WorldWideScience

Sample records for core model nuclear

  1. Nuclear reactor core modelling in multifunctional simulators

    International Nuclear Information System (INIS)

    Puska, E.K.

    1999-01-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  2. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  3. Real-time advanced nuclear reactor core model

    International Nuclear Information System (INIS)

    Koclas, J.; Friedman, F.; Paquette, C.; Vivier, P.

    1990-01-01

    The paper describes a multi-nodal advanced nuclear reactor core model. The model is based on application of modern equivalence theory to the solution of neutron diffusion equation in real time employing the finite differences method. The use of equivalence theory allows the application of the finite differences method to cores divided into hundreds of nodes, as opposed to the much finer divisions (in the order of ten thousands of nodes) where the unmodified method is currently applied. As a result the model can be used for modelling of the core kinetics for real time full scope training simulators. Results of benchmarks, validate the basic assumptions of the model and its applicability to real-time simulation. (orig./HP)

  4. Evaluating nuclear physics inputs in core-collapse supernova models

    Science.gov (United States)

    Lentz, E.; Hix, W. R.; Baird, M. L.; Messer, O. E. B.; Mezzacappa, A.

    Core-collapse supernova models depend on the details of the nuclear and weak interaction physics inputs just as they depend on the details of the macroscopic physics (transport, hydrodynamics, etc.), numerical methods, and progenitors. We present preliminary results from our ongoing comparison studies of nuclear and weak interaction physics inputs to core collapse supernova models using the spherically-symmetric, general relativistic, neutrino radiation hydrodynamics code Agile-Boltztran. We focus on comparisons of the effects of the nuclear EoS and the effects of improving the opacities, particularly neutrino--nucleon interactions.

  5. Nuclear clustering - a cluster core model study

    International Nuclear Information System (INIS)

    Paul Selvi, G.; Nandhini, N.; Balasubramaniam, M.

    2015-01-01

    Nuclear clustering, similar to other clustering phenomenon in nature is a much warranted study, since it would help us in understanding the nature of binding of the nucleons inside the nucleus, closed shell behaviour when the system is highly deformed, dynamics and structure at extremes. Several models account for the clustering phenomenon of nuclei. We present in this work, a cluster core model study of nuclear clustering in light mass nuclei

  6. Modeling of the core of Atucha II nuclear power plant

    International Nuclear Information System (INIS)

    Blanco, Anibal

    2007-01-01

    This work is part of a Nuclear Engineer degree thesis of the Instituto Balseiro and it is carried out under the development of an Argentinean Nuclear Power Plant Simulator. To obtain the best representation of the reactor physical behavior using the state of the art tools this Simulator should couple a 3D neutronics core calculation code with a thermal-hydraulics system code. Focused in the neutronic nature of this job, using PARCS, we modeled and performed calculations of the nuclear power plant Atucha 2 core. Whenever it is possible, we compare our results against results obtained with PUMA (the official core code for Atucha 2). (author) [es

  7. Nuclear characteristic simulation device for reactor core

    International Nuclear Information System (INIS)

    Arakawa, Akio; Kobayashi, Yuji.

    1994-01-01

    In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

  8. Modelling of core protection and monitoring system for PWR nuclear power plant simulator

    International Nuclear Information System (INIS)

    Jung Kun Lee; Byoung Sung Han

    1997-01-01

    A nuclear power plant simulator was developed for Younggwang units 3 and 4 nuclear power plant (YGN Nos 3 and 4) in Korea; it has been in operation on training center since November 1996. The core protection calculator (CPC) and the core operating limit supervisory system (COLSS) for the simulator were also developed. The CPC is a digital computer-based core protection system, which performs on-line calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). It initiates reactor trip when the core conditions exceed designated DNBR or LPD limitations. The COLSS is designed to assist operators by implementing the limiting conditions for operations in the technical specifications. With these systems, it is possible to increase capacity factor and safety of nuclear power plants, because the COLSS data can show accurate operation margin to plant operators and the CPC can protect reactor core. In this study, the function of CPC/COLSS is analyzed in detail, and then simulation model for CPC/COLSS is presented based on the function. Compared with the YGN Nos 3 and 4 plant operation data and CEDIPS/COLSS FORTRAN code test results, the predictions with the model show reasonable results. (Author)

  9. NUCORE - A system for nuclear structure calculations with cluster-core models

    International Nuclear Information System (INIS)

    Heras, C.A.; Abecasis, S.M.

    1982-01-01

    Calculation of nuclear energy levels and their electromagnetic properties, modelling the nucleus as a cluster of a few particles and/or holes interacting with a core which in turn is modelled as a quadrupole vibrator (cluster-phonon model). The members of the cluster interact via quadrupole-quadrupole and pairing forces. (orig.)

  10. State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

    2017-02-15

    A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  11. 3D Core Model for simulation of nuclear power plants: Simulation requirements, model features, and validation

    International Nuclear Information System (INIS)

    Zerbino, H.

    1999-01-01

    In 1994-1996, Thomson Training and Simulation (TT and S) earned out the D50 Project, which involved the design and construction of optimized replica simulators for one Dutch and three German Nuclear Power Plants. It was recognized early on that the faithful reproduction of the Siemens reactor control and protection systems would impose extremely stringent demands on the simulation models, particularly the Core physics and the RCS thermohydraulics. The quality of the models, and their thorough validation, were thus essential. The present paper describes the main features of the fully 3D Core model implemented by TT and S, and its extensive validation campaign, which was defined in extremely positive collaboration with the Customer and the Core Data suppliers. (author)

  12. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated

  13. Scale model study of the seismic response of a nuclear reactor core

    International Nuclear Information System (INIS)

    Dove, R.C.; Dunwoody, W.E.; Rhorer, R.L.

    1983-01-01

    The use of scale models to study the dynamics of a system of graphite core blocks used in certain nuclear reactor designs is described. Scaling laws, material selecton, model instrumentation to measure collision forces, and the response of several models to simulated seismic excitation are covered. The effects of Coulomb friction between the blocks and the clearance gaps between the blocks on the system response to seismic excitation are emphasized

  14. Development of an automated core model for nuclear reactors

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1998-01-01

    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

  15. Nuclear reactor core catcher

    International Nuclear Information System (INIS)

    1977-01-01

    A nuclear reactor core catcher is described for containing debris resulting from an accident causing core meltdown and which incorporates a method of cooling the debris by the circulation of a liquid coolant. (U.K.)

  16. Nuclear reactor core flow baffling

    International Nuclear Information System (INIS)

    Berringer, R.T.

    1979-01-01

    A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

  17. Phase diagram of nuclear 'pasta' and its uncertainties in supernova cores

    International Nuclear Information System (INIS)

    Sonoda, Hidetaka; Watanabe, Gentaro; Sato, Katsuhiko; Yasuoka, Kenji; Ebisuzaki, Toshikazu

    2008-01-01

    We examine the model dependence of the phase diagram of inhomogeneous nulcear matter in supernova cores using the quantum molecular dynamics (QMD). Inhomogeneous matter includes crystallized matter with nonspherical nuclei--''pasta'' phases--and the liquid-gas phase-separating nuclear matter. Major differences between the phase diagrams of the QMD models can be explained by the energy of pure neutron matter at low densities and the saturation density of asymmetric nuclear matter. We show the density dependence of the symmetry energy is also useful to understand uncertainties of the phase diagram. We point out that, for typical nuclear models, the mass fraction of the pasta phases in the later stage of the collapsing cores is higher than 10-20%

  18. Core catcher for nuclear reactor core meltdown containment

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Bowman, F.L.

    1978-01-01

    A bed of graphite particles is placed beneath a nuclear reactor core outside the pressure vessel but within the containment building to catch the core debris in the event of failure of the emergency core cooling system. Spray cooling of the debris and graphite particles together with draining and flooding of coolant fluid of the graphite bed is provided to prevent debris slump-through to the bottom of the bed

  19. A Novel Method To On-Line Monitor Reactor Nuclear Power And In-Core Thermal Environments

    International Nuclear Information System (INIS)

    Liu, Hanying; Miller, Don W.; Li, Dongxu; Radcliff, Thomas D.

    2002-01-01

    For current nuclear power plants, nuclear power can not be directly measured and in-core fuel thermal environments can not be monitored due to the unavailability of an appropriate measurement technology and the inaccessibility of the fuel. If the nuclear deposited power and the in-core thermal conditions (i.e. fuel or coolant temperature and heat transfer coefficient) can be monitored in-situ, then it would play a valuable and critical role in increasing nuclear power, predicting abnormal reactor operation, improving core physical models and reducing core thermal margin so as to implement higher fuel burn-up. Furthermore, the management of core thermal margin and fuel operation may be easier during reactor operation, post-accident or spent fuel storage. On the other hand, for some advanced Generation IV reactors, the sealed and long-lived reactor core design challenges traditional measurement techniques while conventional ex-core detectors and current in-core detectors can not monitor details of the in-core fuel conditions. A method is introduced in this paper that responds to the challenge to measure nuclear power and to monitor the in-core thermal environments, for example, local fuel pin or coolant heat convection coefficient and temperature. In summary, the method, which has been designed for online in-core measurement and surveillance, will be beneficial to advanced plant safety, efficiency and economics by decreasing thermal margin or increasing nuclear power. The method was originally developed for a constant temperature power sensor (CTPS). The CTPS is undergoing design and development for an advanced reactor core to measure in-core nuclear power in measurement mode and to monitor thermal environments in compensation mode. The sensor dynamics was analyzed in compensation mode to determine the environmental temperature and the heat transfer coefficient. Previous research demonstrated that a first order dynamic model is not sufficient to simulate sensor

  20. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1979-01-01

    A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  1. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1979-01-01

    A lateral restraint and control systemm for a nuclear reactor core provides an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit is composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased by an amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  2. The online simulation of core physics in nuclear power plant

    International Nuclear Information System (INIS)

    Zhao Qiang

    2005-01-01

    The three-dimensional power distribution in core is one of the most important status variables of nuclear reactor. In order to monitor the 3-D in core power distribution timely and accurately, the online simulation system of core physics was designed in the paper. This system combines core physics simulation with the data, which is from the plant and reactor instrumentation. The design of the system consists of the hardware part and the software part. The online simulation system consists of a main simulation computer and a simulation operation station. The online simulation system software includes of the real-time simulation support software, the system communication software, the simulation program and the simulation interface software. Two-group and three-dimensional neutron kinetics model with six groups delayed neutrons was used in the real-time simulation of nuclear reactor core physics. According to the characteristics of the nuclear reactor, the core was divided into many nodes. Resolving the neutron equation, the method of separate variables was used. The input data from the plant and reactor instrumentation system consist of core thermal power, loop temperatures and pressure, control rod positions, boron concentration, core exit thermocouple data, Excore detector signals, in core flux detectors signals. There are two purposes using the data, one is to ensure that the model is as close as the current actual reactor condition, and the other is to calibrate the calculated power distribution. In this paper, the scheme of the online simulation system was introduced. Under the real-time simulation support system, the simulation program is being compiled. Compared with the actual operational data, the elementary simulation results were reasonable and correct. (author)

  3. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1978-01-01

    Disclosed is a lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  4. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

    Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

    2008-01-01

    This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

  5. Nuclear core catchers

    International Nuclear Information System (INIS)

    Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

    1976-01-01

    A receptacle is described for taking the molten fragments of a nuclear reactor during a reactor core fusion accident. The receptacle is placed under the reactor. It includes at least one receptacle for the reactor core fragments, with a dome shaped part to distribute the molten fragments and at least one outside layer of alumina bricks around the dome. The characteristic of this receptacle is that the outer layer of bricks contains neutron poison rods which pass through the bricks and protrude in relation to them [fr

  6. Calculation models for a nuclear reactor

    International Nuclear Information System (INIS)

    Tashanii, Ahmed Ali

    2010-01-01

    Determination of different parameters of nuclear reactors requires neutron transport calculations. Due to complicity of geometry and material composition of the reactor core, neutron calculations were performed for simplified models of the real arrangement. In frame of the present work two models were used for calculations. First, an elementary cell model was used to prepare cross section data set for a homogenized-core reactor model. The homogenized-core reactor model was then used to perform neutron transport calculation. The nuclear reactor is a tank-shaped thermal reactor. The semi-cylindrical core arrangement consists of aluminum made fuel bundles immersed in water which acts as a moderator as well as a coolant. Each fuel bundle consists of aluminum cladded fuel rods arranged in square lattices. (author)

  7. Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

    Science.gov (United States)

    Darmawan, R.

    2018-01-01

    Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

  8. Hyper-heuristic applied to nuclear reactor core design

    International Nuclear Information System (INIS)

    Domingos, R P; Platt, G M

    2013-01-01

    The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

  9. On-line core monitoring system based on buckling corrected modified one group model

    International Nuclear Information System (INIS)

    Freire, Fernando S.

    2011-01-01

    Nuclear power reactors require core monitoring during plant operation. To provide safe, clean and reliable core continuously evaluate core conditions. Currently, the reactor core monitoring process is carried out by nuclear code systems that together with data from plant instrumentation, such as, thermocouples, ex-core detectors and fixed or moveable In-core detectors, can easily predict and monitor a variety of plant conditions. Typically, the standard nodal methods can be found on the heart of such nuclear monitoring code systems. However, standard nodal methods require large computer running times when compared with standards course-mesh finite difference schemes. Unfortunately, classic finite-difference models require a fine mesh reactor core representation. To override this unlikely model characteristic we can usually use the classic modified one group model to take some account for the main core neutronic behavior. In this model a course-mesh core representation can be easily evaluated with a crude treatment of thermal neutrons leakage. In this work, an improvement made on classic modified one group model based on a buckling thermal correction was used to obtain a fast, accurate and reliable core monitoring system methodology for future applications, providing a powerful tool for core monitoring process. (author)

  10. The Use of Hidden Markov Models for Anomaly Detection in Nuclear Core Condition Monitoring

    Science.gov (United States)

    Stephen, Bruce; West, Graeme M.; Galloway, Stuart; McArthur, Stephen D. J.; McDonald, James R.; Towle, Dave

    2009-04-01

    Unplanned outages can be especially costly for generation companies operating nuclear facilities. Early detection of deviations from expected performance through condition monitoring can allow a more proactive and managed approach to dealing with ageing plant. This paper proposes an anomaly detection framework incorporating the use of the Hidden Markov Model (HMM) to support the analysis of nuclear reactor core condition monitoring data. Fuel Grab Load Trace (FGLT) data gathered within the UK during routine refueling operations has been seen to provide information relating to the condition of the graphite bricks that comprise the core. Although manual analysis of this data is time consuming and requires considerable expertise, this paper demonstrates how techniques such as the HMM can provide analysis support by providing a benchmark model of expected behavior against which future refueling events may be compared. The presence of anomalous behavior in candidate traces is inferred through the underlying statistical foundation of the HMM which gives an observation likelihood averaged along the length of the input sequence. Using this likelihood measure, the engineer can be alerted to anomalous behaviour, indicating data which might require further detailed examination. It is proposed that this data analysis technique is used in conjunction with other intelligent analysis techniques currently employed to analyse FGLT to provide a greater confidence measure in detecting anomalous behaviour from FGLT data.

  11. Nuclear reactor core stabilizing arrangement

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1976-01-01

    A nuclear reactor core stabilizing arrangement is described wherein a plurality of actuators, disposed in a pattern laterally surrounding a group of elongated fuel assemblies, press against respective contiguous fuel assemblies on the periphery of the group to reduce the clearance between adjacent fuel assemblies thereby forming a more compacted, vibration resistant core structure. 7 claims, 4 drawing figures

  12. Impact of nuclear 'pasta' on neutrino transport in collapsing stellar cores

    International Nuclear Information System (INIS)

    Sonoda, Hidetaka; Watanabe, Gentaro; Sato, Katsuhiko; Takiwaki, Tomoya; Yasuoka, Kenji; Ebisuzaki, Toshikazu

    2007-01-01

    Nuclear 'pasta', nonspherical nuclei in dense matter, is predicted to occur in collapsing supernova cores. We show how pasta phases affect the neutrino transport cross section via weak neutral current using several nuclear models. This is the first calculation of the neutrino opacity of the phases with rod-like and slab-like nuclei taking account of finite temperature effects, which are well described by the quantum molecular dynamics. We also show that pasta phases can occupy 10-20% of the mass of supernova cores in the later stage of the collapse

  13. Improvement of Cycle Dependent Core Model for NPP Simulator

    International Nuclear Information System (INIS)

    Song, J. S.; Koo, B. S.; Kim, H. Y. and others

    2003-11-01

    The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations

  14. Improvement of Cycle Dependent Core Model for NPP Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Song, J. S.; Koo, B. S.; Kim, H. Y. and others

    2003-11-15

    The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations.

  15. Methods and techniques of nuclear in-core fuel management

    International Nuclear Information System (INIS)

    Jong, A.J. de.

    1992-04-01

    Review of methods of nuclear in-core fuel management (the minimal critical mass problem, minimal power peaking) and calculational techniques: reactorphysical calculations (point reactivity models, continuous refueling, empirical methods, depletion perturbation theory, nodal computer programs); optimization techniques (stochastic search, linear programming, heuristic parameter optimization). (orig./HP)

  16. Nuclear reactor core servicing apparatus

    International Nuclear Information System (INIS)

    Andrea, C.

    1977-01-01

    Disclosed is an improved core servicing apparatus for a nuclear reactor of the type having a reactor vessel, a vessel head having a head penetration therethrough, a removable plug adapted to fit in the head penetration, and a core of the type having an array of elongated assemblies. The improved core servicing apparatus comprises a plurality of support columns suspended from the removable plug and extending downward toward the nuclear core, rigid support means carried by each of the support columns, and a plurality of servicing means for each of the support columns for servicing a plurality of assemblies. Each of the plurality of servicing means for each of the support columns is fixedly supported in a fixed array from the rigid support means. Means are provided for rotating the rigid support means and servicing means between condensed and expanded positions. When in the condensed position, the rigid support means and servicing means lie completely within the coextensive boundaries of the plug, and when in the expanded position, some of the rigid support means and servicing means lie without the coextensive boundaries of the plug

  17. Thermal radiation in gas core nuclear reactors for space propulsion

    International Nuclear Information System (INIS)

    Slutz, S.A.; Gauntt, R.O.; Harms, G.A.; Latham, T.; Roman, W.; Rodgers, R.J.

    1994-01-01

    A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept. 15 refs

  18. Development of the core-model implementation technology for YGN1 simulator

    International Nuclear Information System (INIS)

    Hong, J. H.; Lee, M. S.; Lee, Y. K.; Su, I. Y.

    2004-01-01

    The existing core models for the domestic nuclear power plant simulators for PWRs are entirely imported from the foreign simulator vendor. To solve the time-accuracy problem in the poor capabilities in the computer in the early 1990s, several simplifications and assumptions for the neutronics governing equations were indispensible for the realtime calculations of nuclear phenomena in the core region. To overcome the shortages, a new core model based on the MASTER code certified by the domestic regulatory body (KINS) instead of the existing core models is now being developed especially for the realtime core solver for the YGN-1 simulator. This code is named R-MASTER (Realtime MASTER code). Due to the deficiency of the host computer, it is quitely required to run the R-MASTER code on the separate computer with high performance from the host computer on which all the other models than the core model are running. This paper deals with the applied protocols and procedures to guarantee the realtime communication and calculation of the R-MASTER code

  19. Development of Pipeline Database and CAD Model for Selection of Core Security Zone in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Choi, Seong Soo; Kwon, Tae Gyun; Baek, Hun Hyun; Kwon, Min Jin

    2008-07-01

    The objective of the project is to develop the pipeline database which can be used for selection of core security zones considering safety significance of pipes and to develop CAD model for 3-dimensional visualization of core security zones, for the purpose of minimizing damage and loss, enforcing security and protection on important facilities, and improving plant design preparing against emergency situations such as physical terrors in nuclear power plants. In this study, the pipeline database is developed for selection of core security zones considering safety significance of safety class 1 and 2 pipes. The database includes the information on 'pipe-room information-surrogate component' mapping, initiating events which may occur and accident mitigation functions which may be damaged by the pipe failure, and the drawing information related to 2,270 pipe segments of 30 systems. For the 3-dimensional visualization of core security zones, the CAD models on the containment building and the auxiliary building are developed using 3-D MAX tool and the demo program which can visualize the direct-X model converted from the 3-D MAX model is also developed. In addition to this, the coordinate information of all the buildings and their rooms is generated using AUTO CAD tool in order to be used as an input for 3-dimensional browsing of the VIP program

  20. Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

    Directory of Open Access Journals (Sweden)

    Andrea Cerutti

    Full Text Available Hepatitis C virus (HCV infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS, but no nuclear export signal (NES has yet been identified.We show here that the aa(109-133 region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126 in the identified NES or in the sequence encoding the mature core aa(1-173 significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

  1. Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

    Science.gov (United States)

    Cerutti, Andrea; Maillard, Patrick; Minisini, Rosalba; Vidalain, Pierre-Olivier; Roohvand, Farzin; Pecheur, Eve-Isabelle; Pirisi, Mario; Budkowska, Agata

    2011-01-01

    Hepatitis C virus (HCV) infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS), but no nuclear export signal (NES) has yet been identified.We show here that the aa(109-133) region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126) in the identified NES or in the sequence encoding the mature core aa(1-173) significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

  2. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  3. Nuclear core baffling apparatus

    International Nuclear Information System (INIS)

    Cooper, F.W. Jr.; Silverblatt, B.L.; Knight, C.B.; Berringer, R.T.

    1979-01-01

    An apparatus for baffling the flow of reactor coolant fluid into and about the core of a nuclear reactor is described. The apparatus includes a plurality of longitudinally aligned baffle plates with mating surfaces that allow longitudinal growth with temperature increases while alleviating both leakage through the aligned plates and stresses on the components supporting the plates

  4. Method for refuelling a nuclear reactor core

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This invention relates to an improved method for refuelling a nuclear reactor core inside a reactor vessel. The technique allows a substantial reduction in the refuelling time as compared with previously known methods and permits fewer out of core operations and smaller temporary storage space. (U.K.)

  5. SCDAP/RELAP5 lower core plate model

    International Nuclear Information System (INIS)

    Coryell, E.W.; Griffin, F.P.

