of different core design options of SFRs. The code system SAS-SFR was used for this purpose. Results of analyses for cases postulating unavailability of prevention measures as shut-down systems, passive and/or active additional devices show that entering into an energetic power excursion as a consequence of the initiation phase of a ULOF cannot be avoided for those core designs with a cumulative void reactivity feedback larger than zero. However, even for core designs aiming at values of the void reactivity less than zero it is difficult to find system design characteristics which prevent the transient entering into partial core destruction. Further studies of the transient core and system behavior would require codes dedicated to specific aspects of transition phase analyses and of in-vessel material relocation analyses.

  1. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    Lazaro, A., E-mail: [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)


    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

  2. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.


    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  3. Bifurcation analysis of the simplified models of boiling water reactor and identification of global stability boundary

    Pandey, Vikas; Singh, Suneet, E-mail:


    Highlights: • Non-linear stability analysis of nuclear reactor is carried out. • Global and local stability boundaries are drawn in the parameter space. • Globally stable, bi-stable, and unstable regions have been demarcated. • The identification of the regions is verified by numerical simulations. - Abstract: Nonlinear stability study of the neutron coupled thermal hydraulics instability has been carried out by several researchers for boiling water reactors (BWRs). The focus of these studies has been to identify subcritical and supercritical Hopf bifurcations. Supercritical Hopf bifurcation are soft or safe due to the fact that stable limit cycles arise in linearly unstable region; linear and global stability boundaries are same for this bifurcation. It is well known that the subcritical bifurcations can be considered as hard or dangerous due to the fact that unstable limit cycles (nonlinear phenomena) exist in the (linearly) stable region. The linear stability leads to a stable equilibrium in such regions, only for infinitesimally small perturbations. However, finite perturbations lead to instability due to the presence of unstable limit cycles. Therefore, it is evident that the linear stability analysis is not sufficient to understand the exact stability characteristics of BWRs. However, the effect of these bifurcations on the stability boundaries has been rarely discussed. In the present work, the identification of global stability boundary is demonstrated using simplified models. Here, five different models with different thermal hydraulics feedback have been investigated. In comparison to the earlier works, current models also include the impact of adding the rate of change in temperature on void reactivity as well as effect of void reactivity on rate of change of temperature. Using the bifurcation analysis of these models the globally stable region in the parameter space has been identified. The globally stable region has only stable solutions and

  4. Improved Generalized Predictive Control Algorithm with Offline and Online Identification and Its Application to Fixed Bed Reactor

    余世明; 王海清


    An improved generalized predictive control algorithm is presented in this paper by incorporating offline identification into onlie identification.Unlike the existing generalized predictive control algorithms.the proposed approach divides parameters of a predictive model into the time invariant and time-varying ones,which are treated respectively by offline and onlie identification algorithms.Therefore,both the reliability and accuracy of the predictive model are improved,Two simulation examples of control of a fixed bed reactor show that this new algorithm is not only reliable and stable in the case of uncertainties and abnormal distrubances,but also adaptable to slow time varying processes.

  5. RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

    Talley, Darren G.


    This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code shows good agreement between simulation and actual ACRR operations.

  6. Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

    Hu, Rui


    Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developed at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.

  7. Assessment of RELAP5 point kinetic model against reactivity insertion transient in the IAEA 10 MW MTR research reactor

    Hamidouche, T., E-mail: t.hamidouche@crna.d [Division de l' Environnement, de la Surete et des Dechets Radioactifs, Centre de Recherche Nucleaire d' Alger, 02 Boulevard Frantz Fanon, BP 399 Alger RP (Algeria); Bousbia-Salah, A. [DIMNP - University of Pisa, Via Diotisalvi 02, 56126 Pisa (Italy)


    The current study emphasizes an aspect related to the assessment of a model embedded in a computer code. The study concerns more particularly the point neutron kinetics model of the RELAP5/Mod3 code which is worldwide used. The model is assessed against positive reactivity insertion transient taking into account calculations involving thermal-hydraulic feedback as well as transients with no feedback effects. It was concluded that the RELAP5 point kinetics model provides unphysical power evolution trends due most probably to a bug during the programming process.

  8. Large-break loss-of-coolant accident phenomena identification and ranking table (PIRT) for the advanced Candu reactor

    Popov, N.; Snell, V.G.; Sills, H.E.; Langman, V.J.; Boyack, B. [Atomic Energy of Canada Ltd (Canada)


    The Advanced Candu Reactor (ACR) is an evolutionary advancement of the current Candu-6 reactor, aimed at producing electrical power for a capital cost and unit-energy cost significantly less than that of current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by heavy water moderator, as with all Candu reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper is focused on the double-ended guillotine critical inlet header break (CRIHB) loss-of-coolant accident (LOCA) in an ACR reactor, which is considered as a large break LOCA. Large Break LOCA in water-cooled reactors has been used historically as a design basis event by regulators, and it has attracted a very large share of safety analysis and regulatory review. The LBLOCA event covers a wide range of system behaviours and fundamental phenomena. The Phenomena Identification and Ranking Table (PIRT) for LBLOCA therefore provides a good understanding of many of the safety characteristics of the ACR design. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the LOCA phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the final PIRT summary table. (authors)

  9. Effects of transient and non-uniform distribution of heat flux on intensity of heat transfer and burnout conditions in the channels of nuclear reactors

    Vitaly Osmachkin [Russian Research Center ' Kurchatov Institute' 1, Kurchatov sq, Moscow 123182 (Russian Federation)


    Full text of publication follows: The influence of power transient, changes of flow rate, inlet temperatures or pressure in cores of nuclear reactors on heat transfer and burnout conditions in channels depend on rate of such violations. Non-uniform distribution of the heat flux is also important factor for heat transfer and development of crisis phenomenon. Such effects may be significant for NPPs safety. But they have not yet generally accepted interpretation. Steady state approach is often recommended for use in calculations. In the paper a review of experimental observed so-called non-equilibrium effects is presented. The effects of space and time factors are displaying due delay in reformation turbulence intensity, velocity, temperatures or void fraction profiles, water film flow on the surface of heated channels. For estimation of such effect different methods are used. Modern computer codes based on two or three fluids approaches are considered as most effective. But simple and clear correlations may light up the mechanics of effects on heat transfer and improve general understanding of scale and significance of the transient events. In the paper the simplified methods for assessment the influence of lags in the development of distributions of parameters of flow, the relaxation of temporal or space violations are considered. They are compared with more sophisticated approaches. Velocities of disturbance fronts moving along the channels are discussed also. (author)

  10. One-dimensional modeling of radial heat removal during depressurized heatup transients in modular pebble-bed and prismatic high temperature gas-cooled reactors

    Savage, M.G.


    A one-dimensional computational model was developed to evaluate the heat removal capabilities of both prismatic-core and pebble-bed modular HTGRs during depressurized heatup transients. A correlation was incorporated to calculate the temperature- and neutron-fluence-dependent thermal conductivity of graphite. The modified Zehner-Schluender model was used to determine the effective thermal conductivity of a pebble bed, accounting for both conduction and radiation. Studies were performed for prismatic-core and pebble-bed modular HTGRs, and the results were compared to analyses performed by GA and GR, respectively. For the particular modular reactor design studied, the prismatic HTGR peak temperature was 2152.2/sup 0/C at 38 hours following the transient initiation, and the pebble-bed peak temperature was 1647.8/sup 0/C at 26 hours. These results compared favorably with those of GA and GE, with only slight differences caused by neglecting axial heat transfer in a one-dimensional radial model. This study found that the magnitude of the initial power density had a greater effect on the temperature excursion than did the initial temperature.

  11. Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

    Salko, Robert K.

    COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for

  12. A Derivation of Source-based Kinetics Equation with Time Dependent Fission Kernel for Reactor Transient Analyses

    Kim, Song Hyun; Woo, Myeong Hyun; Shin, Chang Ho [Hanyang University, Seoul (Korea, Republic of); Pyeon, Cheol Ho [Kyoto University, Osaka (Japan)


    In this study, a new balance equation to overcome the problems generated by the previous methods is proposed using source-based balance equation. And then, a simple problem is analyzed with the proposed method. In this study, a source-based balance equation with the time dependent fission kernel was derived to simplify the kinetics equation. To analyze the partial variations of reactor characteristics, two representative methods were introduced in previous studies; (1) quasi-statics method and (2) multipoint technique. The main idea of quasistatics method is to use a low-order approximation for large integration times. To realize the quasi-statics method, first, time dependent flux is separated into the shape and amplitude functions, and shape function is calculated. It is noted that the method has a good accuracy; however, it can be expensive as a calculation cost aspect because the shape function should be fully recalculated to obtain accurate results. To improve the calculation efficiency, multipoint method was proposed. The multipoint method is based on the classic kinetics equation with using Green's function to analyze the flight probability from region r' to r. Those previous methods have been used to analyze the reactor kinetics analysis; however, the previous methods can have some limitations. First, three group variables (r{sub g}, E{sub g}, t{sub g}) should be considered to solve the time dependent balance equation. This leads a big limitation to apply large system problem with good accuracy. Second, the energy group neutrons should be used to analyze reactor kinetics problems. In time dependent problem, neutron energy distribution can be changed at different time. It can affect the change of the group cross section; therefore, it can lead the accuracy problem. Third, the neutrons in a space-time region continually affect the other space-time regions; however, it is not properly considered in the previous method. Using birth history of the

  13. Transient Hydraulic Characteristics of Nuclear Reactor Coolant Pump in Variable Flow Transient Process%核主泵变流量过渡过程瞬态水力特性研究

    王秀礼; 袁寿其; 朱荣生; 付强; 俞志君


    For the study on the transient hydraulic characteristics and internal flow mechanism of the nuclear reactor coolant pump in the transient process from design operation conditions to off-design conditions,the variable flow transient characteristics of centrifugal pump impeller passageway were simulated by using CFX software.The results show that during the variable flow transition,the distribution of pressure pulsation of the nuclear reactor coolant pump along the circumference direction is nonuniform.The pressure pulsation trends to rise gradually to reach the maximum value and then fall,basically following a sine-shape changing law.The times of transient pressure fluctuation change are equal to the times of rotor-stator interference between the vane and the guide vane.The closer monitoring point to the intersection surface between the vane and the guide blade is,the greater the pressure fluctuation is.Because of the attack angle,the speed of the impeller passageway first falls and then rises.The guide vane not only transfers the kinetic energy to pressure energy,but also effectively reduces the pressure pulsation amplitude.During the transition to small flow,flow reducing causes the secondary backflow to occur near the outlet of impeller and in turn leads the amplitude of flow velocity variation in the flow channel of impeller to increase with flow decrease.%为研究核主泵从设计工况向非设计工况过渡过程的瞬态水力特性及内部流动机理,应用计算流体力学软件CFX对核主泵叶轮流道内的变流量瞬态流动特性进行数值模拟计算.研究结果表明:变流量过渡时,核主泵的压力脉动沿圆周方向分布并不均匀,其变化趋势是逐渐上升到最大值后又降低,基本呈正弦变化规律,瞬态压力波动变化次数等于叶片与导叶片数之间的动静干涉次数,监测点越靠近叶片与导叶交界面,压力波动越大;由于冲角的存在造成叶轮流道内的速度呈先下降后

  14. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    S.T. Revankar; W. Zhou; Gavin Henderson


    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  15. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)


    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  16. Vibration system identification of Paks and Kozloduy reactor buildings on the basis of the blast test results

    Varpasuo, P


    System identification allows to build mathematical models of a dynamic system based on measured data. System identification is carried out by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The aim of this study is to investigate and model the behavior of complex vibratory systems on the basis of measured excitation and response. The first part of the study describes the theory used in the analysis and the software tools used in the analysis. The second part of the study describes the investigation and modeling of the response of single degree of freedom oscillator excited by sinusoidal and blast excitation. In the third part of the study the system identification of the Kozloduy NPP unit 5 reactor building and Paks NPP unit 1 reactor building is studied and the models are estimated using the method of segmentation of excitation and response. System identification is carried out using MATLAB software by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The types of models used for the were: l) ARX models; 2) ARMAX model; 3) Output-Error (OE) models; 4) Box-Jenkins (BJ) models; 5) State-space models. The model coefficients for different models were calculated using the least-squares and maximum likelihood estimation methods available in MATLAB system identification toolbox. Excitation was in both Paks and Kozloduy case the measured free-field excitation and responses were the vibration responses of the building on the foundation slab level and top of the building. By examining the established models the frequency characteristics of vibration systems were determined with 95 % accuracy and the amplitude response with 80 % accuracy. In case of the steady state response of sinusoidally excited single dof oscillator the modelling gave almost exact results. But in the case of the blast response of the reactor building the obtaining of the

  17. Using reactor network for global identification based on residence time distribution theory

    Hocine, S.; Pibouleau, L.; Azzaro-Pantel, C.; Domenech, S. [Laboratoire de Genie Chimique - UMR 5503 CNRS/ INPT ENSIACET, 31 - Toulouse (France)


    In the ventilation systems, the control of transfer contaminants is one of the principal problems during the design and control phases. The installation of a suitable ventilation system for the control of contaminant transfer is essential in industry, because it makes it possible to detect and to prevent chemical and radiological risks. Research on air distribution in ventilated rooms traditionally involves full-scale experiments, scale -model experiments and application of the computational fluid dynamics (C.F.D.) tools. Most of the time, particularly in our case of large and cluttered enclosures, the predictive approach based on C.F.D. codes can not be used. The solution retained here is the establishment of a model based on the well known residence time distribution. This model is widely used in chemical engineering to treat non-ideal flows. The proposed method is based on the experimental determination of the residence time distribution curve, generally obtained through the response of the system to tracer release. A superstructure involving the set of all the possible solutions corresponding to the physical reactor is then defined, and the model will be selected from this superstructure according to its simulated response. The superstructure is identified as a combination of elementary systems, representing ideal flow patterns, as perfect mixed flows, plug flows, continuous stirred tank reactors, etc. The selected model is derived from the comparison between the simulated response to a stimulus, and the experimental response. The structure and parameters of the model are simultaneously optimized in order to fit the experimental curve with a minimal number of elementary units, constituting a key point for future control purposes of the process. This problem is a dynamic M.I.N.L.P. (Mixed Integer Non Linear Programming) problem with bilinear equality constraints. Generally, these constraints lead to numerical difficulties for reaching an optimum solution (even a

  18. Thermo-hydraulic experiments for the development of a system for identification and classification of transients (SICT)

    Aronne, Ivan Dionysio; Palmieri, Elcio Tadeu; Navarro, Moyses Alberto; Azevedo, Carlos Vicente Goulart de [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mails:;;;; Baptista Filho, Benedito Dias [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail:


    The safety of nuclear power plants has always been a concern when this technology is considered as an option for power generation. As a contribution to the improvement of its safety performance, a System for Identification and Classification of Transients (SICT) is being developed. This system is based in neural networks particularly Self-Organizing Maps and has as goal to assist the operation of nuclear plants. The development of this system has several phases and one of them is the demonstration of the capability of SICT to respond on time for transients being able to warn the operator. This demonstration will be achieved using experiments in a thermo-hydraulic facility - CT1 - in CDTN, having the SICT coupled to it. Before coupling the SICT with CT1 instrumentation it has to be trained to recognize different operational states possible in the installation. This training is performed using results of simulation of experiments with the RELAP5 code, in the same way as the SICT for the Nuclear Power Plant shall be preliminarily trained using results of simulations. This paper presents the description of such facility, with the coupled SICT, the carried out experiments, as well as, their simulations with RELAP5 and the overall performance of SICT. (author)

  19. Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

    Garner, P. L.; Hanan, N. A. (Nuclear Engineering Division)


    The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

  20. Identification of Prognostic Risk Factors for Transient and Persistent Lymphedema after Multimodal Treatment for Breast Cancer.

    Kim, Myungsoo; Shin, Kyung Hwan; Jung, So-Youn; Lee, Seeyoun; Kang, Han-Sung; Lee, Eun Sook; Chung, Seung Hyun; Kim, Yeon-Joo; Kim, Tae Hyun; Cho, Kwan Ho


    The purpose of this study is to identify risk factors for transient lymphedema (TLE) and persistent lymphedema (PLE) following treatment for breast cancer. A total of 1,073 patients who underwent curative breast surgery were analyzed. TLE was defined as one episode of arm swelling that had resolved spontaneously by the next follow-up; arm swelling that persisted over two consecutive examinations was considered PLE. At a median follow-up period of 5.1 years, 370 cases of lymphedema were reported, including 120 TLE (11.2%) and 250 PLE (23.3%). Initial grade 1 swelling was observed in 351 patients, of which 120 were limited to TLE (34%), while the other 231 progressed to PLE (66%). All initial swelling observed in TLE patients was classified as grade 1. In multivariate analysis, chemotherapy with taxane and supraclavicular radiation therapy (SCRT) were associated with development of TLE, whereas SCRT, stage III cancer and chemotherapy with taxane were identified as risk factors for PLE (p PLE based on the number of risk factors were 7:1 (no factor), 1:1 (one factor), 1:2 (two factors), and 1:3 (three factors). One-third of initial swelling events were transient, whereas the other two-thirds of patients experienced PLE. Estimation of TLE and PLE based on known treatment factors could facilitate prediction of this life-long complication.

  1. Identification of a novel group of bacteria in sludge from a deteriorated biological phosphorus removal reactor

    Nielsen, Alex Toftgaard; Liu, Wen-Tso; Filipe, Carlos


    The microbial diversity of a deteriorated biological phosphorus removal reactor was investigated by methods not requiring direct cultivation. The reactor was fed with media containing acetate and high levels of phosphate (P/C weight ratio, 8:100) but failed to completely remove phosphate in the e...... obtained by the PCR-based DGGE method. Further, based on electron microscopy and standard staining microscopic analysis, this novel group was able to accumulate granule inclusions, possibly consisting of polyhydroxyalkanoate, inside the cells....

  2. Identification of relaxation parameter of a physical model of vein from fluid transient experiment

    Hromádka David


    Full Text Available This paper presents a new fluid transient inflation experiment applied on a physical model of vein (short latex tube, 5mm diameter. Aim of experiments is assessment of wall viscous behaviour from attenuated pulsation of the tested sample. Experimental data obtained from dynamic test are compared with numerical simulation and a viscoelastic parameter of Haslach constitutive model is identified. It is verified that the viscoelasticity of wall has a greater impact to the damping of pulsation than the viscosity of water filling the sample and the attached capillary. Volume of sample depends on internal pressure measured by a pressure transducer. The maximum dissipation constitutive model of viscoelastic wall sample was employed for description of viscoelastic behaviour. Frequency of natural oscillation of pressure is determined by inertia of water column within the tested sample and attached capillary and by the tested specimen stiffness. The pressure pulsations are initiated by a sudden pressure drop at water surface.

  3. A novel application of wavelet based SVM to transient phenomena identification of power transformers

    Jazebi, S., E-mail: [Department of Electrical Engineering, Amirkabir University of Technology, Tehran (Iran, Islamic Republic of); Vahidi, B., E-mail: [Department of Electrical Engineering, Amirkabir University of Technology, Tehran (Iran, Islamic Republic of); Jannati, M., E-mail: [Department of Electrical Engineering, Amirkabir University of Technology, Tehran (Iran, Islamic Republic of)


    A novel differential protection approach is introduced in the present paper. The proposed scheme is a combination of Support Vector Machine (SVM) and wavelet transform theories. Two common transients such as magnetizing inrush current and internal fault are considered. A new wavelet feature is extracted which reduces the computational cost and enhances the discrimination accuracy of SVM. Particle swarm optimization technique (PSO) has been applied to tune SVM parameters. The suitable performance of this method is demonstrated by simulation of different faults and switching conditions on a power transformer in PSCAD/EMTDC software. The method has the advantages of high accuracy and low computational burden (less than a quarter of a cycle). The other advantage is that the method is not dependent on a specific threshold. Sympathetic and recovery inrush currents also have been simulated and investigated. Results show that the proposed method could remain stable even in noisy environments.

  4. Time of flight mass spectrometry for quantitative data analysis in fast transient studies using a Temporal Analysis of Products (TAP) reactor.

    Goguet, Alexandre; Hardacre, Christopher; Maguire, Noleen; Morgan, Kevin; Shekhtman, Sergiy O; Thompson, Steve P


    A Time of flight (ToF) mass spectrometer suitable in terms of sensitivity, detector response and time resolution, for application in fast transient Temporal Analysis of Products (TAP) kinetic catalyst characterization is reported. Technical difficulties associated with such application as well as the solutions implemented in terms of adaptations of the ToF apparatus are discussed. The performance of the ToF was validated and the full linearity of the specific detector over the full dynamic range was explored in order to ensure its applicability for the TAP application. The reported TAP-ToF setup is the first system that achieves the high level of sensitivity allowing monitoring of the full 0-200 AMU range simultaneously with sub-millisecond time resolution. In this new setup, the high sensitivity allows the use of low intensity pulses ensuring that transport through the reactor occurs in the Knudsen diffusion regime and that the data can, therefore, be fully analysed using the reported theoretical TAP models and data processing.

  5. Identification of a cis-regulatory element by transient analysis of co-ordinately regulated genes

    Allan Andrew C


    Full Text Available Abstract Background Transcription factors (TFs co-ordinately regulate target genes that are dispersed throughout the genome. This co-ordinate regulation is achieved, in part, through the interaction of transcription factors with conserved cis-regulatory motifs that are in close proximity to the target genes. While much is known about the families of transcription factors that regulate gene expression in plants, there are few well characterised cis-regulatory motifs. In Arabidopsis, over-expression of the MYB transcription factor PAP1 (PRODUCTION OF ANTHOCYANIN PIGMENT 1 leads to transgenic plants with elevated anthocyanin levels due to the co-ordinated up-regulation of genes in the anthocyanin biosynthetic pathway. In addition to the anthocyanin biosynthetic genes, there are a number of un-associated genes that also change in expression level. This may be a direct or indirect consequence of the over-expression of PAP1. Results Oligo array analysis of PAP1 over-expression Arabidopsis plants identified genes co-ordinately up-regulated in response to the elevated expression of this transcription factor. Transient assays on the promoter regions of 33 of these up-regulated genes identified eight promoter fragments that were transactivated by PAP1. Bioinformatic analysis on these promoters revealed a common cis-regulatory motif that we showed is required for PAP1 dependent transactivation. Conclusion Co-ordinated gene regulation by individual transcription factors is a complex collection of both direct and indirect effects. Transient transactivation assays provide a rapid method to identify direct target genes from indirect target genes. Bioinformatic analysis of the promoters of these direct target genes is able to locate motifs that are common to this sub-set of promoters, which is impossible to identify with the larger set of direct and indirect target genes. While this type of analysis does not prove a direct interaction between protein and DNA

  6. Identification of Selected Areas to Support Federal Clean Energy Goals Using Small Modular Reactors

    Belles, Randy [ORNL; Mays, Gary T [ORNL; Omitaomu, Olufemi A [ORNL; Poore III, Willis P [ORNL


    This analysis identifies candidate locations, in a broad sense, where there are high concentrations of federal government agency use of electricity, which are also suitable areas for near-term SMRs. Near-term SMRs are based on light-water reactor (LWR) technology with compact design features that are expected to offer a host of safety, siting, construction, and economic benefits. These smaller plants are ideally suited for small electric grids and for locations that cannot support large reactors, thus providing utilities or governement entities with the flexibility to scale power production as demand changes by adding additional power by deploying more modules or reactors in phases. This research project is aimed at providing methodologies, information, and insights to assist the federal government in meeting federal clean energy goals.

  7. Identification of a novel group of bacteria in sludge from a deteriorated biological phosphorus removal reactor

    Nielsen, Alex Toftgaard; Liu, Wen-Tso; Filipe, Carlos


    The microbial diversity of a deteriorated biological phosphorus removal reactor was investigated by methods not requiring direct cultivation. The reactor was fed with media containing acetate and high levels of phosphate (P/C weight ratio, 8:100) but failed to completely remove phosphate...... in the effluent and showed very limited biological phosphorus removal activity. Denaturing gradient gel electrophoresis (DGGE) of PCR-amplified 16S ribosomal DNA was used to investigate the bacterial diversity. Up to 11 DGGE bands representing at least 11 different sequence types were observed; DNA from the 6...

  8. Neutron activation analysis at the Livermore pool-type reactor for the environmental research program. [Identification of trace element contaminants

    Ragaini, R.C.; Heft, R.E.; Garvis, D.


    Instrumental neutron activation analysis is a technique of trace analysis using measurements of radioactivity induced in the sample by exposure to a source of neutrons. The induced activity is measured by the emitted gamma radiation. Each gamma emitter can then be identified by the energy of the photopeaks produced as the nuclide decays and by the half-life of the neutron-induced activity. A complex computer program GAMANAL has been used to accomplish the major tasks of nuclide identification and quantification. The nuclide data output from GAMANAL is processed by a second computer code NADAC, which develops elemental abundance data from disintegration rates observed. The methods are those employed at the Livermore Pool-Type Reactor in support of the environmental research trace element analysis program. Among the procedures described and discussed are sample preparation, irradiation, analysis, and application of the technique.

  9. Using a combination of weighting factor method and imperialist competitive algorithm to improve speed and enhance process of reloading pattern optimization of VVER-1000 reactors in transient cycles

    Rahmani, Yashar, E-mail: [Department of Physics, Faculty of Engineering, Islamic Azad University, Sari Branch, Sari (Iran, Islamic Republic of); Shahvari, Yaser [Department of Computer Engineering, Payame Noor University (PNU), P.O. Box 19395-3697, Tehran (Iran, Islamic Republic of); Kia, Faezeh [Golestan Institute of Higher Education, Gorgan 49139-83635 (Iran, Islamic Republic of)


    Highlights: • This article was an attempt to optimize reloading pattern of Bushehr VVER-1000 reactor. • A combination of weighting factor method and the imperialist competitive algorithm was used. • The speed of optimization and desirability of the proposed pattern increased considerably. • To evaluate arrangements, a coupling of WIMSD5-B, CITATION-LDI2 and WERL codes was used. • Results reflected the considerable superiority of the proposed method over direct optimization. - Abstract: In this research, an innovative solution is described which can be used with a combination of the new imperialist competitive algorithm and the weighting factor method to improve speed and increase globality of search in reloading pattern optimization of VVER-1000 reactors in transient cycles and even obtain more desirable results than conventional direct method. In this regard, to reduce the scope of the assumed searchable arrangements, first using the weighting factor method and based on values of these coefficients in each of the 16 types of loadable fuel assemblies in the second cycle, the fuel assemblies were classified in more limited groups. In consequence, the types of fuel assemblies were reduced from 16 to 6 and consequently the number of possible arrangements was reduced considerably. Afterwards, in the first phase of optimization the imperialist competitive algorithm was used to propose an optimum reloading pattern with 6 groups. In the second phase, the algorithm was reused for finding desirable placement of the subset assemblies of each group in the optimum arrangement obtained from the previous phase, and thus the retransformation of the optimum arrangement takes place from the virtual 6-group mode to the real mode with 16 fuel types. In this research, the optimization process was conducted in two states. In the first state, it was tried to obtain an arrangement with the maximum effective multiplication factor and the smallest maximum power peaking factor. In

  10. Identification of a Novel Group of Bacteria in Sludge from a Deteriorated Biological Phosphorus Removal Reactor

    Nielsen, Alex T.; Liu, Wen-Tso; Filipe, Carlos; Grady, Leslie; Molin, Søren; Stahl, David A.


    The microbial diversity of a deteriorated biological phosphorus removal reactor was investigated by methods not requiring direct cultivation. The reactor was fed with media containing acetate and high levels of phosphate (P/C weight ratio, 8:100) but failed to completely remove phosphate in the effluent and showed very limited biological phosphorus removal activity. Denaturing gradient gel electrophoresis (DGGE) of PCR-amplified 16S ribosomal DNA was used to investigate the bacterial diversity. Up to 11 DGGE bands representing at least 11 different sequence types were observed; DNA from the 6 most dominant of these bands was further isolated and sequenced. Comparative phylogenetic analysis of the partial 16S rRNA sequences suggested that one sequence type was affiliated with the alpha subclass of the Proteobacteria, one was associated with the Legionella group of the gamma subclass of the Proteobacteria, and the remaining four formed a novel group of the gamma subclass of the Proteobacteria with no close relationship to any previously described species. The novel group represented approximately 75% of the PCR-amplified DNA, based on the DGGE band intensities. Two oligonucleotide rRNA probes for this novel group were designed and used in a whole-cell hybridization analysis to investigate the abundance of this novel group in situ. The bacteria were coccoid and 3 to 4 μm in diameter and represented approximately 35% of the total population, suggesting a relatively close agreement with the results obtained by the PCR-based DGGE method. Further, based on electron microscopy and standard staining microscopic analysis, this novel group was able to accumulate granule inclusions, possibly consisting of polyhydroxyalkanoate, inside the cells. PMID:10049891

  11. Removal of estrogenic compounds from filtered secondary wastewater effluent in a continuous enzymatic membrane reactor. Identification of biotransformation products.