    1999-01-01

    The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. This report describes the justification, theory, implementation, and testing of a new modeling capability which will refine the analysis of the movement of molten material from the core region to the vessel lower head. As molten material moves from the core region through the core support structures it may encounter conditions which will cause it to freeze in the region of the lower core plate, delaying its arrival to the vessel head. The timing of this arrival is significant to reactor safety, because during the time span for material relocation to the lower head, the core may be experiencing steam-limited oxidation. The time at which hot material arrives in a coolant-filled lower vessel head, thereby significantly increasing the steam flow rate through the core region, becomes significant to the progression and timing of a severe accident. This report is a revision of a report INEEL/EXT-00707, entitled ''Preliminary Design Report for SCDAP/RELAP5 Lower Core Plate Model''

  6. A system for obtaining an optimized pre design of nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.

    1989-01-01

    This work proposes a method for obtaing a first design of nuclear reactor cores. It takes into consideration the objectives of the project, physical limits, economical limits and the reactor safety. For this purpose, some simplifications were made in the reactor model: one-energy-group, unidimensional and homogeneous core. The adopted model represents a typical PWR core and the optimized parameters are the fuel thickness, refletor thickness, enrichement and moderating ratio. The objective is to gain a larger residual reactivity at the end of the cycle. This work also presents results for a PWR core. From the results, many conclusions are established: system efficiency, limitations and problems. Also some suggestions are proposed to improve the system performance for futures works. (author) [pt

  7. Automated software analysis of nuclear core discharge data

    International Nuclear Information System (INIS)

    Larson, T.W.; Halbig, J.K.; Howell, J.A.; Eccleston, G.W.; Klosterbuer, S.F.

    1993-03-01

    Monitoring the fueling process of an on-load nuclear reactor is a full-time job for nuclear safeguarding agencies. Nuclear core discharge monitors (CDMS) can provide continuous, unattended recording of the reactor's fueling activity for later, qualitative review by a safeguards inspector. A quantitative analysis of this collected data could prove to be a great asset to inspectors because more information can be extracted from the data and the analysis time can be reduced considerably. This paper presents a prototype for an automated software analysis system capable of identifying when fuel bundle pushes occurred and monitoring the power level of the reactor. Neural network models were developed for calculating the region on the reactor face from which the fuel was discharged and predicting the burnup. These models were created and tested using actual data collected from a CDM system at an on-load reactor facility. Collectively, these automated quantitative analysis programs could help safeguarding agencies to gain a better perspective on the complete picture of the fueling activity of an on-load nuclear reactor. This type of system can provide a cost-effective solution for automated monitoring of on-load reactors significantly reducing time and effort

  8. Nucleoporins as components of the nuclear pore complex core structure and Tpr as the architectural element of the nuclear basket.

    Science.gov (United States)

    Krull, Sandra; Thyberg, Johan; Björkroth, Birgitta; Rackwitz, Hans-Richard; Cordes, Volker C

    2004-09-01

    The vertebrate nuclear pore complex (NPC) is a macromolecular assembly of protein subcomplexes forming a structure of eightfold radial symmetry. The NPC core consists of globular subunits sandwiched between two coaxial ring-like structures of which the ring facing the nuclear interior is capped by a fibrous structure called the nuclear basket. By postembedding immunoelectron microscopy, we have mapped the positions of several human NPC proteins relative to the NPC core and its associated basket, including Nup93, Nup96, Nup98, Nup107, Nup153, Nup205, and the coiled coil-dominated 267-kDa protein Tpr. To further assess their contributions to NPC and basket architecture, the genes encoding Nup93, Nup96, Nup107, and Nup205 were posttranscriptionally silenced by RNA interference (RNAi) in HeLa cells, complementing recent RNAi experiments on Nup153 and Tpr. We show that Nup96 and Nup107 are core elements of the NPC proper that are essential for NPC assembly and docking of Nup153 and Tpr to the NPC. Nup93 and Nup205 are other NPC core elements that are important for long-term maintenance of NPCs but initially dispensable for the anchoring of Nup153 and Tpr. Immunogold-labeling for Nup98 also results in preferential labeling of NPC core regions, whereas Nup153 is shown to bind via its amino-terminal domain to the nuclear coaxial ring linking the NPC core structures and Tpr. The position of Tpr in turn is shown to coincide with that of the nuclear basket, with different Tpr protein domains corresponding to distinct basket segments. We propose a model in which Tpr constitutes the central architectural element that forms the scaffold of the nuclear basket.

  9. Towards an efficient multiphysics model for nuclear reactor dynamics

    Directory of Open Access Journals (Sweden)

    Obaidurrahman K.

    2015-01-01

    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.

  10. Station blackout core damage frequency in an advanced nuclear reactor

    International Nuclear Information System (INIS)

    Carvalho, Luiz Sergio de

    2004-01-01

    Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

  11. A system to obtain an optimized first design of a nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.

    1988-01-01

    This work proposes a method for obtaining a first design of nuclear reactor cores. It takes into consideration the objectives of the project, physical limits, economical limits and the reactor safety. For this purpose, some simplifications were made in the reactor model: one energy-group, one-dimensional and homogeneous core. The adopted model represents a typical PWR core and the optimized parameters are the fuel thickness, reflector thickness, enrichment and moderating ratio. The objective is to gain a larger residual reactivity at the end of the cycle. This work also presents results for a PWR core. From the results, many conclusions are established: system efficiency, limitations and problems. Also some suggestions are proposed to improve the system performance for future works. (autor)

  12. Modelling of a rod bundle under viscous and uncompressible flow by porous media. Applied to nuclear reactor core

    International Nuclear Information System (INIS)

    Ricciardi, Guillaume; Collard, Bruno; Bellizzi, Sergio; Cochelin, Bruno

    2007-01-01

    This study is about the safety of nuclear reactor core submitted to seismic loading. In order to reduce the incertitude margin of the present day codes we propose to develop a numerical code including the non linear behavior of the fluid/structure coupling. The challenge of this work is to find out a tractable model taking the structure complexity into account. In this paper we model the nuclear reactor core mechanical behavior including the dynamics of both fuel assemblies of fluid. Each rod bundle is considered as a deformable porous media, so the velocity field of the fluid and the displacement field of the structure are defined in the whole domain space. Fluid part and structure part are in a first time considered separately, and in second time, the two parts are coupled. The motion equations of the structure are obtained by a Lagrangian formulation, and to allow the fluid structure coupling, the motion equations of the fluid are obtained by an Arbitrary Lagrangian Eulerian formulation. The finite elements method is applied to spatially discretize the equations. Simulations have been performed to analyze the influence of the fluid and structure characteristics, phenomena observed by the experience have been reproduced qualitatively. (author)

  13. In-medium no-core shell model for ab initio nuclear structure calculations

    International Nuclear Information System (INIS)

    Gebrerufael, Eskendr

    2017-01-01

    In this work, we merge two successful ab initio nuclear-structure methods, the no-core shell model (NCSM) and the multi-reference in-medium similarity renormalization group (IM-SRG), to define a novel many-body approach for the comprehensive description of ground and excited states of closed- and open-shell medium-mass nuclei. Building on the key advantages of the two methods - the decoupling of excitations at the many-body level in the IM-SRG, and the exact diagonalization in the NCSM applicable up to medium-light nuclei - their combination enables fully converged no-core calculations for an unprecedented range of nuclei and observables at moderate computational cost. The efficiency and rapid model-space convergence of the new approach make it ideally suited for ab initio studies of ground and low-lying excited states of nuclei up to the medium-mass regime. Interactions constructed within the framework of chiral effective field theory provide an excellent opportunity to describe properties of nuclei from first principles, i.e., rooted in quantum chromodynamics, they overcome the lack of predictive power of phenomenological potentials. The hard core of these interactions causes strong short-range correlations, which we soften by using the similarity-renormalization-group transformation that accelerates the model-space convergence of many-body calculations. Three-nucleon effects, which are mandatory for the correct description of bulk properties of nuclei, are included in our calculations by using the normal-ordered two-body approximation, which has been shown to be sufficient to capture the main effects of the three-nucleon interaction. Using these interactions, we analyze energies of ground and excited states in the carbon and oxygen isotopic chains, where conventional NCSM calculations are still feasible and provide an important benchmark. Furthermore, we study the Hoyle state in 12 C - a three-alpha cluster state that cannot be converged in standard NCSM

  14. Improving the calculated core stability by the core nuclear design optimization

    International Nuclear Information System (INIS)

    Partanen, P.

    1995-01-01

    Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)

  15. Fast three-dimensional core optimization based on modified one-group model

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Fernando S. [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Rio de Janeiro, RJ (Brazil). Dept. GCN-T], e-mail: freire@eletronuclear.gov.br; Martinez, Aquilino S.; Silva, Fernando C. da [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear], e-mail: aquilino@con.ufrj.br, e-mail: fernando@con.ufrj.br

    2009-07-01

    The optimization of any nuclear reactor core is an extremely complex process that consumes a large amount of computer time. Fortunately, the nuclear designer can rely on a variety of methodologies able to approximate the analysis of each available core loading pattern. Two-dimensional codes are usually used to analyze the loading scheme. However, when particular axial effects are present in the core, two-dimensional analysis cannot produce good results and three-dimensional analysis can be required at all time. Basically, in this paper are presented the major advantages that can be found when one use the modified one-group diffusion theory coupled with a buckling correction model in optimization process. The results of the proposed model are very accurate when compared to benchmark results obtained from detailed calculations using three-dimensional nodal codes (author)

  16. Fast three-dimensional core optimization based on modified one-group model

    International Nuclear Information System (INIS)

    Freire, Fernando S.; Martinez, Aquilino S.; Silva, Fernando C. da

    2009-01-01

    The optimization of any nuclear reactor core is an extremely complex process that consumes a large amount of computer time. Fortunately, the nuclear designer can rely on a variety of methodologies able to approximate the analysis of each available core loading pattern. Two-dimensional codes are usually used to analyze the loading scheme. However, when particular axial effects are present in the core, two-dimensional analysis cannot produce good results and three-dimensional analysis can be required at all time. Basically, in this paper are presented the major advantages that can be found when one use the modified one-group diffusion theory coupled with a buckling correction model in optimization process. The results of the proposed model are very accurate when compared to benchmark results obtained from detailed calculations using three-dimensional nodal codes (author)

  17. Thermal barrier and support for nuclear reactor fuel core

    International Nuclear Information System (INIS)

    Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

    1987-01-01

    A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

  18. A genetic algorithm solution for combinatorial problems - the nuclear core reload example

    Energy Technology Data Exchange (ETDEWEB)

    Schirru, R.; Silva, F.C. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia; Pereira, C.M.N.A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Chapot, J.L.C. [FURNAS, Rio de Janeiro, RJ (Brazil)

    1997-12-01

    This paper presents a solution to Traveling Salesman Problem based upon genetic algorithms (GA), using the classic crossover, but avoiding the feasibility problem in offspring individuals, allowing the natural evolution of the GA without introduction of heuristics in the genetic crossover operator. The genetic model presented, that we call the List Model (LM) is based on the encoding and decoding genotype in the way to always generate a phenotype that has a valid structure, over which will be applied the fitness, represented by the total distance. The main purpose of this work was to develop the basis for a new genetic model to be used in the reload of nuclear core of a PWR. In a generic way, this problem can be interpreted as a a search of the optimal combination of N different fuel elements in N nuclear core `holes`, where each combination or load pattern, determines the neutron flux shape and its associate peak factor. The goal is to find out the load pattern that minimizes the peak factor and consequently maximize the useful life of the nuclear fuel. The GA with the List Model was applied to the Angra-1 PWR reload problem and the results are remarkably better than the ones used in the last fuel cycle. (author). 12 refs., 3 figs., 2 tabs.

  19. Real-time simulation of ex-core nuclear instrumentation system

    International Nuclear Information System (INIS)

    Zhao Qiang; Zhang Zhijian; Cao Xinrong

    2005-01-01

    Real-time simulation of ex-core nuclear instrumentation system is an indispensable part of nuclear power plant (NPP) full-scope training simulator. The simulation method, which is based upon the theory of measurement, is introduced in the paper. The fitting formula between the measured data and the three-dimensional neutron flux distribution in the core is established. The fitting parameter is adjusted according to the reactor physical calculation or the experiment of power calibration. The simulation result shows that the method can simulate the ex-core neutron instrumentation system accurately in real-time and meets the needs of NPP full-scope training simulator. (authors)

  20. An assessment of coupling algorithms for nuclear reactor core physics simulations

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, Steven, E-mail: hamiltonsp@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Berrill, Mark, E-mail: berrillma@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Clarno, Kevin, E-mail: clarnokt@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Pawlowski, Roger, E-mail: rppawlo@sandia.gov [Sandia National Laboratories, MS 0316, P.O. Box 5800, Albuquerque, NM 87185 (United States); Toth, Alex, E-mail: artoth@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Kelley, C.T., E-mail: tim_kelley@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Evans, Thomas, E-mail: evanstm@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States)

    2016-04-15

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  1. Fast breeder physics and nuclear core design

    International Nuclear Information System (INIS)

    Marth, W.; Schroeder, R.

    1983-07-01

    This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.) [de

  2. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases

  3. Nuclear waste disposal utilizing a gaseous core reactor

    Science.gov (United States)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  4. Apparatus for controlling nuclear core debris

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    Nuclear reactor apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling.

  5. Thermal margin model for transition core of KSNP

    International Nuclear Information System (INIS)

    Nahm, Kee Yil; Lim, Jong Seon; Park, Sung Kew; Chun, Chong Kuk; Hwang, Sun Tack

    2004-01-01

    The PLUS7 fuel was developed with mixing vane grids for KSNP. For the transition core partly loaded with the PLUS7 fuels, the procedure to set up the optimum thermal margin model of the transition core was suggested by introducing AOPM concept into the screening method which determines the limiting assembly. According to the procedure, the optimum thermal margin model of the first transition core was set up by using a part of nuclear data for the first transition and the homogeneous core with PLUS7 fuels. The generic thermal margin model of PLUS7 fuel was generated with the AOPM of 138%. The overpower penalties on the first transition core were calculated to be 1.0 and 0.98 on the limiting assembly and the generic thermal margin model, respectively. It is not usual case to impose the overpower penalty on reload cores. It is considered that the lack of channel flow due to the difference of pressure drop between PLUS7 and STD fuels results in the decrease of DNBR. The AOPM of the first transition core is evaluated to be about 135% by using the optimum generic thermal margin model which involves the generic thermal margin model and the total overpower penalty. The STD fuel is not included among limiting assembly candidates in the second transition core, because they have much lower pin power than PLUS7 fuels. The reduced number of STD fuels near the limiting assembly candidates the flow from the limiting assembly to increase the thermal margin for the second transition core. It is expected that cycle specific overpower penalties increase the thermal margin for the transition core. Using the procedure to set up the optimum thermal margin model makes sure that the enhanced thermal margin of PLUS7 fuel can be sufficiently applied to not only the homogeneous core but also the transition core

  6. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  7. Sensors for use in nuclear reactor cores

    International Nuclear Information System (INIS)

    Brown, W.L.; Geronime, R.L.

    1978-01-01

    Sensors including radiation detectors and the like for use within the core of nuclear reactors and which are constructed in a manner to provide optimum reliability of the sensor during use are described

  8. Analysis of LWR Full MOX Core Physics Experiments with Major Nuclear Data Libraries

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toru [Japan Nuclear Energy Safety Organization, Tokyo (Japan)

    2007-07-01

    Nuclear Power Engineering Corporation (NUPEC) studied high moderation full MOX cores as a part of advanced LWR core concept studies from 1994 to 2003 supported by the Ministry of Economy, Trade and Industry. In order to obtain the major physics characteristics of such advanced MOX cores, NUPEC carried out core physics experimental programs called MISTRAL and BASALA from 1996 to 2002 in the EOLE critical facility of the Cadarache Center in collaboration with CEA. NUPEC also obtained a part of experimental data of the EPICURE program that CEA had conducted for 30 % Pu recycling in French PWRs. Japan Nuclear Energy Safety Organization(JNES) established in 2003 as an incorporated administrative agency took over the NUPEC's projects for nuclear regulation and has been implementing FUBILA program that is for high burn up BWR full MOX cores. This paper presents an outline of the programs and a summary of the analysis results of the criticality of those experimental cores with major nuclear data libraries.

  9. Spring unit especially intended for a nuclear reactor core

    International Nuclear Information System (INIS)

    Brown, S.J.; Gorholt, Wilhelm.

    1977-01-01

    This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

  10. No-Core Shell Model and Reactions

    International Nuclear Information System (INIS)

    Navratil, P; Ormand, W E; Caurier, E; Bertulani, C

    2005-01-01

    There has been a significant progress in ab initio approaches to the structure of light nuclei. Starting from realistic two- and three-nucleon interactions the ab initio no-core shell model (NCSM) can predict low-lying levels in p-shell nuclei. It is a challenging task to extend ab initio methods to describe nuclear reactions. In this contribution, we present a brief overview of the NCSM with examples of recent applications as well as the first steps taken toward nuclear reaction applications. In particular, we discuss cross section calculations of p+ 6 Li and 6 He+p scattering as well as a calculation of the astrophysically important 7 Be(p, γ) 8 B S-factor

  11. Interfacing high-fidelity core neutronics models to whole plant models

    International Nuclear Information System (INIS)

    McEllin, M.

    1999-01-01

    Until recently available computer power dictated that whole-plant models of nuclear power stations have typically employed simple models of the reactor core which can not match the fidelity of safety-qualified 2-group, 3D neutronics models. As a result the treatment of situations involving strong coupling between the core and the rest of the plant has inevitably been somewhat approximate, requiring conservative modelling assumptions, or manual iteration between cases, to bound worse case scenarios. Such techniques not only place heavy demands on the engineers involved, they may also result in potentially unnecessary operational constraints. Hardware is today no longer the limiting factor, but the cost of developing and validating high-quality software is now such that it appears attractive to build new systems with a wider simulation scope by using existing stand-alone codes as sub-components. This is not always as straightforward as it might at first appear. This paper illustrates some of the pitfalls, and discusses more sophisticated and robust strategies. (author)

  12. A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, E.A. [Department of Mechanical Engineering, University of Texas, Austin, TX (United States); Deinert, M.R. [Theoretical and Applied Mechanics, Cornell University, 219 Kimball Hall, Ithaca, NY 14853 (United States)]. E-mail: mrd6@cornell.edu; Cady, K.B. [Theoretical and Applied Mechanics, Cornell University, 219 Kimball Hall, Ithaca, NY 14853 (United States)

    2006-10-15

    The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

  13. Global physical and numerical stability of a nuclear reactor core

    International Nuclear Information System (INIS)

    Morales-Sandoval, Jaime; Hernandez-Solis, Augusto

    2005-01-01

    Low order models are used to investigate the influence of integration methods on observed power oscillations of some nuclear reactor simulators. The zero-power point reactor kinetics with six-delayed neutron precursor groups are time discretized using explicit, implicit and Crank-Nicholson methods, and the stability limit of the time mesh spacing is exactly obtained by locating their characteristic poles in the z-transform plane. These poles are the s to z mappings of the inhour equation roots and, except for one of them, they show little or no dependence on the integration method. Conditions for stable power oscillations can be also obtained by tracking when steady state output signals resulting from reactivity oscillations in the s-Laplace plane cross the imaginary axis. The dynamics of a BWR core operating at power conditions is represented by a reduced order model obtained by adding three ordinary differential equations, which can model void and Doppler reactivity feedback effects on power, and collapsing all delayed neutron precursors in one group. Void dynamics are modeled as a second order system and fuel heat transfer as a first order system. This model shows rich characteristics in terms of indicating the relative importance of different core parameters and conditions on both numerical and physical oscillations observed by large computer code simulations. A brief discussion of the influence of actual core and coolant conditions on the reduced order model is presented

  14. A new equation of state for core-collapse supernovae based on realistic nuclear forces and including a full nuclear ensemble

    International Nuclear Information System (INIS)

    Furusawa, S; Togashi, H; Nagakura, H; Sumiyoshi, K; Yamada, S; Suzuki, H; Takano, M

    2017-01-01

    We have constructed a nuclear equation of state (EOS) that includes a full nuclear ensemble for use in core-collapse supernova simulations. It is based on the EOS for uniform nuclear matter that two of the authors derived recently, applying a variational method to realistic two- and three-body nuclear forces. We have extended the liquid drop model of heavy nuclei, utilizing the mass formula that accounts for the dependences of bulk, surface, Coulomb and shell energies on density and/or temperature. As for light nuclei, we employ a quantum-theoretical mass evaluation, which incorporates the Pauli- and self-energy shifts. In addition to realistic nuclear forces, the inclusion of in-medium effects on the full ensemble of nuclei makes the new EOS one of the most realistic EOSs, which covers a wide range of density, temperature and proton fraction that supernova simulations normally encounter. We make comparisons with the FYSS EOS, which is based on the same formulation for the nuclear ensemble but adopts the relativistic mean field theory with the TM1 parameter set for uniform nuclear matter. The new EOS is softer than the FYSS EOS around and above nuclear saturation densities. We find that neutron-rich nuclei with small mass numbers are more abundant in the new EOS than in the FYSS EOS because of the larger saturation densities and smaller symmetry energy of nuclei in the former. We apply the two EOSs to 1D supernova simulations and find that the new EOS gives lower electron fractions and higher temperatures in the collapse phase owing to the smaller symmetry energy. As a result, the inner core has smaller masses for the new EOS. It is more compact, on the other hand, due to the softness of the new EOS and bounces at higher densities. It turns out that the shock wave generated by core bounce is a bit stronger initially in the simulation with the new EOS. The ensuing outward propagations of the shock wave in the outer core are very similar in the two simulations, which

  15. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt

    International Nuclear Information System (INIS)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects

  16. An intelligent nuclear reactor core controller for load following operations, using recurrent neural networks and fuzzy systems

    International Nuclear Information System (INIS)

    Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.

    2003-01-01

    In the last decade, the intelligent control community has paid great attention to the topic of intelligent control systems for nuclear plants (core, steam generator...). Papers mostly used approximate and simple mathematical SISO (single-input-single-output) model of nuclear plants for testing and/or tuning of the control systems. They also tried to generalize theses models to a real MIMO (multi-input-multi-output) plant, while nuclear plants are typically of complex nonlinear and multivariable nature with high interactions between their state variables and therefore, many of these proposed intelligent control systems are not appropriate for real cases. In this paper, we designed an on-line intelligent core controller for load following operations, based on a heuristic control algorithm, using a valid and updatable recurrent neural network (RNN). We have used an accurate 3-dimensional core calculation code to represent the real plant and to train the RNN. The results of simulation show that this intelligent controller can control the reactor core during load following operations, using optimum control rod groups manoeuvre and variable overlapping strategy. This methodology represents a simple and reliable procedure for controlling other complex nonlinear MIMO plants, and may improve the responses, comparing to other control systems

  17. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  18. In-core fuel management for nuclear reactor

    International Nuclear Information System (INIS)

    Ross, M.F.; Visner, S.

    1986-01-01

    This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

  19. Nuclear reactor core safety device

    International Nuclear Information System (INIS)

    Colgate, S.A.

    1977-01-01

    The danger of a steam explosion from a nuclear reactor core melt-down can be greatly reduced by adding a gasifying agent to the fuel that releases a large amount of gas at a predetermined pre-melt-down temperature that ruptures the bottom end of the fuel rod and blows the finely divided fuel into a residual coolant bath at the bottom of the reactor. This residual bath should be equipped with a secondary cooling loop

  20. Gas core nuclear rocket feasibility project

    International Nuclear Information System (INIS)

    Howe, S.D.; DeVolder, B.; Thode, L.; Zerkle, D.

    1997-09-01

    The next giant leap for mankind will be the human exploration of Mars. Almost certainly within the next thirty years, a human crew will brave the isolation, the radiation, and the lack of gravity to walk on and explore the Red planet. However, because the mission distances and duration will be hundreds of times greater than the lunar missions, a human crew will face much greater obstacles and a higher risk than those experienced during the Apollo program. A single solution to many of these obstacles is to dramatically decrease the mission duration by developing a high performance propulsion system. The gas core nuclear rocket (GCNR) has the potential to be such a system. The gas core concept relies on the use of fluid dynamic forces to create and maintain a vortex. The vortex is composed of a fissile material which will achieve criticality and produce high power levels. By radiatively coupling to the surrounding fluids, extremely high temperatures in the propellant and, thus, high specific impulses can be generated. The ship velocities enabled by such performance may allow a 9 month round trip, manned Mars mission to be considered. Alternatively, one might consider slightly longer missions in ships that are heavily shielded against the intense Galactic Cosmic Ray flux to further reduce the radiation dose to the crew. The current status of the research program at the Los Alamos National Laboratory into the gas core nuclear rocket feasibility will be discussed

  1. Supporting system for the core restraint of nuclear reactors

    International Nuclear Information System (INIS)

    Kaser, A.

    1973-01-01

    The core restraint of water cooled nuclear reactors which is needed to direct the flow of the coolant through the core can be manufactured only in a moderate wall thickness. Thus, the majority of the loads have to be transmitted to the core barrel which is more rigid. The patent refers to a system of circumferential and vertical support members most of which are free to move relatively to each other, thus reducing thermal stresses during operation. (P.K.)

  2. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  3. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  4. Neutronic analysis of the ford nuclear reactor leu core

    International Nuclear Information System (INIS)

    Raza, S.S.; Hayat, T.

    1989-08-01

    Neutronic analysis of the ford nuclear reactor low enriched uranium core has been carried out to gain confidence in the com puting methodology being used for Pakistan Research Reactor-1 core conversion calculations. The computed value of the effective multiplication factor (Keff) is found to be in good agreement with that quoted by others. (author). 6 figs

  5. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

    2011-01-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  6. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  7. Nuclear-fuel-cycle education: Module 5. In-core fuel management

    International Nuclear Information System (INIS)

    Levine, S.H.

    1980-07-01

    The purpose of this project was to develop a series of educational modules for use in nuclear-fuel-cycle education. These modules are designed for use in a traditional classroom setting by lectures or in a self-paced, personalized system of instruction. This module on in-core fuel management contains information on computational methods and theory; in-core fuel management using the Virginia Polytechnic Institute and State University computer modules; pressurized water reactor in-core fuel management; boiling water reactor in-core fuel management; and in-core fuel management for gas-cooled and fast reactors

  8. Status of core nuclear design technology for future fuel

    International Nuclear Information System (INIS)

    Joo, Hyung Kook; Jung, Hyung Guk; Noh, Jae Man; Kim, Yeong Il; Kim, Taek Kyum; Gil, Choong Sup; Kim, Jung Do; Kim, Young Jin; Sohn, Dong Seong

    1997-01-01

    The effective utilization of nuclear resource is more important factor to be considered in the design of next generation PWR in addition to the epochal consideration on economics and safety. Assuming that MOX fuel can be considered as one of the future fuel corresponding to the above request, the establishment of basic technology for the MOX core design has been performed : : the specification of the technical problem through the preliminary core design and nuclear characteristic analysis of MOX, the development and verification of the neutron library for lattice code, and the acquisition of data to be used for verification of lattice and core analysis codes. The following further studies will be done in future: detailed verification of library E63LIB/A, development of the spectral history effect treatment module, extension of decay chain, development of new homogenization for the MOX fuel assembly. (author). 6 refs., 7 tabs., 2 figs

  9. Recent Developments in No-Core Shell-Model Calculations

    International Nuclear Information System (INIS)

    Navratil, P.; Quaglioni, S.; Stetcu, I.; Barrett, B.R.

    2009-01-01

    We present an overview of recent results and developments of the no-core shell model (NCSM), an ab initio approach to the nuclear many-body problem for light nuclei. In this aproach, we start from realistic two-nucleon or two- plus three-nucleon interactions. Many-body calculations are performed using a finite harmonic-oscillator (HO) basis. To facilitate convergence for realistic inter-nucleon interactions that generate strong short-range correlations, we derive effective interactions by unitary transformations that are tailored to the HO basis truncation. For soft realistic interactions this might not be necessary. If that is the case, the NCSM calculations are variational. In either case, the ab initio NCSM preserves translational invariance of the nuclear many-body problem. In this review, we, in particular, highlight results obtained with the chiral two- plus three-nucleon interactions. We discuss efforts to extend the applicability of the NCSM to heavier nuclei and larger model spaces using importance-truncation schemes and/or use of effective interactions with a core. We outline an extension of the ab initio NCSM to the description of nuclear reactions by the resonating group method technique. A future direction of the approach, the ab initio NCSM with continuum, which will provide a complete description of nuclei as open systems with coupling of bound and continuum states is given in the concluding part of the review.

  10. Recent Developments in No-Core Shell-Model Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Navratil, P; Quaglioni, S; Stetcu, I; Barrett, B R

    2009-03-20

    We present an overview of recent results and developments of the no-core shell model (NCSM), an ab initio approach to the nuclear many-body problem for light nuclei. In this aproach, we start from realistic two-nucleon or two- plus three-nucleon interactions. Many-body calculations are performed using a finite harmonic-oscillator (HO) basis. To facilitate convergence for realistic inter-nucleon interactions that generate strong short-range correlations, we derive effective interactions by unitary transformations that are tailored to the HO basis truncation. For soft realistic interactions this might not be necessary. If that is the case, the NCSM calculations are variational. In either case, the ab initio NCSM preserves translational invariance of the nuclear many-body problem. In this review, we, in particular, highlight results obtained with the chiral two- plus three-nucleon interactions. We discuss efforts to extend the applicability of the NCSM to heavier nuclei and larger model spaces using importance-truncation schemes and/or use of effective interactions with a core. We outline an extension of the ab initio NCSM to the description of nuclear reactions by the resonating group method technique. A future direction of the approach, the ab initio NCSM with continuum, which will provide a complete description of nuclei as open systems with coupling of bound and continuum states is given in the concluding part of the review.

  11. A flow test for calibrating 177 core tubes of 1/5-scale reactor flow model for Yonggwang nuclear units 3 and 4

    International Nuclear Information System (INIS)

    Lee, Byung Jin; Jang, Ho Cheol; Cheong, Jong Sik; Kuh, Jung Eui

    1990-01-01

    A flow test was performed to find out the hydraulic characteristics of every one of 177 core tubes, representing a fuel assembly respectively, as a preparatory step of 1/5 scale reactor flow model test for Yonggwang Nuclear Units (hereafter YGN) 3 and 4. The axial hydraulic resistance of the fuel assembly was simulated in the square core tube with six orifice plates positioned along the tube length; core support structure below each fuel assembly was done in the core upstream geometry section of the test loop. For each core tube the pressure differentials across the inlet, exit orifice plate and overall tube length were measured, along with the flow rates and temperatures of the test fluid. The measured pressure drops were converted to pressure loss or flow metering coefficients. The metering coefficient of the inlet orifice plate was sensitive to the configuration and location of the upstream geometry. The hydraulic resistance of the core tubes were reasonably coincided with a target value and consistent. The polynomial curve fits of the calibrated coefficients for the 177 core tubes were obtained with reasonable data scatters

  12. In core instrumentation for online nuclear heating measurements of material testing reactor

    International Nuclear Information System (INIS)

    Reynard, C.; Andre, J.; Brun, J.; Carette, M.; Janulyte, A.; Merroun, O.; Zerega, Y.; Lyoussi, A.; Bignan, G.; Chauvin, J-P.; Fourmentel, D.; Glayse, W.; Gonnier, C.; Guimbal, P.; Iracane, D.; Villard, J.-F.

    2010-01-01

    The present work focuses on nuclear heating. This work belongs to a new advanced research program called IN-CORE which means 'Instrumentation for Nuclear radiations and Calorimetry Online in REactor' between the LCP (University of Provence-CNRS) and the CEA (French Atomic Energy Commission) - Jules Horowitz Reactor (JHR) program. This program started in September 2009 and is dedicated to the conception and the design of an innovative mobile experimental device coupling several sensors and ray detectors for on line measurements of relevant physical parameters (photonic heating, neutronic flux ...) and for an accurate parametric mapping of experimental channels in the JHR Core. The work presented below is the first step of this program and concerns a brief state of the art related to measurement methods of nuclear heating phenomena in research reactor in general and MTR in particular. A special care is given to gamma heating measurements. A first part deals with numerical codes and models. The second one presents instrumentation divided into various kinds of sensor such as calorimeter measurements and gamma ionization chamber measurements. Their basic principles, characteristics such as metrological parameters, operating mode, disadvantages/advantages, ... are discussed. (author)

  13. Nuclear characteristics evaluation for Kyoto University Research Reactor with low-enriched uranium core

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Ken; Unesaki, Hironobu [Kyoto University Research Reactor Institute, Kumatori-cho Sennan-gun Osaka (Japan)

    2008-07-01

    A project to convert the fuel of Kyoto University Research Reactor (KUR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) is in progress as a part of RERTR program. Prior to the operation of LEU core, the nuclear characteristics of the core have been evaluated to confirm the safety operation. In the evaluation, nuclear parameters, such as the excess reactivity, shut down margin control rod worth, reactivity coefficients, were calculated, and they were compared with the safety limits. The results of evaluation show that the LEU core is able to satisfy the safety requirements for operation, i.e. all the parameters satisfy the safety limits. Consequently, it was confirmed that the LEU fuel core has the proper nuclear characteristics for the safety operation. (authors)

  14. Roles of nuclear weak rates on the evolution of degenerate cores in stars

    Directory of Open Access Journals (Sweden)

    Suzuki Toshio

    2017-01-01

    Full Text Available Electron-capture and β-decay rates in stellar environments are evaluated with the use of new shell-model Hamiltonians for sd-shell and pf-shell nuclei as well as for nuclei belonging to the island of inversion. Important role of the nuclear weak rates on the final evolution of stellar degenerate cores is presented. The weak interaction rates for sd-shell nuclei are calculated to study nuclear Urca processes in O-Ne-Mg cores of stars with 8-10 M⊙ (solar mass and their effects on the final fate of the stars. Nucleosynthesis of iron-group elements in Type Ia supernova explosions are studied with the weak rates for pf-shell nuclei. The problem of the neutron-rich iron-group isotope over-production compared to the solar abundances is shown to be nearly solved with the use of the new rates and explosion model of slow defraglation with delayed detonation. Evaluation of the weak rates is extended to the island of inversion and the region of neutron-rich nuclei near 78Ni, where two major shells contribute to their configurations.

  15. Improvements to the sodium supply system of a nuclear reactor core

    International Nuclear Information System (INIS)

    Chevallier, Rene; Marchais, Christian.

    1981-01-01

    This invention concerns an improvement to the sodium supply system of a nuclear reactor core and, in particular, concerns the area included between the outlet of the primary circulation pumps and the core proper. A simplified structure and a lightening of all this linking area between the circulation pumps and the distribution tank under the core is achieved and this results in a very significant reduction in the risks of deterioration and in a definite increase in the reliability of the reactor. The invention is therefore an improvement to the sodium supply system of the nuclear reactor core vessel with incorporated exchangers, in which the cool sodium, after passing through the primary exchangers, is collected in a ring compartment from whence it is taken up by the pumps and moved to at least one pipe reaching a distribution tank located under the reactor core [fr

  16. Study on in-core fuel management for CNP1500 nuclear power plant

    International Nuclear Information System (INIS)

    Li Dongsheng

    2005-10-01

    CNP1500 is a four-loop PWR nuclear power plant with light water as moderator and coolant. The reactor core is composed of 205 AFA-3GXL fuel assemblies. The active core height at cold is 426.4 cm and equivalent diameter is 347.0 cm. The reactor thermal output is 4250 MW, and average linear power density is 179.5 W/cm. The cycle length of equilibrium cycle core is 470 equivalent full power days. For all cycles, the moderator temperature coefficients at all conditions are negative values, the nuclear enthalpy rise factors F ΔH at hot full power, all control rods out and equilibrium xenon are less than the limit value, the maximum discharge assembly burnup is less 55000 MW·d/tU, and the shutdown margin values at the end of life meet design criteria. The low-leakage core loading reduces radiation damage on pressure vessel and is beneficial to prolong use lifetime of it. The in-core fuel management design scheme and main calculation results for CNP1500 nuclear power plant are presented. (author)

  17. Neutrino-pair emission from nuclear de-excitation in core-collapse supernova simulations

    Science.gov (United States)

    Fischer, T.; Langanke, K.; Martínez-Pinedo, G.

    2013-12-01

    We study the impact of neutrino-pair production from the de-excitation of highly excited heavy nuclei on core-collapse supernova simulations, following the evolution up to several 100 ms after core bounce. Our study is based on the agile-boltztransupernova code, which features general relativistic radiation hydrodynamics and accurate three-flavor Boltzmann neutrino transport in spherical symmetry. In our simulations the nuclear de-excitation process is described in two different ways. At first we follow the approach proposed by Fuller and Meyer [Astrophys. J.AJLEEY0004-637X10.1086/170317 376, 701 (1991)], which is based on strength functions derived in the framework of the nuclear Fermi-gas model of noninteracting nucleons. Second, we parametrize the allowed and forbidden strength distributions in accordance with measurements for selected nuclear ground states. We determine the de-excitation strength by applying the Brink hypothesis and detailed balance. For both approaches, we find that nuclear de-excitation has no effect on the supernova dynamics. However, we find that nuclear de-excitation is the leading source for the production of electron antineutrinos as well as heavy-lepton-flavor (anti)neutrinos during the collapse phase. At sufficiently high densities, the associated neutrino spectra are influenced by interactions with the surrounding matter, making proper simulations of neutrino transport important for the determination of the neutrino-energy loss rate. We find that, even including nuclear de-excitations, the energy loss during the collapse phase is overwhelmingly dominated by electron neutrinos produced by electron capture.

  18. Emergency core cooling systems in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1981-12-01

    This report contains the responses by the Advisory Committee on Nuclear Safety to three questions posed by the Atomic Energy Control Board concerning the need for Emergency Core Cooling Systems (ECCS) in CANDU nuclear power plants, the effectiveness requirement for such systems, and the extent to which experimental evidence should be available to demonstrate compliance with effectiveness standards

  19. Core support structure for nuclear power plants

    International Nuclear Information System (INIS)

    Steinkamp, E.; Tautz, J.; Ries, H.

    1979-01-01

    A core support structure for nuclear power plants includes a grid of mutually crossing bridges and a support ring surrounding the grid and connected to ends of the outer bridges of the grid, the grid being formed of profile rod crosses having legs of given length, respective legs of pairs of adjacent crosses abutting one another endwise to form together a side of the smallest mesh opening of the grid, and weld means for securing the profile rod crosses to one another at the mutually abutting ends of the legs thereof; and method of producing the foregoing core support structure

  20. A new equation of state Based on Nuclear Statistical Equilibrium for Core-Collapse Simulations

    Science.gov (United States)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-09-01

    We calculate a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ~ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect on the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores.

  1. Reactor core modeling practice: Operational requirements, model characteristics, and model validation

    International Nuclear Information System (INIS)

    Zerbino, H.

    1997-01-01

    The physical models implemented in power plant simulators have greatly increased in performance and complexity in recent years. This process has been enabled by the ever increasing computing power available at affordable prices. This paper describes this process from several angles: First the operational requirements which are more critical from the point of view of model performance, both for normal and off-normal operating conditions; A second section discusses core model characteristics in the light of the solutions implemented by Thomson Training and Simulation (TT and S) in several full-scope simulators recently built and delivered for Dutch, German, and French nuclear power plants; finally we consider the model validation procedures, which are of course an integral part of model development, and which are becoming more and more severe as performance expectations increase. As a conclusion, it may be asserted that in the core modeling field, as in other areas, the general improvement in the quality of simulation codes has resulted in a fairly rapid convergence towards mainstream engineering-grade calculations. This is remarkable performance in view of the stringent real-time requirements which the simulation codes must satisfy as well as the extremely wide range of operating conditions that they are called upon to cover with good accuracy. (author)

  2. Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 1. ZPPR analysis results

    International Nuclear Information System (INIS)

    Mantourov, Guennadi

    2001-05-01

    This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess of Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. The data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS and ZPPR fast reactor cores applying JNC core calculation code CITATION. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the first flight collision probability method and subgroup approach. Especially a converting program was written to transmit the prepared effective cross sections to JNC standard PDS files. Then the CITATION code was applied for 3-D XYZ neutronics calculations of BFS and ZPPR JUPITER experiments series cores. The effects of nuclear data library have been studied by comparing the former results based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for ZPPR-9, ZPPR-13A and ZPPR-17A cores are presented. The calculated correction factor in all cases was less than 1.0%. So the uncertainty in C value caused by possible errors in calculation of these corrections is expected to be less than 0.3% in case of ZPPR-13A and ZPPR-17A cores, and rather less for ZPPR-9 core. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC calculation route revealed a large enough discrepancy in k-eff for ZPPR-9 (about 0.6%) and ZPPR-17A (about 0.5%) cores. For BFS-62-1 and BFS-62-2 cores such analysis is in progress. Stretch cell models for both BFS cores were formed and cell calculations using FFCP code have started. Some results of cell calculations are presented. (author)

  3. In-core gamma dosimetry by solid state nuclear track detectors

    International Nuclear Information System (INIS)

    Khan, H.A.

    1980-02-01

    Results are reported of a study undertaken to develop Solid State Nuclear Track Detectors (SSNTD) for the measurement of gamma doses in the megarad region such as those existing in and around a nuclear reactor core. The changes brought about in the track etching parameters and in the ultraviolet and infrared transmittances, have been studied for possible use as gamma dose measuring indices. Effects of various parameters in the core such as neutron flux, beta particles, water, temperature, and gamma ray spectrum have been investigated and found to have only small influence on the proposed gamma dose measuring indices

  4. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Schirru, Roberto; Carvalho da Silva, Fernando; Medeiros, Jose Antonio Carlos Canedo

    2008-01-01

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem

  5. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: alanmmlima@yahoo.com.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br; Carvalho da Silva, Fernando [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: fernando@con.ufrj.br; Medeiros, Jose Antonio Carlos Canedo [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br

    2008-09-15

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem.

  6. The whiteStar development project: Westinghouse's next generation core design simulator and core monitoring software to power the nuclear renaissance

    International Nuclear Information System (INIS)

    Boyd, W. A.; Mayhue, L. T.; Penkrot, V. S.; Zhang, B.

    2009-01-01

    The WhiteStar project has undertaken the development of the next generation core analysis and monitoring system for Westinghouse Electric Company. This on-going project focuses on the development of the ANC core simulator, BEACON core monitoring system and NEXUS nuclear data generation system. This system contains many functional upgrades to the ANC core simulator and BEACON core monitoring products as well as the release of the NEXUS family of codes. The NEXUS family of codes is an automated once-through cross section generation system designed for use in both PWR and BWR applications. ANC is a multi-dimensional nodal code for all nuclear core design calculations at a given condition. ANC predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. BEACON is an advanced core monitoring and support system which uses existing instrumentation data in conjunction with an analytical methodology for on-line generation and evaluation of 3D core power distributions. This new system is needed to design and monitor the Westinghouse AP1000 PWR. This paper describes provides an overview of the software system, software development methodologies used as well some initial results. (authors)

  7. Nuclear power reactor core melt accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

    2015-11-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus-FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day. Following the introduction, which describes the structure of this book and highlights the objectives of R and D on core melt accidents, this book briefly presents the design and operating principles (Chapter 2) and safety principles (Chapter 3) of the reactors currently in operation in France, as well as the main accident scenarios envisaged and studied (Chapter 4). The objective of these chapters is not to provide exhaustive information on these subjects (the reader should refer to the general reference documents listed in the corresponding chapters), but instead to provide the information needed in order to understand, firstly, the general approach adopted in France for preventing and mitigating the consequences of core melt

  8. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Dominguez, Dany S.; Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G.; Lira, Carlos Alberto Brayner de Oliveira

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  9. VHTR core modeling: coupling between neutronic and thermal-hydraulics

    International Nuclear Information System (INIS)

    Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.

    2005-01-01

    Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)

  10. Nuclear reactor core and fuel element therefor

    International Nuclear Information System (INIS)

    Fortescue, P.

    1986-01-01

    This patent describes a nuclear reactor core. This core consists of vertical columns of disengageable fuel elements stacked one atop another. These columns are arranged in side-by-side relationship to form a substantially continuous horizontal array. Each of the fuel elements include a block of refractory material having relatively good thermal conductivity and neutron moderating characteristics. The block has a pair of parallel flat top and bottom end faces and sides which are substantially prependicular to the end faces. The sides of each block is aligned vertically within a vertical column, with the sides of vertically adjacent blocks. Each of the blocks contains fuel chambers, including outer rows containing only fuel chambers along the sides of the block have nuclear fuel material disposed in them. The blocks also contain vertical coolant holes which are located inside the fuel chambers in the outer rows and the fuel chambers which are not located in the outer rows with the fuel chambers and which extend axially completely through from end face to end face and form continuous vertical intracolumn coolant passageways in the reactor core. The blocks have vertical grooves extending along the sides of the blocks form interblock channels which align in groups to form continuous vertical intercolumn coolant passsageways in the reactor core. The blocks are in the form of a regular hexagonal prism with each side of the block having vertical gooves defining one half of one of the coolant interblock channels, six corner edges on the blocks have vertical groves defining one-third of an interblock channel, the vertical sides of the blocks defining planar vertical surfaces

  11. Modular core component support for nuclear reactor

    International Nuclear Information System (INIS)

    Finch, L.M.; Anthony, A.J.

    1975-01-01

    The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

  12. Nuclear detectors for in-core power-reactors

    International Nuclear Information System (INIS)

    Duchene, Jean; Verdant, Robert.