    Lloret, Lucia; Eibes, Gemma; Moreira, M Teresa; Feijoo, Gumersindo; Lema, Juan M


    In the present study, a novel and efficient technology based on the use of an oxidative enzyme was developed to perform the continuous removal of estrogenic compounds from polluted wastewaters. A 2 L enzymatic membrane reactor (EMR) was successfully operated for 100 h with minimal requirements of laccase for the transformation of estrone (E1), 17β-estradiol (E2), and 17α-ethinylestradiol (EE2)from both buffer solution and real wastewater (filtered secondary effluent). When the experiments were performed at high and low concentrations of the target compounds, 4 mg/L and 100 μg/L, not only high removal yields (80-100%) but also outstanding reduction of estrogenicity (about 84-95%) were attained. When the EMR was applied for the treatment of municipal wastewaters with real environmental concentrations of the different compounds (0.29-1.52 ng/L), excellent results were also achieved indicating the high efficiency and potential of the enzymatic reactor system. A second goal of this study relied on the identification of the transformation products to elucidate the catalytic mechanism of estrogens' transformation by laccase. The formation of dimers and trimers of E1, E2, and EE2, as well as the decomposition of E2 into E1 by laccase-catalyzed treatment, has been demonstrated by liquid chromatography atmospheric pressure chemical ionization (LC-APCI) analysis and confirmed by determination of accurate masses through liquid chromatography electrospray time-of-flight mass spectrometry (LC-ESI-TOF). Dimeric products of E2 and EE2 were found even when operating at environmental concentrations. Moreover, the reaction pathways of laccase-catalyzed transformation of E2 were proposed.

  12. Development of System Analysis Code for Pool-Type Fast Reactor Under Transient Operation%池式快堆系统瞬态分析软件开发

    陆道纲; 隋丹婷


    为实现快堆系统分析软件国产化,在已开发的适用于稳态计算的池式快堆系统分析软件SAC-CFR的基础上,进一步开发了系统各部件的瞬态模型、控制系统和保护系统模型、瞬态工况热工水力学的求解逻辑,完成瞬态计算功能的开发.通过对日本文殊快堆45%功率汽机跳闸工况进行建模分析,验证了SAC-CFR用于系统瞬态分析的有效性,为进一步开发非能动余热排出系统分析模型打下了基础.%Aiming at developing system analysis code independently, the system analysis code for pool-type fast reactor in China (SAC-CFR) under transient operation was developed with further development of component transient model, plant control and protection system model, calculation logic for system transient thermal-hydraulic analysis based on the former SAC-CFR version applicable to steady state analysis. The transient started from turbine trip test at 45 % thermal output in the Monju Plant was analyzed with the developed SAC-CFR. A good agreement between the calculated results and the test data was obtained. SAC-CFR is now ready to incorporate passive residual heat removal model for China Experimental Fast Reactor.

  13. Adaptation and implementation of the TRACE code for transient analysis in designs lead cooled fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Lazaro, A.; Ammirabile, L.; Martorell, S.


    Lead-Cooled Fast Reactor (LFR) has been identified as one of promising future reactor concepts in the technology road map of the Generation IVC International Forum (GIF)as well as in the Deployment Strategy of the European Sustainable Nuclear Industrial Initiative (ESNII), both aiming at improved sustainability, enhanced safety, economic competitiveness, and proliferation resistance. This new nuclear reactor concept requires the development of computational tools to be applied in design and safety assessments to confirm improved inherent and passive safety features of this design. One approach to this issue is to modify the current computational codes developed for the simulation of Light Water Reactors towards their applicability for the new designs. This paper reports on the performed modifications of the TRACE system code to make it applicable to LFR safety assessments. The capabilities of the modified code are demonstrated on series of benchmark exercises performed versus other safety analysis codes. (Author)

  14. Development of PARA-ID Code to Simulate Inelastic Constitutive Equations and Their Parameter Identifications for the Next Generation Reactor Designs

    Koo, Gyeong Hoi; Lee, J. H


    The establishment of the inelastic analysis technology is essential issue for a development of the next generation reactors subjected to elevated temperature operations. In this report, the peer investigation of constitutive equations in points of a ratcheting and creep-fatigue analysis is carried out and the methods extracting the constitutive parameters from experimental data are established. To perform simulations for each constitutive model, the PARA-ID (PARAmeter-IDentification) computer program is developed. By using this code, various simulations related with the parameter identification of the constitutive models are carried out.

  15. Identification of the optical and quiescent counterparts to the bright X-ray transient in NGC 6440

    in 't Zand, J J M; Pooley, D; Verbunt, F; Wijnands, R; Lewin, W H G


    After 3 years of quiescence, the globular cluster NGC 6440 exhibited a bright transient X-ray source turning on in August 2001, as noted with the RXTE All-Sky Monitor. We carried out a short target of opportunity observation with the Chandra X-ray Observatory and are able to associate the transient with the brightest of 24 X-ray sources detected during quiescence in July 2000 with Chandra. Furthermore, we securely identify the optical counterpart and determine that the 1998 X-ray outburst in NGC 6440 was from the same object. This is the first time that an optical counterpart to a transient in a globular cluster is securely identified. Since the transient is a type I X-ray burster, it is established that the compact accretor is a neutron star. Thus, this transient provides an ideal case to study the quiescent emission in the optical and X-ray of a transiently accreting neutron star while knowing the distance and reddening accurately. One model that fits the quiescent spectrum is an absorbed power law plus neu...

  16. Equivalent Electro-magnetic Transient Model Based on Dynamic Reluctance for Magnetically Controlled Shunt Reactor%基于动态磁阻的磁控式并联电抗器等效电抗暂态模型

    郑伟杰; 周孝信


    A mathematical model which describes the nonlinear magnetic saturation characteristics of magnetically controlled shunt reactor (MCSR) for extra and ultra voltage was proposed. Based on dynamic reluctance, instantaneous equivalent reactor calculation algorithm was proposed and used to reflect the dynamic characteristics of a shunt reactor system with AC-DC mixed excitation, while inverse-hyperbolic function describing the nonlinear characteristics, damping implicit trapezoidal integration algorithm was used to differentiate the reactor, equivalent electro-magnetic transient model was established accordingly, which is compatible with the electromagnetic transient algorithm. Advanced digital power system simulator (ADPSS) has contained the model, site operation parameters of MCSR in project were used as input in the simulation example, the results confirm the accuracy of model.%提出一种描述超/特高压磁控式并联电抗器(magnetically controlled shunt reactor,MCSR)非线性磁饱和特性的实用化解耦电磁暂态模型.基于动态磁阻的思想,提出瞬时等效电抗的算法,反映了交直流混合励磁条件下饱和等效电抗的实时动态变化特性:以反双曲函数描述非线性磁路特性,对耦合磁路方程进行解耦,并用带阻尼的隐式梯形积分算法对瞬时等效电抗进行差分化,建立电磁暂态模型,能够与电磁暂态计算相兼容.已在电力系统全数字实时仿真装置(advanced digital power system simulator,ADPSS)中编程实现了仿真模型的工程化应用,并搭建标准算例,以工程投运的MCSR现场运行参数为模型输入量,进行仿真,计算结果数值精确,波形吻合,验证了模型的正确性.

  17. Nuclear Reactors

    Hogerton, John


    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  18. Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor; Algoritmo neuro-difuso para la identificacion de la potencia neutronica del reactor Triga Mark III

    Rojas R, E.; Benitez R, J. S. [Instituto Tecnologico de Toluca, Division de Estudios de Posgrado e Investigacion, Av. Tecnologico s/n, Ex-Rancho La Virgen, 50140 Metepec, Estado de Mexico (Mexico); Segovia de los Rios, J. A.; Rivero G, T. [ININ, Gerencia de Ciencias Aplicadas, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail:


    In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

  19. Reactor operation safety information document


    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  20. LMFBR type reactor

    Kawakami, Hiroto


    A reactor container of the present invention has a structure that the reactor container is entirely at the same temperature as that at the inlet of the reactor and, a hot pool is incorporated therein, and the reactor container has is entirely at the same temperature and has substantially uniform temperature follow-up property transiently. Namely, if the temperature at the inlet of the reactor core changes, the temperature of the entire reactor container changes following this change, but no great temperature gradient is caused in the axial direction and no great heat stresses due to axial temperature distribution is caused. Occurrence of thermal stresses caused by the axial temperature distribution can be suppressed to improve the reliability of the reactor container. In addition, since the laying of the reactor inlet pipelines over the inside of the reactor is eliminated, the reactor container is made compact and the heat shielding structures above the reactor and a protection structure of container walls are simplified. Further, secondary coolants are filled to the outside of the reactor container to simplify the shieldings. The combined effects described above can improve economical property and reliability. (N.H.).

  1. Graybox and adaptative dynamic neural network identification models to infer the steady state efficiency of solar thermal collectors starting from the transient condition

    Roberto, Baccoli; Ubaldo, Carlini; Stefano, Mariotti; Roberto, Innamorati; Elisa, Solinas; Paolo, Mura [Institute of Technical Physics of the University of Cagliari, via Marengo 1, 09123 Cagliari (Italy)


    This paper deals with the development of methods for non steady state test of solar thermal collectors. Our goal is to infer performances in steady-state conditions in terms of the efficiency curve when measures in transient conditions are the only ones available. We take into consideration the method of identification of a system in dynamic conditions by applying a Graybox Identification Model and a Dynamic Adaptative Linear Neural Network (ALNN) model. The study targets the solar collector with evacuated pipes, such as Dewar pipes. The mathematical description that supervises the functioning of the solar collector in transient conditions is developed using the equation of the energy balance, with the aim of determining the order and architecture of the two models. The input and output vectors of the two models are constructed, considering the measures of 4 days of solar radiation, flow mass, environment and heat-transfer fluid temperature in the inlet and outlet from the thermal solar collector. The efficiency curves derived from the two models are detected in correspondence to the test and validation points. The two synthetic simulated efficiency curves are compared with the actual efficiency curve certified by the Swiss Institute Solartechnik Puffung Forschung which tested the solar collector performance in steady-state conditions according to the UNI-EN 12975 standard. An acquisition set of measurements of only 4 days in the transient condition was enough to trace through a Graybox State Space Model the efficiency curve of the tested solar thermal collector, with a relative error of synthetic values with respect to efficiency certified by SPF, lower than 0.5%, while with the ALNN model the error is lower than 2.2% with respect to certified one. (author)

  2. Parameter Identification with the Random Perturbation Particle Swarm Optimization Method and Sensitivity Analysis of an Advanced Pressurized Water Reactor Nuclear Power Plant Model for Power Systems

    Li Wang


    Full Text Available The ability to obtain appropriate parameters for an advanced pressurized water reactor (PWR unit model is of great significance for power system analysis. The attributes of that ability include the following: nonlinear relationships, long transition time, intercoupled parameters and difficult obtainment from practical test, posed complexity and difficult parameter identification. In this paper, a model and a parameter identification method for the PWR primary loop system were investigated. A parameter identification process was proposed, using a particle swarm optimization (PSO algorithm that is based on random perturbation (RP-PSO. The identification process included model variable initialization based on the differential equations of each sub-module and program setting method, parameter obtainment through sub-module identification in the Matlab/Simulink Software (Math Works Inc., Natick, MA, USA as well as adaptation analysis for an integrated model. A lot of parameter identification work was carried out, the results of which verified the effectiveness of the method. It was found that the change of some parameters, like the fuel temperature and coolant temperature feedback coefficients, changed the model gain, of which the trajectory sensitivities were not zero. Thus, obtaining their appropriate values had significant effects on the simulation results. The trajectory sensitivities of some parameters in the core neutron dynamic module were interrelated, causing the parameters to be difficult to identify. The model parameter sensitivity could be different, which would be influenced by the model input conditions, reflecting the parameter identifiability difficulty degree for various input conditions.

  3. Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee



    This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process.

  4. Transient analysis through Hilbert spectra of electrochemical noise signals for the identification of localized corrosion of stainless steel

    Homborg, A.M.; Tinga, T.; Zhang, X.; Westing, van E.P.M.; Oonincx, P.J.; Ferrari, G.M.; Wit, de J.H.W.; Mol, J.M.C.


    Hilbert spectra allow identification of instantaneous frequencies that are attributed to specific corrosion mechanisms in electrochemical noise data. The present work proposes to identify and analyze areas of interest in Hilbert spectra, which enables to obtain valuable frequency information from el

  5. Selective enrichment and ultrasensitive identification of trace peptides in proteome analysis using transient capillary isotachophoresis/zone electrophoresis coupled with nano-ESI-MS.

    An, Yanming; Cooper, Jonathan W; Balgley, Brian M; Lee, Cheng S


    Besides the complexity in protein samples of biological origin, probably the greatest challenge presently facing comprehensive proteome analysis is related to the large variation of protein relative abundances (>6 orders of magnitude), having potential biological significance in mammalian systems. As demonstrated in this work, transient capillary ITP/zone electrophoresis (CITP/CZE) provides selective analyte enrichment through electrokinetic stacking and extremely high resolving power toward protein and peptide mixtures. The result of the CITP process is that major components may be diluted, but trace compounds are concentrated. The on-column transition of CITP to CZE minimizes additional band broadening while providing superior analyte resolution. Online coupling of transient CITP/CZE with nano-ESI-MS allows ultrasensitive detection of trace peptides at levels of subnanomolar concentration or subfemtomole mass in complex peptide mixtures. More importantly, selective enrichment of trace peptides enables the identification and sequence analysis of low-abundance peptides co-migrated with highly abundant species at a concentration ratio of 1:500,000. The combined CITP/CZE-nano-ESI-MS system is demonstrated to be at least one to two orders of magnitude more sensitive than that attained in conventional electrophoretic and chromatographic-based proteome technologies over a wide dynamic concentration range, potentially allowing comprehensive analysis of protein profiles within a small cell population and limited tissue samples using conventional mass spectrometers. Furthermore, the speed of CITP/CZE separation and the lack of column equilibration in CITP/CZE not only improve the throughput of proteome analysis, but also facilitate its seamless integration with other separation technologies in a multidimensional protein identification platform.

  6. Test reactor risk assessment methodology

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.


    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor.

  7. Construction and identification of eukaryotic eukaryotic expression plasmid pcdna3.1-bace and its transient expression in cells

    Huilin Gong; Guanjun Zhang; Weijiang Dong


    Objective: To generate eukaryotic expression vector of pcDNA3.1-BACE and obtain its transient expression in COS-7 cells and high expression in the neuroblastoma SK-N-SH cells. Methods: A 1503 bp cDNA fragment was amplified from the total RNA of human neuroblastoma by RT-PCR method and cloned into plasmid pcDNA3.1. The vector was identified by digestion with restriction enzymes BamHI and XhoI and sequenced by Sanger-dideoxy-mediated chain termination. The expression of BACE gene was detected by immunocytochemistry method. Results: The results showed that the cDNAfragment included 1503 bp total coding region. The recombinant eukaryotic cell expression vector of pcDNA3.1-BACE was constructed successfully,and the sequence of insert was identical to the published sequence. The COS-7 cells and the neuroblastoma SK-N-SH cells transfected with the pcDNA3.1-BACE plasmid expressed high level of BACE protein in cytoplasm. Conclusion: The recombinant plasmid pcDNA3.1-BACE can provide very useful tool for researching the reason of Alzheimer's disease and lays the important foundation for preventing the AD laterly.

  8. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor; Aplicacion del algoritmo MOVE para la identificacion de modelos de orden reducido del nucleo de un reactor tipo BWR

    Victoria R, M.A.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail:


    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  9. Thermal Reactor Safety


    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  10. CFD Analysis for Flow Behavior Characteristics in the Upper Plenum during low flow/low pressure transients for the Gas Cooled Fast Reactor (GCFR)

    Piyush Sabharwall; Theron Marshall; Kevan Weaver; Hans Gougar


    Gas coolant at low pressure exhibits poor heat transfer characteristics. This is an area of concern for the passive response targeted by the Generation IV GCFR design. For the first 24 hour period, the decay heat removal for the GCFR design is dependent on an actively powered blower, which also would reduce the temperature in the fuel during transients, before depending on the passive operation. Natural circulation cooling initiates when the blower is stopped for the final phase of the decay heat removal, as under forced convection the core decay heat is adequately cooled by the running blower. The ability of the coolant to flow in the reverse direction or having recirculation, when the blowers are off, necessitates more understanding of the flow behavior characteristics in the upper plenum. The work done here focuses primarily on the period after the blower has been turned off, as the core is adequately cooled when the blowers are running, thus there was no need to carry out the analysis for the first 24 hours. In order to understand the plume behavior for the GCFR upper plenum several cases were run, with air, helium and helium-air mixture. For each case, the FLUENT was used to characterize the steady state velocity vectors and corresponding temperature in the upper plenum under passive decay heat removal conditions. This study will provide better insight into the plume interaction in the upper plenum at low flow and low pressure conditions.

  11. Simple 2,4-diacylphloroglucinols as classic transient receptor potential-6 activators--identification of a novel pharmacophore.

    Leuner, K; Heiser, J H; Derksen, S; Mladenov, M I; Fehske, C J; Schubert, R; Gollasch, M; Schneider, G; Harteneck, C; Chatterjee, S S; Müller, W E


    The naturally occurring acylated phloroglucinol derivative hyperforin was recently identified as the first specific canonical transient receptor potential-6 (TRPC6) activator. Hyperforin is the major antidepressant component of St. John's wort, which mediates its antidepressant-like properties via TRPC6 channel activation. However, its pharmacophore moiety for activating TRPC6 channels is unknown. We hypothesized that the phloroglucinol moiety could be the essential pharmacophore of hyperforin and that its activity profile could be due to structural similarities with diacylglycerol (DAG), an endogenous nonselective activator of TRPC3, TRPC6, and TRPC7. Accordingly, a few 2-acyl and 2,4-diacylphloroglucinols were tested for their hyperforin-like activity profiles. We used a battery of experimental models to investigate all functional aspects of TRPC6 activation, including ion channel recordings, Ca(2+) imaging, neurite outgrowth, and inhibition of synaptosomal uptake. Phloroglucinol itself was inactive in all of our assays, which was also the case for 2-acylphloroglucinols. For TRPC6 activation, the presence of two symmetrically acyl-substitutions with appropriate alkyl chains in the phloroglucinol moiety seems to be an essential prerequisite. Potencies of these compounds in all assays were comparable with that of hyperforin for activating the TRPC6 channel. Finally, using structure-based modeling techniques, we suggest a binding mode for hyperforin to TRPC6. Based on this modeling approach, we propose that DAG is able to activate TRPC3, TRPC6, and TRPC7 because of higher flexibility within the chemical structure of DAG compared with the rather rigid structures of hyperforin and the 2,4-diacylphloroglucinol derivatives.

  12. The background rate of false positives: Combining simulations of gravitational wave events with an unsupervised algorithm for transient identification in crowded image-subtracted data

    Ackley, Kendall; Eikenberry, Stephen; Klimenko, Sergey; LIGO Collaboration


    We are now entering the era of multimessenger gravitational wave (GW) astronomy with the completion of the first observing run of Advanced LIGO. Multiwavelength electromagnetic (EM) emission is expected to accompany gravitational radiation from compact object binary mergers, such as those between neutron stars and stellar-mass black holes, where Advanced LIGO is most sensitive to their detection. Attempting to perform EM follow-up over the 10-100s deg2 error regions will be faced with many challenges, including the identification and removal of O (105) false positive transients that appear as a commotion of background events and as image artifacts in crowded image-subtracted fields. We present an update to our automated unsupervised algorithm including how our pipeline uses the existing coherent WaveBurst pipeline in an attempt to develop optimized EM follow-up schema. Our end-to-end pipeline combines simulated GW events with actual observational data from a number of ground-based optical observatories, including PTF, ROTSE, and DECam. Our performance is reported both in terms of the number of coincident false positives as well as the efficiency of recovery.

  13. Introduction to the neutron kinetics of nuclear power reactors

    Tyror, J G; Grant, P J


    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  14. The TITAN reversed-field-pinch fusion reactor study


    This paper on titan plasma engineering contains papers on the following topics: reversed-field pinch as a fusion reactor; parametric systems studies; magnetics; burning-plasma simulations; plasma transient operations; current drive; and physics issues for compact RFP reactors.

  15. H Reactor

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  16. Selection and identification of a bacterial community able to degrade and detoxify m-nitrophenol in continuous biofilm reactors.

    González, Ana J; Fortunato, María S; Papalia, Mariana; Radice, Marcela; Gutkind, Gabriel; Magdaleno, Anahí; Gallego, Alfredo; Korol, Sonia E


    Nitroaromatics are widely used for industrial purposes and constitute a group of compounds of environmental concern because of their persistence and toxic properties. Biological processes used for decontamination of nitroaromatic-polluted sources have then attracted worldwide attention. In the present investigation m-nitrophenol (MNP) biodegradation was studied in batch and continuous reactors. A bacterial community able to degrade the compound was first selected from a polluted freshwater stream and the isolates were identified by the analysis of the 16S rRNA gene sequence. The bacterial community was then used in biodegradation assays. Batch experiments were conducted in a 2L aerobic microfermentor at 28 °C and with agitation (200 rpm). The influence of abiotic factors in the biodegradation process in batch reactors, such as initial concentration of the compound and initial pH of the medium, was also studied. Continuous degradation of MNP was performed in an aerobic up-flow fixed-bed biofilm reactor. The biodegradation process was evaluated by determining MNP and ammonium concentrations and chemical oxygen demand (COD). Detoxification was assessed by Vibrio fischeri and Pseudokirchneriella subcapitata toxicity tests. Under batch conditions the bacterial community was able to degrade 0.72 mM of MNP in 32 h, with efficiencies higher than 99.9% and 89.0% of MNP and COD removals respectively and with concomitant release of ammonium. When the initial MNP concentration increased to 1.08 and 1.44 mM MNP the biodegradation process was accomplished in 40 and 44 h, respectively. No biodegradation of the compound was observed at higher concentrations. The community was also able to degrade 0.72 mM of the compound at pH 5, 7 and 9. In the continuous process biodegradation efficiency reached 99.5% and 96.8% of MNP and COD removal respectively. The maximum MNP removal rate was 37.9 gm(-3) day(-1). Toxicity was not detected after the biodegradation process.

  17. Identification of trigger factors selecting for polyphosphate- and glycogen-accumulating organisms in aerobic granular sludge sequencing batch reactors.

    Weissbrodt, David G; Schneiter, Guillaume S; Fürbringer, Jean-Marie; Holliger, Christof


    Nutrient removal performances of sequencing batch reactors using granular sludge for intensified biological wastewater treatment rely on optimal underlying microbial selection. Trigger factors of bacterial selection and nutrient removal were investigated in these novel biofilm systems with specific emphasis on polyphosphate- (PAO) and glycogen-accumulating organisms (GAO) mainly affiliated with Accumulibacter and Competibacter, respectively. In a first dynamic reactor operated with stepwise changes in concentration and ratio of acetate and propionate (Ac/Pr) under anaerobic feeding and aerobic starvation conditions and without wasting sludge periodically, propionate favorably selected for Accumulibacter (35% relative abundance) and stable production of granular biomass. A Plackett-Burman multifactorial experimental design was then used to screen in eight runs of 50 days at stable sludge retention time of 15 days for the main effects of COD concentration, Ac/Pr ratio, COD/P ratio, pH, temperature, and redox conditions during starvation. At 95% confidence level, pH was mainly triggering direct Accumulibacter selection and nutrient removal. The overall PAO/GAO competition in granular sludge was statistically equally impacted by pH, temperature, and redox factors. High Accumulibacter abundances (30-47%), PAO/GAO ratios (2.8-8.4), and phosphorus removal (80-100%) were selected by slightly alkaline (pH > 7.3) and lower mesophilic (temperature. In addition to alkalinity, non-limited organic conditions, 3-carbon propionate substrate, sludge age control, and phase length adaptation under alternating aerobic-anoxic conditions during starvation can lead to efficient nutrient-removing granular sludge biofilm systems.

  18. Identifiability of sorption parameters in stirred flow-through reactor experiments and their identification with a Bayesian approach.

    Nicoulaud-Gouin, V; Garcia-Sanchez, L; Giacalone, M; Attard, J C; Martin-Garin, A; Bois, F Y


    This paper addresses the methodological conditions -particularly experimental design and statistical inference- ensuring the identifiability of sorption parameters from breakthrough curves measured during stirred flow-through reactor experiments also known as continuous flow stirred-tank reactor (CSTR) experiments. The equilibrium-kinetic (EK) sorption model was selected as nonequilibrium parameterization embedding the Kd approach. Parameter identifiability was studied formally on the equations governing outlet concentrations. It was also studied numerically on 6 simulated CSTR experiments on a soil with known equilibrium-kinetic sorption parameters. EK sorption parameters can not be identified from a single breakthrough curve of a CSTR experiment, because Kd,1 and k(-) were diagnosed collinear. For pairs of CSTR experiments, Bayesian inference allowed to select the correct models of sorption and error among sorption alternatives. Bayesian inference was conducted with SAMCAT software (Sensitivity Analysis and Markov Chain simulations Applied to Transfer models) which launched the simulations through the embedded simulation engine GNU-MCSim, and automated their configuration and post-processing. Experimental designs consisting in varying flow rates between experiments reaching equilibrium at contamination stage were found optimal, because they simultaneously gave accurate sorption parameters and predictions. Bayesian results were comparable to maximum likehood method but they avoided convergence problems, the marginal likelihood allowed to compare all models, and credible interval gave directly the uncertainty of sorption parameters θ. Although these findings are limited to the specific conditions studied here, in particular the considered sorption model, the chosen parameter values and error structure, they help in the conception and analysis of future CSTR experiments with radionuclides whose kinetic behaviour is suspected.

  19. Reactor Physics

    Ait Abderrahim, A


    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  20. Reactor safeguards

    Russell, Charles R


    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  1. Reactor operation

    Shaw, J


    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  2. Source identification of nitrous oxide emission pathways from a single-stage nitritation-anammox granular reactor

    Ali, Muhammad


    Nitrous oxide (N2O) production pathway in a signal-stage nitritation-anammox sequencing batch reactor (SBR) was investigated based on a multilateral approach including real-time N2O monitoring, N2O isotopic composition analysis, and in-situ analyses of spatial distribution of N2O production rate and microbial populations in granular biomass. N2O emission rate was high in the initial phase of the operation cycle and gradually decreased with decreasing NH4+ concentration. The average emission of N2O was 0.98 ± 0.42% and 1.35 ± 0.72% of the incoming nitrogen load and removed nitrogen, respectively. The N2O isotopic composition analysis revealed that N2O was produced via NH2OH oxidation and NO2− reduction pathways equally, although there is an unknown influence from N2O reduction and/or anammox N2O production. However, the N2O isotopomer analysis could not discriminate the relative contribution of nitrifier denitrification and heterotrophic denitrification in the NO2− reduction pathway. Various in-situ techniques (e.g. microsensor measurements and FISH (fluorescent in-situ hybridization) analysis) were therefore applied to further identify N2O producers. Microsensor measurements revealed that approximately 70% of N2O was produced in the oxic surface zone, where nitrifiers were predominantly localized. Thus, NH2OH oxidation and NO2 reduction by nitrifiers (nitrifier-denitrification) could be responsible for the N2O production in the oxic zone. The rest of N2O (ca. 30%) was produced in the anammox bacteria-dominated anoxic zone, probably suggesting that NO2− reduction by coexisting putative heterotrophic denitrifiers and some other unknown pathway(s) including the possibility of anammox process account for the anaerobic N2O production. Further study is required to identify the anaerobic N2O production pathways. Our multilateral approach can be useful to quantitatively examine the relative contributions of N2O production pathways. Good understanding of the key N2O

  3. Transient osteoporosis.

    Korompilias, Anastasios V; Karantanas, Apostolos H; Lykissas, Marios G; Beris, Alexandros E


    Transient osteoporosis is characterized primarily by bone marrow edema. The disease most commonly affects the hip, knee, and ankle in middle-aged men. Its cause remains unknown. The hallmark that separates transient osteoporosis from other conditions presenting with a bone marrow edema pattern is its self-limited nature. Laboratory tests usually do not contribute to the diagnosis. Plain radiographs may reveal regional osseous demineralization. Magnetic resonance imaging is used primarily for early diagnosis and monitoring disease progression. Early differentiation from more aggressive conditions with long-term sequelae is essential to avoid unnecessary treatment. Clinical entities such as transient osteoporosis of the hip and regional migratory osteoporosis are spontaneously resolving conditions. However, early differential diagnosis and surgical treatment are crucial for the patient with osteonecrosis of the hip or knee.

  4. A model of reactor kinetics

    Thompson, A.S.; Thompson, B.R.


    The analytical model of nuclear reactor transients, incorporating both mechanical and nuclear effects, simulates reactor kinetics. Linear analysis shows the stability borderline for small power perturbations. In a stable system, initial power disturbances die out with time. With an unstable combination of nuclear and mechanical characteristics, initial disturbances persist and may increase with time. With large instability, oscillations of great magnitude occur. Stability requirements set limits on the power density at which particular reactors can operate. The limiting power density depends largely on the product of two terms: the fraction of delayed neutrons and the frictional damping of vibratory motion in reactor core components. As the fraction of delayed neutrons is essentially fixed, mechanical damping largely determines the maximum power density. A computer program, based on the analytical model, calculates and plots reactor power as a nonlinear function of time in response to assigned values of mechanical and nuclear characteristics.

  5. Reactor Neutrinos

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang


    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...


    Untermyer, S.


    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  7. An Iris Mechanism Driven Temperature Control of Solar Thermal Reactors

    Van den Langenbergh, Lode; Ophoff, Cédric; Ozalp, Nesrin


    In spite of their attraction for clean production of fuels and commodities; solar thermal reactors are challenged by the transient nature of solar energy. Control of reactor temperature during transient periods is the key factor to maintain solar reactor performance. Currently, there are few techniques that are being used to accommodate the fluctuations of incoming solar radiation. One of the commonly practiced methods is to adjust the mass flow rate of the feedstock which is very simple to i...


    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.


    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  9. The MWA Transients Survey (MWATS).

    Bell, M.; Murphy, T.; Kaplan, D. L.; Croft, S. D.; Hancock, P.; Rowlinson, A.; Wayth, R.; Gaensler, B.; Hurley-Walker, N.; Offringa, A.; Loi, C.; Bannister, K.; Trott, C.; Marquart, J.


    We propose the continuation of the MWA transients survey to search for and monitor low frequency transient and variable radio sources in the southern sky. This proposal is aimed at commensally utilising data from the GLEAM-X (G0008) project in semester 2017-A. The aim of this commensal data acquisition is to commission long baseline observations for transient science. In particular this will involve studying the impact of the ionosphere on calibration and imaging, and developing the techniques needed to produce science quality data products. The proposed drift scans with LST locking (see G0008 proposal) are particularly exciting as we can test image subtraction for transient and variable identification. This survey is targeted at studying objects such as AGN (intrinsic and extrinsic variability), long duration synchrotron emitters, pulsars and transients of unknown origin. The maps generated from this survey will be analysed with the Variables and Slow Transients (VAST) detection pipeline. The motivation for this survey is as follows: (i) To obtain temporal data on an extremely large and robust sample of low frequency sources to explore and quantify both intrinsic and extrinsic variability; (ii) To search and find new classes of low frequency radio transients that previously remained undetected and obscured from multi-wavelength discovery; (iii) To place rigorous statistics on the occurrence of both transients and variables prior to the Australian SKA era.

  10. Membrane reactor. Membrane reactor

    Shindo, Y.; Wakabayashi, K. (National Chemical Laboratory for Industry, Tsukuba (Japan))


    Many reaction examples were introduced of membrane reactor, to be on the point of forming a new region in the field of chemical technology. It is a reactor to exhibit excellent function, by its being installed with membrane therein, and is generally classified into catalyst function type and reaction promotion type. What firstly belongs to the former is stabilized zirconia, where oxygen, supplied to the cathodic side of membrane with voltage, impressed thereon, becomes O {sup 2 {minus}} to be diffused through the membrane and supplied, as variously activated oxygenous species, on the anodic side. Examples with many advantages can be given such as methane coupling, propylene oxidation, methanating reaction of carbon dioxide, etc. Apart, palladium film and naphion film also belong to the former. While examples of the latter comprise, among others, decomposition of hydrogen sulfide by porous glass film and dehydrogenation of cyclohexane or palladium alloy film, which are expected to be developed and materialized in the industry. 33 refs., 8 figs.

  11. Helias reactor studies

    Beidler, C.D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Grieger, G. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Harmeyer, E. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Kisslinger, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Karulin, N. [Nuclear Fusion Institute, Moscow (Russian Federation); Maurer, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany); Nuehrenberg, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Rau, F. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Sapper, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Wobig, H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)


    The present status of Helias reactor studies is characterised by the identification and investigation of specific issues which result from the particular properties of this type of stellarator. On the technical side these are issues related to the coil system, while physics studies have concentrated on confinement, alpha-particle behaviour and ignition conditions. The usual assumptions have been made in those fields which are common to all toroidal fusion reactors: blanket and shield, refuelling and exhaust, safety and economic aspects. For blanket and shield sufficient space has been provided, a detailed concept will be developed in future. To date more emphasis has been placed on scoping and parameter studies as opposed to fixing a specific set of parameters and providing a detailed point study. One result of the Helias reactor studies is that physical dimensions are on the same order as those of tokamak reactors. However, it should be noticed that this comparison is difficult in view of the large spectrum of tokamak reactors ranging from a small reactor like Aries, to a large device such as SEAFP. The notion that the large aspect ratio of 10 or more in Helias configurations also leads to large reactors is misleading, since the large major radius of 22 m is compensated by the average plasma radius of 1.8 m and the average coil radius of 5 m. The plasma volume of 1400 m{sup 3} is about the same as the ITER reactor and the magnetic energy of the coil system is about the same or even slightly smaller than envisaged in ITER. (orig.)

  12. improving the transient stability of nigerian 330kv transmission ...


    controlled reactor is used to model the SVC and is appropriately sized and located within the network. Transient stability of the .... The mathematical analysis in relation to Nigerian .... iteratively using Newton-Raphson's procedure following the ...

  13. The Detection and Identification of Transient Power Quality Based on Wavelet Transform%基于小波变换的暂态电能质量的检测与识别

    潘从茂; 李凤婷


    The intermittent energy connected to the grid affects the operating characteristics of the power system and power quality, especially the impact of transient power quality should not be overlooked. This paper details Wavelet detection of transient power disturbances basic principle and method based on the wavelet transform which is applied to transient power quality disturbance detection and identification, combined with the example simulation study. Theoretical analysis and simulation results show that this method can achieve a voltage transient power quality disturbances fast, accurate detection, and be able to identify the disturbance type according to the energy function. Research ideas with intermittent energy access system for the detection and recognition of power quality provides an effective and feasible method.%间歇性能源接入电网影响电力系统的运行特性和电能质量,特别是对暂态电能质量的影响不容忽视。本文在详细介绍暂态电能扰动小波检测的基本原理和实现方法的基础上,将小波变换应用于暂态电能质量扰动检测和识别,结合算例进行了仿真研究。理论分析和仿真结果表明该方法能够实现对电压暂态电能质量扰动快速、准确的检测,并能够根据能量函数识别出扰动类型。研究思路可为检测识别含间歇性能源接入系统的电能质量提供一种有效、可行的方法。

  14. A comparative study of kinetics of nuclear reactors

    Obaidurrahman Khalilurrahman


    Full Text Available The paper deals with the study of reactivity initiated transients to investigate major differences in the kinetics behavior of various reactor systems under different operating conditions. The article also states guidelines to determine the safety limits on reactivity insertion rates. Three systems, light water reactors (pressurized water reactors, heavy water reactors (pressurized heavy water reactors, and fast breeder reactors are considered for the sake of analysis. The upper safe limits for reactivity insertion rate in these reactor systems are determined. The analyses of transients are performed by a point kinetics computer code, PKOK. A simple but accurate method for accounting total reactivity feedback in kinetics calculations is suggested and used. Parameters governing the kinetics behavior of the core are studied under different core states. A few guidelines are discussed to project the possible kinetics trends in the next generation reactors.

  15. Naval reactors physics handbook. Volume 3: The physics of intermediate spectrum reactors

    Stehn, J.R. [ed.] [Knolls Atomic Power Lab., Schenectady, NY (United States)


    The present volume has been prepared for persons with some knowledge of the physics of nuclear reactors. It is intended to make available the accumulated physics experience of the Knolls Atomic Power Laboratory in its work on intermediate spectrum reactors. Only those portions have been selected which were deemed to be most useful and significant to other physicists concerned with the problems of reactor design. The volume is divided into four parts which are more or less independent of one another. Part 1 (Chaps. 2--9), Investigation of Reactor Characteristics by Critical Assemblies, reflects the importance of the properties of critical assemblies and of the techniques for obtaining experimental information about such assemblies. Part 2 (Chaps. 10--20), Reactivity Effects Associated with Reactor Operation, details the use of both critical assemblies and reactor theory to make and test predictions of the manner in which the reactivity of an intermediate power reactor will vary during operation. Part 3 (Chaps. 21--26), Heat Generation and Nuclear Materials Problems, considers how reactor heat generation is spread out in space and time, and what nuclear effects result from the presence of beryllium or sodium in the reactor. Part 4 (Chaps. 27--38), Reactor Kinetics and Temperature Coefficients, relates to the transient or near-transient behavior of intermediate reactors.

  16. Multifunctional reactors

    Westerterp, K.R.


    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  17. Study of hydrogen generation plant coupled to high temperature gas cooled reactor

    Brown, Nicholas Robert

    Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Several thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. Eight unique case studies are performed based on a thorough literature review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without SCRAM, (6) total failure of the chemical plant, (7) parametric study of the temperature in an individual reaction chamber, and (8) control rod insertion in the nuclear reactor. Various parametric

  18. A buoyantly-driven shutdown rod concept for passive reactivity control of a Fluoride salt-cooled High-temperature Reactor

    Blandford, Edward D., E-mail: [Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, NM 87131-0001 (United States); Peterson, Per F. [Department of Nuclear Engineering, University of California, Berkeley, CA 94720-1730 (United States)


    Highlights: • We develop a novel buoyantly-driven shutdown rod concept for a FHR. • Shutdown rod system can be actively or passively activated during transients. • Response of the rod was computationally simulated and experimentally validated. • Initial results indicate rod could provide effective transient reactivity control. -- Abstract: This paper presents a novel buoyantly-driven shutdown rod concept for use in Fluoride salt-cooled High-temperature Reactors (FHRs). The baseline design considered here is a 900 MWth modular version of the FHR class called the Pebble Bed Advanced High-Temperature Reactor (PB-AHTR) that uses pebble fuel. Due to the high volumetric heat capacity of the primary coolant, the FHRs operate with a high power density core with a similar average coolant temperature as in modular helium reactors. The reactivity control system for the baseline PB-AHTR uses a novel buoyantly-driven shutdown rod system that can be actively or passively activated during reactor transients. In addition to a traditional active insertion mechanism, the new shutdown rod system is designed to also operate passively, fulfilling the role of a reserve shutdown system. The physical response of the shutdown rod was simulated both computationally and experimentally, using scaling arguments where applicable, with an emphasis on key phenomena identified by a preliminary Phenomena Identification and Ranking Table (PIRT) study. This paper presents results from both the pre-predicted simulation and experimental validation efforts.

  19. Thermal-hydraulic modeling of reactivity accidents in MTR reactors

    Khater Hany


    Full Text Available This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.

  20. Reactor vessel

    Makkee, M.; Kapteijn, F.; Moulijn, J.A


    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and comprising connections for supplying (9, 10, 11) and discharging (13, 14, 15) via the channels (3) gases and/or liquids entering into a reaction with each other and substances formed upon this reactio...

  1. Thermal hydraulics analysis of the Advanced High Temperature Reactor

    Wang, Dean, E-mail: [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)


    Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

  2. Chemical Reactors.

    Kenney, C. N.


    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  3. Reactor Neutrinos

    Soo-Bong Kim


    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.


    Miller, H.I.; Smith, R.C.


    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  5. Reactor Engineering

    Lema, Juan M.; López, Carmen; Eibes, Gemma; Taboada-Puig, Roberto; Moreira, M. Teresa; Feijoo, Gumersindo

    In this chapter, the engineering aspects of processes catalyzed by peroxidases will be presented. In particular, a discussion of the existing technologies that utilize peroxidases for different purposes, such as the removal of recalcitrant compounds or the synthesis of polymers, is analyzed. In the first section, the essential variables controlling the process will be investigated, not only those that are common in any enzymatic system but also those specific to peroxidative reactions. Next, different reactor configurations and operational modes will be proposed, emphasizing their suitability and unsuitability for different systems. Finally, two specific reactors will be described in detail: enzymatic membrane reactors and biphasic reactors. These configurations are especially valuable for the treatment of xenobiotics with high and poor water solubility, respectively.

  6. Reactor Neutrinos

    Lasserre, T.; Sobel, H.W.


    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrin...

  7. High power transient characteristics and capability of NSRR

    Nakamura, Takehiko; Kashima, Yoichi; Yachi, Shigeyasu; Yoshinaga, Makio; Terakado, Yoshibumi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Katanishi, Shoji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment


    In order to study fuel behavior under abnormal transients and accidents, the control system of the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute (JAERI) was modified to achieve high power transients. With this new operational mode, called Shaped Pulse (SP), transients at the maximum power of 10 MW can be conducted for a few seconds. This new operational mode supplements the previous Natural Pulse (NP) operation at the maximum power of 23 GW for milliseconds. For high power transient operation, a simulator using a point kinetic model was developed, and characteristics of the NSRR in the new operational mode were examined through tests and calculations. With the new operational mode, new types of fuel irradiation tests simulating power oscillations of boiling water reactors (BWRs) can be conducted in the NSRR. Reactor characteristics and capability, such as control rod worth, feedback reactivity, and operational limits of the NSRR for SP operations are discussed. (author)

  8. Diagnostic system for identification of accident scenarios in nuclear power plants using artificial neural networks

    Santosh, T.V. [Health, Safety and Environment Group, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)], E-mail:; Srivastava, A.; Sanyasi Rao, V.V.S.; Ghosh, A.K.; Kushwaha, H.S. [Health, Safety and Environment Group, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)


    This paper presents the work carried out towards developing a diagnostic system for the identification of accident scenarios in 220 MWe Indian PHWRs. The objective of this study is to develop a methodology based on artificial neural networks (ANNs), which assists in identifying a transient quickly and suggests the operator to initiate the corrective actions during abnormal operations of the reactor. An operator support system, known as symptom-based diagnostic system (SBDS), has been developed using ANN that diagnoses the transients based on reactor process parameters, and continuously displays the status of the reactor. As a pilot study, the large break loss of coolant accident (LOCA) with and without the emergency core cooling system (ECCS) in reactor headers has been considered. Several break scenarios of large break LOCA have been analyzed. The time-dependent transient data have been generated using the RELAP5 thermal hydraulic code assuming an equilibrium core, which conforms to a realistic estimation. The diagnostic results obtained from the ANN study are satisfactory. These results have been incorporated in the SBDS software for operator assistance. A few important outputs of the SBDS have been discussed in this paper.

  9. Thermal-hydraulic calculation methods for transients and accidents of the reactor cooling system under special consideration of multi-dimensional flows (ATHLET, FLUBOX, CFX). Final report; Thermohydraulische Rechenmethoden zu Transienten und Stoerfaellen im Reaktorkuehlkreislauf unter besonderer Beruecksichtigung mehrdimensionaler Stroemungen (ATHLET, FLUBOX, CFX). Abschlussbericht

    Glaeser, H.; Graf, U.; Herb, J.; and others


    The project RS1184 „Thermal-hydraulic Calculation Methods for Transients and Accidents of the Reactor Cooling System Under Special Consideration of Multi-Dimensional Flows (ATHLET, FLUBOX, CFX)'' consists of four work packages: 1. Further development of the computer code ATHLET 2. Termination of FLUBOX development and development of an ATHLET-internal 3D module 3. Coupling of ATHLET and CFD code ANSYS CFX as well as CFX model development to simulate three-dimensional flows in the reactor coolant system 4. Prediction capability of computer code ATHLET. One of the superior objectives of the project is to improve the prediction capability of the thermal-hydraulic system code ATHLET, including the simulation of multi-dimensional flow in the reactor vessel. The constitutive equations in ATHLET, especially the momentum equations in ATHLET, are written in one-dimensional form. It was planned to develop the 2D/3D module FLUBOX further and couple it with ATHLET. Due to reasons given in chapter 3, the FLUBOX development was terminated. Instead, the decision was made to develop a fast running internal ATHLET-module. The 2D/3D equations for ATHLET have been derived and were implemented. That strategy allows using all ATHLET models and the ATHLET code structure. An additional advantage is that different numerical schemes of different codes, and consequently a loss of efficiency, will be avoided. A second possibility is the coupling of ATHLET with the CFD code ANSYS CFX. Such a coupled code system will be used in those cases when a part of the simulation area is needed to be calculated with high resolution. Such a detailed modelling cannot be provided by ATHLET-3D. A complete representation of the cooling system by a CFD code cannot be performed due to calculation time. In order to calculate the complete system behavior still with ATHLET, that part to be investigated in more detail, will be replaced by a CFX model. Several new models and improvements of existing

  10. Transient Ischemic Attack

    Full Text Available Transient Ischemic Attack TIA , or transient ischemic attack, is a "mini stroke" that occurs when a blood ... The only difference between a stroke and TIA is that with TIA the blockage is transient (temporary). ...

  11. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  12. Time-domain parameter identification of aeroelastic loads by forced-vibration method for response of flexible structures subject to transient wind

    Cao, Bochao

    Slender structures representing civil, mechanical and aerospace systems such as long-span bridges, high-rise buildings, stay cables, power-line cables, high light mast poles, crane-booms and aircraft wings could experience vortex-induced and buffeting excitations below their design wind speeds and divergent self-excited oscillations (flutter) beyond a critical wind speed because these are flexible. Traditional linear aerodynamic theories that are routinely applied for their response prediction are not valid in the galloping, or near-flutter regime, where large-amplitude vibrations could occur and during non-stationary and transient wind excitations that occur, for example, during hurricanes, thunderstorms and gust fronts. The linear aerodynamic load formulation for lift, drag and moment are expressed in terms of aerodynamic functions in frequency domain that are valid for straight-line winds which are stationary or weakly-stationary. Application of the frequency domain formulation is restricted from use in the nonlinear and transient domain because these are valid for linear models and stationary wind. The time-domain aerodynamic force formulations are suitable for finite element modeling, feedback-dependent structural control mechanism, fatigue-life prediction, and above all modeling of transient structural behavior during non-stationary wind phenomena. This has motivated the developing of time-domain models of aerodynamic loads that are in parallel to the existing frequency-dependent models. Parameters defining these time-domain models can be now extracted from wind tunnel tests, for example, the Rational Function Coefficients defining the self-excited wind loads can be extracted using section model tests using the free vibration technique. However, the free vibration method has some limitations because it is difficult to apply at high wind speeds, in turbulent wind environment, or on unstable cross sections with negative aerodynamic damping. In the current

  13. Nuclear Reactor Engineering Analysis Laboratory

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez


    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  14. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation. [PWR and BWR



    This portion of the RELAP4/MOD5 User's Manual presents the details of setting up and entering the reactor model to be evaluated. The input card format and arrangement is presented in depth, including not only cards for data but also those for editing and restarting. Problem initalization including pressure distribution and energy balance is discussed. A section entitled ''User Guidelines'' is included to provide modeling recommendations, analysis and verification techniques, and computational difficulty resolution. The section is concluded with a discussion of the computer output form and format.

  15. Transient global amnesia mimics: Transient epileptic amnesia

    Nicolas Nicastro


    Full Text Available We describe the case of a 79-year-old patient referred for suspected transient global amnesia, after an episode of anterograde amnesia which lasted 90 min. An EEG, performed after the episode, showed bilateral temporal electrographic seizures, orienting the diagnosis toward a transient epileptic amnesia. Transient epileptic amnesia is defined by temporal lobe epilepsy characterized by recurrent transient amnestic episodes of 30–90 min in duration, sometimes associated with olfactory hallucinations or oral automatisms. Response to antiepileptic drugs is excellent. We would like to raise awareness toward this epileptic amnesia when facing atypical or recurrent transient amnestic episodes.

  16. Bioconversion reactor

    McCarty, Perry L.; Bachmann, Andre


    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  17. Individual frequency-hopping radio identification method based on transient characteris-tics of frequency domain%基于频域瞬时特征的跳频电台个体识别方法

    顾晨辉; 王伦文


    The radio communication signals usually turn out to be some fine character differences. In this paper, a method based on the box-counting dimensions and the maximal Lyapunov exponents of the individual FH radio is presented according to the fine character differences. Firstly, the transient frequency of the FH signals is extracted based on the improved Prony algorithm. Secondly, the transient characteristics including the maximal Lyapunov exponents and the box dimensions are computed. Finally, individual identification of the different FH radio based on the method of the Constructive Neural Network is realized. The ex-perimental results have shown that this method is efficiency.%通信电台发射的信号通常表现出一定的细微特征差异,针对这种细微特征差异,提出了基于最大Lyapunov指数和盒维数的跳频电台个体识别方法。基于改进的Prony算法,提取样本信号跳变时刻的瞬时频率,分离并定量计算其最大Lyapunov指数和盒维数等瞬时特征,采用基于构造型神经网络的分类方法实现不同跳频电台的个体识别。实际数据的实验结果验证了算法的有效性。

  18. Identification of valid reference genes for the normalization of RT-qPCR expression studies in human breast cancer cell lines treated with and without transient transfection.

    Lin-Lin Liu

    Full Text Available Reverse transcription-quantitative polymerase chain reaction (RT-qPCR is a powerful technique for examining gene expression changes during tumorigenesis. Target gene expression is generally normalized by a stably expressed endogenous reference gene; however, reference gene expression may differ among tissues under various circumstances. Because no valid reference genes have been documented for human breast cancer cell lines containing different cancer subtypes treated with transient transfection, we identified appropriate and reliable reference genes from thirteen candidates in a panel of 10 normal and cancerous human breast cell lines under experimental conditions with/without transfection treatments with two transfection reagents. Reference gene expression stability was calculated using four algorithms (geNorm, NormFinder, BestKeeper and comparative delta Ct, and the recommended comprehensive ranking was provided using geometric means of the ranking values using the RefFinder tool. GeNorm analysis revealed that two reference genes should be sufficient for all cases in this study. A stability analysis suggests that 18S rRNA-ACTB is the best reference gene combination across all cell lines; ACTB-GAPDH is best for basal breast cancer cell lines; and HSPCB-ACTB is best for ER+ breast cancer cells. After transfection, the stability ranking of the reference gene fluctuated, especially with Lipofectamine 2000 transfection reagent in two subtypes of basal and ER+ breast cell lines. Comparisons of relative target gene (HER2 expression revealed different expressional patterns depending on the reference genes used for normalization. We suggest that identifying the most stable and suitable reference genes is critical for studying specific cell lines under certain circumstances.

  19. Identification of Non-stationary Excitation and Analysis of Transient Radiation Noise on Steering Engine%舵机非稳态激励识别及瞬态声辐射分析

    林长刚; 邹明松; 焦慧锋; 刘朋


    When an underwater vehicle changes directions by the steering engine, the opening pro-cess of steering engine generates a high transient radiation noise. So before a steering engine is in-stalled in an underwater vehicle, it is necessary to estimate transient radiation noise of the steering engine. In this paper, based on the vibration test of the steering engine for an underwater vehicle, the transient maximum and mean excitation forces acting on the positions of connections between steering engine and experimental setup are calculated indirectly by least square method of load i-dentification in frequency domain and STFT signal processing method. In addition, the accuracy and feasibility of results are verified. Cylindrical shells with ribs are basic structures which are used in underwater vehicles. By taking excitation forces results as an approximate input, a cylindrical shell whose two tips are simply supported is chosen as calculation model of a cabin of underwater vehi-cle, and the transient maximum and mean radiation noise of shells are calculated separately to esti-mate the steering engine’s characteristics of noise sources. A non-stationary STFT signal processing method to evaluate transient radiation noise of steering engine is provided.%当水下航行器利用舵机改变航向时,舵机的开启会产生很大的瞬态辐射噪声。因此,在舵机安装之前,有必要对舵机的源特性进行评估。文章基于某水下航行器的舵机振动测试试验,应用频域载荷识别中的最小二乘法方法和短时傅里叶变换的信号处理技术,对舵机操舵过程中舵机与试验台架连接的机脚点处的瞬态最大激励力和整体平均激励力分别进行了间接估算,并将估算结果的准确性和可行性进行了验证。考虑到加肋圆柱壳是水下航行器的基本结构形式,文中将估算的激励力结果作为辐射声计算的近似输入,以一个两端简支的加肋圆柱壳体作

  20. Application of an asymmetric helical tube reactor for fast identification of gene transcripts of pathogenic viruses by micro flow-through PCR.

    Hartung, R; Brösing, A; Sczcepankiewicz, G; Liebert, U; Häfner, N; Dürst, M; Felbel, J; Lassner, D; Köhler, J M


    We have established a fast PCR-based micro flow-through process consisting of a helical constructed tube reactor. By this approach we can detect transcripts of measles and human papilloma virus (HPV) by continuous flow allowing for reverse transcription (RT) and amplification of cDNA. The micro reaction system consisted of two columnar reactors for thermostating the different reaction zones of the RT process and the amplification. The PCR reactor was built by asymmetric heating sections thus realizing different residence times and optimal conditions for denaturation, annealing and elongation. The system concept is based on low electrical power consumption (50-120 W) and is suited for portable diagnostic applications. The samples were applied in form of micro fluidic segments with single volumes between 65 and 130 nL injected into an inert carrier liquid inside a Teflon FEP tube with an inner diameter of 0.5 mm. Optimal amplification for template lengths of 292 bp (lambda-DNA), 127 bp (measles virus) and 95 bp (HPV) was achieved by maximal cycle times of 75 s.

  1. Dynamic model of Fast Breeder Test Reactor

    Vaidyanathan, G., E-mail: [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Kasinathan, N.; Velusamy, K. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)


    Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.

  2. Sonochemical Reactors.

    Gogate, Parag R; Patil, Pankaj N


    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  3. Current interruption transients calculation

    Peelo, David F


    Provides an original, detailed and practical description of current interruption transients, origins, and the circuits involved, and how they can be calculated Current Interruption Transients Calculationis a comprehensive resource for the understanding, calculation and analysis of the transient recovery voltages (TRVs) and related re-ignition or re-striking transients associated with fault current interruption and the switching of inductive and capacitive load currents in circuits. This book provides an original, detailed and practical description of current interruption transients, origins,

  4. Coupled high fidelity thermal hydraulics and neutronics for reactor safety simulations

    Vincent A. Mousseau; Hongbin Zhang; Haihua Zhao


    This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un’SCRAM’ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role.

  5. Online stress corrosion crack and fatigue usages factor monitoring and prognostics in light water reactor components: Probabilistic modeling, system identification and data fusion based big data analytics approach

    Mohanty, Subhasish M. [Argonne National Lab. (ANL), Argonne, IL (United States); Jagielo, Bryan J. [Argonne National Lab. (ANL), Argonne, IL (United States); Oakland Univ., Rochester, MI (United States); Iverson, William I. [Argonne National Lab. (ANL), Argonne, IL (United States); Univ. of Illinois at Urbana-Champaign, Champaign, IL (United States); Bhan, Chi Bum [Argonne National Lab. (ANL), Argonne, IL (United States); Pusan National Univ., Busan (Korea, Republic of); Soppet, William S. [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin M. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken N. [Argonne National Lab. (ANL), Argonne, IL (United States)


    Nuclear reactors in the United States account for roughly 20% of the nation's total electric energy generation, and maintaining their safety in regards to key component structural integrity is critical not only for long term use of such plants but also for the safety of personnel and the public living around the plant. Early detection of damage signature such as of stress corrosion cracking, thermal-mechanical loading related material degradation in safety-critical components is a necessary requirement for long-term and safe operation of nuclear power plant systems.

  6. Numerical Simulation of Thermal Stratification in Pressurized Water Reactor Pressurizer Surge Line under Transient Condition%瞬态工况下压水堆稳压器波动管热分层现象数值模拟

    郭超; 温丽晶; 刘宇生; 张盼; 马帅


    T he thermal stratification under transient condition in pressurizer surge line of Qinshan Phase Ⅱ extension nuclear power project (2 × 650 MW PWR generator 4) was investigated by computational fluid dynamics program ANSYS/CFX .The whole and cross‐sectional thermal stratification transient analysis models for the pressurizer surge line were established ,and the heat stratified flow and heat transfer of the surge line were studied .The way of temperature growth is different between high‐and low‐temperature fluid layers in the same cross section . T he fluid temperature distribution has great difference in different cross sections , but the temperature difference first increases and then decreases in every cross section .The research results can provide a basis for subsequent analysis of thermal stress and lifespan .%利用计算流体动力学软件ANSYS/CFX ,对秦山核电二期扩建工程2×650 MW压水堆核电站四号机组核岛厂房的稳压器波动管进行了三维全尺寸非稳态计算。建立了波动管整体和不同截面的热分层瞬态,对管内热分层流动与换热进行了研究。研究结果表明:同一截面内高温层流体和低温层流体的升温方式不同;不同截面位置的管内流动温度分布特性差别较大,但均呈现分层流体温差先增大后减小的趋势。计算结果可为后续波动管热应力分析及寿命评价提供一定基础。

  7. Microprocessor tester for the treat upgrade reactor trip system

    Lenkszus, F.R.; Bucher, R.G.


    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.

  8. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)


    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

  9. Remark on the applicability of Campbelling techniques for transient signals

    Elter, Zs., E-mail: [Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Göteborg (Sweden); CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 Saint-Paul-lez-Durance (France); Pázsit, I. [Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Göteborg (Sweden); Jammes, C. [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 Saint-Paul-lez-Durance (France)


    The signals of fission chambers are traditionally evaluated with Campbelling methods at medium count rates. Lately there has been a growing interest in the extension of Campbelling methods to cover a wider range of count rates. These methods are applied to measure the neutron flux in the stationary state of the reactor. However, there has not been any attempt to generalize these techniques for transient reactor states. This short note is devoted to a discussion of this question. It is shown through analytic and numerical calculations that for practical reasons the traditional, stationary Campbelling methods can be applied for transient scenarios as well.