    1979-12-01

    Nuclear reactor control is commonly obtained through neutronic measurements, ex-core and in-core. In large size reactors flux instabilities may take place. For a good monitoring of them, local in-core power measurements become particularly useful. This paper intends to review the questions about neutronic sensors with could be used in-core. A historical account about methods is given first, from early power reactors with brief description of each system. Sensors presently used (ionization fission chambers, self-powered detectors) are then considered and also those which could be developped such as gamma thermometers. Their physical basis, main characteristics and operation modes are detailed. Preliminary tests and works needed for an extension of their life-time are indicated. As an example present irradiation tests at the CEA are then proposed. Two tables will help comparing the characteristics of each type in terms of its precise purpose: fuel monitoring, safety or power control. Finally a table summarizes the kind of sensors mounted on working power reactors and another one is a review of characteristics for some detectors from obtainable commercial sheets [fr

  13. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

    Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

  14. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M J; Oyinloye, J O; Chambers, I [Electrowatt Consulting Engineers and Scientists, Warrington, Cheshire (United Kingdom); Scott, C K [Atlantic Nuclear Services, Fredericton, NB (Canada); Omar, A M [Atomic Energy Control Board, Ottawa, ON (Canada)

    1991-12-31

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs.

  15. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    International Nuclear Information System (INIS)

    Lewis, M.J.; Oyinloye, J.O.; Chambers, I.; Scott, C.K.; Omar, A.M.

    1990-01-01

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs

  16. A nuclear reactor core fuel reload optimization using Artificial-Ant-Colony Connective Networks

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Schirru, Roberto

    2005-01-01

    A Pressurized Water Reactor core must be reloaded every time the fuel burnup reaches a level when it is not possible to sustain nominal power operation. The nuclear core fuel reload optimization consists in finding a burned-up and fresh-fuel-assembly pattern that maximizes the number of full operational days. This problem is NP-hard, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Besides that, the problem is non-linear and its search space is highly discontinual and multimodal. In this work a parallel computational system based on Ant Colony System (ACS) called Artificial-Ant-Colony Networks is introduced to solve the nuclear reactor core fuel reload optimization problem. ACS is a system based on artificial agents that uses the reinforcement learning technique and was originally developed to solve the Traveling Salesman Problem, which is conceptually similar to the nuclear fuel reload problem. (author)

  17. Mathematical model for the preliminary analysis of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239a

    Energy Technology Data Exchange (ETDEWEB)

    Grey, J.; Chow, S.

    1976-06-30

    The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. Such a concept could be particularly useful for missions which require both relatively high acceleration (e.g., for planetocentric maneuvers) and high performance at low acceleration (e.g., on heliocentric trajectories or for trajectory shaping). The first volume develops the mathematical model of the system.

  18. A benchmark for coupled thermohydraulics system/three-dimensional neutron kinetics core models

    International Nuclear Information System (INIS)

    Kliem, S.

    1999-01-01

    During the last years 3D neutron kinetics core models have been coupled to advanced thermohydraulics system codes. These coupled codes can be used for the analysis of the whole reactor system. Although the stand-alone versions of the 3D neutron kinetics core models and of the thermohydraulics system codes generally have a good verification and validation basis, there is a need for additional validation work. This especially concerns the interaction between the reactor core and the other components of a nuclear power plant (NPP). In the framework of the international 'Atomic Energy Research' (AER) association on VVER Reactor Physics and Reactor Safety, a benchmark for these code systems was defined. (orig.)

  19. Conversion of the core of the TRIGA Mark III reactor at the Mexican Nuclear Centre

    International Nuclear Information System (INIS)

    Moran Lopez, J.M.; Lucatero, M.A.; Reyes Andrade, B.; Rivero Gutierrez, T.; Sainz Mejia, E.

    1990-01-01

    It was decided to convert the core of the TRIGA MARK III reactor at the Mexican Nuclear Centre run by the National Nuclear Institute because of problems detected during the operation, such as a lack of excess reactivity for operation at nominal power over long periods and difficulties in the maintenance and calibration of the control panel. In order to compensate for the lack of excess reactivity the fuel elements taken to the highest burnup were replaced by fresh elements acquired for this purpose. The latter, however, had a different enrichment, and this necessitated a detailed analysis of the neutronic and thermohydraulic behaviour of the reactor with a view to determining a mixed core configuration which would meet safe operation requirements. In conducting the thermohydraulic analysis, a natural convection coolant flow model was developed to determine coolant velocity and pressure drop patterns within the core. The heat transfer equations were solved and it was found that the hottest fuel element did not attain critical heat flux conditions. In loading this core it was also necessary to analyse procedures and to consider the possible effects of reaching criticality with fuel elements having different enrichments. The loading procedure is described, as is the measurement system and the results obtained. In order to resolve the calibration and maintenance problems, a new, more advanced control panel was designed with conventional and nuclear detection systems and modern components

  20. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  1. A computational model for thermal fluid design analysis of nuclear thermal rockets

    International Nuclear Information System (INIS)

    Given, J.A.; Anghaie, S.

    1997-01-01

    A computational model for simulation and design analysis of nuclear thermal propulsion systems has been developed. The model simulates a full-topping expander cycle engine system and the thermofluid dynamics of the core coolant flow, accounting for the real gas properties of the hydrogen propellant/coolant throughout the system. Core thermofluid studies reveal that near-wall heat transfer models currently available may not be applicable to conditions encountered within some nuclear rocket cores. Additionally, the possibility of a core thermal fluid instability at low mass fluxes and the effects of the core power distribution are investigated. Results indicate that for tubular core coolant channels, thermal fluid instability is not an issue within the possible range of operating conditions in these systems. Findings also show the advantages of having a nonflat centrally peaking axial core power profile from a fluid dynamic standpoint. The effects of rocket operating conditions on system performance are also investigated. Results show that high temperature and low pressure operation is limited by core structural considerations, while low temperature and high pressure operation is limited by system performance constraints. The utility of these programs for finding these operational limits, optimum operating conditions, and thermal fluid effects is demonstrated

  2. Calculational model used in the analysis of nuclear performance of the Light Water Breeder Reactor (LWBR) (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Freeman, L.B. (ed.)

    1978-08-01

    The calculational model used in the analysis of LWBR nuclear performance is described. The model was used to analyze the as-built core and predict core nuclear performance prior to core operation. The qualification of the nuclear model using experiments and calculational standards is described. Features of the model include: an automated system of processing manufacturing data; an extensively analyzed nuclear data library; an accurate resonance integral calculation; space-energy corrections to infinite medium cross sections; an explicit three-dimensional diffusion-depletion calculation; a transport calculation for high energy neutrons; explicit accounting for fuel and moderator temperature feedback, clad diameter shrinkage, and fuel pellet growth; and an extensive testing program against experiments and a highly developed analytical standard.

  3. Nuclear Human Resources Development Program using Educational Core Simulator

    International Nuclear Information System (INIS)

    Choi, Yu Sun; Hong, Soon Kwan

    2015-01-01

    KHNP-CRI(Korea Hydro and Nuclear Power Co.-Central Research Institute) has redesigned the existing Core Simulator(CoSi) used as a sort of training tools for reactor engineers in operating nuclear power plant to support Nuclear Human Resources Development (NHRD) Program focusing on the nuclear department of Dalat university in Vietnam. This program has been supported by MOTIE in Korea and cooperated with KNA(Korea Nuclear Association for International Cooperation) and HYU(Hanyang University) for enhancing the nuclear human resources of potential country in consideration with Korean Nuclear Power Plant as a next candidate energy sources. KHNP-CRI has provided Edu-CoSi to Dalat University in Vietnam in order to support Nuclear Human Resources Development Program in Vietnam. Job Qualification Certificates Program in KHNP is utilized to design a training course for Vietnamese faculty and student of Dalat University. Successfully, knowhow on lecturing the ZPPT performance, training and maintaining Edu-CoSi hardware are transferred by several training courses which KHNP-CRI provides

  4. Nuclear Human Resources Development Program using Educational Core Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yu Sun; Hong, Soon Kwan [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    KHNP-CRI(Korea Hydro and Nuclear Power Co.-Central Research Institute) has redesigned the existing Core Simulator(CoSi) used as a sort of training tools for reactor engineers in operating nuclear power plant to support Nuclear Human Resources Development (NHRD) Program focusing on the nuclear department of Dalat university in Vietnam. This program has been supported by MOTIE in Korea and cooperated with KNA(Korea Nuclear Association for International Cooperation) and HYU(Hanyang University) for enhancing the nuclear human resources of potential country in consideration with Korean Nuclear Power Plant as a next candidate energy sources. KHNP-CRI has provided Edu-CoSi to Dalat University in Vietnam in order to support Nuclear Human Resources Development Program in Vietnam. Job Qualification Certificates Program in KHNP is utilized to design a training course for Vietnamese faculty and student of Dalat University. Successfully, knowhow on lecturing the ZPPT performance, training and maintaining Edu-CoSi hardware are transferred by several training courses which KHNP-CRI provides.

  5. Nuclear design and analysis report for KALIMER breakeven core conceptual design

    International Nuclear Information System (INIS)

    Kim, Sang Ji; Song, Hoon; Lee, Ki Bog; Chang, Jin Wook; Hong, Ser Gi; Kim, Young Gyun; Kim, Yeong Il

    2002-04-01

    During the phase 2 of LMR design technology development project, the breakeven core configuration was developed with the aim of the KALIMER self-sustaining with regard to the fissile material. The excess fissile material production is limited only to the extent of its own requirement for sustaining its planned power operation. The average breeding ratio is estimated to be 1.05 for the equilibrium core and the fissile plutonium gain per cycle is 13.9 kg. The nuclear performance characteristics as well as the reactivity coefficients have been analyzed so that the design evaluation in other activity areas can be made. In order to find out a realistic heavy metal flow evolution and investigate cycle-dependent nuclear performance parameter behaviors, the startup and transition cycle loading strategies are developed, followed by the startup core physics analysis. Driver fuel and blankets are assumed to be shuffled at the time of each reload. The startup core physics analysis has shown that the burnup reactivity swing, effective delayed neutron fraction, conversion ratio and peak linear heat generation rate at the startup core lead to an extreme of bounding physics data for safety analysis. As an outcome of this study, a whole spectrum of reactor life is first analyzed in detail for the KALIMER core. It is experienced that the startup core analysis deserves more attention than the current design practice, before the core configuration is finalized based on the equilibrium cycle analysis alone.

  6. Core access system for nuclear reactor

    International Nuclear Information System (INIS)

    Andrea, C.

    1977-01-01

    Disclosed is an improved nuclear reactor arrangement to facilitate both through-the-head instrumentation and insertion and removal of assemblies from the nuclear core. The arrangement is of the type including a reactor vessel head comprising a large rotatable cover having a plurality of circular openings therethrough, a plurality of upwardly extending nozzles mounted on the upper surface of a large cover, and a plurality of upwardly extending skirts mounted on a large cover about the periphery or boundary of the circular openings; a plurality of small plugs for each of the openings in the large cover, the plugs also having nozzles mounted on the upper surface thereof, and drive mechanisms mounted on top of some of the nozzles and having means extending therethrough into the reactor vessel, the drive mechanisms and nozzles extending above the elevation of the upwardly extending skirts

  7. Evaluation of the Troxler Model 4640 Thin Lift Nuclear Density Gauge. Research report (Interim)

    International Nuclear Information System (INIS)

    Solaimanian, M.; Holmgreen, R.J.; Kennedy, T.W.

    1990-07-01

    The report describes the results of a research study to determine the effectiveness of the Troxler Model 4640 Thin Lift Nuclear Density Gauge. The densities obtained from cores and the nuclear density gauge from seven construction projects were compared. The projects were either newly constructed or under construction when the tests were performed. A linear regression technique was used to investigate how well the core densities could be predicted from nuclear densities. Correlation coefficients were determined to indicate the degree of correlation between the core and nuclear densities. Using a statistical analysis technique, the range of the mean difference between core and nuclear measurements was established for specified confidence levels for each project. Analysis of the data indicated that the accuracy of the gauge is material dependent. While relatively acceptable results were obtained with limestone mixtures, the gauge did not perform satisfactorily with mixtures containing siliceous aggregate

  8. Thermohydraulic modeling of nuclear thermal rockets: The KLAXON code

    International Nuclear Information System (INIS)

    Hall, M.L.; Rider, W.J.; Cappiello, M.W.

    1992-01-01

    The hydrogen flow from the storage tanks, through the reactor core, and out the nozzle of a Nuclear Thermal Rocket is an integral design consideration. To provide an analysis and design tool for this phenomenon, the KLAXON code is being developed. A shock-capturing numerical methodology is used to model the gas flow (the Harten, Lax, and van Leer method, as implemented by Einfeldt). Preliminary results of modeling the flow through the reactor core and nozzle are given in this paper

  9. Development of Reactor Core Model based on Optimal Analysis for Shinhanul no. 1, 2 Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung-min [Korea Hydro Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    As one of the outputs of 'Development of the Shin Hanul Nuclear Plant(SHN) 1,2 Simulator' project which is being done by KHNP Central Research Institute, the SHN1,2 Simulator is being developed including the KNICS methodology and advanced Alarm Systems first applied to the Nuclear Power Plant in Korea, and the SHN 1,2 simulator adopts the virtually stimulated HMI(Human-Machine Interface) for the non-safety MMIS system, whose key-programs are identical to those applied to the real SHN 1,2 plants. The purpose of this paper is to develop localization core model by integrating the Simulator system with the Simulator core model though technology agreement of KAERI. To develop ShinHanul 1 and 2 reactor core simulator model, KHNP and KAERI create MASTER-SIM model and tried validation. And calculations of MASSIM{sub S}S program for MASTER{sub S}IM validation, are within tolerance range. Test has not yet been completed. And many verification will be conducted MASTER-SIM software is expected to be the highest economic software and satisfy international simulator standards.

  10. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling

    International Nuclear Information System (INIS)

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail

  11. A fermionic molecular dynamics technique to model nuclear matter

    International Nuclear Information System (INIS)

    Vantournhout, K.; Jachowicz, N.; Ryckebusch, J.

    2009-01-01

    Full text: At sub-nuclear densities of about 10 14 g/cm 3 , nuclear matter arranges itself in a variety of complex shapes. This can be the case in the crust of neutron stars and in core-collapse supernovae. These slab like and rod like structures, designated as nuclear pasta, have been modelled with classical molecular dynamics techniques. We present a technique, based on fermionic molecular dynamics, to model nuclear matter at sub-nuclear densities in a semi classical framework. The dynamical evolution of an antisymmetric ground state is described making the assumption of periodic boundary conditions. Adding the concepts of antisymmetry, spin and probability distributions to classical molecular dynamics, brings the dynamical description of nuclear matter to a quantum mechanical level. Applications of this model vary from investigation of macroscopic observables and the equation of state to the study of fundamental interactions on the microscopic structure of the matter. (author)

  12. Comparative assessment of out-of-core nuclear thermionic power systems

    International Nuclear Information System (INIS)

    Estabrook, W.C.; Koenig, D.R.; Prickett, W.Z.

    1975-01-01

    The hardware selections available for fabrication of a nuclear electric propulsion stage for planetary exploration were explored. The investigation was centered around a heat-pipe-cooled, fast-spectrum nuclear reactor for an out-of-core power conversion system with sufficient detail for comparison with the in-core system studies completed previously. A survey of competing power conversion systems still indicated that the modular reliability of thermionic converters makes them the desirable choice to provide the 240-kWe end-of-life power for at least 20,000 full power hours. The electrical energy will be used to operate a number of mercury ion bombardment thrusters with a specific impulse in the range of about 4,000-5,000 seconds. (Author)

  13. Vibration tests on some models of PEC reactor core elements

    International Nuclear Information System (INIS)

    Bonacina, G.; Castoldi, A.; Zola, M.; Cecchini, F.; Martelli, A.; Vincenzi, D.

    1982-01-01

    This paper describes the aims of the experimental tests carried out at ISMES, within an agreement with the Department of Fast Reactors of ENEA, on some models of the elements of PEC Fast Nuclear Reactor Core in the frame of the activities for the seismic verification of the PEC core. The seismic verification is briefly described with particular attention to the problems arising from the shocks among the various elements during an earthquake, as well as the computer code used, the purpose and the techniques used to perform tests, some results and the first comparison between the theory and the experimental data

  14. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  15. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Mazaira, Leorlen Y.R.; Dominguez, Dany S.; Hernandez, Carlos R.G.

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  16. AGR core models and their application to HTRs and RBMKs

    International Nuclear Information System (INIS)

    Baylis, Samuel

    2014-01-01

    EDF Energy operates 14 AGRs, commissioned between 1976 and 1989. The graphite moderators of these gas cooled reactors are subjected to a number of ageing processes under fast neutron irradiation in a high temperature CO2 environment. As the graphite ages, continued safe operation requires an advanced whole-core modeling capability to enable accurate assessments of the core’s ability to fulfil fundamental nuclear safety requirements. This is also essential in evaluating the reactor's remaining economic lifetime, and similar assessments are useful for HTRs in the design stage. A number of computational and physical models of AGR graphite cores have been developed or are in development, allowing simulation of the reactors in normal, fault and seismic conditions. Many of the techniques developed are applicable to other graphite moderated reactors. Modeling of the RBMK allows validation against a core in a more advanced state of ageing than the AGRs, while there is also an opportunity to adapt the models for high temperature reactors. As an example, a finite element model of the HTR-PM side reflector based on rigid bodies and nonlinear springs is developed, allowing rapid assessments of distortion in the structure to be made. A model of the RBMK moderator has also been produced using an established AGR code based on similar methods. In addition, this paper discusses the limitations of these techniques and the development of more complex core models that address these limitations, along with the lessons that can be applied to HTRs. (author)

  17. No-Core Shell Model for A = 47 and A = 49

    Energy Technology Data Exchange (ETDEWEB)

    Vary, J P; Negoita, A G; Stoica, S

    2006-11-13

    We apply the no-core shell model to the nuclear structure of odd-mass nuclei straddling {sup 48}Ca. Starting with the NN interaction, that fits two-body scattering and bound state data, we evaluate the nuclear properties of A = 47 and A = 49 nuclei while preserving all the underlying symmetries. Due to model space limitations and the absence of three-body interactions, we incorporate phenomenological interaction terms determined by fits to A = 48 nuclei in a previous effort. Our modified Hamiltonian produces reasonable spectra for these odd-mass nuclei. In addition to the differences in single-particle basis states, the absence of a single-particle Hamiltonian in our no-core approach complicates comparisons with valence effective NN interactions. We focus on purely off-diagonal two-body matrix elements since they are not affected by ambiguities in the different roles for one-body potentials and we compare selected sets of fp-shell matrix elements of our initial and modified Hamiltonians in the harmonic oscillator basis with those of a recent model fp-shell interaction, the GXPF1 interaction of Honma et al. While some significant differences emerge from these comparisons, there is an overall reasonably good correlation between our off-diagonal matrix elements and those of GXPF1.

  18. Rules for design of nuclear graphite core components - some considerations and approaches

    International Nuclear Information System (INIS)

    Svalbonas, V.; Stilwell, T.C.; Zudans, Z.

    1978-01-01

    The use of graphite as a structural element presents unusual problems both for the designer and stress analysist. When the structure happens to be a nuclear reactor core, these problems are significantly magnified both by the environment and the attendant safety requirements. In the high temperature gas reactor (HTGR) core a large number of elements are constructed of nuclear graphite. This paper discusses the attendant difficulties, and presents some approaches, for ASME code safety-consistent design and analysis. The statistical scatter of material properties, which complicates even the definitions of allowable stress, as well as the brittle, anisotropic, inhomogeneous nature of the graphite was considered. The study of this subject was undertaken under contract to the U.S. Nuclear Regulatory Commission. (Auth.)

  19. Real time thermal hydraulic model for high temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

    2013-01-01

    A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model. (authors)

  20. Review of the SCDAP/RELAP5/MOD3.1 code structure and core T/H model before core damage

    International Nuclear Information System (INIS)

    Kim, See Darl; Kim, Dong Ha

    1998-04-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code is being developed at the INEL under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. NRC. As The current time, the SCDAP/RELAP5/MOD3.1 code is the result of merging the RELAP5/MOD3 and SCDAP models. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. Major purpose of the report is to provide information about the characteristics of SCDAP/RELAP5/MOD3.1 core T/H models for an integrated severe accident computer code being developed under the mid/long-term project. This report analyzes the overall code structure which consists of the input processor, transient controller, and plot file handler. The basic governing equations to simulate the thermohydraulics of the primary system are also described. As the focus is currently concentrated in the core, core nodalization parameters of the intact geometry and the phenomenological subroutines for the damaged core are summarized for the future usage. In addition, the numerical approach for the heat conduction model is investigated along with heat convection model. These studies could provide a foundation for input preparation and model improvement. (author). 6 refs., 3 tabs., 4 figs

  1. Improvement of Axial Reflector Cross Section Generation Model for PWR Core Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Cheon Bo; Lee, Kyung Hoon; Cho, Jin Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This paper covers the study for improvement of axial reflector XS generation model. In the next section, the improved 1D core model is represented in detail. Reflector XS generated by the improved model is compared to that of the conventional model in the third section. Nuclear design parameters generated by these two XS sets are also covered in that section. Significant of this study is discussed in the last section. Two-step procedure has been regarded as the most practical approach for reactor core designs because it offers core design parameters quite rapidly within acceptable range. Thus this approach is adopted for SMART (System-integrated Modular Advanced Reac- Tor) core design in KAERI with the DeCART2D1.1/ MASTER4.0 (hereafter noted as DeCART2D/ MASTER) code system. Within the framework of the two-step procedure based SMART core design, various researches have been studied to improve the core design reliability and efficiency. One of them is improvement of reflector cross section (XS) generation models. While the conventional FA/reflector two-node model used for most core designs to generate reflector XS cannot consider the actual configuration of fuel rods that intersect at right angles to axial reflectors, the revised model reflects the axial fuel configuration by introducing the radially simplified core model. The significance of the model revision is evaluated by observing HGC generated by DeCART2D, reflector XS, and core design parameters generated by adopting the two models. And it is verified that about 30 ppm CBC error can be reduced and maximum Fq error decreases from about 6 % to 2.5 % by applying the revised model. Error of AO and axial power shapes are also reduced significantly. Therefore it can be concluded that the simplified 1D core model improves the accuracy of the axial reflector XS and leads to the two-step procedure reliability enhancement. Since it is hard for core designs to be free from the two-step approach, it is necessary to find

  2. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  3. Gas core nuclear thermal rocket engine research and development in the former USSR

    International Nuclear Information System (INIS)

    Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G.; Gurfink, M.M.

    1992-09-01

    Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept

  4. Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

    International Nuclear Information System (INIS)

    Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

    1980-01-01

    Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

  5. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

  6. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  7. Fundamental design bases for independent core cooling in Swedish nuclear power reactors

    International Nuclear Information System (INIS)

    Jelinek, Tomas

    2015-01-01

    New regulations on design and construction of nuclear power plants came into force in 2005. The need of an independent core cooling system and if the regulations should include such a requirement was discussed. The Swedish Radiation Safety authority (SSM) decided to not include such a requirement because of open questions about the water balance and started to investigate the consequences of an independent core cooling system. The investigation is now finished and SSM is also looking at the lessons learned from the accident in Fukushima 2011. One of the most important measures in the Swedish national action plan is the implementation of an independent core cooling function for all Swedish power plants. SSM has investigated the basic design criteria for such a function where some important questions are the level of defence in depth and the acceptance criteria. There is also a question about independence between the levels of defence in depth that SSM have included in the criteria. Another issue that has to be taken into account is the complexity of the system and the need of automation where independence and simplicity are very strong criteria. In the beginning of 2014 a memorandum was finalized regarding fundamental design bases for independent core cooling in Swedish nuclear power reactors. A decision based on this memorandum with an implementation plan will be made in the first half of 2014. Sweden is also investigating the possibility to have armed personnel on site, which is not allowed currently. The result from the investigation will have impact on the possibility to use mobile equipment and the level of protection of permanent equipment. In this paper, SSM will present the memorandum for design bases for independent core cooling in Swedish nuclear power reactors that was finalized in March 20147 that also describe SSM's position regarding independence and automation of the independent core cooling function. This memorandum describes the Swedish

  8. In-core nuclear fuel management optimization of VVER1000 using perturbation theory

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser

    2011-01-01

    In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain fuel integrity. Because of the numerous possible patterns of the fuel assemblies in the reactor core, finding the best configuration is so important and complex. Different methods for optimization of fuel loading pattern in the core have been introduced so far. In this study, a software is programmed in C ⧣ language to find an order of the fuel loading pattern of the VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process lunches by considering the initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. It shall be noticed that the designed algorithm is performed by just shuffling the fuel assemblies. The obtained results by employing the mentioned method on a typical reactor reveal that this method has a high precision in achieving a pattern with an allowable radial power peaking factor. (author)

  9. Human Cytomegalovirus Nuclear Capsids Associate with the Core Nuclear Egress Complex and the Viral Protein Kinase pUL97.

    Science.gov (United States)

    Milbradt, Jens; Sonntag, Eric; Wagner, Sabrina; Strojan, Hanife; Wangen, Christina; Lenac Rovis, Tihana; Lisnic, Berislav; Jonjic, Stipan; Sticht, Heinrich; Britt, William J; Schlötzer-Schrehardt, Ursula; Marschall, Manfred

    2018-01-13

    The nuclear phase of herpesvirus replication is regulated through the formation of regulatory multi-component protein complexes. Viral genomic replication is followed by nuclear capsid assembly, DNA encapsidation and nuclear egress. The latter has been studied intensely pointing to the formation of a viral core nuclear egress complex (NEC) that recruits a multimeric assembly of viral and cellular factors for the reorganization of the nuclear envelope. To date, the mechanism of the association of human cytomegalovirus (HCMV) capsids with the NEC, which in turn initiates the specific steps of nuclear capsid budding, remains undefined. Here, we provide electron microscopy-based data demonstrating the association of both nuclear capsids and NEC proteins at nuclear lamina budding sites. Specifically, immunogold labelling of the core NEC constituent pUL53 and NEC-associated viral kinase pUL97 suggested an intranuclear NEC-capsid interaction. Staining patterns with phospho-specific lamin A/C antibodies are compatible with earlier postulates of targeted capsid egress at lamina-depleted areas. Important data were provided by co-immunoprecipitation and in vitro kinase analyses using lysates from HCMV-infected cells, nuclear fractions, or infectious virions. Data strongly suggest that nuclear capsids interact with pUL53 and pUL97. Combined, the findings support a refined concept of HCMV nuclear trafficking and NEC-capsid interaction.

  10. Human Cytomegalovirus Nuclear Capsids Associate with the Core Nuclear Egress Complex and the Viral Protein Kinase pUL97

    Directory of Open Access Journals (Sweden)

    Jens Milbradt

    2018-01-01

    Full Text Available The nuclear phase of herpesvirus replication is regulated through the formation of regulatory multi-component protein complexes. Viral genomic replication is followed by nuclear capsid assembly, DNA encapsidation and nuclear egress. The latter has been studied intensely pointing to the formation of a viral core nuclear egress complex (NEC that recruits a multimeric assembly of viral and cellular factors for the reorganization of the nuclear envelope. To date, the mechanism of the association of human cytomegalovirus (HCMV capsids with the NEC, which in turn initiates the specific steps of nuclear capsid budding, remains undefined. Here, we provide electron microscopy-based data demonstrating the association of both nuclear capsids and NEC proteins at nuclear lamina budding sites. Specifically, immunogold labelling of the core NEC constituent pUL53 and NEC-associated viral kinase pUL97 suggested an intranuclear NEC-capsid interaction. Staining patterns with phospho-specific lamin A/C antibodies are compatible with earlier postulates of targeted capsid egress at lamina-depleted areas. Important data were provided by co-immunoprecipitation and in vitro kinase analyses using lysates from HCMV-infected cells, nuclear fractions, or infectious virions. Data strongly suggest that nuclear capsids interact with pUL53 and pUL97. Combined, the findings support a refined concept of HCMV nuclear trafficking and NEC-capsid interaction.

  11. Modelling guidelines for core exit temperature simulations with system codes

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)

    2015-05-15

    Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.

  12. Modeling of reflood of severely damaged reactor core

    International Nuclear Information System (INIS)

    Bachrata, A.

    2012-01-01

    The TMI-2 accident and recently Fukushima accident demonstrated that the nuclear safety philosophy has to cover accident sequences involving massive core melt in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although severe accidents are low likelihood and might be caused only by multiple failures, accident management is implemented for controlling their course and mitigating their consequences. In case of severe accident, the fuel rods may be severely damaged and oxidized. Finally, they collapse and form a debris bed on core support plate. Removal of decay heat from a damaged core is a challenging issue because of the difficulty for water to penetrate inside a porous medium. The reflooding (injection of water into core) may be applied only if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ from original rod bundle geometry and will resemble to the severe damaged core observed in TMI-2. The higher temperatures and smaller hydraulic diameters in a porous medium make the coolability more difficult than for intact fuel rods under typical loss of coolant accident conditions. The modeling of this kind of hydraulic and heat transfer is a one of key objectives of this. At IRSN, part of the studies is realized using an European thermo-hydraulic computer code for severe accident analysis ICARE-CATHARE. The objective of this thesis is to develop a 3D reflood model (implemented into ICARE-CATHARE) that is able to treat different configurations of degraded core in a case of severe accident. The proposed model is characterized by treating of non-equilibrium thermal between the solid, liquid and gas phase. It includes also two momentum balance equations. The model is based on a previously developed model but is improved in order to take into account intense boiling regimes (in particular

  13. WNP-2 core model upgrade

    International Nuclear Information System (INIS)

    Golightly, C.E.; Ravindranath, T.K.; Belblidia, L.A.; O'Farrell, D.; Andersen, P.S.

    2006-01-01

    The paper describes the core model upgrade of the WNP-2 training simulator and the reasons for the upgrade. The core model as well as the interface with the rest of the simulator are briefly described . The paper also describes the procedure that will be used by WNP-2 to update the simulator core data after future core reloads. Results from the fully integrated simulator are presented. (author)

  14. Fuel assembly and nuclear reactor core

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

    1995-01-01

    The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

  15. Device for measuring flow rate in a nuclear reactor core

    International Nuclear Information System (INIS)

    Hamano, Jiro.

    1980-01-01

    Purpose: To always calculate core flow rate automatically and accurately in BWR type nuclear power plants. Constitution: Jet pumps are provided to the recycling pump and to the inside of the pressure vessel of a nuclear reactor. The jet pumps comprise a plurality of calibrated jet pumps for forcively convecting the coolants and a plurality of not calibrated jet pumps in order to cool the heat generated in the reactor core. The difference in the pressures between the upper and the lower portions in both of the jet pumps is measured by difference pressure transducers. Further, a thermo-sensitive element is provided to measure the temperature of recycling water at the inlet of the recycling pump. The output signal from the difference pressure transducer is inputted to a process computer, calculated periodically based on predetermined calculation equations, compensated for the temperature by a recycling water temperature signal and outputted as a core flow rate signal to a recoder. The signal is also used for the power distribution calculation in the process computer and the minimum limit power ratio as the thermal limit value for the fuels is outputted. (Furukawa, Y.)

  16. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model

    International Nuclear Information System (INIS)

    Oliveira, A.C.J.G. de.

    1988-12-01

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs

  17. Model stars with degenerate dwarf cores and helium-burning shells - A stationary-burning approximation

    Energy Technology Data Exchange (ETDEWEB)

    Iben, I. Jr.; Tutukov, A.V. (Illinois Univ., Urbana (USA); Astronomicheskii Sovet, Moscow (USSR))

    1989-07-01

    The characteristics of model stars consisting of a degenerate dwarf core and an envelope which is burning a nuclear fuel or fuels in its interior are explored. The models are relevant to stars which are accreting matter from a companion, to single stars in late stages of evolution, to stripped noninteracting remnants of binary star evolution, and to merging and merged degenerate dwarfs. For any given mass and choice of nuclear fuels, a sequence of models is constructed which differ with respect to the mass of the degenerate core and the envelope characteristics. Each sequence has at least three distinct branches: a degenerate dwarf branch along which envelope mass increases with decreasing luminosity, a plateau branch characterized by a very small envelope mass and by a nearly constant luminosity which reaches the maximum achievable value for the sequence, and an asymptotic giant branch which is at the lowest temperatures achievable and along which envelope mass decreases with increasing luminosity. 78 refs.

  18. Model stars with degenerate dwarf cores and helium-burning shells - A stationary-burning approximation

    International Nuclear Information System (INIS)

    Iben, I. Jr.; Tutukov, A.V.

    1989-01-01

    The characteristics of model stars consisting of a degenerate dwarf core and an envelope which is burning a nuclear fuel or fuels in its interior are explored. The models are relevant to stars which are accreting matter from a companion, to single stars in late stages of evolution, to stripped noninteracting remnants of binary star evolution, and to merging and merged degenerate dwarfs. For any given mass and choice of nuclear fuels, a sequence of models is constructed which differ with respect to the mass of the degenerate core and the envelope characteristics. Each sequence has at least three distinct branches: a degenerate dwarf branch along which envelope mass increases with decreasing luminosity, a plateau branch characterized by a very small envelope mass and by a nearly constant luminosity which reaches the maximum achievable value for the sequence, and an asymptotic giant branch which is at the lowest temperatures achievable and along which envelope mass decreases with increasing luminosity. 78 refs

  19. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    International Nuclear Information System (INIS)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-01-01

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ∼ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  20. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    Science.gov (United States)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-11-01

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ~ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  1. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    Energy Technology Data Exchange (ETDEWEB)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki [Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan) and Advanced Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Numazu College of Technology, Ooka 3600, Numazu, Shizuoka 410-8501 (Japan); Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan)

    2012-11-12

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to {approx} 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  2. The development of direct core monitoring in Nuclear Electric plc

    International Nuclear Information System (INIS)

    Curtis, R.F.; Jones, S. Reed, J.; Wickham, A.J.

    1996-01-01

    Monitoring of graphite behaviour in Nuclear Electric Magnox and AGR reactors is necessary to support operating safety cases and to ensure that reactor operation is optimized to sustain the necessary core integrity for the economic life of the reactors. The monitoring programme combines studies for pre-characterized ''installed'' samples with studies on samples trepanned from within the cores and also with studies of core and channel geometry using specially designed equipment. Nuclear Electric has two trepanning machines originally designed for Magnox-reactor work which have been used for a substantial programme over many years. They have recently been upgraded to improve sampling speed, safety and versatility - the last being demonstrated by their adaptation for a recently-won contract associated with decommissioning the Windscale piles. Radiological hazards perceived when the AGR trepanning system was designed resulted in very cumbersome equipment. This has worked well but has been inconvenient in operation. The development of a smaller and improved system for deploying the equipment is now reported. Channel dimension monitoring equipment is discussed in detail with examples of data recovered from both Magnox and AGR cores. A resolution of ± 2 of arc (tilt) and ± 0.01 mm change in diameter in attainable. It is also theoretically possible to establish brick stresses by measuring geometry changes which result from trepanning. Current development work on a revolving scanning laser rangefinder which will enable the measurement of diameters to a resolution of 0.001 mm will also be discussed. This paper also discusses non-destructive techniques for crack detection employing ultrasound or resistance networks, the use of special manipulators to deliver inspection and repair equipment and recent developments to install displacement monitors in peripheral regions of the cores, to aid the understanding of the interaction of the restraint system with the core - the region

  3. Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport

    International Nuclear Information System (INIS)

    Hecker, H.C.

    1984-04-01

    This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U 233 -Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

  4. Development of 3D ferromagnetic model of tokamak core withstrong toroidal asymmetry

    Czech Academy of Sciences Publication Activity Database

    Markovič, Tomáš; Gryaznevich, M.; Ďuran, Ivan; Svoboda, V.; Pánek, Radomír

    96-97, October (2015), s. 302-305 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] R&D Projects: GA ČR GAP205/11/2341; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : tokamak * ferromagnetic core * model of ferromagnet * integral method * tokamak GOLEM Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://www.sciencedirect.com/science/article/pii/S0920379615002100

  5. Web-based Core Design System Development

    International Nuclear Information System (INIS)

    Moon, So Young; Kim, Hyung Jin; Yang, Sung Tae; Hong, Sun Kwan

    2011-01-01

    The selection of a loading pattern is one of core design processes in the operation of a nuclear power plant. A potential new loading pattern is identified by selecting fuels that to not exceed the major limiting factors of the design and that satisfy the core design conditions for employing fuel data from the existing loading pattern of the current operating cycle. The selection of a loading pattern is also related to the cycle plan of an operating nuclear power plant and must meet safety and economic requirements. In selecting an appropriate loading pattern, all aspects, such as input creation, code runs and result processes are processed as text forms manually by a designer, all of which may be subject to human error, such as syntax or running errors. Time-consuming results analysis and decision-making processes are the most significant inefficiencies to avoid. A web-based nuclear plant core design system was developed here to remedy the shortcomings of an existing core design system. The proposed system adopts the general methodology of OPR1000 (Korea Standard Nuclear Power Plants) and Westinghouse-type plants. Additionally, it offers a GUI (Graphic User Interface)-based core design environment with a user-friendly interface for operators. It reduces human errors related to design model creation, computation, final reload core model selection, final output confirmation, and result data validation and verification. Most significantly, it reduces the core design time by more than 75% compared to its predecessor

  6. Seismic response of a block-type nuclear reactor core

    International Nuclear Information System (INIS)

    Dove, R.C.; Bennett, J.G.; Merson, J.L.

    1976-05-01

    An analytical model is developed to predict seismic response of large gas-cooled reactor cores. The model is used to investigate scaling laws involved in the design of physical models of such cores, and to make parameter studies

  7. Ab Initio Symmetry-Adapted No-Core Shell Model

    International Nuclear Information System (INIS)

    Draayer, J P; Dytrych, T; Launey, K D

    2011-01-01

    A multi-shell extension of the Elliott SU(3) model, the SU(3) symmetry-adapted version of the no-core shell model (SA-NCSM), is described. The significance of this SA-NCSM emerges from the physical relevance of its SU(3)-coupled basis, which – while it naturally manages center-of-mass spuriosity – provides a microscopic description of nuclei in terms of mixed shape configurations. Since typically configurations of maximum spatial deformation dominate, only a small part of the model space suffices to reproduce the low-energy nuclear dynamics and hence, offers an effective symmetry-guided framework for winnowing of model space. This is based on our recent findings of low-spin and high-deformation dominance in realistic NCSM results and, in turn, holds promise to significantly enhance the reach of ab initio shell models.

  8. Nuclear cardiology core syllabus of the European Association of Cardiovascular Imaging (EACVI).

    Science.gov (United States)

    Gimelli, Alessia; Neglia, Danilo; Schindler, Thomas H; Cosyns, Bernard; Lancellotti, Patrizio; Kitsiou, Anastasia

    2015-04-01

    The European Association of Cardiovascular Imaging (EACVI) Core Syllabus for Nuclear Cardiology is now available online. The syllabus lists key elements of knowledge in nuclear cardiology. It represents a framework for the development of training curricula and provides expected knowledge-based learning outcomes to the nuclear cardiology trainees. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2015. For permissions please email: journals.permissions@oup.com.

  9. Muscle spindles exhibit core lesions and extensive degeneration of intrafusal fibers in the Ryr1I4895T/wt mouse model of core myopathy

    International Nuclear Information System (INIS)

    Zvaritch, Elena; MacLennan, David H.

    2015-01-01

    Muscle spindles from the hind limb muscles of adult Ryr1 I4895T/wt (IT/+) mice exhibit severe structural abnormalities. Up to 85% of the spindles are separated from skeletal muscle fascicles by a thick layer of connective tissue. Many intrafusal fibers exhibit degeneration, with Z-line streaming, compaction and collapse of myofibrillar bundles, mitochondrial clumping, nuclear shrinkage and pyknosis. The lesions resemble cores observed in the extrafusal myofibers of this animal model and of core myopathy patients. Spindle abnormalities precede those in extrafusal fibers, indicating that they are a primary pathological feature in this murine Ryr1-related core myopathy. Muscle spindle involvement, if confirmed for human core myopathy patients, would provide an explanation for an array of devastating clinical features characteristic of these diseases and provide novel insights into the pathology of RYR1-related myopathies. - Highlights: • Muscle spindles exhibit structural abnormalities in a mouse model of core myopathy. • Myofibrillar collapse and mitochondrial clumping is observed in intrafusal fibers. • Myofibrillar degeneration follows a pattern similar to core formation in extrafusal myofibers. • Muscle spindle abnormalities are a part of the pathological phenotype in the mouse model of core myopathy. • Direct involvement of muscle spindles in the pathology of human RYR1-related myopathies is proposed

  10. Analysis of ex-core detector response measured during nuclear ship Mutsu land-loaded core critical experiment

    International Nuclear Information System (INIS)

    Itagaki, M.; Abe, J.I.; Kuribayashi, K.

    1987-01-01

    There are some cases where the ex-core neutron detector response is dependent not only on the fission source distribution in a core but also on neutron absorption in the borated water reflector. For example, an unexpectedly large response variation was measured during the nuclear ship Mutsu land-loaded core critical experiment. This large response variation is caused largely by the boron concentration change associated with the change in control rod positioning during the experiment. The conventional Crump-Lee response calculation method has been modified to take into account this boron effect. The correction factor in regard to this effect has been estimated using the one-dimensional transport code ANISN. The detector response variations obtained by means of this new calculation procedure agree well with the measured values recorded during the experiment

  11. The Impact of the Nuclear Equation of State in Core Collapse Supernovae

    Science.gov (United States)

    Baird, M. L.; Lentz, E. J.; Hix, W. R.; Mezzacappa, A.; Messer, O. E. B.; Liebendoerfer, M.; TeraScale Supernova Initiative Collaboration

    2005-12-01

    One of the key ingredients to the core collapse supernova mechanism is the physics of matter at or near nuclear density. Included in simulations as part of the Equation of State (EOS), nuclear repulsion experienced at high densities are responsible for the bounce shock, which initially causes the outer envelope of the supernova to expand, as well as determining the structure of the newly formed proto-neutron star. Recent years have seen renewed interest in this fundamental piece of supernova physics, resulting in several promising candidate EOS parameterizations. We will present the impact of these variations in the nuclear EOS using spherically symmetric, Newtonian and General Relativistic neutrino transport simulations of stellar core collapse and bounce. This work is supported in part by SciDAC grants to the TeraScale Supernovae Initiative from the DOE Office of Science High Energy, Nuclear, and Advanced Scientific Computing Research Programs. Oak Ridge National Laboratory is managed by UT-Battelle, LLC, for U.S. Department of Energy under contract DEAC05-00OR22725

  12. Enhanced core monitoring system for Browns Ferry Nuclear Plant

    International Nuclear Information System (INIS)

    Lindsey, R.S.

    1980-01-01

    A system of computer hardware and software is being developed to supplement the process computers at Browns Ferry Nuclear Plant in the area of reactor core monitoring. All data stored in the process computers will be made available through a data link to an onsite minicomputer which will store and edit the data for engineering and operations personnel. Important core parameters will be effectively displayed on color graphic CRT terminals using techniques such as blinking, shading, and color coding to emphasize significant values. This data will also be made available to Tennessee Valley Authority's Chattanooga central office support groups through a data network between existing computers

  13. Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2008-01-01

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

  14. Aerosol core nuclear reactor for space-based high energy/power nuclear-pumped lasers

    International Nuclear Information System (INIS)

    Prelas, M.A.; Boody, F.P.; Zediker, M.S.

    1987-01-01

    An aerosol core reactor concept can overcome the efficiency and/or chemical activity problems of other fuel-reactant interface concepts. In the design of a laser using the nuclear energy for a photon-intermediate pumping scheme, several features of the aerosol core reactor concept are attractive. First, the photon-intermediate pumping concept coupled with photon concentration methods and the aerosol fuel can provide the high power densities required to drive high energy/power lasers efficiently (about 25 to 100 kW/cu cm). Secondly, the intermediate photons should have relatively large mean free paths in the aerosol fuel which will allow the concept to scale more favorably. Finally, the aerosol core reactor concept can use materials which should allow the system to operate at high temperatures. An excimer laser pumped by the photons created in the fluorescer driven by a self-critical aerosol core reactor would have reasonable dimensions (finite cylinder of height 245 cm and radius of 245 cm), reasonable laser energy (1 MJ in approximately a 1 millisecond pulse), and reasonable mass (21 kg uranium, 8280 kg moderator, 460 kg fluorescer, 450 kg laser medium, and 3233 kg reflector). 12 references

  15. The effects of radiation on aluminium alloys in the core of energy nuclear reactors

    International Nuclear Information System (INIS)

    Petrossian, V.G.

    1995-01-01

    One of the attractive directions in the worldwide practice of nuclear installations is the replacement of expensive zirconium alloy with more cheap materials, particularly aluminium allo. For Heat Supply Nuclear Plants (HSNP) with approximately 473 K core temperatures, the use of heat-resistant aluminium alloys seems to be reasonable. The present work is concerned with the studies on radiation effects on aluminium alloy, and interaction between the alloy and coolant in the reactor core. (author). 2 refs., 3 figs., 1 tab

  16. Optimization analysis of the nuclear fuel cycle transition to the last core

    International Nuclear Information System (INIS)

    Rebollo, L.; Blanco, J.