  10. Alternatives Analysis for the Resumption of Transient Testing Program

    Lee Nelson


    An alternatives analysis was performed for resumption of transient testing. The analysis considered eleven alternatives – including both US international facilities. A screening process was used to identify two viable alternatives from the original eleven. In addition, the alternatives analysis includes a no action alternative as required by the National Environmental Policy Act (NEPA). The alternatives considered in this analysis included: 1. Restart the Transient Reactor Test Facility (TREAT) 2. Modify the Annular Core Research Reactor (ACRR) which includes construction of a new hot cell and installation of a new hodoscope. 3. No Action

  11. Mechanical modelling of transient- to- failure SFR fuel cladding

    Feria, F.; Herranz, L. E.


    The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

  12. Transient drainage summary report



    This report summarizes the history of transient drainage issues on the Uranium Mill Tailings Remedial Action (UMTRA) Project. It defines and describes the UMTRA Project disposal cell transient drainage process and chronicles UMTRA Project treatment of the transient drainage phenomenon. Section 4.0 includes a conceptual cross section of each UMTRA Project disposal site and summarizes design and construction information, the ground water protection strategy, and the potential for transient drainage.

  13. Localization and identification of transient power quality disturbances based on generalized S-transform%基于广义S变换的暂态电能质量扰动定位与识别

    刘奇; 周雒维; 卢伟国


    S-transform has fixed time-frequency resolution, leading to poor results of localizing transient power quality disturbances. A new method to localize the disturbances is proposed based on generalized S-transform. The method detects mutation peak in the high frequency time-amplitude curve, in order to increase localization accuracy. At first, modulus time-frequency matrix is calculated by generalized S-transform, then the disturbances' start-stop time is localized using the high frequency time-amplitude curve, and four identification features are extracted according to maximum frequency spectrum curve, fundamental frequency amplitude curve and the localization results. At last, automatic classification of disturbance signals is performed by use of a rule-based decision tree. Simulation results show that the proposed localization method is simple and intuitive, with high accuracy. The number of identification features is small and they are effective with good classification results.This work is supported by National Natural Science Foundation of China (No. 51077137).%S变换由于时频分辨率固定,从而导致定位暂态电能质量扰动的效果差.提出一种基于广义S变换的扰动定位新方法,利用高频处时间幅值曲线的突变点峰值进行定位检测,以提高扰动的定位精度.首先通过广义S变换得到扰动信号的模时频矩阵,然后利用高频处时间幅值曲线定位扰动的起止时刻,再根据最大频谱曲线、基频幅值曲线与定位结果提取四个识别特征量,最后基于分类规则树方法实现扰动信号的自动分类.仿真结果表明,所提出的定位方法简单直观,精度较高;提取的识别特征量少而有效,分类效果良好.

  14. Hybrid adsorptive membrane reactor

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)


    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  15. D and DR Reactors

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  16. Hybrid adsorptive membrane reactor

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.


    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  17. Transient Diagnosis and Prognosis for Secondary System in Nuclear Power Plants

    Sangjun Park


    Full Text Available This paper introduces the development of a transient monitoring system to detect the early stage of a transient, to identify the type of the transient scenario, and to inform an operator with the remaining time to turbine trip when there is no operator's relevant control. This study focused on the transients originating from a secondary system in nuclear power plants (NPPs, because the secondary system was recognized to be a more dominant factor to make unplanned turbine-generator trips which can ultimately result in reactor trips. In order to make the proposed methodology practical forward, all the transient scenarios registered in a simulator of a 1,000 MWe pressurized water reactor were archived in the transient pattern database. The transient patterns show plant behavior until turbine-generator trip when there is no operator's intervention. Meanwhile, the operating data periodically captured from a plant computer is compared with an individual transient pattern in the database and a highly matched section among the transient patterns enables isolation of the type of transient and prediction of the expected remaining time to trip. The transient pattern database consists of hundreds of variables, so it is difficult to speedily compare patterns and to draw a conclusion in a timely manner. The transient pattern database and the operating data are, therefore, converted into a smaller dimension using the principal component analysis (PCA. This paper describes the process of constructing the transient pattern database, dealing with principal components, and optimizing similarity measures.

  18. Transient diagnosis and prognosis for secondary system in nuclear power plants

    Park, Sang Jin; Park, Jin Kyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Heo, Gyun Young [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)


    This paper introduces the development of a transient monitoring system to detect the early stage of a transient, to identify the type of the transient scenario, and to inform an operator with the remaining time to turbine trip when there is no operator's relevant control. This study focused on the transients originating from a secondary system in nuclear power plants (NPPs), because the secondary system was recognized to be a more dominant factor to make unplanned turbine-generator trips which can ultimately result in reactor trips. In order to make the proposed methodology practical forward, all the transient scenarios registered in a simulator of a 1,000 MWe pressurized water reactor were archived in the transient pattern database. The transient patterns show plant behavior until turbine-generator trip when there is no operator's intervention. Meanwhile, the operating data periodically captured from a plant computer is compared with an individual transient pattern in the database and a highly matched section among the transient patterns enables isolation of the type of transient and prediction of the expected remaining time to trip. The transient pattern database consists of hundreds of variables, so it is difficult to speedily compare patterns and to draw a conclusion in a timely manner. The transient pattern database and the operating data are, therefore, converted into a smaller dimension using the principal component analysis (PCA). This paper describes the process of constructing the transient pattern database, dealing with principal components, and optimizing similarity measures.


    Hakan Ozaltun & Herman Shen


    This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

  20. A laboratory flow reactor with gas particle separation and on-line MS/MS for product identification in atmospherically important reactions

    J. F. Bennett


    Full Text Available A system to study the gas and particle phase products from gas phase hydrocarbon oxidation is described. It consists of a gas phase photochemical flow reactor followed by a diffusion membrane denuder to remove gases from the reacted products, or a filter to remove the particles. Chemical analysis is performed by an atmospheric pressure chemical ionization (APCI triple quadrupole mass spectrometer. A diffusion membrane denuder is shown to remove trace gases to below detectable limits so the particle phase can be studied. The system was tested by examining the products of the oxidation of m-xylene initiated by HO radicals. Dimethylphenol was observed in both the gas and particle phases although individual isomers could not be identified. Two furanone isomers, 5-methyl-2(3Hfuranone and 3-methyl-2(5Hfuranone were identified in the particulate phase, but the isobaric product 2,5 furandione was not observed. One isomer of dimethyl-nitrophenol was identified in the particle phase but not in the gas phase.

  1. A laboratory flow reactor with gas particle separation and on-line MS/MS for product identification in atmospherically important reactions

    J. F. Bennett


    Full Text Available A system to study the gas and particle phase products from gas phase hydrocarbon oxidation is described. It consists of a gas phase photochemical flow reactor followed by a diffusion membrane denuder to remove gases from the reacted products, or a filter to remove the particles. Chemical analysis is performed by an atmospheric pressure chemical ionization (APCI triple quadrupole mass spectrometer. A diffusion membrane denuder is shown to remove trace gases to below detectable limits so the particle phase can be studied. The system was tested by examining the products of the oxidation of m-xylene initiated by HO radicals. Dimethylphenol was observed in both the gas and particle phases although individual isomers could not be identified. Two furanone isomers, 5-methyl-2(3Hfuranone and 3-methyl-2(5Hfuranone were identified in the particulate phase, but the isobaric product 2,5 furandione was not observed. One isomer of dimethyl-nitrophenol was identified in the particle phase but not in the gas phase.

  2. Improved Generalized Predictive Control Algorithm with Offline and Online Identification and Its Application to Fixed Bed Reactor%离线辨识和在线辨识相结合的广义预测控制算法在固定床反应器温度控制中的应用

    余世明; 王海青


    An improved generalized predictive control algorithm is presented in this paper by incorporating offlineidentification into online identification. Unlike the existing generalized predictive control algorithms, the proposedapproach divides parameters of a predictive model into the time invariant and time-varying ones, which are treatedrespectively by offiine and online identification algorithms. Therefore, both the reliability and accuracy of thepredictive model are improved. Two simulation examples of control of a fixed bed reactor show that this newalgorithm is not only reliable and stable in the case of uncertainties and abnormal disturbances, but also adaptableto slow time varying processes.

  3. Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

    Fletcher, C. D.

    The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes.

  4. Transient absorption probe of intermolecular triplet excimer of naphthalene in fluid solutions: Identification of the species based on comparison to the intramolecular triplet excimers of covalently-linked dimers

    Wang, X.; Kofron, W.G.; Kong, S.; Rajesh, C.S.; Modarelli, D.A.; Lim, E.C.


    The authors report here the observation of the laser-induced transient absorption spectrum of intermolecular triplet excimers of naphthalene in fluid solution. This assignment is confirmed by comparison to the transient absorption spectra of the intramolecular triplet excimers of covalently linked dimers of naphthalene and quinoxaline.

  5. Transient study for a fully compensated export cable in large offshore wind farms

    Arana Aristi, Iván; Johnsen, I; Holbøll, Joachim


    This paper presents the results of switching transient simulations using PSCAD on Walney Offshore Windfarm 1. The model of WOW1 was created based on as-build information from the main components like cables, reactors, filters and switchgear. Energizing operations of the export cable, reactor...

  6. Nuclear reactor neutron shielding

    Speaker, Daniel P; Neeley, Gary W; Inman, James B


    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  7. Reactor and method of operation

    Wheeler, John A.


    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  8. Modeling of Reactor Kinetics and Dynamics

    Matthew Johnson; Scott Lucas; Pavel Tsvetkov


    In order to model a full fuel cycle in a nuclear reactor, it is necessary to simulate the short time-scale kinetic behavior of the reactor as well as the long time-scale dynamics that occur with fuel burnup. The former is modeled using the point kinetics equations, while the latter is modeled by coupling fuel burnup equations with the kinetics equations. When the equations are solved simultaneously with a nonlinear equation solver, the end result is a code with the unique capability of modeling transients at any time during a fuel cycle.

  9. Linear inverse problem of the reactor dynamics

    Volkov, N. P.


    The aim of this work is the study transient processes in nuclear reactors. The mathematical model of the reactor dynamics excluding reverse thermal coupling is investigated. This model is described by a system of integral-differential equations, consisting of a non-stationary anisotropic multispeed kinetic transport equation and a delayed neutron balance equation. An inverse problem was formulated to determine the stationary part of the function source along with the solution of the direct problem. The author obtained sufficient conditions for the existence and uniqueness of a generalized solution of this inverse problem.

  10. Power distribution control of CANDU reactors based on modal representation of reactor kinetics

    Xia, Lingzhi, E-mail: [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Jiang, Jin, E-mail: [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Luxat, John C., E-mail: [Department of Engineering Physics, McMaster University, Hamilton, Ontario L8S 4L7 (Canada)


    Highlights: • Linearization of the modal synthesis model of neutronic kinetic equations for CANDU reactors. • Validation of the linearized dynamic model through closed-loop simulations by using the reactor regulating system. • Design of a LQR state feedback controller for CANDU core power distribution control. • Comparison of the results of this new controller against those of the conventional reactor regulation system. - Abstract: Modal synthesis representation of a neutronic kinetic model for a CANDU reactor core has been utilized in the analysis and synthesis for reactor control systems. Among all the mode shapes, the fundamental mode of the power distribution, which also coincides with the desired reactor power distribution during operation, is used in the control system design. The nonlinear modal models are linearized around desired operating points. Based on the linearized model, linear quadratic regulator (LQR) control approach is used to synthesize a state feedback controller. The performance of this controller has been evaluated by using the original nonlinear models under load-following conditions. It has been demonstrated that the proposed reactor control system can produce more uniform power distribution than the traditional reactor regulation systems (RRS); in particular, it is more effective in compensating the Xenon induced transients.

  11. Numerical simulation and analysis of axial instabilities occurrence and development in turbomachines. Application to a break transient in a helium nuclear reactor; Simulation numerique et analyse du declenchement et du developpement des instabilites axiales dans les turbomachines: application a un transitoire de breche dans un reacteur nucleaire a helium

    Tauveron, N


    The subject of the present work was to develop models able to simulate axial instabilities occurrence and development in multistage turbomachines. The construction of a 1D unsteady axisymmetric model of internal flow in a turbomachine (at the scale of the row) has followed different steps: generation of steady correlations, adapted to different regimes (off-design conditions, low mass flowrate, negative mass flow rate); building of a model able to describe transient behaviour; use of implicit time schemes adapted to long transients; validation of the model in comparison of experimental investigations, measurements and numerical results from the bibliography. This model is integrated in a numerical tool, which has the capacity to describe the gas dynamics in a complete circuit containing different elements (ducts, valves, plenums). Thus, the complete model can represent the coupling between local and global phenomena, which is a very important mechanism in axial instability occurrence and development. An elementary theory has also been developed, based on a generalisation of Greitzer's model. These models, which were validated on various configurations, have provided complementary elements for the validation of the complete model. They have also allowed a more comprehensive description of physical phenomena at stake in instability occurrence and development by quantifying various effects (inertia, compressibility, performance levels) and underlying the main phenomena (in particular the collapse and recovery kinetics of the plenum), which were the only retained in the final elementary theory. The models were first applied to academic configurations (compression system), and then to an innovative industrial project: a helium cooled fast nuclear reactor with a Brayton cycle. The use of the models have brought comprehensive elements to surge occurrence due to a break event. It has been shown that surge occurrence is highly dependent of break location and that surge

  12. Transient Voltage Recorder

    Medelius, Pedro J. (Inventor); Simpson, Howard J. (Inventor)


    A voltage transient recorder can detect lightning induced transient voltages. The recorder detects a lightning induced transient voltage and adjusts input amplifiers to accurately record transient voltage magnitudes. The recorder stores voltage data from numerous monitored channels, or devices. The data is time stamped and can be output in real time, or stored for later retrieval. The transient recorder, in one embodiment, includes an analog-to-digital converter and a voltage threshold detector. When an input voltage exceeds a pre-determined voltage threshold, the recorder stores the incoming voltage magnitude and time of arrival. The recorder also determines if its input amplifier circuits clip the incoming signal or if the incoming signal is too low. If the input data is clipped or too low, the recorder adjusts the gain of the amplifier circuits to accurately acquire subsequent components of the lightning induced transients.



    definitions of what it means to be transient. The purpose of this technical report is to provide a background of the issues related to transient...In this section, we will attempt to identify these parameters and provide preliminary definitions for categories of transience behavior. Transient...programmable lifetimes. Ideally , such materials with expiration dates will deconstruct themselves into harmless and invisible remnants. The technology base

  14. Validation of coupled Relap5-3D code in the analysis of RBMK-1500 specific transients

    Evaldas, Bubelis; Algirdas, Kaliatka; Eugenijus, Uspuras [Lithuanian Energy Institute, Kaunas (Lithuania)


    This paper deals with the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. These transients include: measurements of void and fast power reactivity coefficients, change of graphite cooling conditions and reactor power reduction transients. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is very unique and important from the gap between fuel channel and the graphite bricks model validation point of view. The measurement results, obtained during this transient, allowed to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. Reactor power reduction is a regular operation procedure during the entire lifetime of the reactor. In all cases it starts by either a scram or a power reduction signal activation by the reactor control and protection system or by an operator. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modelling of the neutronic processes taking place in RBMK- 1500 reactor core. And finally, the performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500

  15. A spatial kinetic model for simulating VVER-1000 start-up transient

    Kashi, Samira [Department of Nuclear Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Moghaddam, Nader Maleki, E-mail: [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology, Tehran (Iran, Islamic Republic of); Shahriari, Majid [Department of Nuclear Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)


    Research highlights: > A spatial kinetic model of a VVER-1000 reactor core is presented. > The reactor power is tracked using the point kinetic equations from 100 W to 612 kW. > The lamped parameter approximation is used for solving the energy balance equations. > The value of reactivity related to feedback effects of core elements is calculated. > The main neutronic parameters during the transient are calculated. - Abstract: An accurate prediction of reactor core behavior in transients depends on how much it could be possible to exactly determine the thermal feedbacks of the core elements such as fuel, clad and coolant. In short time transients, results of these feedbacks directly affect the reactor power and determine the reactor response. Such transients are commonly happened during the start-up process which makes it necessary to carefully evaluate the detail of process. Hence this research evaluates a short time transient occurring during the start up of VVER-1000 reactor. The reactor power was tracked using the point kinetic equations from HZP state (100 W) to 612 kW. Final power (612 kW) was achieved by withdrawing control rods and resultant excess reactivity was set into dynamic equations to calculate the reactor power. Since reactivity is the most important part in the point kinetic equations, using a Lumped Parameter (LP) approximation, energy balance equations were solved in different zones of the core. After determining temperature and total reactivity related to feedbacks in each time step, the exact value of reactivity is obtained and is inserted into point kinetic equations. In reactor core each zone has a specific temperature and its corresponding thermal feedback. To decrease the effects of point kinetic approximations, these partial feedbacks in different zones are superposed to show an accurate model of reactor core dynamics. In this manner the reactor point kinetic can be extended to the whole reactor core which means 'Reactor spatial

  16. The Zwicky Transient Facility

    Bellm, Eric C


    The Zwicky Transient Facility (ZTF) is a next-generation optical synoptic survey that builds on the experience and infrastructure of the Palomar Transient Factory (PTF). Using a new 47 deg$^2$ survey camera, ZTF will survey more than an order of magnitude faster than PTF to discover rare transients and variables. I describe the survey and the camera design. Searches for young supernovae, fast transients, counterparts to gravitational-wave detections, and rare variables will benefit from ZTF's high cadence, wide area survey.

  17. Uncertainty quantification approaches for advanced reactor analyses.

    Briggs, L. L.; Nuclear Engineering Division


    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  18. Future Transient Testing of Advanced Fuels

    Jon Carmack


    The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

  19. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Bragg-Sitton, Shannon M.


    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .

  20. Effects of transient temperature conditions on the divergence of activated sludge bacterial community structure and function.

    Nadarajah, Nalina; Allen, D Grant; Fulthorpe, Roberta R


    The effect of temperature fluctuations on bacterial community structure and function in lab-scale sequencing batch reactors treating bleached kraft mill effluent was investigated. An increase in temperature from 30 to 45 degrees C caused shifts in both bacterial community structure and function. Triplicate reactors were highly similar for 40 days following startup. After the temperature shift, their community structure and function started to diverge from each other and from the control. A multi-response permutation procedure confirmed that the variability in community structure between transient and control reactors were greater than that among the triplicate transient reactors. The fact that these disturbances manifest themselves in different ways in apparently identical reactors suggests a high degree of variability between replicate systems.

  1. Multiphysics simulation of fast transients with the FINIX fuel behaviour module

    Ikonen Timo


    Full Text Available FINIX is a recently developed fuel behaviour module that is designed to provide “simple but sufficient” descriptions of the most essential fuel behaviour phenomena in multiphysics simulations. In such simulations, it is possible to obtain significant improvement in the feedback to neutronics or thermal hydraulics modelling even with a relatively simple fuel performance model. In this work, FINIX is used as an internal fuel behaviour module both in reactor physics and in reactor dynamics codes to simulate coupled behaviour in fast transient scenarios. With the Monte Carlo reactor physics code Serpent we model a prompt transient in a VVER-1000 pin cell, and with the reactor dynamics code HEXTRAN, a control rod ejection accident in a VVER-440 reactor.


    Artur Wodołażski


    Full Text Available This paper presents a model of dynamics control for continuous stirred tank reactor (CSTR in methanol synthesis in a three-phase system. The reactor simulation was carried out for steady and transient state. Efficiency ratio to achieve maximum performance of the product per reactor unit volume was calculated. Reactor dynamics simulation in closed loop allowed to received data for tuning PID controller (proportional-integral-derivative. The results of the regulation process allow to receive data for optimum reactor production capacity, along with local hot spots eliminations or temperature runaway.

  3. Analysis of an Earthquake-Initiated-Transient in a PBR

    A. M. Ougouag; J. Ortensi; H. Hiruta


    One of the Design Basis Accidents (DBA) for a Pebble Bed Reactor has been identified as the “Safe shutdown earthquake with core conduction cooling to passive mode of Reactor Cavity Cooling System.” A new methodology to analyze this particular DBA has been developed at the Idaho National Laboratory (INL). During the seismic event the reactor core experiences the densification of the pebbles, which produce small reactivity insertions due to the effective fuel densification. In addition, a decrease in the active core height results in the relative withdrawal of the control rods, which are assumed to remain stationary during the transient. The methodology relies on the dynamic re-meshing of the core during the transient to capture the local packing fraction changes and their corresponding effects on temperature and reactivity. The core re-meshing methodology is based on the velocity profiles of the pebbles in the core, which were obtained with the INL’s pebble mechanics code PEBBLES. The methodology has been added to the coupled code system CYNOD-THERMIX-KONVEK. The reactor power calculation is further improved with the use of the new advanced TRISO fuel model to better approximate the temperatures in the fuel kernels. During the transient the core is brought back to a safe condition by the strong Doppler feedback from local temperature increases.

  4. Reactor Physics Programme

    De Raedt, C


    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  5. Transient Ischemic Attack

    Full Text Available ... TIA , or transient ischemic attack, is a "mini stroke" that occurs when a blood clot blocks an ... a short time. The only difference between a stroke and TIA is that with TIA the blockage ...

  6. Transient Ischemic Attack

    Full Text Available ... TIA , or transient ischemic attack, is a "mini stroke" that occurs when a blood clot blocks an ... a short time. The only difference between a stroke and TIA is that with TIA the blockage ...

  7. Searches for radio transients

    Bhat, N D R


    Exploration of the transient Universe is an exciting and fast-emerging area within radio astronomy. Known transient phenomena range in time scales from sub-nanoseconds to years or longer, thus spanning a huge range in time domain and hinting a rich diversity in their underlying physical processes. Transient phenomena are likely locations of explosive or dynamic events and they offer tremendous potential to uncover new physics and astrophysics. A number of upcoming next-generation radio facilities and recent advances in computing and instrumentation have provided a much needed impetus for this field which has remained a relatively uncharted territory for the past several decades. In this paper we focus mainly on the class of phenomena that occur on very short time scales (i.e. from $\\sim$ milliseconds to $\\sim$ nanoseconds), known as {\\it fast transients}, the detections of which involve considerable signal processing and data management challenges, given the high time and frequency resolutions required in the...

  8. Transient Ischemic Attack

    Full Text Available ... Ischemic Attack TIA , or transient ischemic attack, is a "mini stroke" that occurs when a blood clot blocks an artery for a short time. The only difference between a stroke ...

  9. Advanced Demonstration and Test Reactor Options Study

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)


    Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy

  10. A transient, quadratic nodal method for triangular-Z geometry

    DeLorey, T.F.


    Many systematically-derived nodal methods have been developed for Cartesian geometry due to the extensive interest in Light Water Reactors. These methods typically model the transverse-integrated flux as either an analytic or low order polynomial function of position within the node. Recently, quadratic nodal methods have been developed for R-Z and hexagonal geometry. A static and transient quadratic nodal method is developed for triangular-Z geometry. This development is particularly challenging because the quadratic expansion in each node must be performed between the node faces and the triangular points. As a consequence, in the 2-D plane, the flux and current at the points of the triangles must be treated. Quadratic nodal equations are solved using a non-linear iteration scheme, which utilizes the corrected, mesh-centered finite difference equations, and forces these equations to match the quadratic equations by computing discontinuity factors during the solution. Transient nodal equations are solved using the improved quasi-static method, which has been shown to be a very efficient solution method for transient problems. Several static problems are used to compare the quadratic nodal method to the Coarse Mesh Finite Difference (CMFD) method. The quadratic method is shown to give more accurate node-averaged fluxes. However, it appears that the method has difficulty predicting node leakages near reactor boundaries and severe material interfaces. The consequence is that the eigenvalue may be poorly predicted for certain reactor configurations. The transient methods are tested using a simple analytic test problem, a heterogeneous heavy water reactor benchmark problem, and three thermal hydraulic test problems. Results indicate that the transient methods have been implemented correctly.

  11. Attrition reactor system

    Scott, Charles D.; Davison, Brian H.


    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  12. Transient multivariable sensor evaluation

    Vilim, Richard B.; Heifetz, Alexander


    A method and system for performing transient multivariable sensor evaluation. The method and system includes a computer system for identifying a model form, providing training measurement data, generating a basis vector, monitoring system data from sensor, loading the system data in a non-transient memory, performing an estimation to provide desired data and comparing the system data to the desired data and outputting an alarm for a defective sensor.

  13. Transient multivariable sensor evaluation

    Vilim, Richard B.; Heifetz, Alexander


    A method and system for performing transient multivariable sensor evaluation. The method and system includes a computer system for identifying a model form, providing training measurement data, generating a basis vector, monitoring system data from sensor, loading the system data in a non-transient memory, performing an estimation to provide desired data and comparing the system data to the desired data and outputting an alarm for a defective sensor.

  14. Transient Simulation of the Multi-SERTTA Experiment with MAMMOTH

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); deHart, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States)


    This work details the MAMMOTH reactor physics simulations of the Static Environment Rodlet Transient Test Apparatus (SERTTA) conducted at Idaho National Laboratory in FY-2017. TREAT static-environment experiment vehicles are being developed to enable transient testing of Pressurized Water Reactor (PWR) type fuel specimens, including fuel concepts with enhanced accident tolerance (Accident Tolerant Fuels, ATF). The MAMMOTH simulations include point reactor kinetics as well as spatial dynamics for a temperature-limited transient. The strongly coupled multi-physics solutions of the neutron flux and temperature fields are second order accurate both in the spatial and temporal domains. MAMMOTH produces pellet stack powers that are within 1.5% of the Monte Carlo reference solutions. Some discrepancies between the MCNP model used in the design of the flux collars and the Serpent/MAMMOTH models lead to higher power and energy deposition values in Multi-SERTTA unit 1. The TREAT core results compare well with the safety case computed with point reactor kinetics in RELAP5-3D. The reactor period is 44 msec, which corresponds to a reactivity insertion of 2.685% delta k/k$. The peak core power in the spatial dynamics simulation is 431 MW, which the point kinetics model over-predicts by 12%. The pulse width at half the maximum power is 0.177 sec. Subtle transient effects are apparent at the beginning insertion in the experimental samples due to the control rod removal. Additional difference due to transient effects are observed in the sample powers and enthalpy. The time dependence of the power coupling factor (PCF) is calculated for the various fuel stacks of the Multi-SERTTA vehicle. Sample temperatures in excess of 3100 K, the melting point UO$_2$, are computed with the adiabatic heat transfer model. The planned shaped-transient might introduce additional effects that cannot be predicted with PRK models. Future modeling will be focused on the shaped-transient by improving the

  15. Dynamic simulation of a sodium-cooled fast reactor power plant

    Shinaishin, M.A.M.


    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

  16. Mitigation method of thermal transient stress by a total analysis of thermal hydraulic and structural phenomena

    Kasahara, Naoto [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Jinbo, Masakazu [Toshiba Co., Tokyo (Japan); Hosogai, Hiromi [Joyo Industry Co., Ltd., Tokai, Ibaraki (Japan)


    This study proposes a mitigation method of thermal transient loads in fast reactor components by utilizing relationships among plant system parameters and resulting thermal stresses. Conventional design procedure against thermal transient loads has two independent steps: thermal hydraulic analysis to determine conservative thermal transient conditions considering variation of the system parameters and structural analysis to check structural integrity under given conditions. On the other hand, a total analysis procedure of thermal hydraulic and structural phenomena can grasp the relationship among system parameters and thermal stresses. It enables the mitigation of thermal transient loads by adjusting system parameters. (author)

  17. Light water reactor safety

    Pershagen, B


    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  18. Nuclear reactor physics

    Stacey, Weston M


    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  19. Assessment of the enhanced DHRS configuration for MYRRHA reactor

    Bubelis, E.; Jaeger, W. [KIT, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bandini, G. [ENEA, via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Alemberti, A.; Palmero, M. [ANSALDO, Corso Perrone 25, 16152 Genova (Italy)


    Highlights: • Innovative decay heat removal system (DHRS). • Heavy liquid metal cooled reactor. • Avoiding of lead bismuth eutectic (LBE) freezing. • Numerical assessment and proof of operational principles of innovative DHRS. - Abstract: This paper deals with the assessment of an innovative decay heat removal system for the MYRRHA reactor, based on the analysis of the selected transients with two different system codes. The application to liquid metal cooled reactors has the disadvantage of adding overcooling transients to the transient spectrum. Under these circumstances, freezing of the coolant can occur if no corrective or operator actions are taken in the medium and long term. Therefore, ANSALDO Nucleare invented an enhanced decay heat removal system which avoids the risk of freezing. The numerical assessment and proof of operational principles are performed by KIT and ENEA. The simulation results show that the freezing can be avoided. Moreover, both institutions calculate similar behavior during overcooling transients. This study will help to implement the novel decay heat removal system and the overall safety philosophy of innovative reactor concepts.

  20. Development of a preliminary PIRT (Phenomena Identification and Ranking Table) of thermal-hydraulic phenomena for SMART

    Chung, Bub Dong; Lee, Won Jae; Kim, Hee Cheol; Song, Jin Ho; Sim, Suk Ku [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    The work reported in this paper identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in SMART (System-integrated Modular Advanced ReacTor) which is under development at KAERI. The result of this effort is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary Phenomena Identification and Ranking Table (PIRT) has been developed based on the experts` knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental tests needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART. 8 refs., 4 tabs (Author)

  1. Spinning fluids reactor

    Miller, Jan D; Hupka, Jan; Aranowski, Robert


    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  2. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.