    2001-01-01

    The Zorita NPP was the first Spanish commercial nuclear reactor connected to the grid. It is a 160 MW one loop PWR, Westinghouse design, owned by UFG, in operation since 1968. The configuration of the reactor core is based on 69 fuel elements type 14 x 14, the standard reload of the present equilibrium cycle being based on 16 fuel elements with 3.6% enrichment in 235 U. In order to properly plan the nuclear fuel management of the transition cycles to its end of life, presently foreseen by 2008, an based on the non-reprocessing option required by the policy of the Spanish Administration, a technical-economical optimization analysis has been performed. As a result, a fuel management strategy has been defined looking for getting simultaneously the minimum integral fuel cost of the transition from the present equilibrium cycle to the last core, as well as the minimum residual worth of the fuel remaining in the core after the final outage. Based on the ''lessons learned'' derived from the study, the time margin for the decision making has been determined, and a planning of the nuclear fuel supply for the transition reloads, specifying both the number of fuel elements and their enrichment in 235 U, as been prepared. Finally, based on the calculated economical worth of the partially burned fuel of the last core, after the end of its operation cycle, a financial cover for yearly compensation from now on of the foreseen final lost has been elaborated. Most of the conceptual conclusions obtained are applicable to the other commercial nuclear reactors in operation owned by UFG, so that they are understood to be of general interest and broad application to commercial PWR. (author)

  17. Application of the pertubation theory to a two channels model for sensitivity calculations in PWR cores

    International Nuclear Information System (INIS)

    Oliveira, A.C.J.G. de; Andrade Lima, F.R. de

    1989-01-01

    The present work is an application of the perturbation theory (Matricial formalism) to a simplified two channels model, for sensitivity calculations in PWR cores. Expressions for some sensitivity coefficients of thermohydraulic interest were developed from the proposed model. The code CASNUR.FOR was written in FORTRAN to evaluate these sensitivity coefficients. The comparison between results obtained from the matrical formalism of pertubation theory with those obtained directly from the two channels model, makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations. (author) [pt

  18. Studies on the inhomogeneous core density of a fluidized bed nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Van der Hagen, T.H.J.J.; Van Dam, H.; Hoogenboom, J.E.; Khotylev, V.A. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.; Harteveld, W.; Mudde, R.F.

    1997-12-31

    Results are reported on the expected time dependent core density profile of a fluidized-bed nuclear fission reactor. Core densities have been measured in a test facility by the gamma-transmission technique. Bubble and particle-cluster sizes, positions, velocities and frequencies could be determined. Neutronic studies have been performed on the influence of core voids on reactivity using Monte-Carlo and neutron-transport codes. Fuel-particle importance has been determined. Point-kinetic parameters have been calculated for linking reactivity perturbations to power fluctuations. (author)

  19. Nuclear structure calculations in $^{20}$Ne with No-Core Configuration-Interaction model

    OpenAIRE

    Konieczka, Maciej; Satuła, Wojciech

    2016-01-01

    Negative parity states in $^{20}$Ne and Gamow-Teller strength distribution for the ground-state beta-decay of $^{20}$Na are calculated for the very first time using recently developed No-Core Configuration-Interaction model. The approach is based on multi-reference density functional theory involving isospin and angular-momentum projections. Advantages and shortcomings of the method are briefly discussed.

  20. Core Technology Development of Nuclear spin polarization

    International Nuclear Information System (INIS)

    Yoo, Byung Duk; Gwon, Sung Ok; Kwon, Duck Hee; Lee, Sung Man

    2009-12-01

    In order to study nuclear spin polarization, we need several core technologies such as laser beam source to polarize the nuclear spin, low pressured helium cell development whose surface is essential to maintain polarization otherwise most of the polarized helium relaxed in short time, development of uniform magnetic field system which is essential for reducing relaxation, efficient vacuum system, development of polarization measuring system, and development of pressure raising system about 1000 times. The purpose of this study is to develop resonable power of laser system, that is at least 5 watt, 1083 nm, 4GHz tuneable. But the limitation of this research fund enforce to develop amplifying system into 5 watt with 1 watt system utilizing laser-diod which is already we have in stock. We succeeded in getting excellent specification of fiber laser system with power of 5 watts, 2 GHz linewidth, more than 80 GHz tuneable

  1. Results of an analysis of in-core measurements during the first core cycle of the Greifswald nuclear power plant, unit 3

    International Nuclear Information System (INIS)

    Gehre, G.

    1982-01-01

    First results of an analysis of flux and temperature values obtained from the in-core system in the third unit of the Greifswald nuclear power plant during the first core cycle are presented. The analysis has been performed with the aid of the computer code INCA. Possibilities and limits of this code are shown. (author)

  2. Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

    International Nuclear Information System (INIS)

    Friess, Friederike Renate

    2017-01-01

    According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

  3. Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Friess, Friederike Renate

    2017-07-12

    According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

  4. Calculations and selection of a TRIGA core for the Nuclear Reactor IAN-R1

    International Nuclear Information System (INIS)

    Castiblanco, L.A.; Sarta, J.A.

    1997-01-01

    The Reactor Group used the code WIMS reduced to five groups of energy, together with the code CITATION, and evaluated four configurations for a core, according to the grid actually installed. The four configurations were taken from the two proposals presented to the Instituto de Ciencias Nucleares y Energias Alternativas by General Atomics Company. In this paper, the Authors selected the best configuration according to the performance of flux distribution and excess reactivity, for a TRIGA core to be installed in the Nuclear Reactor IAN-R1

  5. Three-Body Nuclear Forces from a Matrix Model

    CERN Document Server

    Hashimoto, Koji

    2010-01-01

    We compute three-body nuclear forces at short distances by using the nuclear matrix model of holographic QCD proposed in our previous paper with P. Yi. We find that the three-body forces at short distances are repulsive for (a) aligned three neutrons with averaged spins, and (b) aligned proton-proton-neutron / proton-neutron-neutron. These indicate that in dense states of neutrons such as cores of neutron stars, or in Helium-3 / tritium nucleus, the repulsive forces are larger than the ones estimated from two-body forces only.

  6. Nonlinear Model of Tape Wound Core Transformers

    Directory of Open Access Journals (Sweden)

    A. Vahedi

    2015-03-01

    Full Text Available Recently, tape wound cores due to their excellent magnetic properties, are widely used in different types of transformers. Performance prediction of these transformers needs an accurate model with ability to determine flux distribution within the core and magnetic loss. Spiral structure of tape wound cores affects the flux distribution and always cause complication of analysis. In this paper, a model based on reluctance networks method is presented for analysis of magnetic flux in wound cores. Using this model, distribution of longitudinal and transverse fluxes within the core can be determined. To consider the nonlinearity of the core, a dynamic hysteresis model is included in the presented model. Having flux density in different points of the core, magnetic losses can be calculated. To evaluate the validity of the model, results are compared with 2-D FEM simulations. In addition, a transformer designed for series-resonant converter and simulation results are compared with experimental measurements. Comparisons show accuracy of the model besides simplicity and fast convergence

  7. Nuclear equation of state for core-collapse supernova simulations with realistic nuclear forces

    Energy Technology Data Exchange (ETDEWEB)

    Togashi, H., E-mail: hajime.togashi@riken.jp [Nishina Center for Accelerator-Based Science, Institute of Physical and Chemical Research (RIKEN), 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Nakazato, K. [Faculty of Arts and Science, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Takehara, Y.; Yamamuro, S.; Suzuki, H. [Department of Physics, Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan); Takano, M. [Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Department of Pure and Applied Physics, Graduate School of Advanced Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan)

    2017-05-15

    A new table of the nuclear equation of state (EOS) based on realistic nuclear potentials is constructed for core-collapse supernova numerical simulations. Adopting the EOS of uniform nuclear matter constructed by two of the present authors with the cluster variational method starting from the Argonne v18 and Urbana IX nuclear potentials, the Thomas–Fermi calculation is performed to obtain the minimized free energy of a Wigner–Seitz cell in non-uniform nuclear matter. As a preparation for the Thomas–Fermi calculation, the EOS of uniform nuclear matter is modified so as to remove the effects of deuteron cluster formation in uniform matter at low densities. Mixing of alpha particles is also taken into account following the procedure used by Shen et al. (1998, 2011). The critical densities with respect to the phase transition from non-uniform to uniform phase with the present EOS are slightly higher than those with the Shen EOS at small proton fractions. The critical temperature with respect to the liquid–gas phase transition decreases with the proton fraction in a more gradual manner than in the Shen EOS. Furthermore, the mass and proton numbers of nuclides appearing in non-uniform nuclear matter with small proton fractions are larger than those of the Shen EOS. These results are consequences of the fact that the density derivative coefficient of the symmetry energy of our EOS is smaller than that of the Shen EOS.

  8. Utilization of cross-section covariance data in FBR core nuclear design and cross-section adjustment

    International Nuclear Information System (INIS)

    Ishikawa, Makoto

    1994-01-01

    In the core design of large fast breeder reactors (FBRs), it is essentially important to improve the prediction accuracy of nuclear characteristics from the viewpoint of both reducing cost and insuring reliability of the plant. The cross-section errors, that is, covariance data are one of the most dominant sources for the prediction uncertainty of the core parameters, therefore, quantitative evaluation of covariance data is indispensable for FBR core design. The first objective of the present paper is to introduce how the cross-section covariance data are utilized in the FBR core nuclear design works. The second is to delineate the cross-section adjustment study and its application to an FBR design, because this improved design method markedly enhances the needs and importance of the cross-section covariance data. (author)

  9. A review of MAAP4 code structure and core T/H model

    International Nuclear Information System (INIS)

    Song, Yong Mann; Park, Soo Yong

    1998-03-01

    The modular accident analysis program (MAAP) version 4 is a computer code that can simulate the response of LWR plants during severe accident sequences and includes models for all of the important phenomena which might occur during accident sequences. In this report, MAAP4 code structure and core thermal hydraulic (T/H) model which models the T/H behavior of the reactor core and the response of core components during all accident phases involving degraded cores are specifically reviewed and then reorganized. This reorganization is performed via getting the related models together under each topic whose contents and order are same with other two reports for MELCOR and SCDAP/RELAP5 to be simultaneously published. Major purpose of the report is to provide information about the characteristics of MAAP4 core T/H models for an integrated severe accident computer code development being performed under the one of on-going mid/long-term nuclear developing project. The basic characteristics of the new integrated severe accident code includes: 1) Flexible simulation capability of primary side, secondary side, and the containment under severe accident conditions, 2) Detailed plant simulation, 3) Convenient user-interfaces, 4) Highly modularization for easy maintenance/improvement, and 5) State-of-the-art model selection. In conclusion, MAAP4 code has appeared to be superior for 3) and 4) items but to be somewhat inferior for 1) and 2) items. For item 5), more efforts should be made in the future to compare separated models in detail with not only other codes but also recent world-wide work. (author). 17 refs., 1 tab., 12 figs

  10. Fuel/propellant mixing in an open-cycle gas core nuclear rocket engine

    International Nuclear Information System (INIS)

    Guo, X.; Wehrmeyer, J.A.

    1997-01-01

    A numerical investigation of the mixing of gaseous uranium and hydrogen inside an open-cycle gas core nuclear rocket engine (spherical geometry) is presented. The gaseous uranium fuel is injected near the centerline of the spherical engine cavity at a constant mass flow rate, and the hydrogen propellant is injected around the periphery of the engine at a five degree angle to the wall, at a constant mass flow rate. The main objective is to seek ways to minimize the mixing of uranium and hydrogen by choosing a suitable injector geometry for the mixing of light and heavy gas streams. Three different uranium inlet areas are presented, and also three different turbulent models (k-var-epsilon model, RNG k-var-epsilon model, and RSM model) are investigated. The commercial CFD code, FLUENT, is used to model the flow field. Uranium mole fraction, axial mass flux, and radial mass flux contours are obtained. copyright 1997 American Institute of Physics

  11. Modeling Stress Strain Relationships and Predicting Failure Probabilities For Graphite Core Components

    Energy Technology Data Exchange (ETDEWEB)

    Duffy, Stephen [Cleveland State Univ., Cleveland, OH (United States)

    2013-09-09

    This project will implement inelastic constitutive models that will yield the requisite stress-strain information necessary for graphite component design. Accurate knowledge of stress states (both elastic and inelastic) is required to assess how close a nuclear core component is to failure. Strain states are needed to assess deformations in order to ascertain serviceability issues relating to failure, e.g., whether too much shrinkage has taken place for the core to function properly. Failure probabilities, as opposed to safety factors, are required in order to capture the bariability in failure strength in tensile regimes. The current stress state is used to predict the probability of failure. Stochastic failure models will be developed that can accommodate possible material anisotropy. This work will also model material damage (i.e., degradation of mechanical properties) due to radiation exposure. The team will design tools for components fabricated from nuclear graphite. These tools must readily interact with finite element software--in particular, COMSOL, the software algorithm currently being utilized by the Idaho National Laboratory. For the eleastic response of graphite, the team will adopt anisotropic stress-strain relationships available in COMSO. Data from the literature will be utilized to characterize the appropriate elastic material constants.

  12. Modeling Stress Strain Relationships and Predicting Failure Probabilities For Graphite Core Components

    International Nuclear Information System (INIS)

    Duffy, Stephen

    2013-01-01

    This project will implement inelastic constitutive models that will yield the requisite stress-strain information necessary for graphite component design. Accurate knowledge of stress states (both elastic and inelastic) is required to assess how close a nuclear core component is to failure. Strain states are needed to assess deformations in order to ascertain serviceability issues relating to failure, e.g., whether too much shrinkage has taken place for the core to function properly. Failure probabilities, as opposed to safety factors, are required in order to capture the bariability in failure strength in tensile regimes. The current stress state is used to predict the probability of failure. Stochastic failure models will be developed that can accommodate possible material anisotropy. This work will also model material damage (i.e., degradation of mechanical properties) due to radiation exposure. The team will design tools for components fabricated from nuclear graphite. These tools must readily interact with finite element software--in particular, COMSOL, the software algorithm currently being utilized by the Idaho National Laboratory. For the eleastic response of graphite, the team will adopt anisotropic stress-strain relationships available in COMSO. Data from the literature will be utilized to characterize the appropriate elastic material constants.

  13. Cycle length maximization in PWRs using empirical core models

    International Nuclear Information System (INIS)

    Okafor, K.C.; Aldemir, T.

    1987-01-01

    The problem of maximizing cycle length in nuclear reactors through optimal fuel and poison management has been addressed by many investigators. An often-used neutronic modeling technique is to find correlations between the state and control variables to describe the response of the core to changes in the control variables. In this study, a set of linear correlations, generated by two-dimensional diffusion-depletion calculations, is used to find the enrichment distribution that maximizes cycle length for the initial core of a pressurized water reactor (PWR). These correlations (a) incorporate the effect of composition changes in all the control zones on a given fuel assembly and (b) are valid for a given range of control variables. The advantage of using such correlations is that the cycle length maximization problem can be reduced to a linear programming problem

  14. The Westinghouse BEACON on-line core monitoring system

    International Nuclear Information System (INIS)

    Buechel, Robert J.; Boyd, William A.; Casadei, Alberto L.

    1995-01-01

    BEACON (Best Estimate Analysis of Core Operations - Nuclear), a core monitoring and operational support package developed by Westinghouse, has been installed at many operating PWRs worldwide. The BEACON system is a real-time monitoring system which can be used in plants with both fixed and movable incore detector systems and utilizes an on-line nodal model combined with core instrumentation data to provide continuous core power distribution monitoring. In addition, accurate core-predictive capabilities utilizing a full core nodal model updated according to plant operating history can be made to provide operational support. Core history information is kept and displayed to help operators anticipate core behavior and take pro-active control actions. The BEACON system has been licensed by the U.S. Nuclear Regulatory Commission for direct, continuous monitoring of DNBR and peak linear heat rate. This allows BEACON to be integrated into the plant technical specifications to permit significant relaxation of operating limitations defined by conventional technical specifications. (author). 4 refs, 2 figs, 1 tab

  15. Core of a fast neutron nuclear reactor

    International Nuclear Information System (INIS)

    Giacometti, Christian; Mougniot, J.-C.; Ravier, Jean.

    1974-01-01

    The fast neutron nuclear reactor described includes an internal area in fissile material completely enclosed in an area of fertile material forming the outside blanket. The internal fissile area is provided with housings exclusively filled with fertile material forming one or more inside blankets. In this core the internal blankets are shaped like rings vertically separating superimposed rings of fissile material. The blanket of material nearest to the periphery is circumscribed externally by a contour having an indented shape on its straight section so as to increase the contact area between this blanket and the external blanket [fr

  16. Sensors for use in nuclear reactor cores

    International Nuclear Information System (INIS)

    1980-01-01

    A neutron sensor is described for use in nuclear reactor cores which does not require external power but merely an emitter, a collector and an insulator material between the two to generate an electric current that is indicative of the intensity of the radiation. The sensor is manufactured in such a way that brazed joints or spices are avoided and the insulation material used may be of relatively low density of compaction and will center the emitter and the lead wire with respect to the outer sheath or tube without deformation or varying geometry of the center wire or emitter. (UK)

  17. Sensors for use in nuclear reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-21

    A neutron sensor is described for use in nuclear reactor cores which does not require external power but merely an emitter, a collector and an insulator material between the two to generate an electric current that is indicative of the intensity of the radiation. The sensor is manufactured in such a way that brazed joints or spices are avoided and the insulation material used may be of relatively low density of compaction and will center the emitter and the lead wire with respect to the outer sheath or tube without deformation or varying geometry of the center wire or emitter.

  18. Observation of the state of the nuclear reactor core by means of non-linear observation algorithms

    International Nuclear Information System (INIS)

    Maciel Palacio, F.E.; Espana, M.D.

    1990-01-01

    A combined, variable-adaptive structure, non-linear observer was designed in order to observe the state of the nuclear reactor core, based on the Absolute Stability Theory. The observer was proved under noise and modelling error conditions. Successful results were obtained in the observation of the states in both cases, showing clear improvement in the observation due to the application of adaptive and variable structure ideas. (Author) [es

  19. The 3-dimensional core model DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Mittag, S.; Rohde, U.

    1999-01-01

    Analyzing the safety margins in transients and accidents of nuclear reactors 3-dimensional models of the core were used to avoid conservative assumptions needed for point kinetics or 1-dimensional models. Therefore, the 3D code DYN3D has been developed for the analysis of reactivity initiated accidents (RIA) in thermal nuclear reactors. The power distributions are calculated with the help of nodal expansion methods (NEM) for hexagonal and Cartesian geometry. The fuel rod model and the thermohydraulic part provide fuel temperatures, coolant temperatures and densities as well as boron concentrations for the calculation of feedback effects on the basis of cross section libraries generated by cell codes. Safety relevant parameters like maximum fuel and cladding temperatures, critical heat flux and degree of cladding oxidation are estimated. DYN3D can analyze RIA initiated by moved control rods and/or perturbations of the coolant flow. Stationary and transient boundary conditions for the coolant flow, the core inlet temperatures and boron concentrations at the core inlet have to be given. For analyzing more complex transients the code DYN3D is coupled with the plant model ATHLET of the GRS. The extensive validation work accomplished for DYN3D is presented in several examples. Some applications of the code are described. (orig.) [Deutsch] Die Verwendung 3-dimensionaler Kernmodelle zur Untersuchung der Sicherheitsreserven bei Uebergangsprozessen und Stoerfaellen in Kernreaktoren vermeidet konservative Annahmen, die bei der Benutzung des Punktmodells oder 1-dimensionaler Modelle erforderlich sind. Aus diesen Gruenden wurde das 3-dimensionale Rechenprogramm DYN3D fuer die Untersuchung von Reaktivitaetsstoerfaellen in thermischen Reaktoren entwickelt. Die Leistungsverteilung wird mit nodalen Methoden fuer hexagonale oder kartesische Geometrie berechnet. Das Brennstabmodell und der thermohydraulische Teil von DYN3D liefert die Brennstofftemperaturen, Kuehlmitteltemperaturen

  20. Validating neural-network refinements of nuclear mass models

    Science.gov (United States)

    Utama, R.; Piekarewicz, J.

    2018-01-01

    Background: Nuclear astrophysics centers on the role of nuclear physics in the cosmos. In particular, nuclear masses at the limits of stability are critical in the development of stellar structure and the origin of the elements. Purpose: We aim to test and validate the predictions of recently refined nuclear mass models against the newly published AME2016 compilation. Methods: The basic paradigm underlining the recently refined nuclear mass models is based on existing state-of-the-art models that are subsequently refined through the training of an artificial neural network. Bayesian inference is used to determine the parameters of the neural network so that statistical uncertainties are provided for all model predictions. Results: We observe a significant improvement in the Bayesian neural network (BNN) predictions relative to the corresponding "bare" models when compared to the nearly 50 new masses reported in the AME2016 compilation. Further, AME2016 estimates for the handful of impactful isotopes in the determination of r -process abundances are found to be in fairly good agreement with our theoretical predictions. Indeed, the BNN-improved Duflo-Zuker model predicts a root-mean-square deviation relative to experiment of σrms≃400 keV. Conclusions: Given the excellent performance of the BNN refinement in confronting the recently published AME2016 compilation, we are confident of its critical role in our quest for mass models of the highest quality. Moreover, as uncertainty quantification is at the core of the BNN approach, the improved mass models are in a unique position to identify those nuclei that will have the strongest impact in resolving some of the outstanding questions in nuclear astrophysics.

  1. Review of turbulence modelling for numerical simulation of nuclear reactor thermal-hydraulics

    International Nuclear Information System (INIS)

    Bernard, J.P.; Haapalehto, T.

    1996-01-01

    The report deals with the modelling of turbulent flows in nuclear reactor thermal-hydraulic applications. The goal is to give tools and knowledge about turbulent flows and their modelling in practical applications for engineers, and especially nuclear engineers. The emphasize is on the theory of turbulence, the existing different turbulence models, the state-of-art of turbulence in research centres, the available models in the commercial code CFD-FLOW3D, and the latest applications of turbulence modelling in nuclear reactor thermal-hydraulics. It turns out that it is difficult to elaborate an universal turbulence model and each model has its advantages and drawbacks in each application. However, the increasing power of computers can permit the emergence of new methods of turbulence modelling such as Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) which open new horizons in this field. These latter methods are beginning to be available in commercial codes and are used in different nuclear applications such as 3-D modelling of the nuclear reactor cores and the steam generators. (orig.) (22 refs.)

  2. Westinghouse Nuclear Core Design Training Center - a design simulator

    International Nuclear Information System (INIS)

    Altomare, S.; Pritchett, J.; Altman, D.