    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  3. Observations of transients and pulsars with LOFAR international stations

    Serylak, Maciej; Williams, Chris; Armour, Wes


    The LOw FRequency ARray - LOFAR is a new radio telescope that is moving the science of radio pulsars and transients into a new phase. Its design places emphasis on digital hardware and flexible software instead of mechanical solutions. LOFAR observes at radio frequencies between 10 and 240 MHz where radio pulsars and many transients are expected to be brightest. Radio frequency signals emitted from these objects allow us to study the intrinsic pulsar emission and phenomena such as propagation effects through the interstellar medium. The design of LOFAR allows independent use of its stations to conduct observations of known bright objects, or wide field monitoring of transient events. One such combined software/hardware solution is called the Advanced Radio Transient Event Monitor and Identification System (ARTEMIS). It is a backend for both targeted observations and real-time searches for millisecond radio transients which uses Graphical Processing Unit (GPU) technology to remove interstellar dispersion and d...

  4. Improved statistical confirmation of margins for setpoints and transients

    Nutt, W.T. [Framatome ANP Richland, INC., WA (United States)


    Framatome ANP Richland, Inc. has developed an integrated, automated, statistical methodology for Pressurized Water Reactors (PWRs). Margins for transients and calculated trips are confirmed using several new applications of probability theory. The methods used for combining statistics reduces the conservatisms inherent in conventional methods and avoids the numerical limitations and time constraints imposed by Monte Carlo techniques. The new methodology represents the state of the art in the treatment of uncertainties for reactor protection systems. It all but eliminates concerns with the calculated trips for PWRs and by improving the margin for all transients will allow for far more aggressive peaking limits and fuel management schemes. The automated nature of the bulk of this process saves Framatome ANP time and effort, minimizes the potential for errors and makes the analysis for all cycles and plants consistent. The enhanced margins remove analytical limitations from the customer and allow for more economical operation of the plant. (authors)

  5. Study on secondary shutdown systems in Tehran research reactor

    Jalali, H.R.; Fadaei, A.H., E-mail:; Gharib, M.


    Highlights: • A study was undertaken to summarize the techniques for secondary shutdown systems (SSS). • Neutronic calculation performed for proposed systems as SSS. • Dumping the heavy water stored in the reflector vessel is capable to shut down reactor. • Neutronic and transient calculation was done for validating the selected SSS. • All calculation shown that this system has advantages in safety and neutron economy. - Abstract: One important safety aspect of any research reactor is the ability to shut down the reactor. Usually, research reactors, currently in operation, have a single shutdown system based on the simultaneous insertion of the all control rods into the reactor core through gravity. Nevertheless, the International Atomic Energy Agency currently recommends use of two shutdown systems which are fully independent from each other to guarantee secure shutdown when one of them fails. This work presents an investigative study into secondary shutdown systems, which will be an important safety component in the research reactor and will provide another alternative way to shut down the reactor emergently. As part of this project, a study was undertaken to summarize the techniques that are currently used at world-wide research reactors for recognizing available techniques to consider in research reactors. Removal of the reflector, removal of the fuels, change in critical shape of reactor core and insertion of neutron absorber between the core and reflector are selected as possible techniques in mentioned function. In the next step, a comparison is performed for these methods from neutronic aspects. Then, chosen method is studied from the transient behavior point of view. Tehran research reactor which is a 5 MW open-pool reactor selected as a case study and all calculations are carried out for it. It has 5 control rods which serve the purpose of both reactivity control and shutdown of reactor under abnormal condition. Results indicated that heavy

  6. Reactor Vessel Surveillance Program for Advanced Reactor

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo


    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  7. TREAT Transient Analysis Benchmarking for the HEU Core

    Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)


    This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used to determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term “reported” values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core’s performance.

  8. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Koroush Shirvan


    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  9. Transient lingual papillitis.

    Kornerup, Ida M; Senye, Mireya; Peters, Edmund


    A case of recurrent, clinically innocuous, but painful papules involving the tongue dorsum of a 25-year-old man is presented. The lesions were interpreted to represent a transient lingual papillitis. This a poorly understood, but benign and self-limited condition involving the tongue fungiform papillae, which does not appear to be widely recognized.

  10. Transient Heat Conduction

    Rode, Carsten


    Analytical theory of transient heat conduction.Fourier's law. General heat conducation equation. Thermal diffusivity. Biot and Fourier numbers. Lumped analysis and time constant. Semi-infinite body: fixed surface temperature, convective heat transfer at the surface, or constant surface heat flux...

  11. Transient tachypnea - newborn

    ... or reabsorbing it. The first few breaths a baby takes after delivery fill the lungs with air and help to ... goes away within 24 to 48 hours after delivery. In most cases, babies who have had transient tachypnea have no further ...

  12. On Detecting Transients

    Belanger, G


    Transient phenomena are interesting and potentially highly revealing of details about the processes under observation and study that could otherwise go unnoticed. It is therefore important to maximise the sensitivity of the method used to identify such events. In this article we present a general procedure based on the use of the likelihood function for identifying transients that is particularly suited for real-time applications because it requires no grouping or pre-processing of the data. The method is optimal in the sense that all the information that is available in the data is used in the statistical decision making process, and is suitable for a wide range of applications. We here consider those most common in astrophysics which involve searching for transient sources, events or features in images, time series, energy spectra and power spectra, and demonstrate the use of the method in the cases of a transient in a time series or in a power spectrum. We derive a fit statistic that is ideal for fitting a...

  13. The LOFAR Transients Pipeline

    Swinbank, J.; Staley, T.; Molenaar, G.; Rol, E.; Rowlinson, A.; Scheers, L.H.A.; Spreeuw, H.; Bell, M.E.; Broderick, J.; Carbone, D.; Garsden, H.; Horst, A. van der; Law, C.J.; Wise, M.W.; Breton, R.P.; Cendes, Y.; Corbel, S.; Eisloeffel, J.; Falcke, H.; Fender, R.P.; Griessmeier, J.-M.; Hessels, J.W.T.; Stappers, B.W.; Stewart, A.; Wijers, R.A.M.J.; Wijnands, R.; Zarka, P.


    Current and future astronomical survey facilities provide a remarkably rich opportunity for transient astronomy, combining unprecedented fields of view with high sensitivity and the ability to access previously unexplored wavelength regimes. This is particularly true of LOFAR, a recently-commissione

  14. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 2: Accident and Thermal Fluids Analysis PIRTs

    Ball, Sydney J [ORNL; Corradini, M. [University of Wisconsin; Fisher, Stephen Eugene [ORNL; Gauntt, R. [Sandia National Laboratories (SNL); Geffraye, G. [CEA, France; Gehin, Jess C [ORNL; Hassan, Y. [Texas A& M University; Moses, David Lewis [ORNL; Renier, John-Paul [ORNL; Schultz, R. [Idaho National Laboratory (INL); Wei, T. [Argonne National Laboratory (ANL)


    An accident, thermal fluids, and reactor physics phenomena identification and ranking process was conducted by a panel of experts on the next generation nuclear plant (NGNP) design (consideration given to both pebble-bed and prismatic gas-cooled reactor configurations). Safety-relevant phenomena, importance, and knowledge base were assessed for the following event classes: (1) normal operation (including some reactor physics aspects), (2) general loss of forced circulation (G-LOFC), (3) pressurized loss-of-forced circulation (P-LOFC), (4) depressurized loss-of-forced circulation (D-LOFC), (5) air ingress (following D-LOFC), (6) reactivity transients - including anticipated transients without scram (ATWS), (7) processes coupled via intermediate heat exchanger (IHX) (IHX failure with molten salt), and (8) steam/water ingress. The panel's judgment of the importance ranking of a given phenomenon (or process) was based on the effect it had on one or more figures of merit or evaluation criteria. These included public and worker dose, fuel failure, and primary (and other safety) system integrity. The major phenomena of concern that were identified and categorized as high importance combined with medium to low knowledge follow: (1) core coolant bypass flows (normal operation), (2) power/flux profiles (normal operation), (3) outlet plenum flows (normal operation), (4) reactivity-temperature feedback coefficients for high-plutonium-content cores (normal operation and accidents), (5) fission product release related to the transport of silver (normal operation), (6)emissivity aspects for the vessel and reactor cavity cooling system (G-LOFC), (7) reactor vessel cavity air circulation and heat transfer (G-LOFC), and (8)convection/radiation heating of upper vessel area (P-LOFC).

  15. SNTP program reactor design

    Walton, Lewis A.; Sapyta, Joseph J.


    The Space Nuclear Thermal Propulsion (SNTP) program is evaluating the feasibility of a particle bed reactor for a high-performance nuclear thermal rocket engine. Reactors operating between 500 MW and 2,000 MW will produce engine thrusts ranging from 20,000 pounds to 80,000 pounds. The optimum reactor arrangement depends on the power level desired and the intended application. The key components of the reactor have been developed and are being tested. Flow-to-power matching considerations dominate the thermal-hydraulic design of the reactor. Optimal propellant management during decay heat cooling requires a three-pronged approach. Adequate computational methods exist to perform the neutronics analysis of the reactor core. These methods have been benchmarked to critical experiment data.

  16. Hybrid reactors. [Fuel cycle

    Moir, R.W.


    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  17. Fast Spectrum Reactors

    Todd, Donald; Tsvetkov, Pavel


    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  18. Compressive Transient Imaging

    Sun, Qilin


    High resolution transient/3D imaging technology is of high interest in both scientific research and commercial application. Nowadays, all of the transient imaging methods suffer from low resolution or time consuming mechanical scanning. We proposed a new method based on TCSPC and Compressive Sensing to achieve a high resolution transient imaging with a several seconds capturing process. Picosecond laser sends a serious of equal interval pulse while synchronized SPAD camera\\'s detecting gate window has a precise phase delay at each cycle. After capturing enough points, we are able to make up a whole signal. By inserting a DMD device into the system, we are able to modulate all the frames of data using binary random patterns to reconstruct a super resolution transient/3D image later. Because the low fill factor of SPAD sensor will make a compressive sensing scenario ill-conditioned, We designed and fabricated a diffractive microlens array. We proposed a new CS reconstruction algorithm which is able to denoise at the same time for the measurements suffering from Poisson noise. Instead of a single SPAD senor, we chose a SPAD array because it can drastically reduce the requirement for the number of measurements and its reconstruction time. Further more, it not easy to reconstruct a high resolution image with only one single sensor while for an array, it just needs to reconstruct small patches and a few measurements. In this thesis, we evaluated the reconstruction methods using both clean measurements and the version corrupted by Poisson noise. The results show how the integration over the layers influence the image quality and our algorithm works well while the measurements suffer from non-trival Poisson noise. It\\'s a breakthrough in the areas of both transient imaging and compressive sensing.

  19. Irradiation rigs in material testing reactor

    Rozenblum, F.; Gonnier, C.; Bignan, G. [CEA, Research Centers of Saclay and Cadarache (France)


    Osiris is a research reactor with a thermal power of 70 MW. It is a light-water reactor, open-core pool type, the principal aim of which is to carry out tests and irradiate structural materials and fuel elements of nuclear power plants under a high flux of neutrons, and to produce radioisotopes. Osiris operates around 200 days a year, in cycles of varying lengths from 3 to 4 weeks. A shutdown of about 10 days between two cycles allows reloading the core with fuel. Mainly 2 types of irradiation device are present: capsules for materials irradiation (CHOUCA and IRMA devices) and fuels irradiation loops (GRIFFONOS and ISABELLE). Although Osiris is still providing experiments of very good quality, it is facing obsolescence due to its ageing. Osiris is planned to be shut down during next decade. Consequently, it has been decided to launch the construction of the Jules Horowitz Reactor (JHR) in Cadarache. JHR is a water cooled reactor which provides the necessary flexibility and accessibility to manage several highly instrumented experiments, reproducing different reactor environments (water, gas or liquid metal loops), generating transient regimes (key for safety). The JHR facility includes the reactor building, including core, cooling system and the experimental bunkers connected to the core through pool wall penetrations and the auxiliary building, including pools and hot cells necessary for the experimental irradiation process. JHR core is optimised to produce high fast neutron flux to study structural material ageing and high thermal neutrons flux for fuel experiments. The conception of this first fleet of devices integrates the operational experience accumulated by the existing MTR and specifically the Osiris one

  20. LMFBR type reactor

    Kanbe, Mitsuru


    An LMFBR type reactor comprises a plurality of reactor cores in a reactor container. Namely, a plurality of pot containing vessels are disposed in the reactor vessel and a plurality of reactor cores are formed in a state where an integrated-type fuel assembly is each inserted to a pot, and a coolant pipeline is connected to each of the pot containing-vessel to cool the reactor core respectively. When fuels are exchanged, the integrated-type fuel assembly is taken out together with the pot from the reactor vessel in a state where the integrated-type fuel assembly is immersed in the coolants in the pot as it is. Accordingly, coolants are supplied to each of the pot containing-vessel connected with the coolant pipeline and circulate while cooling the integrated-type fuel assembly for every pot. Then, when the fuels are exchanged, the integrated type fuel assembly is taken out to the outside of the reactor together with the pot by taking up the pot from the pot-containing vessel. Then, neutron economy is improved to thereby improve reactor power and the breeding ratio. (N.H.)

  1. INVAP's Research Reactor Designs

    Eduardo Villarino


    Full Text Available INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors.

  2. Multi purpose research reactor

    Raina, V.K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail:; Sasidharan, K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sengupta, Samiran [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, Tej [Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)


    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.

  3. Fission product transport and behavior during two postulated loss of flow transients in the air

    Adams, J.P.; Carboneau, M.L.


    This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10{sup {minus}5 }and 10{sup {minus}7} per reactor year for LCP15 and LPP9, respectively.

  4. Fission product transport and behavior during two postulated loss of flow transients in the air

    Adams, J.P.; Carboneau, M.L.


    This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10{sup {minus}5 }and 10{sup {minus}7} per reactor year for LCP15 and LPP9, respectively.

  5. A Study on the Kinetic Characteristics of Transmutation Process Reactor

    Chung, Chang Hyun; You, Young Woo; Cho, Jae seon; Huh, Chang Wook; Kim, Doh Hyung [Seoul National University, Seoul (Korea, Republic of)


    The purpose of this study is to examine the transient heat transfer characteristics of liquid mental as the coolant used in accelerator-driven transmutation process reactor which is related the disposal of high-level radioactive nuclide. At current stage, the accelerator-driven transmutation process is investigated as the most appropriate method among many transmutation process methods. In this study, previous research works are investigated especially about the thermal hydraulics and kinetic behavior of coolant material including heat transfer of coolant in transmutation process reactor. A study on the heat transfer characteristics of liquid metal is performed based on the thermal hydraulic kinetic characteristics of liquid metal reactor which uses liquid metal coolant. Based on this study, the most appropriate material for the coolant of transmutation reactor will be recommended. 53 refs., 15 tabs., 33 figs. (author)

  6. SBWR Model for Steady-State and Transient Analysis

    Gilberto Espinosa-Paredes


    Full Text Available This paper presents a model of a simplified boiling water reactor (SBWR to analyze the steady-state and transient behavior. The SBWR model is based on approximations of lumped and distributed parameters to consider neutronics and natural circulation processes. The main components of the model are vessel dome, downcomer, lower plenum, core (channel and fuel, upper plenum, pressure, and level controls. Further consideration of the model is the natural circulation path in the internal circuit of the reactor, which governs the safety performance of the SBWR. To demonstrate the applicability of the model, the predictions were compared with plant data, manufacturer_s predictions, and RELAP5 under steady-state and transient conditions of a typical BWR. In steady-state conditions, the profiles of the main variables of the SBWR core such as superficial velocity, void fraction, temperatures, and convective heat transfer coefficient are presented and analyzed. The transient behavior of SBWR was analyzed during the closure of all main steam line isolation valves (MSIVs. Our results in this transient show that the cooling system due to natural circulation in the SBWR is around 70% of the rated core flow. According to the results shown here, one of the main conclusions of this work is that the simplified model could be very helpful in the licensing process.

  7. Coolant mixing in pressurized water reactors

    Hoehne, T.; Grunwald, G.


    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  8. PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

    Kirk, W.L. (comp.)


    The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters.

  9. Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model

    Baudron, Anne-Marie A -M; Maday, Yvon; Riahi, Mohamed Kamel; Salomon, Julien


    We present a parareal in time algorithm for the simulation of neutron diffusion transient model. The method is made efficient by means of a coarse solver defined with large time steps and steady control rods model. Using finite element for the space discretization, our implementation provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch-Maurer-Werner (LMW) benchmark [1].

  10. Coherent Transient Systems Evaluation


    manuscript is submitted for publication with the understanding that the United States Government is authorized to reproduce and distribute reprints...for governmental purposes. 1.0 Introduction The continuous optical correlator presented here is based on the phenomena of coherent transients, also...Gating the Continuous Processor Programming the continuous processor is accomplished by illuminati , n, the material with ,.’ modulated light pulses: a

  11. The Rapid Transient Surveyor

    Baranec, Christoph; Tonry, John; Wright, Shelley; Tully, R. Brent; Lu, Jessica R.; Takamiya, Marianne Y.; Hunter, Lisa


    The next decade of astronomy will be dominated by large area surveys (see the detailed discussion in the Astro-2010 Decadal survey and NRC's recent OIR System Report). Ground-based optical transient surveys, e.g., LSST, ZTF and ATLAS and space-based exoplanet, supernova, and lensing surveys such as TESS and WFIRST will join the Gaia all-sky astrometric survey in producing a flood of data that will enable leaps in our understanding of the universe. There is a critical need for further characterization of these discoveries through high angular resolution images, deeper images, spectra, or observations at different cadences or periods than the main surveys. Such follow-up characterization must be well matched to the particular surveys, and requires sufficient additional observing resources and time to cover the extensive number of targets.We describe plans for the Rapid Transient Surveyor (RTS), a permanently mounted, rapid-response, high-cadence facility for follow-up characterization of transient objects on the U. of Hawai'i 2.2-m telescope on Maunakea. RTS will comprise an improved robotic laser adaptive optics system, based on the prototype Robo-AO system (formerly at the Palomar 1.5-m and now at the Kitt Peak 2.2-m telescope), with simultaneous visible and near-infrared imagers as well as a near-infrared integral field spectrograph (R~100, λ = 850 - 1830 nm, 0.15″ spaxels, 8.7″×6.0″ FoV). RTS will achieve an acuity of ~0.07″ in visible wavelengths and automated detection and characterization of astrophysical transients during a sustained observing campaign will yield the necessary statistics to precisely map dark matter in the local universe.

  12. Transient Astrophysics Probe

    Camp, Jordan


    Transient Astrophysics Probe (TAP), selected by NASA for a funded Concept Study, is a wide-field high-energy transient mission proposed for flight starting in the late 2020s. TAP’s main science goals, called out as Frontier Discovery areas in the 2010 Decadal Survey, are time-domain astrophysics and counterparts of gravitational wave (GW) detections. The mission instruments include unique imaging soft X-ray optics that allow ~500 deg2 FoV in each of four separate modules; a high sensitivity, 1 deg2 FoV soft X-ray telescope based on single crystal silicon optics; a passively cooled, 1 deg2 FoV Infrared telescope with bandpass 0.6-3 micron; and a set of ~8 small NaI gamma-ray detectors. TAP will observe many events per year of X-ray transients related to compact objects, including tidal disruptions of stars, supernova shock breakouts, neutron star bursts and superbursts, and high redshift Gamma-Ray Bursts. Perhaps most exciting is TAP’s capability to observe X-ray and IR counterparts of GWs involving stellar mass black holes detected by LIGO/Virgo, and possibly X-ray counterparts of GWs from supermassive black holes, detected by LISA and Pulsar Timing Arrays.

  13. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)


    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  14. Reactor Materials Research

    Van Walle, E


    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  15. Space Nuclear Reactor Engineering

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)


    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  16. Light water reactor program

    Franks, S.M.


    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  17. Status of French reactors

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)


    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  18. Nuclear reactor design


    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  19. Mirror reactor surface study

    Hunt, A. L.; Damm, C. C.; Futch, A. H.; Hiskes, J. R.; Meisenheimer, R. G.; Moir, R. W.; Simonen, T. C.; Stallard, B. W.; Taylor, C. E.


    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included.

  20. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    Primm, Trent [ORNL; Gehin, Jess C [ORNL


    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  1. Transient Classification in LIGO data using Difference Boosting Neural Networks

    Mukund, Nikhil; Kandhasamy, Shivaraj; Mitra, Sanjit; Philip, Ninan Sajeeth


    Detection and classification of transients in data from gravitational wave detectors are crucial for efficient searches for true astrophysical events and identification of noise sources. We present a hybrid method for classification of short duration transients seen in gravitational wave data using both supervised and unsupervised machine learning techniques. To train the classifiers we use the relative wavelet energy and the corresponding entropy obtained by applying one-dimensional wavelet decomposition on the data. The prediction accuracy of the trained classifier on 9 simulated classes of gravitational wave transients and also LIGO's sixth science run hardware injections are reported. Targeted searches for a couple of known classes of non-astrophysical signals in the first observational run of Advanced LIGO data are also presented. The ability to accurately identify transient classes using minimal training samples makes the proposed method a useful tool for LIGO detector characterization as well as search...

  2. Transient classification in LIGO data using difference boosting neural network

    Mukund, N.; Abraham, S.; Kandhasamy, S.; Mitra, S.; Philip, N. S.


    Detection and classification of transients in data from gravitational wave detectors are crucial for efficient searches for true astrophysical events and identification of noise sources. We present a hybrid method for classification of short duration transients seen in gravitational wave data using both supervised and unsupervised machine learning techniques. To train the classifiers, we use the relative wavelet energy and the corresponding entropy obtained by applying one-dimensional wavelet decomposition on the data. The prediction accuracy of the trained classifier on nine simulated classes of gravitational wave transients and also LIGO's sixth science run hardware injections are reported. Targeted searches for a couple of known classes of nonastrophysical signals in the first observational run of Advanced LIGO data are also presented. The ability to accurately identify transient classes using minimal training samples makes the proposed method a useful tool for LIGO detector characterization as well as searches for short duration gravitational wave signals.

  3. Real-Time Detection of Optical Transients with RAPTOR

    Borozdin, K N; Galassi, M; McGowan, K; Starr, D; Vestrand, W T; White, R; Wozniak, P R; Wren, J; Borozdin, Konstantin; Brumby, Steven; Galassi, Mark; Gowan, Katherine Mc; Starr, Dan; White, Robert; Wozniak, Przemyslaw; Wren, James


    Fast variability of optical objects is an interesting though poorly explored subject in modern astronomy. Real-time data processing and identification of transient celestial events in the images is very important for such study as it allows rapid follow-up with more sensitive instruments. We discuss an approach which we have developed for the RAPTOR project, a pioneering closed-loop system combining real-time transient detection with rapid follow-up. RAPTOR's data processing pipeline is able to identify and localize an optical transient within seconds after the observation. The testing we performed so far have been confirming the effectiveness of our method for the optical transient detection. The software pipeline we have developed for RAPTOR can easily be applied to the data from other experiments.

  4. Slurry reactor design studies

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))


    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  5. Thermal transient anemometer

    Bailey, James L.; Vresk, Josip


    A thermal transient anemometer having a thermocouple probe which is utilized to measure the change in temperature over a period of time to provide a measure of fluid flow velocity. The thermocouple probe is located in the fluid flow path and pulsed to heat or cool the probe. The cooling of the heated probe or the heating of the cooled probe from the fluid flow over a period of time is measured to determine the fluid flow velocity. The probe is desired to be locally heated near the tip to increase the efficiency of devices incorporating the probe.

  6. DSN Transient Observatory

    Kuiper, T. B. H.; Monroe, R. M.; White, L. A.; Garcia Miro, C.; Levin, S. M.; Majid, W. A.; Soriano, M.


    The Deep Space Network (DSN) Transient Observatory (DTO) is a signal processing facility that can monitor up to four DSN downlink bands for astronomically interesting signals. The monitoring is done commensally with reception of deep space mission telemetry. The initial signal processing is done with two CASPERa ROACH1 boards, each handling one or two baseband signals. Each ROACH1 has a 10 GBe interface with a GPU-equipped Debian Linux workstation for additional processing. The initial science programs include monitoring Mars for electrostatic discharges, radio spectral lines, searches for fast radio bursts and pulsars and SETI. The facility will be available to the scientific community through a peer review process.

  7. Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)

    Agrawal, A.K.


    Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation.

  8. Evaluation of fatigue vessel in the Sta. M Garona NPP : real transients and design transients; Evaluacion de la fatiga en la vasija de CN Santa M de Garona: transitorios reales frente a transitorios de diseno

    Martin, J.; Gorrochategui, I.


    The number of transient that control the fatigue of its reactor pressure vessel is included in the Sta. M Garona NPP Technical Specifications, being the different transients described in the design specification of the corresponding component. In this work, on the one hand, the description of the design transients with their corresponding real ones is compared and, on the other hand, the number of occurrences and the number of transients originally estimated is also compared. In both cases the influence of the difference between design and reality in the fatigue usage is discussed. (Author)

  9. Fast Breeder Reactor studies

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.


    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  10. Gas cooled fast reactor



    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  11. Microfluidic electrochemical reactors

    Nuzzo, Ralph G [Champaign, IL; Mitrovski, Svetlana M [Urbana, IL


    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  12. 76 FR 37852 - Advisory Committee on Reactor Safeguards; Notice of Meeting


    ... Modular Reactor Issue Identification and Ranking Process (Open)--The Committee will hear presentations by and hold discussions with representatives of the NRC staff regarding the small modular reactor issue... From the Federal Register Online via the Government Publishing Office ] NUCLEAR...

  13. Integrated systems analysis of the PIUS reactor

    Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others


    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  14. Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)


    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm2 and temporary heat flux limit of 600 W/cm2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

  15. Transient regional osteoporosis

    F. Trotta


    Full Text Available Transient osteoporosis of the hip and regional migratory osteoporosis are uncommon and probably underdiagnosed bone diseases characterized by pain and functional limitation mainly affecting weight-bearing joints of the lower limbs. These conditions are usually self-limiting and symptoms tend to abate within a few months without sequelae. Routine laboratory investigations are unremarkable. Middle aged men and women during the last months of pregnancy or in the immediate post-partum period are principally affected. Osteopenia with preservation of articular space and transitory edema of the bone marrow provided by magnetic resonance imaging are common to these two conditions, so they are also known by the term regional transitory osteoporosis. The appearance of bone marrow edema is not specific to regional transitory osteoporosis but can be observed in several diseases, i.e. trauma, reflex sympathetic dystrophy, avascular osteonecrosis, infections, tumors from which it must be differentiated. The etiology of this condition is unknown. Pathogenesis is still debated in particular the relationship with reflex sympathetic dystrophy, with which regional transitory osteoporosis is often identified. The purpose of the present review is to remark on the relationship between transient osteoporosis of the hip and regional migratory osteoporosis with particular attention to the bone marrow edema pattern and relative differential diagnosis.

  16. Transient Black Hole Binaries

    Belloni, T M


    The last two decades have seen a great improvement in our understand- ing of the complex phenomenology observed in transient black-hole binary systems, especially thanks to the activity of the Rossi X-Ray Timing Explorer satellite, com- plemented by observations from many other X-ray observatories and ground-based radio, optical and infrared facilities. Accretion alone cannot describe accurately the intricate behavior associated with black-hole transients and it is now clear that the role played by different kinds of (often massive) outflows seen at different phases of the outburst evolution of these systems is as fundamental as the one played by the accretion process itself. The spectral-timing states originally identified in the X-rays and fundamentally based on the observed effect of accretion, have acquired new importance as they now allow to describe within a coherent picture the phenomenology observed at other wave- length, where the effects of ejection processes are most evident. With a particular focu...

  17. Characterizing Nanoscale Transient Communication.

    Chen, Yifan; Anwar, Putri Santi; Huang, Limin; Asvial, Muhamad


    We consider the novel paradigm of nanoscale transient communication (NTC), where certain components of the small-scale communication link are physically transient. As such, the transmitter and the receiver may change their properties over a prescribed lifespan due to their time-varying structures. The NTC systems may find important applications in the biomedical, environmental, and military fields, where system degradability allows for benign integration into life and environment. In this paper, we analyze the NTC systems from the channel-modeling and capacity-analysis perspectives and focus on the stochastically meaningful slow transience scenario, where the coherence time of degeneration Td is much longer than the coding delay Tc. We first develop novel and parsimonious models to characterize the NTC channels, where three types of physical layers are considered: electromagnetism-based terahertz (THz) communication, diffusion-based molecular communication (DMC), and nanobots-assisted touchable communication (TouchCom). We then revisit the classical performance measure of ϵ-outage channel capacity and take a fresh look at its formulations in the NTC context. Next, we present the notion of capacity degeneration profile (CDP), which describes the reduction of channel capacity with respect to the degeneration time. Finally, we provide numerical examples to demonstrate the features of CDP. To the best of our knowledge, the current work represents a first attempt to systematically evaluate the quality of nanoscale communication systems deteriorating with time.

  18. Outlet plenum flow stratification studies for the Clinch River Breeder Reactor Plant

    Novendstern, E.H.; Woods, M.D.; Andreychek, T.S.; Flannigan, L.J.; Carr, J.A.