    1992-01-01

    The emergence of more powerful computing technology enables nuclear design calculations to be done on workstations. This shift to workstation usage has already had a profound effect in the training area. In 1991, the Westinghouse Electric Corporation's Commercial Nuclear Fuel Division (CNFD) developed and implemented a Nuclear Core Design Training Center (CDTC), a new concept in on-the-job training. The CDTC provides controlled on-the-job training in a structured classroom environment. It alllows one trainer, with the use of a specially prepared training facility, to provide full-scope, hands-on training to many trainees at one time. Also, the CDTC system reduces the overall cycle time required to complete the total training experience while also providing the flexibility of individual training in selected modules of interest. This paper provides descriptions of the CDTC and the respective experience gained in the application of this new concept

  3. A nuclear reactor core fuel reload optimization using Artificial-Ant-Colony Connective Networks; Recarga de reatores nucleares utilizando redes conectivas de colonias de formigas artificiais

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Schirru, Roberto [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: alan@lmp.ufrj.br; schirru@lmp.ufrj.br

    2005-07-01

    A Pressurized Water Reactor core must be reloaded every time the fuel burnup reaches a level when it is not possible to sustain nominal power operation. The nuclear core fuel reload optimization consists in finding a burned-up and fresh-fuel-assembly pattern that maximizes the number of full operational days. This problem is NP-hard, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Besides that, the problem is non-linear and its search space is highly discontinual and multimodal. In this work a parallel computational system based on Ant Colony System (ACS) called Artificial-Ant-Colony Networks is introduced to solve the nuclear reactor core fuel reload optimization problem. ACS is a system based on artificial agents that uses the reinforcement learning technique and was originally developed to solve the Traveling Salesman Problem, which is conceptually similar to the nuclear fuel reload problem. (author)

  4. Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 2. BFS-62 analysis results

    International Nuclear Information System (INIS)

    Mantourov, Guennadi

    2001-11-01

    This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. Data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS-62 and ZPPR JUPIER series fast reactor cores, applying JNC core calculation code CITATION-FBR. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the First Flight Collision Probability method and subgroup approach. Especially, two converting programs were written to transmit the prepared effective cross sections to JNC standard PDS files to let then the CITATION code be applied for 3-D HEXZ neutronics calculations of the investigated cores. The effects of nuclear data library have been studied by comparing the results calculated using ABBN-93 nuclear data library with the former ones obtained in JNC based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for 4 BFS-62 cores as well as for 3 ZPPR JUPITER experiment series cores ZPPR-9, ZPPR-13A and ZPPR-17A are presented. The comparison results for reaction rates distributions for 2 BFS-62 uranium loaded cores are included too. The calculated correction factors applied in all cases were less than 1.0%. The estimated uncertainty in k-effective C values caused by possible errors in calculation of the applied corrections is about 0.3% in case of BFS-62 and ZPPR MOX cores, and is about 0.2% for BFS-62 uranium-loaded cores. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC's calculation route for k-effective results is about 0.3% for ZPPR and BFS-62 cores with plutonium. As for BFS uranium-loaded cores (BFS-62-1 and BFS-62-2) the nuclear data library effect is about 0.1%. Next the sensitivity analysis was applied. It shown that the main contributors to the nuclear data library effect

  5. Understanding Core-Collapse Supernovae

    Science.gov (United States)

    Hix, W. R.; Lentz, E. J.; Baird, M.; Messer, O. E. B.; Mezzacappa, A.; Lee, C.-T.; Bruenn, S. W.; Blondin, J. M.; Marronetti, P.

    2010-03-01

    Our understanding of core-collapse supernovae continues to improve as better microphysics is included in increasingly realistic neutrino-radiationhydrodynamic simulations. Recent multi-dimensional models with spectral neutrino transport, which slowly develop successful explosions for a range of progenitors between 12 and 25 solar mass, have motivated changes in our understanding of the neutrino reheating mechanism. In a similar fashion, improvements in nuclear physics, most notably explorations of weak interactions on nuclei and the nuclear equation of state, continue to refine our understanding of how supernovae explode. Recent progresses on both the macroscopic and microscopic effects that affect core-collapse supernovae are discussed.

  6. An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

    International Nuclear Information System (INIS)

    Clark, J.S.; Walton, J.T.; Mcguire, M.L.

    1992-07-01

    Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines. 11 refs

  7. Feasibility study of passive gamma spectrometry of molten core material from Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy - low-volatile FP and special nuclear material inventory analysis and fundamental characteristics of gamma-rays from fuel debris

    International Nuclear Information System (INIS)

    Sagara, Hiroshi; Tomikawa, Hirofumi; Watahiki, Masaru; Kuno, Yusuke

    2014-01-01

    The technologies applied to the analysis of the Three Mile Island accident were examined in a feasibility study of gamma spectrometry of molten core material from the Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy. The focus is on low-volatile fission products and heavy metal inventory analysis, and the fundamental characteristics of gamma-rays from fuel debris with respect to passive measurements. The inventory ratios of the low-volatile lanthanides, "1"5"4Eu and "1"4"4Ce, to special nuclear materials were evaluated by the entire core inventories in units 1, 2, and 3 with an estimated uncertainty of 9%-13% at the 1σ level for homogenized molten fuel material. The uncertainty is expected to be larger locally owing to the use of the irradiation cycle averaging approach. The ratios were also evaluated as a function of burnup for specific fuel debris with an estimated uncertainty of 13%-25% at the 1σ level for units 1 and 2, and most of the fuels in unit 3, although the uncertainty regarding the separated mixed oxide fuel in unit 3 would be significantly higher owing to the burnup dependence approach. Source photon spectra were also examined and cooling-time-dependent data sets were prepared. The fundamental characteristics of high-energy gamma-rays from fuel debris were investigated by a bare-sphere model transport calculation. Mass attenuation coefficients of fuel debris were evaluated to be insensitive to its possible composition in a high-energy region. The leakage photon ratio was evaluated using a variety of parameters, and a significant impact was confirmed for a certain size of fuel debris. Its correlation was summarized with respect to the leakage photopeak ratio of source "1"5"4Eu. Finally, a preliminary study using a hypothetical canister model of fuel debris based on the experience at Three Mile Island was presented, and future plans were introduced. (author)

  8. Contribution to the modelling of flows and heat transfers during the reflooding phase of a PWR core

    International Nuclear Information System (INIS)

    Colas, D.

    1984-01-01

    This thesis contributes to modelise thermohydraulic phenomena occuring in a pressurized water nuclear reactor core during the reflood phase of a LOCA. The reference accident and phenomena occuring during reflooding are described as well as flow regime and heat transfer proposed models. With these models, we developed a code to compute fluid conditions and fuel rods temperatures in a reactor core chanel. In order to test this code, results of computation are compared with experiments (FLECHT Skewed Tests) and a conclusion is drawn [fr

  9. α Centauri A as a potential stellar model calibrator: establishing the nature of its core

    Science.gov (United States)

    Nsamba, B.; Monteiro, M. J. P. F. G.; Campante, T. L.; Cunha, M. S.; Sousa, S. G.

    2018-05-01

    Understanding the physical process responsible for the transport of energy in the core of α Centauri A is of the utmost importance if this star is to be used in the calibration of stellar model physics. Adoption of different parallax measurements available in the literature results in differences in the interferometric radius constraints used in stellar modelling. Further, this is at the origin of the different dynamical mass measurements reported for this star. With the goal of reproducing the revised dynamical mass derived by Pourbaix & Boffin, we modelled the star using two stellar grids varying in the adopted nuclear reaction rates. Asteroseismic and spectroscopic observables were complemented with different interferometric radius constraints during the optimisation procedure. Our findings show that best-fit models reproducing the revised dynamical mass favour the existence of a convective core (≳ 70% of best-fit models), a result that is robust against changes to the model physics. If this mass is accurate, then α Centauri A may be used to calibrate stellar model parameters in the presence of a convective core.

  10. Multigroup models of the convective epoch in core collapse supernovae

    International Nuclear Information System (INIS)

    Swesty, F Douglas; Myra, Eric S

    2005-01-01

    Understanding the explosion mechanism of core collapse supernovae is a problem that has plagued nuclear astrophysicists since the first computational models of this phenomenon were carried out in the 1960s. Our current theories of this violent phenomenon center around multi-dimensional effects involving radiation-hydrodynamic flows of hot, dense matter and neutrinos. Modeling these multi-dimensional radiative flows presents a computational challenge that will continue to stress high-performance computing beyond the teraflops to the petaflop level. In this paper we describe a few of the scientific discoveries that we have made via terascale computational simulations of supernovae under the auspices of the SciDAC-funded Terascale Supernova Initiative

  11. Comment on the in-core measurement in the WWER nuclear power plant

    International Nuclear Information System (INIS)

    Krett, V.; Dach, K.; Erben, O.

    1985-01-01

    The activity of the Nuclear Research Institute (NRI) Rez in the field of in-core measurement sensors is described in the paper. The results of comparison and calibration experiments realized on the WWR-S research reactor at the NRI are presented. Measurements with fission calorimeters and SPN detectors carried out in the framework of diagnostic fuel assembly program of WWER NPP reactors are described. Noise measurements with detectors of in-core instrumentation of diagnostic fuel assemblies are also mentioned. Comparison experiments on the WWER-440 NPP reactor are described and the method of function verification of neutron sensors of the in-core control system of these reactors is given. (author)

  12. Core monitoring with analytical model adaption

    International Nuclear Information System (INIS)

    Linford, R.B.; Martin, C.L.; Parkos, G.R.; Rahnema, F.; Williams, R.D.

    1992-01-01

    The monitoring of BWR cores has evolved rapidly due to more capable computer systems, improved analytical models and new types of core instrumentation. Coupling of first principles diffusion theory models such as applied to design to the core instrumentation has been achieved by GE with an adaptive methodology in the 3D Minicore system. The adaptive methods allow definition of 'leakage parameters' which are incorporated directly into the diffusion models to enhance monitoring accuracy and predictions. These improved models for core monitoring allow for substitution of traversing in-core probe (TIP) and local power range monitor (LPRM) with calculations to continue monitoring with no loss of accuracy or reduction of thermal limits. Experience in small BWR cores has shown that with one out of three TIP machines failed there was no operating limitation or impact from the substitute calculations. Other capabilities exist in 3D Monicore to align TIPs more accurately and accommodate other types of system measurements or anomalies. 3D Monicore also includes an accurate predictive capability which uses the adaptive results from previous monitoring calculations and is used to plan and optimize reactor maneuvers/operations to improve operating efficiency and reduce support requirements

  13. Core power distribution measurement and data processing in Daya Bay Nuclear Power Station

    International Nuclear Information System (INIS)

    Zhang Hong

    1997-01-01

    For the first time in China, Daya Bay Nuclear Power Station applied the advanced technology of worldwide commercial pressurized reactors to the in-core detectors, the leading excore six-chamber instrumentation for precise axial power distribution, and the related data processing. Described in this article are the neutron flux measurement in Daya Bay Nuclear Power Station, and the detailed data processing

  14. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  15. ORIGEN2.1 Cycle Specific Calculation of Krsko Nuclear Power Plant Decay Heat and Core Inventory

    International Nuclear Information System (INIS)

    Vukovic, J.; Grgic, D.; Konjarek, D.

    2010-01-01

    This paper presents ORIGEN2.1 computer code calculation of Krsko Nuclear Power Plant core for Cycle 24. The isotopic inventory, core activity and decay heat are calculated in one run for the entire core using explicit depletion and decay of each fuel assembly. Separate pre-ori application which was developed is utilized to prepare corresponding ORIGEN2.1 inputs. This application uses information on core loading pattern to determine fuel assembly specific depletion history using 3D burnup which is obtained from related PARCS computer code calculation. That way both detailed single assembly calculations as well as whole core inventory calculations are possible. Because of the immense output of the ORIGEN2.1, another application called post-ori is used to retrieve and plot any calculated property on the basis of nuclide, element, summary isotope or group of elements for activation products, actinides and fission products segments. As one additional possibility, with the post-ori application it is able to calculate radiotoxicity from calculated ORIGEN2.1 inventory. The results which are obtained using the calculation model of ORIGEN2.1 computer code are successfully compared against corresponding ORIGEN-S computer code results.(author).

  16. Development of Core Monitoring System for Nuclear Power Plants (I)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.H.; Kim, Y.B.; Park, M.G; Lee, E.K.; Shin, H.C.; Lee, D.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    1.Object and Necessity of the Study -The main objectives of this study are (1)conversion of APOLLO version BEACON system to HP-UX version core monitoring system, (2)provision of the technical bases to enhance the in-house capability of developing more advanced core monitoring system. 2.Results of the Study - In this study, the revolutionary core monitoring technologies such as; nodal analysis and isotope depletion calculation method, advanced schemes for power distribution control, and treatment of nuclear databank were established. The verification and validation work has been successfully performed by comparing the results with those of the design code and measurement data. The advanced graphic user interface and plant interface method have been implemented to ensure the future upgrade capability. The Unix shell scripts and system dependent software are also improved to support administrative functions of the system. (author). 14 refs., 112 figs., 52 tabs.

  17. Models of the earth's core

    Science.gov (United States)

    Stevenson, D. J.

    1981-01-01

    Combined inferences from seismology, high-pressure experiment and theory, geomagnetism, fluid dynamics, and current views of terrestrial planetary evolution lead to models of the earth's core with five basic properties. These are that core formation was contemporaneous with earth accretion; the core is not in chemical equilibrium with the mantle; the outer core is a fluid iron alloy containing significant quantities of lighter elements and is probably almost adiabatic and compositionally uniform; the more iron-rich inner solid core is a consequence of partial freezing of the outer core, and the energy release from this process sustains the earth's magnetic field; and the thermodynamic properties of the core are well constrained by the application of liquid-state theory to seismic and labroatory data.

  18. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  19. The core-quasiparticle model for odd-odd nuclei and applications to candidates for gamma-ray lasers

    International Nuclear Information System (INIS)

    Strottman, D.D.

    1988-01-01

    A reliable estimate of the properties of isomers that may be viable candidates for a gamma-ray laser requires the use of the most accurate save functions possible. The majority of models that have been used to estimate the properties of isomers are applicable to only selected regions of the nuclear mass table. In particular, the Bohr-Mottelson model of odd-A and odd-odd nuclei will fail if the even-even core is not strongly deformed or if the deformations are changing strongly as a function of mass. This paper reports how the problem is overcome in a new core- quasiparticle model for odd-odd nuclei. The model introduces the pairing interaction ab initio; the odd-A states are mixtures of particle and hole states. The core may be soft towards deformation or axial asymmetry and may change rapidly as a function of mass. Thus, the model is ideally suited for application to the region of transitional nuclei such as the Te, La, and Os regions

  20. Conceptual Nuclear Design Of Two Models Of Research Reactor Proposed For Vietnam

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Huynh Ton Nghiem; Le Vinh Vinh; Vo Doan Hai Dang

    2007-01-01

    The joint study on the development of a new research reactor model for Vietnam was done. The KAERI (Korea Atomic Energy Research Institute) experts and DNRI (Dalat Nuclear Research Institute) researchers developed an advanced HANARO reactor (AHR), a 20-MW open-tank-in-pool type reactor, upward cooled and moderated by light water, reflected by heavy water and rod type fuel assemblies used. Based on the AHR model, a MTR reactor with plate fuel assemblies was developed. Computer codes named MCNP and MVP/BURN were used. Major analyses have been done for the relevant nuclear design parameters such as the neutron flux and power distributions, reactivity coefficients, control rod worth, etc. in both with clean, unperturbed core and equilibrium core condition. In case of AHR model, calculation results using MVP/BURN and MCNP codes were compared with the results using HELIOS and MCNP codes by KAERI experts and they are in a good agreement. (author)

  1. RSMASS-D nuclear thermal propulsion and bimodal system mass models

    Science.gov (United States)

    King, Donald B.; Marshall, Albert C.

    1997-01-01

    Two relatively simple models have been developed to estimate reactor, radiation shield, and balance of system masses for a particle bed reactor (PBR) nuclear thermal propulsion concept and a cermet-core power and propulsion (bimodal) concept. The approach was based on the methodology developed for the RSMASS-D models. The RSMASS-D approach for the reactor and shield sub-systems uses a combination of simple equations derived from reactor physics and other fundamental considerations along with tabulations of data from more detailed neutron and gamma transport theory computations. Relatively simple models are used to estimate the masses of other subsystem components of the nuclear propulsion and bimodal systems. Other subsystem components include instrumentation and control (I&C), boom, safety systems, radiator, thermoelectrics, heat pipes, and nozzle. The user of these models can vary basic design parameters within an allowed range to achieve a parameter choice which yields a minimum mass for the operational conditions of interest. Estimated system masses are presented for a range of reactor power levels for propulsion for the PBR propulsion concept and for both electrical power and propulsion for the cermet-core bimodal concept. The estimated reactor system masses agree with mass predictions from detailed calculations with xx percent for both models.

  2. Protein kinases responsible for the phosphorylation of the nuclear egress core complex of human cytomegalovirus.

    Science.gov (United States)

    Sonntag, Eric; Milbradt, Jens; Svrlanska, Adriana; Strojan, Hanife; Häge, Sigrun; Kraut, Alexandra; Hesse, Anne-Marie; Amin, Bushra; Sonnewald, Uwe; Couté, Yohann; Marschall, Manfred

    2017-10-01

    Nuclear egress of herpesvirus capsids is mediated by a multi-component nuclear egress complex (NEC) assembled by a heterodimer of two essential viral core egress proteins. In the case of human cytomegalovirus (HCMV), this core NEC is defined by the interaction between the membrane-anchored pUL50 and its nuclear cofactor, pUL53. NEC protein phosphorylation is considered to be an important regulatory step, so this study focused on the respective role of viral and cellular protein kinases. Multiply phosphorylated pUL50 varieties were detected by Western blot and Phos-tag analyses as resulting from both viral and cellular kinase activities. In vitro kinase analyses demonstrated that pUL50 is a substrate of both PKCα and CDK1, while pUL53 can also be moderately phosphorylated by CDK1. The use of kinase inhibitors further illustrated the importance of distinct kinases for core NEC phosphorylation. Importantly, mass spectrometry-based proteomic analyses identified five major and nine minor sites of pUL50 phosphorylation. The functional relevance of core NEC phosphorylation was confirmed by various experimental settings, including kinase knock-down/knock-out and confocal imaging, in which it was found that (i) HCMV core NEC proteins are not phosphorylated solely by viral pUL97, but also by cellular kinases; (ii) both PKC and CDK1 phosphorylation are detectable for pUL50; (iii) no impact of PKC phosphorylation on NEC functionality has been identified so far; (iv) nonetheless, CDK1-specific phosphorylation appears to be required for functional core NEC interaction. In summary, our findings provide the first evidence that the HCMV core NEC is phosphorylated by cellular kinases, and that the complex pattern of NEC phosphorylation has functional relevance.

  3. The density jump at the inner core boundary using underground nuclear explosion records

    International Nuclear Information System (INIS)

    Krasnoshchekov, D.N.; Ovchinnikov, V.M.

    2001-01-01

    This paper presents the estimation of the minimum jump value using experimental wave forms reflected from the boundary between the Earth core and mantle (PcP) and the one between the inner and outer core (PKiKP) at a distance of 6 deg. Digital seismic records of underground nuclear tests conducted at the Semipalatinsk test site in 70s by Zerenda-Vostochny-Chkalovo seismic array have been used. (author)

  4. Impact of correlations between core configurations for the evaluation of nuclear data uncertainty propagation for reactivity

    International Nuclear Information System (INIS)

    Frosio, T.; Bonaccorsi, T.; Blaise, P.

    2017-01-01

    The precise estimation of Pearson correlations, also called 'representativity' coefficients, between core configurations is a fundamental quantity for properly assessing the nuclear data (ND) uncertainties propagation on integral parameters such as k-eff, power distributions, or reactivity coefficients. In this paper, a traditional adjoint method is used to propagate ND uncertainty on reactivity and reactivity coefficients and estimate correlations between different states of the core. We show that neglecting those correlations induces a loss of information in the final uncertainty. We also show that using approximate values of Pearson does not lead to an important error of the model. This calculation is made for reactivity at the beginning of life and can be extended to other parameters during depletion calculations. (authors)

  5. Nuclear piston engine and pulsed gaseous core reactor power systems

    International Nuclear Information System (INIS)

    Dugan, E.T.

    1976-01-01

    The investigated nuclear piston engines consist of a pulsed, gaseous core reactor enclosed by a moderating-reflecting cylinder and piston assembly and operate on a thermodynamic cycle similar to the internal combustion engine. The primary working fluid is a mixture of uranium hexafluoride, UF 6 , and helium, He, gases. Highly enriched UF 6 gas is the reactor fuel. The helium is added to enhance the thermodynamic and heat transfer characteristics of the primary working fluid and also to provide a neutron flux flattening effect in the cylindrical core. Two and four-stroke engines have been studied in which a neutron source is the counterpart of the sparkplug in the internal combustion engine. The piston motions which have been investigated include pure simple harmonic, simple harmonic with dwell periods, and simple harmonic in combination with non-simple harmonic motion. The results of the conducted investigations indicate good performance potential for the nuclear piston engine with overall efficiencies of as high as 50 percent for nuclear piston engine power generating units of from 10 to 50 Mw(e) capacity. Larger plants can be conceptually designed by increasing the number of pistons, with the mechanical complexity and physical size as the probable limiting factors. The primary uses for such power systems would be for small mobile and fixed ground-based power generation (especially for peaking units for electrical utilities) and also for nautical propulsion and ship power

  6. Interface between Core/TH Model and Simulator for OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Do Hyun; Lee, Myeong Soo; Hong, Jin Hyuk; Lee, Seung Ho; Suh, Jung Kwan [KEPRI, Daejeon (Korea, Republic of)

    2009-05-15

    OPR1000 simulator for ShinKori-Unit 1, which will be operated at 2815MWt of thermal core power, is being developed while the ShinKori-Unit 1 and 2 is being built. OPR1000 simulator adopted the RELAP5 R/T code, which is the adaptation of RELAP5 and NESTLE codes to run in real-time mode with graphical visualization, to model Nuclear Steam Supply System (NSSS) Thermal-Hydraulics (TH) and Reactor Core. The RELAP5 is an advanced, best estimate, reactor TH simulation code developed at Idaho National Engineering and Environment Laboratory(INEEL) and the NESTLE is a true two-energy group neutronics code that computes the neutron flux and power for each node at every time step. As a simulator environment, the 3KEYMASTER{sup TM}, a commercial environment tool of WSC is used.