    The transient temperature behavior during a simulated reactor trip was studied in a 1/3 sector, 0.55 scale model of the reactor outlet plenum of the CRBRP. Buoyancy effects were simulated using water and salt solution. Effects of variations in Richardson Number (ratio of buoyancy/inertia forces) and Froude Number (ratio of inertia/viscous forces) were evaluated. Effects of geometrical changes of a component in the outlet plenum called the Upper Internals Structure on transient temperature response were studied; both height of the structure and leakage under this component were experimentally varied. The test results confirm that flow stratification occurs in the outlet plenum following a reactor trip. The colder, denser fluid issuing from the core assemblies during the transient fills the lower portion of the plenum while the hotter fluid is trapped in the region above the outlet nozzles.

  19. System Identification

    Keesman, K.J.


    Summary System Identification Introduction.- Part I: Data-based Identification.- System Response Methods.- Frequency Response Methods.- Correlation Methods.- Part II: Time-invariant Systems Identification.- Static Systems Identification.- Dynamic Systems Identification.- Part III: Time-varying

  20. Fusion Reactor Materials

    Decreton, M


    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  1. New reactor type proposed


    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  2. Reactor Neutrino Spectra

    Hayes, A C


    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  3. Reactor BR2. Introduction

    Gubel, P


    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.


    M. Rizaal


    are to learn the change of thermal power caused by the change of suspended core height and coolant flow rate, and also to learn the inherent safety when loss of heat sink condition prevailed. The Core was modelled on steady condition by using Standard Reactor Analysis Code (SRAC to obtain neutron flux, group constants, delayed neutron fraction, delayed neutron precursor decay constants, and several core parameters. These data will be used as initial value on the transient calculations. Transient analysis was conducted on the following conditions: coolant flow rate changes, suspended core height changes and loss of heat sink occours. The calculated result showed that when the coolant flow rate is 50% decreased, thermal power of FBNR is 28% decreased. When suspended core height is 30% decreased, thermal power of FBNR is 17% decreased. Meanwhile, thermal power at loss of heat sink condition is 76% decreased. Therefore, the adjustment of suspended core height and coolant flow rate can control thermal power of FBNR, and FBNR’s inherent safety is reliable at loss of heat sink condition. Keywords: FBNR, transient, power, flow rate, suspended core

  5. Investigation of effective factors of transient thermal stress of the MONJU-System components

    Inoue, Masaaki; Hirayama, Hiroshi; Kimura, Kimitaka; Jinbo, M. [Toshiba Corp., Kawasaki, Kanagawa (Japan)


    Transient thermal stress of each system Component in the fast breeder reactor is an uncertain factor on it's structural design. The temperature distribution in a system component changes over a wide range in time and in space. An unified evaluation technique of thermal, hydraulic, and structural analysis, in which includes thermal striping, temperature stratification, transient thermal stress and the integrity of the system components, is required for the optimum design of tho fast reactor plant. Thermal boundary conditions should be set up by both the transient thermal stress analysis and the structural integrity evaluation of each system component. The reasonable thermal boundary conditions for the design of the MONJU and a demonstration fast reactor, are investigated. The temperature distribution analysis models and the thermal boundary conditions on the Y-piece structural parts of each system component, such as reactor vessel, intermediate heat exchanger, primary main circulation pump, steam generator, superheater and upper structure of reactor core, are illustrated in the report. (M. Suetake)

  6. Future Reactor Experiments

    He, Miao


    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measure...

  7. Reactor Neutrino Experiments

    Cao, Jun


    Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measu...

  8. Department of Reactor Technology

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  9. INVAP's Research Reactor Designs

    Eduardo Villarino; Alicia Doval


    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper ...

  10. Calculating transient rates from surveys

    Carbone, Dario; Wijers, Ralph A M J; Rowlinson, Antonia


    We have developed a method to determine the transient surface density and transient rate for any given survey, using Monte-Carlo simulations. This method allows us to determine the transient rate as a function of both the flux and the duration of the transients in the whole flux-duration plane rather than one or a few points as currently available methods do. It is applicable to every survey strategy that is monitoring the same part of the sky, regardless the instrument or wavelength of the survey, or the target sources. We have simulated both top-hat and Fast Rise Exponential Decay light curves, highlighting how the shape of the light curve might affect the detectability of transients. Another application for this method is to estimate the number of transients of a given kind that are expected to be detected by a survey, provided that their rate is known.

  11. Calculating transient rates from surveys

    Carbone, D.; van der Horst, A. J.; Wijers, R. A. M. J.; Rowlinson, A.


    We have developed a method to determine the transient surface density and transient rate for any given survey, using Monte Carlo simulations. This method allows us to determine the transient rate as a function of both the flux and the duration of the transients in the whole flux-duration plane rather than one or a few points as currently available methods do. It is applicable to every survey strategy that is monitoring the same part of the sky, regardless the instrument or wavelength of the survey, or the target sources. We have simulated both top-hat and Fast Rise Exponential Decay light curves, highlighting how the shape of the light curve might affect the detectability of transients. Another application for this method is to estimate the number of transients of a given kind that are expected to be detected by a survey, provided that their rate is known.

  12. Small Liquid Metal Cooled Reactor Safety Study

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N


    The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S requires negative void worth throughout

  13. The reactor antineutrino anomalies

    Haser, Julia; Buck, Christian; Lindner, Manfred [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany)


    Major discoveries were made in the past few years in the field of neutrino flavour oscillation. Nuclear reactors produce a clean and intense flux of electron antineutrinos and are thus an essential neutrino source for the determination of oscillation parameters. Most currently the reactor antineutrino experiments Double Chooz, Daya Bay and RENO have accomplished to measure θ{sub 13}, the smallest of the three-flavour mixing angles. In the course of these experiments two anomalies emerged: (1) the reanalysis of the reactor predictions revealed a deficit in experimentally observed antineutrino flux, known as the ''reactor antineutrino anomaly''. (2) The high precision of the latest generation of neutrino experiments resolved a spectral shape distortion relative to the expected energy spectra. Both puzzles are yet to be solved and triggered new experimental as well as theoretical studies, with the search for light sterile neutrinos as most popular explanation for the flux anomaly. This talk outlines the two reactor antineutrino anomalies. Discussing possible explanations for their occurrence, recent and upcoming efforts to solve the reactor puzzles are highlighted.

  14. Moon base reactor system

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.


    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  15. The Zwicky Transient Facility

    Kulkarni, Shrinivas R.


    The Zwicky Transient Facility (ZTF) has been designed with a singular focus: a systematic exploration of the night sky at a magnitude level well suited for spectral classification and follow up with the existing class of 4-m to 10-m class telescopes. ZTF is the successor to the Palomar Transient Factory (PTF). The discovery engine for ZTF is a 47 square degree camera (realized through 16 e2V monolithic CCDs) that fills the entire focal plane of the 48-inch Oschin telescope of the Palomar Observatory. Single 30-s epoch sensitivity is about 20.5 in g and R bands. The Infarared Processing & Analysis Center (IPAC) is the data center for ZTF. ZTF is a public-private partnership with equal contributions from a consortium of world-wide partners and an NSF MSIP grant. Forty percent of ZTF time is set aside for two major community surveys: a 3-day cadence survey of high latitudes (to mimic LSST) and a time domain survey of the entire Northern Galactic plane. We expect first light in February 2017 and begin a 3-year survey starting summer of 2017. The first year will be spent on building up deep reference images of the sky (a must for transient surveys). During the second year IPAC will deliver near archival quality photometric products within 12 hours of observations. By comparison to reference images photometric alerts will be sent out. Year 3 will see the near real-time release of image differencing products. A Community Science Advisory Committee (CSAC), chaired by S. Ridgway (NOAO), has been set up to both advise the PI and to ensure that the US community's interests are well served. Astronomers interested in getting a head start on ZTF may wish to peruse the data releases from PTF. Young people (or young at heart) may wish to attend the annual summer school on PTF/ZTF (August, Caltech campus). The Principal Investigator (PI) for the project is S. Kulkarni and the Project Scientist is Eric Bellm.For further details please consult

  16. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

    Was, Gary S.


    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems.

  17. Study of reactor plant disturbed cooling condition modes caused by the VVER reactor secondary circuit

    V.I. Belozerov


    Based on the RELAP-5, TRAC, and TRACE software codes, reactor plant cooling condition malfunction modes caused by the VVER-1000 secondary circuit were simulated and investigated. Experimental data on the mode with the turbine-generator stop valve closing are presented. The obtained dependences made it possible to determine the maximum values of pressure and temperature in the circulation circuit as well as estimate the Minimum Critical Heat Flux Ratio (MCHFR. It has been found that, if any of the initial events occurs, safety systems are activated according to the set points; transient processes are stabilized in time; and the Critical Heat Flux (CHF limit is provided. Therefore, in the event of emergency associated with the considered modes, the reactor plant safety will be ensured.

  18. The joy of transient chaos

    Tél, Tamás [Institute for Theoretical Physics, Eötvös University, and MTA-ELTE Theoretical Physics Research Group, Pázmány P. s. 1/A, Budapest H-1117 (Hungary)


    We intend to show that transient chaos is a very appealing, but still not widely appreciated, subfield of nonlinear dynamics. Besides flashing its basic properties and giving a brief overview of the many applications, a few recent transient-chaos-related subjects are introduced in some detail. These include the dynamics of decision making, dispersion, and sedimentation of volcanic ash, doubly transient chaos of undriven autonomous mechanical systems, and a dynamical systems approach to energy absorption or explosion.

  19. Development of transient initiating event frequencies for use in probabilistic risk assessments

    Mackowiak, D.P.; Gentillon, C.D.; Smith, K.L.


    Transient initiating event frequencies are an essential input to the analysis process of a nuclear power plant probabilistic risk assessment. These frequencies describe events causing or requiring scrams. This report documents an effort to validate and update from other sources a computer-based data file developed by the Electric Power Research Institute (EPRI) describing such events at 52 United States commercial nuclear power plants. Operating information from the United States Nuclear Regulatory Commission on 24 additional plants from their date of commercial operation has been combined with the EPRI data, and the entire data base has been updated to add 1980 through 1983 events for all 76 plants. The validity of the EPRI data and data analysis methodology and the adequacy of the EPRI transient categories are examined. New transient initiating event frequencies are derived from the expanded data base using the EPRI transient categories and data display methods. Upper bounds for these frequencies are also provided. Additional analyses explore changes in the dominant transients, changes in transient outage times and their impact on plant operation, and the effects of power level and scheduled scrams on transient event frequencies. A more rigorous data analysis methodology is developed to encourage further refinement of the transient initiating event frequencies derived herein. Updating the transient event data base resulted in approx.2400 events being added to EPRI's approx.3000-event data file. The resulting frequency estimates were in most cases lower than those reported by EPRI, but no significant order-of-magnitude changes were noted. The average number of transients per year for the combined data base is 8.5 for pressurized water reactors and 7.4 for boiling water reactors.

  20. Transient osteoporosis of pregnancy.

    Maliha, George; Morgan, Jordan; Vrahas, Mark


    Transient osteoporosis of pregnancy (TOP) is a rare yet perhaps under-reported condition that has affected otherwise healthy pregnancies throughout the world. The condition presents suddenly in the third trimester of a usually uneventful pregnancy and progressively immobilizes the mother. Radiographic studies detect drastic loss of bone mass, elevated rates of turnover in the bone, and oedema in the affected portion. Weakness of the bone can lead to fractures during delivery and other complications for the mother. Then, within weeks of labour, symptoms and radiological findings resolve. Aetiology is currently unknown, although neural, vascular, haematological, endocrine, nutrient-deficiency, and other etiologies have been proposed. Several treatments have also been explored, including simple bed rest, steroids, bisphosphonates, calcitonin, induced termination of pregnancy, and surgical intervention. The orthopedist plays an essential role in monitoring the condition (and potential complications) as well as ensuring satisfactory outcomes for both the mother and newborn.

  1. Transient Detection and Classification

    Becker, Andrew C


    I provide an incomplete inventory of the astronomical variability that will be found by next-generation time-domain astronomical surveys. These phenomena span the distance range from near-Earth satellites to the farthest Gamma Ray Bursts. The surveys that detect these transients will issue alerts to the greater astronomical community; this decision process must be extremely robust to avoid a slew of ``false'' alerts, and to maintain the community's trust in the surveys. I review the functionality required of both the surveys and the telescope networks that will be following them up, and the role of VOEvents in this process. Finally, I offer some ideas about object and event classification, which will be explored more thoroughly by other articles in these proceedings.

  2. Stability of Ignition Transients

    V.E. Zarko


    Full Text Available The problem of ignition stability arises in the case of the action of intense external heat stimuli when, resulting from the cut-off of solid substance heating, momentary ignition is followed by extinction. Physical pattern of solid propellant ignition is considered and ignition criteria available in the literature are discussed. It is shown that the above mentioned problem amounts to transient burning at a given arbitrary temperature distribution in the condensed phase. A brief survey of published data on experimental and theoretical studies on ignition stability is offered. The comparison between theory and experiment is shown to prove qualitatively the efficiency of the phenomenological approach in the theory. However, the methods of mathematical simulation as well as those of experimental studying of ignition phenomenon, especially at high fluxes, need to be improved.

  3. Transient heliosheath modulation

    Quenby, J. J.; Webber, W. R.


    Voyager 1 has explored the solar wind-interstellar medium interaction region between the terminal shock and heliopause, following the intensity distribution of Galactic cosmic ray protons above 200 MeV energy. Before this component reached the expected galactic flux level at 121.7 au from the Sun, four episodes of rapid intensity change occurred with a behaviour similar to that found in Forbush Decreases in the inner Solar system, rather than that expected from a mechanism related to models for the long-term modulation found closer to the Sun. Because the mean solar wind flow is both expected and observed to be perpendicular to the radial direction close to the heliopause, an explanation is suggested in terms of transient radial flows related to possible heliopause boundary flapping. It is necessary that the radial flows are of the order either of the sound speed found for conditions downstream of the terminal shock or of the fluctuations found near the boundary by the Voyager 1 Low Energy Charged Particle detector and that the relevant cosmic ray diffusion perpendicular to the mean field is controlled by `slab' fluctuations accounting for about 20 per cent of the total power in the field variance. However, additional radial drift motion related to possible north to south gradients in the magnetic field may allow the inclusion of some diffusion according to the predictions of a theory based upon the presence of 2D turbulence. The required field gradients may arise due to field variation in the field carried by solar plasma flow deflected away from the solar equatorial plane. Modulation amounting to a total 30 per cent drop in galactic intensity requires explanation by a combination of transient effects.

  4. Recent neutronics developments for reactor safety studies with SIMMER code at KIT

    Rineiski, A.; Marchetti, M.; Andriolo, L.; Gabrielli, F.


    The SIMMER family of codes is applied for safety studies of sodium fast reactors and reactors of other types. Both neutronics and fluid-dynamics parts of SIMMER are under development. In the paper new neutronics capabilities are presented. In particular developments for neutron transport solvers and a new technique for taking into account thermal expansion effects are described. These new capabilities facilitate 3D simulations and improve accuracy of modelling for the initiation transient phase during a hypothetical severe accident.

  5. A controllability study of TRUMOX fuel for load following operations in a CANDU-900 reactor

    Trudell, D.A., E-mail: [McMaster Univ., Hamilton, Ontario (Canada)


    Using a core model of a generic CANDU-900 reactor in RFSP-IST, load following simulations have been performed to assess the controllability of the reactor due to Xenon transients. Week long load following simulations have been performed with daily power cycles 12 hours in duration. Simulations have shown that Natural Uranium fuel can be safely cycled between 100 and 90% Full Power without adjuster rod movement while TRUMOX fuel can be safely cycled between 100 and 85% Full Power. (author)

  6. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    Komen, E.M.J.; Bogaard, J.P.A. van den


    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

  7. Degradation Mechanisms of Colloidal Organic Matter in Biofilm Reactors

    Larsen, Tove; Harremoës, Poul


    The degradation mechanisms of colloidal organic matter in biofilm reactors have been studied in an idealized laboratory reactor system with soluble starch as a model substrate. Batch tests and experiments with different reactor configurations have shown that for this specific substrate, bulk liquid......-diffusible organic matter in a biofilm reactor. DH depends on the combined volumetric and surface hydraulic loading rate, Q2/(AV). In full-scale wastewater treatment plants, the degradation mechanism presented in this paper can explain important differences between the performance of trickling filters and RBC...... hydrolysis is the mechanism for transforming non-diffusible organic matter into biofilm diffusible substrate. A simplified mathematical description has led to the identification of the degree of hydrolysis, DH, as the parameter expressing the major difference between degradation of diffusible and non...

  8. Transient Safety Analysis of Fast Spectrum TRU Burning LWRs with Internal Blankets

    Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Zazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hill, Bob [Argonne National Lab. (ANL), Argonne, IL (United States)


    The objective of this proposal was to perform a detailed transient safety analysis of the Resource-Renewable BWR (RBWR) core designs using the U.S. NRC TRACE/PARCS code system. This project involved the same joint team that has performed the RBWR design evaluation for EPRI and therefore be able to leverage that previous work. And because of their extensive experience with fast spectrum reactors and parfait core designs, ANL was also part the project team. The principal outcome of this project was the development of a state-of-the-art transient analysis capability for GEN-IV reactors based on Monte Carlo generated cross sections and the US NRC coupled code system TRACE/PARCS, and a state-of-the-art coupled code assessment of the transient safety performance of the RBWR.

  9. Accident analysis of heavy water cooled thorium breeder reactor

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki


    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition

  10. Integrated control system responses during load rejection transient for Lungmen ABWR Plant

    Ma Shaoshih, E-mail: [Department of Engineering and System Science, National Tsing Hua University, 101, Section 2, Kuang Fu Rd., HsinChu City 30013, Taiwan (China); Institute of Nuclear Energy Research, Atomic Energy Council, ROC, P.O. Box 3-3, Lungtan, Taoyuan 32546, Taiwan (China); Shih Chunkuan [Department of Engineering and System Science, National Tsing Hua University, 101, Section 2, Kuang Fu Rd., HsinChu City 30013, Taiwan (China); Yuann Yngruey [Institute of Nuclear Energy Research, Atomic Energy Council, ROC, P.O. Box 3-3, Lungtan, Taoyuan 32546, Taiwan (China)


    Highlights: > The Lungmen RETRAN model was used to evaluate the load rejection with bypass start-up test. > The purpose is to verify the capability of the plant withstand a load rejection without scram. > The responses of the control systems are important to help mitigate the transient impact. > The analysis show that the reactor is not scrammed due to the proper responses of the control systems. > As long as bypass capacity functions in time, there is no significant pressurization in transient. - Abstract: The RETRAN model of Lungmen ABWR Plant was used to evaluate the load rejection with bypass transient of the start-up test program. There are three major control systems implemented in Lungmen RETRAN Model, including recirculation flow control, feedwater control, steam bypass and pressure control system. The responses of the control systems are important to help the reactor operate smoothly in the test, and credited to mitigate the transient impact. A brief description about the Lungmen RETRAN Model and the design feature of control systems is included in this article. The predicted analysis of load rejection test was conducted by using RETRAN-3D code to verify the dynamic responses of the reactor and control systems to protective trips in the generator, and also to demonstrate the capability of the plant to withstand against a generator load rejection without reactor scram. The analysis results show that the reactor has not been scrammed due to the proper control system responses, and sustained stable operation under the condition of the minimum station house load. The sensitivity studies show that there will be no significant pressurization in load rejection transient as long as activating bypass capacity functions in time. While, the sensitivity study cases of other settings in the mitigation facilities have no outstanding impact on the system response.

  11. Transient Go: A Mobile App for Transient Astronomy Outreach

    Crichton, D.; Mahabal, A.; Djorgovski, S. G.; Drake, A.; Early, J.; Ivezic, Z.; Jacoby, S.; Kanbur, S.


    Augmented Reality (AR) is set to revolutionize human interaction with the real world as demonstrated by the phenomenal success of `Pokemon Go'. That very technology can be used to rekindle the interest in science at the school level. We are in the process of developing a prototype app based on sky maps that will use AR to introduce different classes of astronomical transients to students as they are discovered i.e. in real-time. This will involve transient streams from surveys such as the Catalina Real-time Transient Survey (CRTS) today and the Large Synoptic Survey Telescope (LSST) in the near future. The transient streams will be combined with archival and latest image cut-outs and other auxiliary data as well as historical and statistical perspectives on each of the transient types being served. Such an app could easily be adapted to work with various NASA missions and NSF projects to enrich the student experience.

  12. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    P. Delmolino


    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  13. The Halden reactor, a facility open to the international nuclear community; Halden, un reacteur ouvert a la communaute internationale

    Vitanza, C. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency (OECD/NEA), 75 - Paris (France)


    The Halden test reactor is a boiling-type reactor moderated and cooled by heavy water, it yields a thermal power of 20 MW. The reactor operates for 2 periods of about 100 days each year. The Halden reactor has been in operation for more than 45 years and is the largest OECD-NEA project, it carries out the OECD joint program and bilateral contract work. Its experimental programs are supported by about 100 organisations in 20 countries. The fuel and materials programs for the years to come focus on the following main areas: -) fuel high burn-up capabilities in normal operating conditions, -) fuel response to transients aiming at generating experimental data on the behaviour of high burn-up fuels in short duration transients and on phenomena occurring during loss of coolant accident and coolant flow oscillations, -) cladding corrosion and water chemistry issues, and -) pressure vessel embrittlement and irradiation assisted stress corrosion cracking of reactor internals. (A.C.)

  14. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  15. A diagnostic system for identifying accident conditions in a nuclear reactor

    Santhosh, T.V., E-mail: [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India); Kumar, M.; Thangamani, I.; Srivastava, A.; Dutta, A.; Verma, V.; Mukhopadhyay, D.; Ganju, S.; Chatterjee, B.; Rao, V.V.S.S.; Lele, H.G.; Ghosh, A.K. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)


    Research highlights: Neural networks based diagnostic system has been developed to identify transients quickly, estimate the source-term and assist the operator to take corrective actions during abnormal situations in 220 MWe PHWRs. The transient data for the break scenarios ranging from 20% to 200% has been generated using RELAP5 and CONTRAN codes. 32 break scenarios of large break LOCA in inlet and outlet reactor headers with and without ECCS have been analyzed using artificial neural networks. A few break scenarios were directly predicted without being trained earlier. Test results obtained from ANN are within the acceptable range. - Abstract: The objective of this study is to develop a system, which assists the operator in identifying an accident quickly using ANNs that diagnoses the accidents based on reactor process parameters, and continuously displays the status of the nuclear reactor. A large database of transient data of reactor process parameters has been generated for reactor core, containment, environmental dispersion and radiological dose to train the ANNs. These data have been generated using various codes e.g., RELAP5-thermal-hydraulics code for the core. The present version of this system is capable of identifying large break LOCA scenarios of 220 MWe Indian PHWRs. The system has been designed to provide the necessary information to the operator to handle emergency situations when the reactor is operating. The diagnostic results obtained from ANNs study are satisfactory.

  16. Preparation macroconstants to simulate the core of VVER-1000 reactor

    Seleznev, V. Y.


    Dynamic model is used in simulators of VVER-1000 reactor for training of operating staff and students. As a code for the simulation of neutron-physical characteristics is used DYNCO code that allows you to perform calculations of stationary, transient and emergency processes in real time to a different geometry of the reactor lattices [1]. To perform calculations using this code, you need to prepare macroconstants for each FA. One way of getting macroconstants is to use the WIMS code, which is based on the use of its own 69-group macroconstants library. This paper presents the results of calculations of FA obtained by the WIMS code for VVER-1000 reactor with different parameters of fuel and coolant, as well as the method of selection of energy groups for further calculation macroconstants.

  17. Transient cases analyses of the TRIGA IPR-R1 using thermal hydraulic and neutron kinetic coupled codes

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Scari, Maria E., E-mail:, E-mail:, E-mail:, E-mail:, E-mail: [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail:, E-mail: [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear


    Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)

  18. Deterministic and Probabilistic Analysis against Anticipated Transient Without Scram

    Choi, Sun Mi; Kim, Ji Hwan [KHNP Central Research Institute, Daejeon (Korea, Republic of); Seok, Ho [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)


    An Anticipated Transient Without Scram (ATWS) is an Anticipated Operational Occurrences (AOOs) accompanied by a failure of the reactor trip when required. By a suitable combination of inherent characteristics and diverse systems, the reactor design needs to reduce the probability of the ATWS and to limit any Core Damage and prevent loss of integrity of the reactor coolant pressure boundary if it happens. This study focuses on the deterministic analysis for the ATWS events with respect to Reactor Coolant System (RCS) over-pressure and fuel integrity for the EU-APR. Additionally, this report presents the Probabilistic Safety Assessment (PSA) reflecting those diverse systems. The analysis performed for the ATWS event indicates that the NSSS could be reached to controlled and safe state due to the addition of boron into the core via the EBS pump flow upon the EBAS by DPS. Decay heat is removed through MSADVs and the auxiliary feedwater. During the ATWS event, RCS pressure boundary is maintained by the operation of primary and secondary safety valves. Consequently, the acceptance criteria were satisfied by installing DPS and EBS in addition to the inherent safety characteristics.

  19. PFR fuel cladding transient test results and analysis

    Cannon, N. S.; Hunter, C. W.; Kear, K. L.; Wood, M. H.


    Fuel Cladding Transient Tests (FCTT) were performed on M316 cladding specimens obtained from mixed-oxide fuel pins irradiated in the Prototype Fast Reactor (PFR) to burnups of 4 and 9 atom percent. In these tests, specimens of fuel cladding were pressurized and heated until failure occurred. Samples of cladding from PFR fuel pins exhibited generally greater strength and ductility than specimens from Experimental Breeder Reactor-II (EBR-II) mixed-oxide fuel pins tested under similar conditions. Apparently, the PFR cladding properties were not degraded by a fuel adjacency effect (FAE) observed in fuel pin cladding from EBR-II irradiations. A recently developed model of grain boundary cavity growth was used to predict the results of the tests conducted on PFR cladding. It was found that the predicted failure temperatures for the relevant internal pressures were in good agreement with experimental failure temperatures.

  20. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    Scarlat, Raluca Olga

    This dissertation treats system design, modeling of transient system response, and characterization of individual phenomena and demonstrates a framework for integration of these three activities early in the design process of a complex engineered system. A system analysis framework for prioritization of experiments, modeling, and development of detailed design is proposed. Two fundamental topics in thermal-hydraulics are discussed, which illustrate the integration of modeling and experimentation with nuclear reactor design and safety analysis: thermal-hydraulic modeling of heat generating pebble bed cores, and scaled experiments for natural circulation heat removal with Boussinesq liquids. The case studies used in this dissertation are derived from the design and safety analysis of a pebble bed fluoride salt cooled high temperature nuclear reactor (PB-FHR), currently under development in the United States at the university and national laboratories level. In the context of the phenomena identification and ranking table (PIRT) methodology, new tools and approaches are proposed and demonstrated here, which are specifically relevant to technology in the early stages of development, and to analysis of passive safety features. A system decomposition approach is proposed. Definition of system functional requirements complements identification and compilation of the current knowledge base for the behavior of the system. Two new graphical tools are developed for ranking of phenomena importance: a phenomena ranking map, and a phenomena identification and ranking matrix (PIRM). The functional requirements established through this methodology were used for the design and optimization of the reactor core, and for the transient analysis and design of the passive natural circulation driven decay heat removal system for the PB-FHR. A numerical modeling approach for heat-generating porous media, with multi-dimensional fluid flow is presented. The application of this modeling


    Ren Chunhui; Wei Ping; Lou Zhiyou; Xiao Xianci


    In this letter, the communication transmitter transient signals are analyzed based on the time-variant hierarchy exponents of multifractal analysis. The species of optimized sample set is selected as the template of transmitter identification, so that the individual communication transmitter identification can be realized. The turn-on signals of four transmitters are used in the simulation. The experimental results show that the multifractal character of transmitter transient signals is an effective character of individual transmitter identification.

  2. Towards an efficient multiphysics model for nuclear reactor dynamics

    Obaidurrahman K.


    Full Text Available Availability of fast computer resources nowadays has facilitated more in-depth modeling of complex engineering systems which involve strong multiphysics interactions. This multiphysics modeling is an important necessity in nuclear reactor safety studies where efforts are being made worldwide to combine the knowledge from all associated disciplines at one place to accomplish the most realistic simulation of involved phenomenon. On these lines coupled modeling of nuclear reactor neutron kinetics, fuel heat transfer and coolant transport is a regular practice nowadays for transient analysis of reactor core. However optimization between modeling accuracy and computational economy has always been a challenging task to ensure the adequate degree of reliability in such extensive numerical exercises. Complex reactor core modeling involves estimation of evolving 3-D core thermal state, which in turn demands an expensive multichannel based detailed core thermal hydraulics model. A novel approach of power weighted coupling between core neutronics and thermal hydraulics presented in this work aims to reduce the bulk of core thermal calculations in core dynamics modeling to a significant extent without compromising accuracy of computation. Coupled core model has been validated against a series of international benchmarks. Accuracy and computational efficiency of the proposed multiphysics model has been demonstrated by analyzing a reactivity initiated transient.

  3. LMFBR type reactor

    Shimizu, Takeshi; Iida, Masaaki; Moriki, Yasuyuki


    A reactor core is divided into a plurality of coolants flowrate regions, and electromagnetic pumps exclusively used for each of the flowrate regions are disposed to distribute coolants flowrates in the reactor core. Further, the flowrate of each of the electromagnetic pumps is automatically controlled depending on signals from a temperature detector disposed at the exit of the reactor core, so that the flowrate of the region can be controlled optimally depending on the burning of reactor core fuels. Then, the electromagnetic pumps disposed for every divided region are controlled respectively, so that the coolants flowrate distribution suitable to each of the regions can be attained. Margin for fuel design is decreased, fuels are used effectively, as well as an operation efficiency can be improved. Moreover, since the electromagnetic pump has less flow resistance compared with a mechanical type pump, and flow resistance of the reactor core flowrate control mechanism is eliminated, greater circulating flowrate can be ensured after occurrence of accident in a natural convection using a buoyancy of coolants utilizable for after-heat removal as a driving force. (N.H.).

  4. U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

    Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.


    The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies.

  5. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Bragg-Sitton, Shannon M.; Morton, T. J.


    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  6. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Chaouadi, R


    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported.

  7. A transient, Hex-Z nodal code corrected by discontinuity factors. Volume 1: The transient nodal code; Final report

    Shatilla, Y.A.M.; Henry, A.F.


    This document constitutes Volume 1 of the Final Report of a three-year study supported by the special Research Grant Program for Nuclear Energy Research set up by the US Department of Energy. The original motivation for the work was to provide a fast and accurate computer program for the analysis of transients in heavy water or graphite-moderated reactors being considered as candidates for the New Production Reactor. Thus, part of the funding was by way of pass-through money from the Savannah River Laboratory. With this intent in mind, a three-dimensional (Hex-Z), general-energy-group transient, nodal code was created, programmed, and tested. In order to improve accuracy, correction terms, called {open_quotes}discontinuity factors,{close_quotes} were incorporated into the nodal equations. Ideal values of these factors force the nodal equations to provide node-integrated reaction rates and leakage rates across nodal surfaces that match exactly those edited from a more exact reference calculation. Since the exact reference solution is needed to compute the ideal discontinuity factors, the fact that they result in exact nodal equations would be of little practical interest were it not that approximate discontinuity factors, found at a greatly reduced cost, often yield very accurate results. For example, for light-water reactors, discontinuity factors found from two-dimensional, fine-mesh, multigroup transport solutions for two-dimensional cuts of a fuel assembly provide very accurate predictions of three-dimensional, full-core power distributions. The present document (volume 1) deals primarily with the specification, programming and testing of the three-dimensional, Hex-Z computer program. The program solves both the static (eigenvalue) and transient, general-energy-group, nodal equations corrected by user-supplied discontinuity factors.

  8. Transient regional osteoporosis.

    Cano-Marquina, Antonio; Tarín, Juan J; García-Pérez, Miguel-Ángel; Cano, Antonio


    Transient regional osteoporosis (TRO) is a disease that predisposes to fragility fracture in weight bearing joints of mid-life women and men. Pregnant women may also suffer the process, usually at the hip. The prevalence of TRO is lower than the systemic form, associated with postmenopause and advanced age, but may be falsely diminished by under-diagnosis. The disease may be uni- or bilateral, and may migrate to distinct joints. One main feature of TRO is spontaneous recovery. Pain and progressive limitation in the functionality of the affected joint(s) are key symptoms. In the case of the form associated with pregnancy, difficulties in diagnosis derive from the relatively young age at presentation and from the clinical overlapping with the frequent aches during gestation. Densitometric osteoporosis in the affected region is not always present, but bone marrow edema, with or without joint effusion, is detected by magnetic resonance. There are not treatment guidelines, but the association of antiresorptives to symptomatic treatment seems to be beneficial. Surgery or other orthopedic interventions can be required for specific indications, like hip fracture, intra-medullary decompression, or other.

  9. Applied hydraulic transients

    Chaudhry, M Hanif


    This book covers hydraulic transients in a comprehensive and systematic manner from introduction to advanced level and presents various methods of analysis for computer solution. The field of application of the book is very broad and diverse and covers areas such as hydroelectric projects, pumped storage schemes, water-supply systems, cooling-water systems, oil pipelines and industrial piping systems. Strong emphasis is given to practical applications, including several case studies, problems of applied nature, and design criteria. This will help design engineers and introduce students to real-life projects. This book also: ·         Presents modern methods of analysis suitable for computer analysis, such as the method of characteristics, explicit and implicit finite-difference methods and matrix methods ·         Includes case studies of actual projects ·         Provides extensive and complete treatment of governed hydraulic turbines ·         Presents design charts, desi...

  10. The Rapid Transient Surveyor

    Baranec, Christoph; Wright, Shelley A; Tonry, John; Tully, R Brent; Szapudi, István; Takamiya, Marianne; Hunter, Lisa; Riddle, Reed; Chen, Shaojie; Chun, Mark


    The Rapid Transient Surveyor (RTS) is a proposed rapid-response, high-cadence adaptive optics (AO) facility for the UH 2.2-m telescope on Maunakea. RTS will uniquely address the need for high-acuity and sensitive near-infrared spectral follow-up observations of tens of thousands of objects in mere months by combining an excellent observing site, unmatched robotic observational efficiency, and an AO system that significantly increases both sensitivity and spatial resolving power. We will initially use RTS to obtain the infrared spectra of ~4,000 Type Ia supernovae identified by the Asteroid Terrestrial-Impact Last Alert System over a two year period that will be crucial to precisely measuring distances and mapping the distribution of dark matter in the z < 0.1 universe. RTS will comprise an upgraded version of the Robo-AO laser AO system and will respond quickly to target-of-opportunity events, minimizing the time between discovery and characterization. RTS will acquire simultaneous-multicolor images with a...

  11. The rapid transient surveyor

    Baranec, C.; Lu, J. R.; Wright, S. A.; Tonry, J.; Tully, R. B.; Szapudi, I.; Takamiya, M.; Hunter, L.; Riddle, R.; Chen, S.; Chun, M.


    The Rapid Transient Surveyor (RTS) is a proposed rapid-response, high-cadence adaptive optics (AO) facility for the UH 2.2-m telescope on Maunakea. RTS will uniquely address the need for high-acuity and sensitive near-infrared spectral follow-up observations of tens of thousands of objects in mere months by combining an excellent observing site, unmatched robotic observational efficiency, and an AO system that significantly increases both sensitivity and spatial resolving power. We will initially use RTS to obtain the infrared spectra of 4,000 Type Ia supernovae identified by the Asteroid Terrestrial-Impact Last Alert System over a two year period that will be crucial to precisely measuring distances and mapping the distribution of dark matter in the z efficiency prism integral field unit spectrograph: R = 70-140 over a total bandpass of 840-1830nm with an 8.7" by 6.0" field of view (0.15" spaxels). The AO correction boosts the infrared point-source sensitivity of the spectrograph against the sky background by a factor of seven for faint targets, giving the UH 2.2-m the H-band sensitivity of a 5.7-m telescope without AO.

  12. Effect of automatic recirculation flow control on the transient response for Lungmen ABWR plant

    Tzang, Y.-C., E-mail: [National Tsing Hua University, Department of Engineering and System Science, Hsinchu 30013, Taiwan (China); Chiang, R.-F.; Ferng, Y.-M.; Pei, B.-S. [National Tsing Hua University, Department of Engineering and System Science, Hsinchu 30013, Taiwan (China)


    In this study the automatic mode of the recirculation flow control system (RFCS) for the Lungmen ABWR plant has been modeled and incorporated into the basic RETRAN-02 system model. The integrated system model is then used to perform the analyses for the two transients in which the automatic RFCS is involved. The two transients selected are: (1) one reactor internal pump (RIP) trip, and (2) loss of feedwater heating. In general, the integrated system model can predict well the response of key system parameters, including neutron flux, steam dome pressure, heat flux, RIP flow, core inlet flow, feedwater flow, steam flow, and reactor water level. The transients are also analyzed for manual RFCS case, between the automatic RFCS and the manual RFCS cases, comparisons of the transient response for the key system parameter show that the difference of transient response can be clearly identified. Also, the results show that the DELTACPR (delta critical power ratio) for the transients analyzed may not be less limiting for the automatic RFCS case under certain combination of control system settings.

  13. TRAC analyses of severe overcooling transients for the Oconee-1 PWR

    Ireland, J R [comp.


    This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs.

  14. Operation of Reactor


    3.1 Annual Report of SPR Operation Chu Shaochu Having overseen by National Nuclear Safety Administration and specialists, the reactor restarted up successfully after Safety renovation on April 16, 1996. In August 1996 the normal operation of SPR was approved by the authorities of Naitonal Nuclear Safety Administration. 1 Operation status In 1996, the reactor operated safely for 40 d and the energy released was about 137.3 MW·d. The operation status of SPR is shown in table 1. The reactor started up to higher power (power more than 1 MW) and lower power (for physics experiments) 4 times and 14 times respectively. Measurement of control rod efficiency and other measurement tasks were 2 times and 5 times respectively.

  15. Nuclear Rocket Engine Reactor

    Lanin, Anatoly


    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  16. Thermionic Reactor Design Studies

    Schock, Alfred


    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  17. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri


    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  18. Systems analysis of the CANDU 3 Reactor

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)


    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  19. An Overview of Reactor Concepts, a Survey of Reactor Designs.


    Public Affairs Office and is releasaole to the National Technical Information Services (NTIS). At NTIS, it will be available to the general public...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept...reactor core at the top and discharged at the Dotton while the reactor is in operation. The discharged fuel can then b inspected to see if it can De used

  20. Oscillatory flow chemical reactors

    Slavnić Danijela S.


    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  1. Influence of main variables modifications on accident transient based on AP1000-like MELCOR model

    Malicki, M.; Pieńkowski, L.


    Analysis of Severe Accidents (SA) is one of the most important parts of nuclear safety researches. MELCOR is a validated system code for severe accident analysis and as such it was used to obtain presented results. Analysed AP1000 model is based on publicly available data only. Sensitivity analysis was done for the main variables of primary reactor coolant system to find their influence on accident transient. This kind of analysis helps to find weak points of reactor design and the model itself. Performed analysis is a base for creation of Small Modular Reactor (SMR) generic model which will be the next step of the investigation aiming to estimate safety level of different reactors. Results clearly help to establish a range of boundary conditions for main the variables in future SMR model.

  2. Perspectives on reactor safety

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)


    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  3. Reactor Materials Research

    Van Walle, E


    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  4. Fusion Reactor Materials

    Decreton, M


    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

  5. Code Coupling for Multi-Dimensional Core Transient Analysis

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)


    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  6. Thermohydraulic and nuclear modeling of natural fission reactors

    Viggato, Jason Charles

    Experimental verification of proposed nuclear waste storage schemes in geologic repositories is not possible, however, a natural analog exists in the form of ancient natural reactors that existed in uranium-rich ores. Two billion years ago, the enrichment of natural uranium was high enough to allow a sustained chain reaction in the presence of water as a moderator. Several natural reactors occurred in Gabon, Africa and were discovered in the early 1970's. These reactors operated at low power levels for hundreds of thousands of years. Heated water generated from the reactors also leached uranium from the surrounding rock strata and deposited it in the reactor cores. This increased the concentration of uranium in the core over time and served to "refuel" the reactor. This has strong implications in the design of modern geologic repositories for spent nuclear fuel. The possibility of accidental fission events in man-made repositories exists and the geologic evidence from Oklo suggests how those events may progress and enhance local concentrations of uranium. Based on a review of the literature, a comprehensive code was developed to model the thermohydraulic behavior and criticality conditions that may have existed in the Oklo reactor core. A two-dimensional numerical model that incorporates modeling of fluid flow, temperatures, and nuclear fission and subsequent heat generation was developed for the Oklo natural reactors. The operating temperatures ranged from about 456 K to about 721 K. Critical reactions were observed for a wide range of concentrations and porosity values (9 to 30 percent UO2 and 10 to 20 percent porosity). Periodic operation occurred in the computer model prediction with UO2 concentrations of 30 percent in the core and 5 percent in the surrounding material. For saturated conditions and 30 percent porosity, the model predicted temperature transients with a period of about 5 hours. Kuroda predicted 3 to 4 hour durations for temperature transients

  7. Improved Prediction of the Temperature Feedback in TRISO-Fueled Reactors

    Javier Ortensi; Abderrafi M. Ougouag


    The Doppler feedback mechanism is a major contributor to the passive safety of gas-cooled, graphite-moderated high temperature reactors that use fuel based on Tristructural-Isotropic coated particles. It follows that the correct prediction of the magnitude and time-dependence of this feedback effect is essential to the conduct of safety analyses for these reactors. We present a fuel conduction model for obtaining better estimates of the temperature feedback during moderate and fast transients. The fuel model has been incorporated in the CYNOD-THERMIX-KONVEK suite of coupled codes as a single TRISO particle within each calculation cell. The heat generation rate is scaled down from the neutronic solution and a Dirichlet boundary condition is imposed as the bulk graphite temperature from the thermal-hydraulic solution. This simplified approach yields similar results to those obtained with more complex methods, requiring multi-TRISO calculations within one control volume, but with much less computational effort. An analysis of the hypothetical total control ejection in the PBMR-400 design verifies the performance of the code during fast transients. In addition, the analysis of the earthquake-initiated event in the PBMR-400 design verifies the performance of the code during slow transients. These events clearly depict the improvement in the predictions of the fuel temperature, and consequently, of the power escalations. In addition, a brief study of the potential effects of particle layer de-bonding on the transient behavior of high temperature reactors is included. Although the formation of a gap occurs under special conditions its consequences on the dynamic behavior of the reactor should be analyzed. The presence of a gap in the fuel can cause some unusual reactor behavior during fast transients, but still the reactor shuts down due to the strong feedback effects.

  8. Optimization and simplification of the concept of non-moderated Thorium Molten Salt Reactor

    Merle-Lucotte, Elsa; Heuer, Daniel; Allibert, Michel; Doligez, Xavier; Ghetta, Veronique; Le Brun, Christian [LPSC-IN2P3-CNRS/UJF/INPG, LPSC 53 avenue des Martyrs, 38026 Grenoble Cedex (France)


    Molten salt reactors, in the configuration presented here and called Thorium Molten Salt Reactor (TMSR), are particularly well suited to fulfil the criteria defined by the Generation IV forum, and may be operated in simplified and safe conditions in the Th/{sup 233}U fuel cycle with fluoride salts. The characteristics of the non-moderated TMSR based on a fast neutron spectrum are detailed in this paper: we aimed at designing an optimised TMSR with the simplest configuration. Using a simple kinetic-point model, we analyze the reactor's transient as the total reactivity margins are introduced in the core. We thus confirm, beyond the classical advantages of molten salt reactors, the satisfactory behaviour of the TMSR in terms of safety and the excellent level of stability which can be achieved in such reactors. (authors)

  9. Natural and mixed convection in the cylindrical pool of TRIGA reactor

    Henry, R.; Tiselj, I.; Matkovič, M.


    Temperature fields within the pool of the JSI TRIGA MARK II nuclear research reactor were measured to collect data for validation of the thermal hydraulics computational model of the reactor tank. In this context temperature of the coolant was measured simultaneously at sixty different positions within the pool during steady state operation and two transients. The obtained data revealed local peculiarities of the cooling water dynamics inside the pool and were used to estimate the coolant bulk velocity above the reactor core. Mixed natural and forced convection in the pool were simulated with a Computational Fluid Dynamics code. A relatively simple CFD model based on Unsteady RANS turbulence model was found to be sufficient for accurate prediction of the temperature fields in the pool during the reactor operation. Our results show that the simple geometry of the TRIGA pool reactor makes it a suitable candidate for a simple natural circulation benchmark in cylindrical geometry.

  10. High Flux Isotope Reactor (HFIR)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  11. Reactor operation environmental information document

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.


    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  12. Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

    Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.


    In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

  13. Description and identification of difficulties arising from the application of a cleaning process in operating conditions for the treatment of components used on liquid metal fast reactors (LMFR). A technical designed approach to avoid these situations.

    Rodriguez, G; Karpov, A V; Nalimov, Y P


    The cleaning process is one of the major maintenance operation for liquid metal fast reactors (LMFRs), both in operation and in their decommissioning stage. Russian and French cleaning processes are briefly described, including problems which have arisen during the processes. It appears that the cause of these problems is always connected to bad draining of the component, resulting in a vigorous reaction between vapour or liquid water and the bulk of sodium. From this discussion, the paper makes major recommendations for the efficient and safe cleaning of sodium wetted components, and proposes several processes which should be developed in order to deal with difficult situations, for example the removal of large amounts of undrainable sodium.

  14. Helioseismic Effects of Energetic Transients

    Ashok Ambastha


    Photospheric and chromospheric signatures related to large, energetic transients such as flares and CMEs, have been extensively reported during the last several years. In addition, energetic solar transients are expected to cause helioseismic effects. Some of the recent results are reviewed here; in particular, the helioseismic effects of the powerful flares in superactive region, NOAA 10486, including the 4B/X17 superflare of October 28, 2003. We also examine the temporal variations of power in low- modes during the period May 1995–October 2005, and compare with daily, disk-integrated flare- and CME-indices to infer the effect of transients on the scale of whole solar disk.

  15. Electromagnetic transients in power cables

    da Silva, Filipe Faria


    From the more basic concepts to the most advanced ones where long and laborious simulation models are required, Electromagnetic Transients in Power Cables provides a thorough insight into the study of electromagnetic transients and underground power cables. Explanations and demonstrations of different electromagnetic transient phenomena are provided, from simple lumped-parameter circuits to complex cable-based high voltage networks, as well as instructions on how to model the cables.Supported throughout by illustrations, circuit diagrams and simulation results, each chapter contains exercises,

  16. Synthesis of Model Based Robust Stabilizing Reactor Power Controller for Nuclear Power Plant

    Arshad Habib Malik


    Full Text Available In this paper, a nominal SISO (Single Input Single Output model of PHWR (Pressurized Heavy Water Reactor type nuclear power plant is developed based on normal moderator pump-up rate capturing the moderator level dynamics using system identification technique. As the plant model is not exact, therefore additive and multiplicative uncertainty modeling is required. A robust perturbed plant model is derived based on worst case model capturing slowest moderator pump-up rate dynamics and moderator control valve opening delay. Both nominal and worst case models of PHWR-type nuclear power plant have ARX (An Autoregressive Exogenous structures and the parameters of both models are estimated using recursive LMS (Least Mean Square optimization algorithm. Nominal and worst case discrete plant models are transformed into frequency domain for robust controller design purpose. The closed loop system is configured into two port model form and H? robust controller is synthesized. The H?controller is designed based on singular value loop shaping and desired magnitude of control input. The selection of desired disturbance attenuation factor and size of the largest anticipated multiplicative plant perturbation for loop shaping of H? robust controller form a constrained multi-objective optimization problem. The performance and robustness of the proposed controller is tested under transient condition of a nuclear power plant in Pakistan and found satisfactory.

  17. Supergiant Fast X-ray Transients

    Sidoli, Lara


    The phenomenology of a subclass of High Mass X-ray Binaries hosting a blue supergiant companion, the so-called Supergiant Fast X-ray Transients (SFXTs), is reviewed. Their number is growing, mainly thanks to the discoveries performed by the INTEGRAL satellite, then followed by soft X-rays observations (both aimed at refining the source position and at monitoring the source behavior) leading to the optical identification of the blue supergiant nature of the donor star. Their defining properties are a transient X-ray activity consisting of sporadic, fast and bright flares, (each with a variable duration between a few minutes and a few hours), reaching 1E36-1E37 erg/s. The quiescence is at a luminosity of 1E32 erg/s, while their more frequent state consists of an intermediate X-ray emission of 1E33-1E34 erg/s (1-10 keV). Only the brightest flares are detected by INTEGRAL (>17 keV) during short pointings, with no detected persistent emission. The physical mechanism driving the short outbursts is still debated, al...


    Whitham, G.K.; Smith, R.R.


    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  19. Nuclear Reactors and Technology

    Cason, D.L.; Hicks, S.C. [eds.


    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.


    King, L.D.P.


    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.


    Jonke, A.A.; Graae, J.E.A.; Levitz, N.M.


    A multistage fluidized bed reactor is described in which each of a number of stages is arranged with respect to an associated baffle so that a fluidizing gas flows upward and a granular solid downward through the stages and baffles, whereas the granular solid stopsflowing downward when the flow of fluidizing gas is shut off.

  2. Fusion reactor materials



    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  3. Integral Fast Reactor concept

    Till, C.E.; Chang, Y.I.


    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  4. The First Reactor.

    Department of Energy, Washington, DC.

    On December 2, 1942, in a racquet court underneath the West Stands of Stagg Field at the University of Chicago, a team of scientists led by Enrico Fermi created the first controlled, self-sustaining nuclear chain reaction. This updated and revised story of the first reactor (or "pile") is based on postwar interviews (as told to Corbin…

  5. Chromatographic and Related Reactors.


    special information about effects of surface heteroge- neity in the methanation reaction. Studies of an efficient multicolumn assembly for measuring...of organic basic catalysts such as pyridine and 4-methylpicoline. It was demonstrated that the chromatographic reactor gave special information about...Programmed Reaction to obtain special information about surface heterogeneity in the methanation reaction. Advantages of stopped flow over steady state

  6. New concepts for shaftless recycle reactors

    Berty, J.M.; Berty, I.J.


    Berty Reaction Engineers, Ltd. (BREL) is developing two new laboratory recycle reactors, the ROTOBERTY and the TURBOBERTY. These new reactors are basically improved versions of the original Berty reactor. To understand why the reactors have the features that they do, it is first necessary to briefly review laboratory reactors in general and specifically the original Berty reactor.

  7. Brazilian multipurpose reactor



    The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U{sub 3}Si{sub 2} dispersion-type Al having a density of up to 3.5 gU/cm{sup 3} and enrichment of 19.75% by weight of {sup 23{sup 5}}U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive

  8. Primary system thermal hydraulics of future Indian fast reactors

    Velusamy, K., E-mail: [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)


    Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

  9. Modeling Chemical Reactors I: Quiescent Reactors

    Michoski, C E; Schmitz, P G


    We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(...

  10. Transient or permanent fisheye views

    Jakobsen, Mikkel Rønne; Hornbæk, Kasper


    , about the benefits and limitations of transient visualizations. We describe an experiment that compares the usability of a fisheye view that participants could call up temporarily, a permanent fisheye view, and a linear view: all interfaces gave access to source code in the editor of a widespread......Transient use of information visualization may support specific tasks without permanently changing the user interface. Transient visualizations provide immediate and transient use of information visualization close to and in the context of the user’s focus of attention. Little is known, however...... programming environment. Fourteen participants performed varied tasks involving navigation and understanding of source code. Participants used the three interfaces for between four and six hours in all. Time and accuracy measures were inconclusive, but subjective data showed a preference for the permanent...

  11. Transient heating of moving objects

    E.I. Baida


    Full Text Available A mathematical model of transient and quasistatic heating of moving objects by various heat sources is considered. The mathematical formulation of the problem is described, examples of thermal calculation given.

  12. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Roth, R. J.


    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  13. Transient fuel behavior of preirradiated PWR fuels under reactivity initiated accident conditions

    Fujishiro, Toshio; Yanagisawa, Kazuaki; Ishijima, Kiyomi; Shiba, Koreyuki


    Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

  14. Chip-based device for parallel sorting, amplification, detection, and identification of nucleic acid subsequences

    Beer, Neil Reginald; Colston, Jr, Billy W.


    An apparatus for chip-based sorting, amplification, detection, and identification of a sample having a planar substrate. The planar substrate is divided into cells. The cells are arranged on the planar substrate in rows and columns. Electrodes are located in the cells. A micro-reactor maker produces micro-reactors containing the sample. The micro-reactor maker is positioned to deliver the micro-reactors to the planar substrate. A microprocessor is connected to the electrodes for manipulating the micro-reactors on the planar substrate. A detector is positioned to interrogate the sample contained in the micro-reactors.

  15. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)


    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  16. Reactor monitoring using antineutrino detectors

    Bowden, N. S.


    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

  17. Reactor vessel support system. [LMFBR

    Golden, M.P.; Holley, J.C.


    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  18. Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Center


    This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

  19. Space radiation studies at the White Sands Missile Range Fast Burst Reactor

    Delapaz, A.


    The operation of the White Sands Missile Range Fast Burst Reactor is discussed. Space radiation studies in radiobiology, dosimetry, and transient radiation effects on electronic systems and components are described. Proposed modifications to increase the capability of the facility are discussed.

  20. Fault diagnosis with the Aladdin transient classifier

    Roverso, Davide


    The purpose of Aladdin is to assist plant operators in the early detection and diagnosis of faults and anomalies in the plant that either have an impact on the plant performance, or that could lead to a plant shutdown or component damage if allowed to go unnoticed. The kind of early fault detection and diagnosis performed by Aladdin is aimed at allowing more time for decision making, increasing the operator awareness, reducing component damage, and supporting improved plant availability and reliability. In this paper we describe in broad lines the Aladdin transient classifier, which combines techniques such as recurrent neural network ensembles, Wavelet On-Line Pre-processing (WOLP), and Autonomous Recursive Task Decomposition (ARTD), in an attempt to improve the practical applicability and scalability of this type of systems to real processes and machinery. The paper focuses then on describing an application of Aladdin to a Nuclear Power Plant (NPP) through the use of the HAMBO experimental simulator of the Forsmark 3 boiling water reactor NPP in Sweden. It should be pointed out that Aladdin is not necessarily restricted to applications in NPPs. Other types of power plants, or even other types of processes, can also benefit from the diagnostic capabilities of Aladdin.

  1. Methanogenesis in Thermophilic Biogas Reactors

    Ahring, Birgitte Kiær


    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...... against Methanothrix soehngenii or Methanothrix CALS-I in any of the thermophilic biogas reactors examined. Studies using 2-14C-labeled acetate showed that at high concentrations (more than approx. 1 mM) acetate was metabolized via the aceticlastic pathway, transforming the methyl-group of acetate...

  2. LOFA and RIA analysis of the Indonesian Multipurpose research reactor RSG-GAS (1)

    Endiah Puji Hastuti; Hudi Hastowo; Iman Kuntoro [Center for Multipurpose Reactor, National Atomic Energy Agency (BATAN), Puspiptek, Serpong, Tangerang (Indonesia)


    Investigation on accident of the Indonesian Multipurpose research reactor RSG-GAS has been performed by computer simulation technique. Two groups of transients were considered, namely transient due to loss of primary cooling system (LOFA) and power excursion due to reactivity insertion (RIA). In such a transient condition, the Common Mode Failure (CMF) is considered and it will induce a situation so called unprotected transient or Anticipated Transient Without Scram (ATWS). RELAP5, PARET-ANL and EUREKA-2RR computer packages have been applied for these analyses. Simulations result done using these computer packages showed that in the occurrence of LOFA and RIA, failure on fuel elements is limited to the region with the highest power factor. (author)

  3. Neutron flux reduction programs for reactor pressure vessel

    Yoo, C.S. [Korea Atomic Energy Research Inst. KAERI, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, B.C. [Korea Reactor Integrity Surveillance Technology KRIST, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)


    The objective of this work is to implement various fast neutron flux reduction programs on the belt-line region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RT PTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the belt-line region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RT PTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RT PTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RT PTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library. (authors)

  4. Modeling of two-phase flow instabilities during startup transients utilizing RAMONA-4B methodology

    Paniagua, J.; Rohatgi, U.S.; Prasad, V.


    RAMONA-4B code is currently under development for simulating thermal hydraulic instabilities that can occur in Boiling Water Reactors (BWRs) and the Simplified Boiling Water Reactor (SBWR). As one of the missions of RAMONA-4B is to simulate SBWR startup transients, where geysering or condensation-induced instability may be encountered, the code needs to be assessed for this application. This paper outlines the results of the assessments of the current version of RAMONA-4B and the modifications necessary for simulating the geysering or condensation-induced instability. The test selected for assessment are the geysering tests performed by Prof Aritomi (1993).

  5. Gaia transient detection efficiency: hunting for nuclear transients

    Blagorodnova, Nadejda; Harrison, Diana L; Koposov, Sergey; Mattila, Seppo; Campbell, Heather; Walton, Nicholas A; Wyrzykowski, Lukasz


    We present a study of the detectability of transient events associated with galaxies for the Gaia European Space Agency astrometric mission. We simulated the on-board detections, and on-ground processing for a mock galaxy catalogue to establish the properties required for the discovery of transient events by Gaia, specifically tidal disruption events (TDEs) and supernovae (SNe). Transients may either be discovered by the on-board detection of a new source or by the brightening of a previously known source. We show that Gaia transients can be identified as new detections on-board for offsets from the host galaxy nucleus of 0.1--0.5,arcsec, depending on magnitude and scanning angle. The Gaia detection system shows no significant loss of SNe at close radial distances to the nucleus. We used the detection efficiencies to predict the number of transients events discovered by Gaia. For a limiting magnitude of 19, we expect around 1300 SNe per year: 65% SN Ia, 28% SN II and 7% SN Ibc, and ~20 TDEs per year.

  6. Structural integrity of nuclear reactor pressure vessels

    Knott, John F.


    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  7. The reactor Cabri; La pile cabri

    Ailloud, J.; Millot, J.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires


    It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m{sup 3}/h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under

  8. Compact fusion reactors

    CERN. Geneva


    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.


    Wheeler, J.A.


    A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

  10. Integrated Microfluidic Reactors.

    Lin, Wei-Yu; Wang, Yanju; Wang, Shutao; Tseng, Hsian-Rong


    Microfluidic reactors exhibit intrinsic advantages of reduced chemical consumption, safety, high surface-area-to-volume ratios, and improved control over mass and heat transfer superior to the macroscopic reaction setting. In contract to a continuous-flow microfluidic system composed of only a microchannel network, an integrated microfluidic system represents a scalable integration of a microchannel network with functional microfluidic modules, thus enabling the execution and automation of complicated chemical reactions in a single device. In this review, we summarize recent progresses on the development of integrated microfluidics-based chemical reactors for (i) parallel screening of in situ click chemistry libraries, (ii) multistep synthesis of radiolabeled imaging probes for positron emission tomography (PET), (iii) sequential preparation of individually addressable conducting polymer nanowire (CPNW), and (iv) solid-phase synthesis of DNA oligonucleotides. These proof-of-principle demonstrations validate the feasibility and set a solid foundation for exploring a broad application of the integrated microfluidic system.

  11. Reactor Neutrino Spectra

    Hayes, A. C.; Vogel, Petr


    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these spectra and their associated uncertainties is crucial for neutrino oscillation studies. The spectra used to date have been determined either by converting measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that make up the spectra, using modern databases as input. The uncertainties in the subdominant corrections to β-decay plague both methods, and we ...


    Greenstreet, B.L.


    A system for maintaining the alignment of moderator block structures in reactors is presented. Integral restraining grids are placed between each layer of blocks in the moderator structure, at the top of the uppermost layer, and at the bottom of the lowermost layer. Slots are provided in the top and bottom surfaces of the moderator blocks so as to provide a keying action with the grids. The grids are maintained in alignment by vertical guiding members disposed about their peripheries. (AEC)

  13. Transient tests on blower trip and rod removal at the HTR-10

    Hu Shouyin [Institute of Nuclear Energy Technology, Tsinghua University, P.O. Box 1021, Beijing 102201 (China)]. E-mail:; Wang Ruipian [Institute of Nuclear Energy Technology, Tsinghua University, P.O. Box 1021, Beijing 102201 (China); Gao Zuying [Institute of Nuclear Energy Technology, Tsinghua University, P.O. Box 1021, Beijing 102201 (China)


    Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes. Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results. Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 deg. C during two tests. The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW. The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased.

  14. FLODIS: a computer model to determine the flow distribution and thermal response of the Fort St. Vrain reactor

    Paul, D.D.


    FLODIS is a combined heat transfer and fluid flow analysis calculation written specifically for the core of the Fort St. Vrain reactor. It is a lumped-node representation of the 37 refueling regions in the active core. Heat conduction to the coolant and in the axial direction is represented; however, the effect of conduction between refueling regions is not included. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Flow distributions following reactor trip and a reduction in coolant pressure and flow are determined assuming that the orifice coefficients remain constant. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Results are presented for the evaluation of the transient for the Fort St. Vrain reactor following depressurization and cooling with two circulators operating at 8000 rpm.

  15. Application of a diffusion-reaction kinetic model for the removal of 4-chlorophenol in continuous tank reactors.

    Murcia, M D; Gómez, M; Bastida, J; Hidalgo, A M; Montiel, M C; Ortega, S


    A continuous tank reactor was used to remove 4-chlorophenol from aqueous solutions, using immobilized soybean peroxidase and hydrogen peroxide. The influence of operational variables (enzyme and substrate concentrations and spatial time) on the removal efficiency was studied. By using the kinetic law and the intrinsic kinetic parameters obtained in a previous work with a discontinuous tank reactor, the mass-balance differential equations of the transient state reactor model were solved and the theoretical conversion values were calculated. Several experimental series were used to obtain the values of the remaining model parameters by numerical calculation and using an error minimization algorithm. The model was checked by comparing the results obtained in some experiments (not used for the determination of the parameters) and the theoretical ones. The good concordance between the experimental and calculated conversion values confirmed that the design model can be used to predict the transient behaviour of the reactor.


    Heckman, T.P.


    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  17. The OPAL reactor

    Miller, R.; Irwin, T. [Australian Nuclear Science and Technology Organisation, Sydney (Australia); Ordonez, J.P. [INVAP SE, Bariloche (Argentina)


    The project to provide a replacement for Australia's HIFAR reactor began with governmental approval in September 1997 and reached its latest milestone with the achievement of the first full power operation of the OPAL reactor in November 2006. OPAL is a pool-type reactor with a thermal power of 20 MW and a fuel enrichment maximum of 20 per cent. This has been a successful project for both ANSTO (Australian Nuclear Science and Technology Organisation) and the contractor INVAP SE. This project was characterised by extensive interaction with the project's stake-holders during project definition and the use of a performance-based turnkey contract which gave the contractor the maximum opportunity to optimise the design to achieve performance and cost effectiveness. The contactor provided significant in-house resources as well as capacity to manage an international team of suppliers and sub-contractors. A key contributor to the project's successful outcomes has been the development and maintenance of an excellent working relationship between ANSTO and INVAP project teams. Commissioning was undertaken in accordance with the IAEA recommended stages. This paper presents the approaches used to define the project requirements, to choose the supplier and to deliver the project. The main results of hot commissioning are reviewed and the problems encountered examined. Operational experience since hot commissioning is also reviewed.

  18. Nuclear research reactors in Brazil

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)


    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  19. Nuclear research reactors in Brazil

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)


    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  20. Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

    Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

  1. Recent development of transient electronics

    Huanyu Cheng


    Full Text Available Transient electronics are an emerging class of electronics with the unique characteristic to completely dissolve within a programmed period of time. Since no harmful byproducts are released, these electronics can be used in the human body as a diagnostic tool, for instance, or they can be used as environmentally friendly alternatives to existing electronics which disintegrate when exposed to water. Thus, the most crucial aspect of transient electronics is their ability to disintegrate in a practical manner and a review of the literature on this topic is essential for understanding the current capabilities of transient electronics and areas of future research. In the past, only partial dissolution of transient electronics was possible, however, total dissolution has been achieved with a recent discovery that silicon nanomembrane undergoes hydrolysis. The use of single- and multi-layered structures has also been explored as a way to extend the lifetime of the electronics. Analytical models have been developed to study the dissolution of various functional materials as well as the devices constructed from this set of functional materials and these models prove to be useful in the design of the transient electronics.

  2. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    Choi, Jae Young; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of)


    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity.

  3. Petroleum Degradation in Soil by Thermophilic Bacteria with Biopile Reactor

    Astri Nugroho


    Full Text Available Crude oil degradation has been carried out using biopile reactor in TPH concentration of 5%, 10% and 15%. The thermophilic microorganism used from isolation result and identification are Aeromonas salmonicida, Bacillus pantothenticus, and Stenotrophomonas maltophilia. Biodegrade of biopile reactor done by various concentration Total Petroleum Hydrocarbon (TPH, Total Plate Count (TPC, and Volatile Suspended Solid (VSS per day during 30 day. Biodegrade kinetic parameter calculated are m, mm, Y, Yt, Yobs, Kd, Ks from TPH concentration decision, TPC and VSS in every microorganism with t (observation time that is 0 hour to 168 hour. Crude oil separation efficiency in a biople reactor shows that the largest separation occurs on a starting TPH concentrate of 15% which was 61.8% later on followed on a starting TPH concentrate of 10% and 5% which was as much as 61% and 48.4%.

  4. Searches for Exotic Transient Signals with a Global Network of Optical Magnetometers for Exotic Physics

    Pustelny, S


    In this letter, we describe a novel scheme for searching for physics beyond the Standard Model. The idea is based on correlation of time-synchronized readouts of distant ($\\gtrsim$100~km) optical magnetometers. Such an approach limits hard-to-identify local transient noise, providing the system with unique capabilities of identification of global transient events. Careful analysis of the signal can reveal the nature of the events (e.g., its nonmagnetic origin), which opens avenues for new class of exotic-physics searches (searches for global transient exotic spin couplings) and tests of yet unverified theoretical models.

  5. Thermionic Reactor Design Studies

    Schock, Alfred


    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic

  6. The effectiveness of using the calculated braking current for longitudinal differential protection of 110 - 750 kV shunt reactors

    Vdovin, S. A. [JSC ' E and E' (Russian Federation); Shalimov, A. S. [LLC Selekt Co. (Russian Federation)


    The use of the function of effective current braking of the longitudinal differential protection of shunt reactors to offset current surges, which enables the sensitivity of differential protection to be increased when there are short circuits with low damage currents, is considered. It is shown that the use of the calculated braking characteristic enables the reliability of offset protection from transients to be increased when the reactor is connected, which is accompanied by the flow of asymmetric currents containing an aperiodic component.

  7. Electromagnetic Transients in Power Cables

    Silva, Filipe Faria Da; Bak, Claus Leth

    of electromagnetic phenomena associated to their operation, among them electromagnetic transients, increased as well. Transient phenomena have been studied since the beginning of power systems, at first using only analytical approaches, which limited studies to more basic phenomena; but as computational tools became...... concerning HVAC cables. An important topic that is not covered in this book is measurements protocols/ methods. The protocols used when performing measurements on a cable depend on what is to be measured, the available equipment and accessibility. Readers interested in the topic are referred to search....... The chapter ends by proposing a systematic method that can be used when doing the insulation co-ordination study for a line, as well as the modelling requirements, both modelling depth and modelling detail of the equipment, for the study of the different types of transients followed by a step-by-step generic...

  8. Transient Faults in Computer Systems

    Masson, Gerald M.


    A powerful technique particularly appropriate for the detection of errors caused by transient faults in computer systems was developed. The technique can be implemented in either software or hardware; the research conducted thus far primarily considered software implementations. The error detection technique developed has the distinct advantage of having provably complete coverage of all errors caused by transient faults that affect the output produced by the execution of a program. In other words, the technique does not have to be tuned to a particular error model to enhance error coverage. Also, the correctness of the technique can be formally verified. The technique uses time and software redundancy. The foundation for an effective, low-overhead, software-based certification trail approach to real-time error detection resulting from transient fault phenomena was developed.

  9. Cohabitation Duration and Transient Domesticity.

    Golub, Andrew; Reid, Megan; Strickler, Jennifer; Dunlap, Eloise


    Research finds that many impoverished urban Black adults engage in a pattern of partnering and family formation involving a succession of short cohabitations yielding children, a paradigm referred to as transient domesticity. Researchers have identified socioeconomic status, cultural adaptations, and urbanicity as explanations for aspects of this pattern. We used longitudinal data from the 2001 Survey of Income and Program Participation to analyze variation in cohabitation and marriage duration by race/ethnicity, income, and urban residence. Proportional hazards regression indicated that separation risk is greater among couples that are cohabiting, below 200% of the federal poverty line, and Black but is not greater among urban dwellers. This provides empirical demographic evidence to support the emerging theory of transient domesticity and suggests that both socioeconomic status and race explain this pattern. We discuss the implications of these findings for understanding transient domesticity and make recommendations for using the Survey of Income and Program Participation to further study this family formation paradigm.

  10. Transient critical heat flux and blowdown heat-transfer studies

    Leung, J.C.


    Objective of this study is to give a best-estimate prediction of transient critical heat flux (CHF) during reactor transients and hypothetical accidents. To accomplish this task, a predictional method has been developed. Basically it involves the thermal-hydraulic calculation of the heated core with boundary conditions supplied from experimental measurements. CHF predictions were based on the instantaneous ''local-conditions'' hypothesis, and eight correlations (consisting of round-tube, rod-bundle, and transient correlations) were tested against most recent blowdown heat-transfer test data obtained in major US facilities. The prediction results are summarized in a table in which both CISE and Biasi correlations are found to be capable of predicting the early CHF of approx. 1 s. The Griffith-Zuber correlation is credited for its prediction of the delay CHF that occurs in a more tranquil state with slowly decaying mass velocity. In many instances, the early CHF can be well correlated by the x = 1.0 criterion; this is certainly indicative of an annular-flow dryout-type crisis. The delay CHF occurred at near or above 80% void fraction, and the success of the modified Zuber pool-boiling correlation suggests that this CHF is caused by flooding and pool-boiling type hydrodynamic crisis.

  11. The Effect of Current-Limiting Reactors on the Tripping of Short Circuits in High-Voltage Electrical Equipment

    Volkov, M. S.; Gusev, Yu. P., E-mail:; Monakov, Yu. V.; Cho, Gvan Chun [National Research University “Moscow Power Engineering Institute,” (Russian Federation)


    The insertion of current-limiting reactors into electrical equipment operating at a voltage of 110 and 220 kV produces a change in the parameters of the transient recovery voltages at the contacts of the circuit breakers for disconnecting short circuits, which could be the reason for the increase in the duration of the short circuit, damage to the electrical equipment and losses in the power system. The results of mathematical modeling of the transients, caused by tripping of the short circuit in a reactive electric power transmission line are presented, and data are given on the negative effect of a current-limiting resistor on the rate of increase and peak value of the transient recovery voltages. Methods of ensuring the standard requirements imposed on the parameters of the transient recovery voltages when using current-limiting reactors in the high-voltage electrical equipment of power plants and substations are proposed and analyzed.

  12. Transient Testing of Nuclear Fuels and Materials in the United States

    Wachs, Daniel M.


    The United States has established that transient irradiation testing is needed to support advanced light water reactors fuel development. The U.S. Department of Energy (DOE) has initiated an effort to reestablish this capability. Restart of the Transient Testing Reactor (TREAT) facility located at the Idaho National Laboratory (INL) is being considered for this purpose. This effort would also include the development of specialized test vehicles to support stagnant capsule and flowing loop tests as well as the enhancement of postirradiation examination capabilities and remote device assembly capabilities at the Hot Fuel Examination Facility. It is anticipated that the capability will be available to support testing by 2018, as required to meet the DOE goals for the development of accident-tolerant LWR fuel designs.

  13. Thermal reactor safety


    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

  14. New reactors for laboratory studies

    Berty, J.M.


    Recent developments in design of laboratory and bench-scale reactors reflect mostly the developments in reaction engineering; that is the improved understanding of physical and chemical rate limiting processes, their interactions, and their effects on commercial-scale reactor performance. Whether a laboratory reactor is used to study the fundamentals of a commercial process or for pure scientific interest, it is important to know what physical or chemical process is limiting or influencing the rate and selectivity. To clarify this, a definition is required of the regime where physical influences exist, and study the intrinsic kinetics at conditions where physical processes do not affect the rate. Reactors are illustrated whose design was influenced by the above considerations. These reactors produce results which are independent of the reactors in which they were measured, and which can be scaled up with up-to-date reaction engineering techniques.

  15. Transient osteoporosis of the hip.

    Mirza, Rabeea; Ishaq, Saliha; Amjad, Hira


    Transient Osteoporosis of Hip (TOH) is an uncommon disorder of idiopathic nature, particularly in the Asian population. It has been described to mostly occur in middle aged men and women in their third trimester of pregnancy. A distinctive hallmark of this condition is that it is self limiting and resolves in a few months. The patient presents to the physician with pain on movement and impaired mobility of the affected joint, developing without any history of trauma. MRI is the main diagnostic tool. We report herein a case of a forty five year old male, who developed transient osteoporosis of the hip, and was managed conservatively.

  16. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.


    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  17. Spiral-shaped disinfection reactors

    Ghaffour, Noreddine


    This disclosure includes disinfection reactors and processes for the disinfection of water. Some disinfection reactors include a body that defines an inlet, an outlet, and a spiral flow path between the inlet and the outlet, in which the body is configured to receive water and a disinfectant at the inlet such that the water is exposed to the disinfectant as the water flows through the spiral flow path. Also disclosed are processes for disinfecting water in such disinfection reactors.

  18. Turning points in reactor design

    Beckjord, E.S.


    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  19. Acceptability of reactors in space

    Buden, D.


    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

  20. Hydrogen Production in Fusion Reactors

    Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M; Uenosono, C.


    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated.

  1. Fast reactor programme in India

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar


    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  2. Transient visual evoked neuromagnetic responses: Identification of multiple sources

    Aine, C.; George, J.; Medvick, P.; Flynn, E.; Bodis-Wollner, I.; Supek, S.


    Neuromagnetic measurements and associated modeling procedures must be able to resolve multiple sources in order to localize and accurately characterize the generators of visual evoked neuromagnetic activity. Workers have identified at least 11 areas in the macaque, throughout occipital, parietal, and temporal cortex, which are primarily or entirely visual in function. The surface area of the human occipital lobe is estimated to be 150--250cm. Primary visual cortex covers approximately 26cm/sup 2/ while secondary visual areas comprise the remaining area. For evoked response amplitudes typical of human MEG data, one report estimates that a two-dipole field may be statistically distinguishable from that of a single dipole when the separation is greater than 1--2 cm. Given the estimated expanse of cortex devoted to visual processes, along with this estimate of resolution limits it is likely that MEG can resolve sources associated with activity in multiple visual areas. Researchers have noted evidence for the existence of multiple sources when presenting visual stimuli in a half field; however, they did not attempt to localize them. We have examined numerous human MEG field patterns resulting from different visual field placements of a small sinusoidal grating which suggest the existence of multiple sources. The analyses we have utilized for resolving multiple sources in these studies differ depending on whether there was evidence of (1) synchronous activation of two spatially discrete sources or (2) two discrete asynchronous sources. In some cases we have observed field patterns which appear to be adequately explained by a single source changing its orientation and location across time. 4 refs., 2 figs.

  3. Transient Splitting of Conoscopic Isogyres of a Uniaxial Nematic

    Kim, Young-Ki; Senuk, Bohdan; Tortora, Luana; Sprunt, Samuel; Lehmann, Matthias; Lavrentovich, Oleg D.


    The phase identification is often based on conoscopic observations of homeotropic cells: A uniaxial nematic produces a pattern with crossed isogyres, while the biaxial nematic shows a split of isogyres. We demonstrate that the splitting of isogyres occurs even when the material remains in the uniaxial nematic phase. In particular, in the bent core material J35, splitting of isogyres is caused by change of the temperature. The effect is transient and the isogyres return to a uniaxial (crossed) configuration after a certain time that depends on sample thickness, temperature, and rate of temperature change; the time varies from a few seconds to tens of hours. The transient splitting is caused by the temperature-induced material flow that triggers a (uniaxial) director tilt in the cell. The flows and the director tilt are demonstrated by the CARS microscopy and fluorescent confocal polarizing microscopy (FCPM). This transient effect is general and can be observed even in E7 and 5CB. The effect should be considered in textural identifications of potential biaxial nematic materials.

  4. Transient critical heat flux under flow coast-down in vertical annulus with non-uniform heat flux distribution

    Moon, S.K.; Chun, S.Y.; Choi, K.Y.; Yang, S.K. [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)


    An experimental study on transient critical heat flux (CHF) under flow coast-down has been performed for water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady state CHF. The transient CHF experiments have been performed for three kinds of flow transient modes based on the coast-down data of the Kori 3/4 nuclear power plant reactor coolant pump. Most of the CHFs occurred in the annular-mist flow regime. Thus, it means that the possible CHF mechanism might be the liquid film dryout in the annular-mist flow regime. For flow transient mode with the smallest flow reduction rate, the time-to-CHF is the largest. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to-CHF becomes large as the heat flux decreases. Usually, the critical mass flux is large for slow flow reduction. There is a pressure effect on the ratio of the transient CHF data to steady state CHF data. Some conventional correlations show relatively better CHF prediction results for high system pressure, high quality and slow transient modes than for low system pressure, low quality and fast transient modes. (author)

  5. The k-[epsilon] modeling of deboration transients in a PWR

    Oosterkamp, W.J.; Termaat, K.P.; Verhagen, F.C.M. (N.V. KEMA, Arnhem (Netherlands))


    The potential for reactivity accidents is receiving more attention after the Chernobyl disaster. Boron dilution transients are one class of reactivity accidents possible in pressurized water reactors (PWRs). Severe boron dilution reactivity accidents can only occur when three conditions are met: (1) a source of nonborated water is attaached to the primary system; (2) conditions are such that this nonborated water accumulates undetected outside the core; and (3) the nonborated water is rapidly moved into the core.

  6. Neutrino Oscillation Studies with Reactors

    Vogel, Petr; Zhang, Chao


    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  7. Enhancement of proteolysis through the silica-gel-derived microfluidic reactor.

    Liu, Yun; Qu, Haiyun; Xue, Yan; Wu, Zhonglin; Yang, Pengyuan; Liu, Baohong


    An on-chip enzymatic reactor providing rapid protein digestion is presented. Trypsin-embedding stationary phase within the microchannel has been prepared by the sol-gel method. Such a microfluidic reactor has been used for low-level protein digestion at 16 fmol per analysis. The analytical potential of the microreactor combined with the strong cation exchange and RPLC ESI-MS/MS for the identification of real samples from the cytoplasma of the human liver tissue has been demonstrated.

  8. Improved lumped models for transient combined convective and radiative cooling of a two-layer spherical fuel element

    Silva, Alice Cunha da; Su, Jian, E-mail:, E-mail: [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil)


    The High Temperature Gas cooled Reactor (HTGR) is a fourth generation thermal nuclear reactor, graphite-moderated and helium cooled. The HTGRs have important characteristics making essential the study of these reactors, as well as its fuel element. Examples of these are: high thermal efficiency,low operating costs and construction, passive safety attributes that allow implication of the respective plants. The Pebble Bed Modular Reactor (PBMR) is a HTGR with spherical fuel elements that named the reactor. This fuel element is composed by a particulate region with spherical inclusions, the fuel UO2 particles, dispersed in a graphite matrix and a convective heat transfer by Helium happens on the outer surface of the fuel element. In this work, the transient heat conduction in a spherical fuel element of a pebble-bed high temperature reactor was studied in a transient situation of combined convective and radiative cooling. Improved lumped parameter model was developed for the transient heat conduction in the two-layer composite sphere subjected to combined convective and radiative cooling. The improved lumped model was obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of the two-layer spherical fuel element was analyzed to illustrate the applicability of the proposed lumped model, with respect to die rent values of the Biot number, the radiation-conduction parameter, the dimensionless thermal contact resistance, the dimensionless inner diameter and coating thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the improved lumped model, with H2,1/H1,1/H0,0 approximation yielded significant improvement of average temperature prediction over the classical lumped model. (author)

  9. Westinghouse Small Modular Reactor nuclear steam supply system design

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)


    generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)

  10. Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

    Mireles, Omar R.; Houts, Michael G.


    Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

  11. Accelerator based fusion reactor

    Liu, Keh-Fei; Chao, Alexander Wu


    A feasibility study of fusion reactors based on accelerators is carried out. We consider a novel scheme where a beam from the accelerator hits the target plasma on the resonance of the fusion reaction and establish characteristic criteria for a workable reactor. We consider the reactions d+t\\to n+α,d+{{}3}{{H}\\text{e}}\\to p+α , and p+{{}11}B\\to 3α in this study. The critical temperature of the plasma is determined from overcoming the stopping power of the beam with the fusion energy gain. The needed plasma lifetime is determined from the width of the resonance, the beam velocity and the plasma density. We estimate the critical beam flux by balancing the energy of fusion production against the plasma thermo-energy and the loss due to stopping power for the case of an inert plasma. The product of critical flux and plasma lifetime is independent of plasma density and has a weak dependence on temperature. Even though the critical temperatures for these reactions are lower than those for the thermonuclear reactors, the critical flux is in the range of {{10}22}-{{10}24}~\\text{c}{{\\text{m}}-2}~{{\\text{s}}-1} for the plasma density {ρt}={{10}15}~\\text{c}{{\\text{m}}-3} in the case of an inert plasma. Several approaches to control the growth of the two-stream instability are discussed. We have also considered several scenarios for practical implementation which will require further studies. Finally, we consider the case where the injected beam at the resonance energy maintains the plasma temperature and prolongs its lifetime to reach a steady state. The equations for power balance and particle number conservation are given for this case.

  12. Biparticle fluidized bed reactor

    Scott, C.D.


    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.


    Snell, A.H.


    This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

  14. Solution of the reactor point kinetics equations by MATLAB computing

    Singh Sudhansu S.


    Full Text Available The numerical solution of the point kinetics equations in the presence of Newtonian temperature feedback has been a challenging issue for analyzing the reactor transients. Reactor point kinetics equations are a system of stiff ordinary differential equations which need special numerical treatments. Although a plethora of numerical intricacies have been introduced to solve the point kinetics equations over the years, some of the simple and straightforward methods still work very efficiently with extraordinary accuracy. As an example, it has been shown recently that the fundamental backward Euler finite difference algorithm with its simplicity has proven to be one of the most effective legacy methods. Complementing the back-ward Euler finite difference scheme, the present work demonstrates the application of ordinary differential equation suite available in the MATLAB software package to solve the stiff reactor point kinetics equations with Newtonian temperature feedback effects very effectively by analyzing various classic benchmark cases. Fair accuracy of the results implies the efficient application of MATLAB ordinary differential equation suite for solving the reactor point kinetics equations as an alternate method for future applications.

  15. Preliminary accident analysis of Flexblue® underwater reactor

    Haratyk Geoffrey


    Full Text Available Flexblue® is a subsea-based, transportable, small modular reactor delivering 160 MWe. Immersion provides the reactor with an infinite heat sink – the ocean – around the metallic hull. The reference design includes a loop-type PWR with two horizontal steam generators. The safety systems are designed to operate passively; safety functions are fulfilled without operator action and external electrical input. Residual heat is removed through four natural circulation loops: two primary heat exchangers immersed in safety tanks cooled by seawater and two emergency condensers immersed in seawater. In case of a primary piping break, a two-train safety injection system is actuated. Each train includes a core makeup tank, an accumulator and a safety tank at low pressure. To assess the capability of these features to remove residual heat, the reactor and its safety systems have been modelled using thermal-hydraulics code ATHLET with conservative assumptions. The results of simulated transients for three typical PWR accidents are presented: a turbine trip with station blackout, a large break loss of coolant accident and a small break loss of coolant accident. The analyses show that the safety criteria are respected and that the reactor quickly reaches a safe shutdown state without operator action and external power.

  16. Development of an automated core model for nuclear reactors

    Mosteller, R.D.


    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input.

  17. Model biases in high-burnup fast reactor simulations

    Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)


    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  18. The Agile Alert System For Gamma-Ray Transients

    Bulgarelli, A; Gianotti, F; Tavani, M; Parmiggiani, N; Fioretti, V; Chen, A W; Vercellone, S; Pittori, C; Verrecchia, F; Lucarelli, F; Santolamazza, P; Fanari, G; Giommi, P; Beneventano, D; Argan, A; Trois, A; Scalise, E; Longo, F; Pellizzoni, A; Pucella, G; Colafrancesco, S; Conforti, V; Tempesta, P; Cerone, M; Sabatini, P; Annoni, G; Valentini, G; Salotti, L


    In recent years, a new generation of space missions offered great opportunities of discovery in high-energy astrophysics. In this article we focus on the scientific operations of the Gamma-Ray Imaging Detector (GRID) onboard the AGILE space mission. The AGILE-GRID, sensitive in the energy range of 30 MeV-30 GeV, has detected many gamma-ray transients of galactic and extragalactic origins. This work presents the AGILE innovative approach to fast gamma-ray transient detection, which is a challenging task and a crucial part of the AGILE scientific program. The goals are to describe: (1) the AGILE Gamma-Ray Alert System, (2) a new algorithm for blind search identification of transients within a short processing time, (3) the AGILE procedure for gamma-ray transient alert management, and (4) the likelihood of ratio tests that are necessary to evaluate the post-trial statistical significance of the results. Special algorithms and an optimized sequence of tasks are necessary to reach our goal. Data are automatically ...

  19. High-order polynomial expansions for reactor kinetics

    Molina, J.L. [Instituto Balseiro, San Carlos de Bariloche (Argentina); Jatuff, F.E. [Investigacion Aplicada SE (INVAP), San Carlos de Bariloche (Argentina)


    Laguerre, Hermite and Legendre polynomial bases were studied for high order time expansions of reactor kinetics solutions. A theorem showing an exponential majoring function for the solution of bounded reactivity transients introduce Laguerre, Hermite and Legendre polynomials for semi-infinite, infinite and finite time domains, respectively. The numerical solutions were obtained by means of the construction of an error estimator and its minimization using a conventional variational method. Some point reactor kinetics problems with exact solution were tested. The results showed a numerical monotone convergent behavior and accuracy, but problem-dependent efficiency caused by the extremely large expansion orders (more than 200 terms) needed in the studied bases for the cases with large reactivity insertions. (author) 13 refs., 5 figs., 4 tabs.

  20. Transient filament stretching rheometer II

    Kolte, Mette Irene; Rasmussen, Henrik K.; Hassager, Ole


    The Lagrangian sspecification is used to simulate the transient stretching filament rheometer. Simulations are performed for dilute PIB-solutions modeled as a four mode Oldroyd-B fluid and a semidilute PIB-solution modeled as a non-linear single integral equation. The simulations are compared...