  7. Interface between Core/TH Model and Simulator for OPR1000

    International Nuclear Information System (INIS)

    Hwang, Do Hyun; Lee, Myeong Soo; Hong, Jin Hyuk; Lee, Seung Ho; Suh, Jung Kwan

    2009-01-01

    OPR1000 simulator for ShinKori-Unit 1, which will be operated at 2815MWt of thermal core power, is being developed while the ShinKori-Unit 1 and 2 is being built. OPR1000 simulator adopted the RELAP5 R/T code, which is the adaptation of RELAP5 and NESTLE codes to run in real-time mode with graphical visualization, to model Nuclear Steam Supply System (NSSS) Thermal-Hydraulics (TH) and Reactor Core. The RELAP5 is an advanced, best estimate, reactor TH simulation code developed at Idaho National Engineering and Environment Laboratory(INEEL) and the NESTLE is a true two-energy group neutronics code that computes the neutron flux and power for each node at every time step. As a simulator environment, the 3KEYMASTER TM , a commercial environment tool of WSC is used

  8. Modelling of the Molten Core Concrete Interaction (MCCI)

    International Nuclear Information System (INIS)

    Guillaume, M.

    2008-01-01

    Severe accidents of nuclear power plants are very unlikely to occur, yet it is necessary to be able to predict the evolution of the accident. In some situations, heat generation due to the disintegration of fission products could lead to the melting of the core. If the molten core falls on the floor of the building, it would provoke the melting of the concrete floor. The objective of the studies is to calculate the melting rate of the concrete floor. The work presented in this report is in the continuity of the segregation phase model of Seiler and Froment. It is based on the results of the ARTEMIS experiments. Firstly, we have developed a new model to simulate the transfers within the interfacial area. The new model explains how heat is transmitted to concrete: by conduction, convection and latent heat generation. Secondly, we have modified the coupled modelling of the pool and the interfacial area. We have developed two new models: the first one is the 'liquidus model', whose main hypothesis is that there is no resistance to solute transfer between the pool and the interfacial area. The second one is 'the thermal resistance model', whose main hypothesis is that there is no solute transfer and no dissolution of the interfacial area. The second model is able to predict the evolution of the pool temperature and the melting rate in the tests 3 and 4, with the condition that the obstruction time of the interfacial area is about 10 5 s. The model is not able to explain precisely the origin of this value. The liquidus model is able to predict correctly the evolution of the pool temperature and the melting rate in the tests 2 and 6. (author) [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

    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)

  11. Nuclear physics: the core of matter, the fuel of stars

    International Nuclear Information System (INIS)

    Schiffer, J.P.

    1999-01-01

    Dramatic progress has been made in all branches of physics since the National Research Council's 1986 decadal survey of the field. The Physics in a New Era series explores these advances and looks ahead to future goals. The series includes assessments of the major subfields and reports on several smaller subfields, and preparation has begun on an overview volume on the unity of physics, its relationships to other fields, and its contributions to national needs. Nuclear Physics is the latest volume of the series. The book describes current activity in understanding nuclear structure and symmetries, the behavior of matter at extreme densities, the role of nuclear physics in astrophysics and cosmology, and the instrumentation and facilities used by the field. It makes recommendations on the resources needed for experimental and theoretical advances in the coming decade. Nuclear physics addresses the nature of matter making up 99.9 percent of the mass of our everyday world. It explores the nuclear reactions that fuel the stars, including our Sun, which provides the energy for all life on Earth. The field of nuclear physics encompasses some 3,000 experimental and theoretical researchers who work at universities and national laboratories across the United States, as well as the experimental facilities and infrastructure that allow these researchers to address the outstanding scientific questions facing us. This report provides an overview of the frontiers of nuclear physics as we enter the next millennium, with special attention to the state of the science in the United States.The current frontiers of nuclear physics involve fundamental and rapidly evolving issues. One is understanding the structure and behavior of strongly interacting matter in terms of its basic constituents, quarks and gluons, over a wide range of conditions - from normal nuclear matter to the dense cores of neutron stars, and to the Big Bang that was the birth of the universe. Another is to describe

  12. Uncertainly propagation analysis for Yonggwang nuclear unit 4 by McCARD/MASTER core analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Dong Hyuk; Shim, Hyung Jin; Kim, Chang Hyo [Seoul National University, Seoul (Korea, Republic of)

    2014-06-15

    This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor (k{sub eff}), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

  13. Baryon-Baryon Interactions ---Nijmegen Extended-Soft-Core Models---

    Science.gov (United States)

    Rijken, T. A.; Nagels, M. M.; Yamamoto, Y.

    We review the Nijmegen extended-soft-core (ESC) models for the baryon-baryon (BB) interactions of the SU(3) flavor-octet of baryons (N, Lambda, Sigma, and Xi). The interactions are basically studied from the meson-exchange point of view, in the spirit of the Yukawa-approach to the nuclear force problem [H. Yukawa, ``On the interaction of Elementary Particles I'', Proceedings of the Physico-Mathematical Society of Japan 17 (1935), 48], using generalized soft-core Yukawa-functions. These interactions are supplemented with (i) multiple-gluon-exchange, and (ii) structural effects due to the quark-core of the baryons. We present in some detail the most recent extended-soft-core model, henceforth referred to as ESC08, which is the most complete, sophisticated, and successful interaction-model. Furthermore, we discuss briefly its predecessor the ESC04-model [Th. A. Rijken and Y. Yamamoto, Phys. Rev. C 73 (2006), 044007; Th. A. Rijken and Y. Yamamoto, Ph ys. Rev. C 73 (2006), 044008; Th. A. Rijken and Y. Yamamoto, nucl-th/0608074]. For the soft-core one-boson-exchange (OBE) models we refer to the literature [Th. A. Rijken, in Proceedings of the International Conference on Few-Body Problems in Nuclear and Particle Physics, Quebec, 1974, ed. R. J. Slobodrian, B. Cuec and R. Ramavataram (Presses Universitè Laval, Quebec, 1975), p. 136; Th. A. Rijken, Ph. D. thesis, University of Nijmegen, 1975; M. M. Nagels, Th. A. Rijken and J. J. de Swart, Phys. Rev. D 17 (1978), 768; P. M. M. Maessen, Th. A. Rijken and J. J. de Swart, Phys. Rev. C 40 (1989), 2226; Th. A. Rijken, V. G. J. Stoks and Y. Yamamoto, Phys. Rev. C 59 (1999), 21; V. G. J. Stoks and Th. A. Rijken, Phys. Rev. C 59 (1999), 3009]. All ingredients of these latter models are also part of ESC08, and so a description of ESC08 comprises all models so far in principle. The extended-soft-core (ESC) interactions consist of local- and non-local-potentials due to (i) one-boson-exchanges (OBE), which are the members of nonets of

  14. Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

    Science.gov (United States)

    Huddar, Lakshana Ravindranath

    pebble-bed test section was designed, constructed and performed. Oil was pumped through a test section filled with randomly packed copper spheres. The temperature of the oil was pulsed at a constant frequency, which caused a temperature difference between the pebbles and the oil. An excellent match was found between the measured heat transfer coefficients and the literature. This data provides an essential closure parameter for multiphysics modeling of the PB-FHR. Using frequency response techniques in scaled experiments is an innovative approach for extracting dynamic responses to coolant-structure interactions. Finally, an integrated model of the passive decay heat removal system was presented using Flownex and the simulations compared to experimental data. A good match was found with the data, which was within 14%. The work presented in this dissertation shows fundamental details on heat transfer in the PB-FHR core using experimental data and simulations, leading us closer to developing advanced nuclear reactors that can later be commercialized. Advanced nuclear reactors such as the PB-FHR have immense potential in reducing greenhouse gas emissions and combating climate change while being exceedingly safe and providing reliable electricity.

  15. Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

    International Nuclear Information System (INIS)

    Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

    2001-01-01

    The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

  16. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

  17. Multiregional coupled conduction--convection model for heat transfer in an HTGR core

    International Nuclear Information System (INIS)

    Giles, G.E. Jr.; Childs, K.W.; Sanders, J.P.

    1978-01-01

    HEXEREI is a three-dimensional, coupled conduction-convection heat transfer and multichannel fluid dynamic analysis computer code with both steady-state and transient capabilities. The program was developed to provide thermal-fluid dynamic analysis of a core following the general design for high-temperature gas-cooled reactors (HTGRs); its purpose was to provide licensing evaluations for the U.S. Nuclear Regulatory Commission. In order to efficiently model the HTGR core, the nodal geometry of HEXEREI was chosen as a regular hexagonal array perpendicular to the axis of and bounded by a right circular cylinder. The cylindrical nodal geometry surrounds the hexagonal center portion of the mesh; these two different types of nodal geometries must be connected by interface nodes to complete the accurate modeling of the HTGR core. HEXEREI will automatically generate a nodal geometry that will accurately model a complex assembly of hexagonal and irregular prisms. The accuracy of the model was proven by a comparison of computed values with analytical results for steady-state and transient heat transfer problems. HEXEREI incorporates convective heat transfer to the coolant in many parallel axial flow channels. Forced and natural convection (which permits different flow directions in parallel channels) is included in the heat transfer and fluid dynamic models. HEXEREI incorporates a variety of steady-state and transient solution techniques that can be matched with a particular problem to minimize the computational time. HEXEREI was compared with a code of similar capabilities that was based on a Cartesian mesh. This code modeled only one specific core design, and the mesh spacing was closer than that generated by HEXEREI. Good agreement was obtained with the detail provided by the representations

  18. Construction and utilization of linear empirical core models for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Okafor, K.C.

    1988-01-01

    An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k ∞ profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results

  19. Models of the earth's core

    International Nuclear Information System (INIS)

    Stevenson, D.J.

    1981-01-01

    The combination of seismology, high pressure experiment and theory, geomagnetism, fluid dynamics, and current views of terrestrial planetary evolution lead to strong constraints on core models. The synthesis presented here is devoted to the defense of the following properties: (1) core formation was contemporaneous with earth accretion; (2) the outer, liquid core is predominately iron but cannot be purely iron; (3) the inner core-outer core boundary represents a thermodynamic equilibrium between a liquid alloys and a predominanately iron solid; (4) thermodynamic and transport properties of outer core can be estimated from liquid-state theories; and (5) the outer core is probably adiabatic and uniform in composition. None of these propositions are universally accepted by geophysicists. But, the intent of this paper is to present a coherent picture which explains most of the data with the fewest ad hoc assumptions. Areas in which future progress is both essential and likely are geo- and cosmochronology, seismological determinations of core structure, fluid dynamics of the core and mantle, and condensed matter physics

  20. Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2011-01-01

    Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

  1. Introduction of virtual detectors for core monitoring system of korean standard nuclear power plant

    International Nuclear Information System (INIS)

    Eun, Ki Lee.; Yong, Hee Kim.; Jybe, Ho Cha.; Moon, Ghu Park.

    2000-01-01

    A novel algorithm known as the virtual detector method (VDM) is introduced to reconstruct the axial power shape (APS) for the on-line core monitoring system of the Korean Standard Nuclear Power Plant (KSNP). A pure statistical method (SM) is also introduced and the results are compared with the currently implemented five-mode Fourier fitting method (FFM). VDM adopts nine virtual detector informations coupled with a regression model based on the Alternating Conditional Expectation (ACE) algorithm. VDM uses Fourier fitting with the information of nine virtual detectors expanded from the currently implemented FFM, which uses five-level detector information. By introducing virtual detectors, we can increase the number of axial detectors, and thus expect the computational errors of APS to be reduced. The two methods (SM and VDM) are applied to in-core mapping data from six cycles of Yong Gwang nuclear power plant Units 3 and 4. For ∼ 3500 cases of APSs extracted from a cycle of operation which is simulated by a three-dimensional nodal code, the accuracy of the three methods (SM, VDM, FFM) is compared. The average root mean square (RMS) error and average of axial peaking error of SM and VDM resulted in reduction of more than 50 % and 70 %, respectively, relative to FFM. VDM and SM also show more realistic axial profiles and predict more accurate axial peaking than FFM. These improvements can contribute to a larger thermal margin. SM shows the most accurate results for all cases. VDM can almost obtain the same results as SM, and using far fewer computation steps. VDM can be a useful tool for precisely reconstructing axial power shapes in a core monitoring system. (authors)

  2. Development of a parallel genetic algorithm using MPI and its application in a nuclear reactor core. Design optimization

    International Nuclear Information System (INIS)

    Waintraub, Marcel; Pereira, Claudio M.N.A.; Baptista, Rafael P.

    2005-01-01

    This work presents the development of a distributed parallel genetic algorithm applied to a nuclear reactor core design optimization. In the implementation of the parallelism, a 'Message Passing Interface' (MPI) library, standard for parallel computation in distributed memory platforms, has been used. Another important characteristic of MPI is its portability for various architectures. The main objectives of this paper are: validation of the results obtained by the application of this algorithm in a nuclear reactor core optimization problem, through comparisons with previous results presented by Pereira et al.; and performance test of the Brazilian Nuclear Engineering Institute (IEN) cluster in reactors physics optimization problems. The experiments demonstrated that the developed parallel genetic algorithm using the MPI library presented significant gains in the obtained results and an accentuated reduction of the processing time. Such results ratify the use of the parallel genetic algorithms for the solution of nuclear reactor core optimization problems. (author)

  3. Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

    Science.gov (United States)

    Kazeminezhad, F.; Anghaie, S.

    2008-01-01

    Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

  4. Solid charged-core model of ball lightning

    Science.gov (United States)

    Muldrew, D. B.

    2010-01-01

    In this study, ball lightning (BL) is assumed to have a solid, positively-charged core. According to this underlying assumption, the core is surrounded by a thin electron layer with a charge nearly equal in magnitude to that of the core. A vacuum exists between the core and the electron layer containing an intense electromagnetic (EM) field which is reflected and guided by the electron layer. The microwave EM field applies a ponderomotive force (radiation pressure) to the electrons preventing them from falling into the core. The energetic electrons ionize the air next to the electron layer forming a neutral plasma layer. The electric-field distributions and their associated frequencies in the ball are determined by applying boundary conditions to a differential equation given by Stratton (1941). It is then shown that the electron and plasma layers are sufficiently thick and dense to completely trap and guide the EM field. This model of BL is exceptional in that it can explain all or nearly all of the peculiar characteristics of BL. The ES energy associated with the core charge can be extremely large which can explain the observations that occasionally BL contains enormous energy. The mass of the core prevents the BL from rising like a helium-filled balloon - a problem with most plasma and burning-gas models. The positively charged core keeps the negatively charged electron layer from diffusing away, i.e. it holds the ball together; other models do not have a mechanism to do this. The high electrical charges on the core and in the electron layer explains why some people have been electrocuted by BL. Experiments indicate that BL radiates microwaves upon exploding and this is consistent with the model. The fact that this novel model of BL can explain these and other observations is strong evidence that the model should be taken seriously.

  5. NCS--a software for visual modeling and simulation of PWR nuclear power plant control system

    International Nuclear Information System (INIS)

    Cui Zhenhua

    1998-12-01

    The modeling and simulation of nuclear power plant control system has been investigated. Some mathematical models for rapid and accurate simulation are derived, including core models, pressurizer model, steam generator model, etc. Several numerical methods such as Runge-Kutta Method and Treanor Method are adopted to solve the above system models. In order to model the control system conveniently, a block diagram-oriented visual modeling platform is designed. And the Discrete Similarity Method is used to calculate the control system models. A corresponding simulating software, NCS, is developed for researching on the control systems of commercial nuclear power plant. And some satisfactory results are obtained. The research works will be of referential and applying value to design and analysis of nuclear power plant control system

  6. Progress and problems in modelling HTR core dynamics

    International Nuclear Information System (INIS)

    Scherer, W.; Gerwin, H.

    1991-01-01

    In recent years greater effort has been made to establish theoretical models for HTR core dynamics. At KFA Juelich the TINTE (TIme dependent Neutronics and TEmperatures) code system has been developed, which is able to model the primary circuit of an HTR plant using modern numerical techniques and taking into account the mutual interference of the relevant physical variables. The HTR core is treated in 2-D R-Z geometry for both nucleonics and thermo-fluid-dynamics. 2-energy-group diffusion theory is used in the nuclear part including 6 groups of delayed neutron precursors and 14 groups of decay heat producers. Local and non-local heat sources are incorporated, thus simulating gamma ray transport. The thermo-fluid-dynamics module accounts for heterogeneity effects due to the pebble bed structure. Pipes and other components of the primary loop are modelled in 1-D geometry. Forced convection may be treated as well as natural convection in case of blower breakdown accidents. Validation of TINTE has started using the results of a comprehensive experimental program that has been performed at the Arbeitsgemeinschaft Versuchsreaktor GmbH (AVR) high temperature pebble bed reactor at Juelich. In the frame of this program power transients were initiated by varying the helium blower rotational speed or by moving the control rods. In most cases a good accordance between experiment and calculation was found. Problems in modelling the special AVR reactor geometry in two dimensions are described and suggestions for overcoming the uncertainties of experimentally determined control rod reactivities are given. The influence of different polynomial expansions of xenon cross sections to long term transients is discussed together with effects of burnup during that time. Up to now the TINTE code has proven its general applicability to operational core transients of HTR. The effects of water ingress on reactivity, fuel element corrosion and cooling gas properties are now being

  7. One dimensional reactor core model

    International Nuclear Information System (INIS)

    Kostadinov, V.; Stritar, A.; Radovo, M.; Mavko, B.

    1984-01-01

    The one dimensional model of neutron dynamic in reactor core was developed. The core was divided in several axial nodes. The one group neutron diffusion equation for each node is solved. Feedback affects of fuel and water temperatures is calculated. The influence of xenon, boron and control rods is included in cross section calculations for each node. The system of equations is solved implicitly. The model is used in basic principle Training Simulator of NPP Krsko. (author)

  8. Out-of-core nuclear fuel cycle optimization utilizing an engineering workstation

    International Nuclear Information System (INIS)

    Turinsky, P.J.; Comes, S.A.

    1986-01-01

    Within the past several years, rapid advances in computer technology have resulted in substantial increases in their performance. The net effect is that problems that could previously only be executed on mainframe computers can now be executed on micro- and minicomputers. The authors are interested in developing an engineering workstation for nuclear fuel management applications. An engineering workstation is defined as a microcomputer with enhanced graphics and communication capabilities. Current fuel management applications range from using workstations as front-end/back-end processors for mainframe computers to completing fuel management scoping calculations. More recently, interest in using workstations for final in-core design calculations has appeared. The authors have used the VAX 11/750 minicomputer, which is not truly an engineering workstation but has comparable performance, to complete both in-core and out-of-core fuel management scoping studies. In this paper, the authors concentrate on our out-of-core research. While much previous work in this area has dealt with decisions concerned with equilibrium cycles, the current project addresses the more realistic situation of nonequilibrium cycles

  9. Evaluation of nuclear power plant component failure probability and core damage probability using simplified PSA model

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2000-01-01

    It is anticipated that the change of frequency of surveillance tests, preventive maintenance or parts replacement of safety related components may cause the change of component failure probability and result in the change of core damage probability. It is also anticipated that the change is different depending on the initiating event frequency or the component types. This study assessed the change of core damage probability using simplified PSA model capable of calculating core damage probability in a short time period, which is developed by the US NRC to process accident sequence precursors, when various component's failure probability is changed between 0 and 1, or Japanese or American initiating event frequency data are used. As a result of the analysis, (1) It was clarified that frequency of surveillance test, preventive maintenance or parts replacement of motor driven pumps (high pressure injection pumps, residual heat removal pumps, auxiliary feedwater pumps) should be carefully changed, since the core damage probability's change is large, when the base failure probability changes toward increasing direction. (2) Core damage probability change is insensitive to surveillance test frequency change, since the core damage probability change is small, when motor operated valves and turbine driven auxiliary feed water pump failure probability changes around one figure. (3) Core damage probability change is small, when Japanese failure probability data are applied to emergency diesel generator, even if failure probability changes one figure from the base value. On the other hand, when American failure probability data is applied, core damage probability increase is large, even if failure probability changes toward increasing direction. Therefore, when Japanese failure probability data is applied, core damage probability change is insensitive to surveillance tests frequency change etc. (author)

  10. Nuclear Statistical Equilibrium for compact stars: modelling the nuclear energy functional

    International Nuclear Information System (INIS)

    Aymard, Francois

    2015-01-01

    The core collapse supernova is one of the most powerful known phenomena in the universe. It results from the explosion of very massive stars after they have burnt all their fuel. The hot compact remnant, the so-called proto-neutron star, cools down to become an inert catalyzed neutron star. The dynamics and structure of compact stars, that is core collapse supernovae, proto-neutron stars and neutron stars, are still not fully understood and are currently under active research, in association with astrophysical observations and nuclear experiments. One of the key components for modelling compact stars concerns the Equation of State. The task of computing a complete realistic consistent Equation of State for all such stars is challenging because a wide range of densities, proton fractions and temperatures is spanned. This thesis deals with the microscopic modelling of the structure and internal composition of baryonic matter with nucleonic degrees of freedom in compact stars, in order to obtain a realistic unified Equation of State. In particular, we are interested in a formalism which can be applied both at sub-saturation and super-saturation densities, and which gives in the zero temperature limit results compatible with the microscopic Hartree-Fock-Bogoliubov theory with modern realistic effective interactions constrained on experimental nuclear data. For this purpose, we present, for sub-saturated matter, a Nuclear Statistical Equilibrium model which corresponds to a statistical superposition of finite configurations, the so-called Wigner-Seitz cells. Each cell contains a nucleus, or cluster, embedded in a homogeneous electron gas as well as a homogeneous neutron and proton gas. Within each cell, we investigate the different components of the nuclear energy of clusters in interaction with gases. The use of the nuclear mean-field theory for the description of both the clusters and the nucleon gas allows a theoretical consistency with the treatment at saturation

  11. 77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2012-05-23

    ... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [Docket No. PRM-50-105; NRC-2012-0056] In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission... of operating licenses for nuclear power plants (``NPP'') to operate NPPs with in-core thermocouples...

  12. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

    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.

  13. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    Nagano, M.; Sakurai, S.; Yamaguchi, H.

    1997-01-01

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  14. A porous medium approach for the fluid structure interaction modelling of a water pressurized nuclear reactor core fuel assemblies: simulation and experimentation

    International Nuclear Information System (INIS)

    Ricciardi, G.

    2008-10-01

    The designing of a pressurized water reactor core subjected to seismic loading, is a major concern of the nuclear industry. We propose, in this PhD report, to establish the global behaviour equations of the core, in term of a porous medium. Local equations of fluid and structure are space averaged on a control volume, thus we define an equivalent fluid and an equivalent structure, of which unknowns are defined on the whole space. The non-linear fuel assemblies behaviour is modelled by a visco-elastic constitutive law. The fluid-structure coupling is accounted for by a body force, the expression of that force is based on empirical formula of fluid forces acting on a tube subject to an axial flow. The resulting equations are solved using a finite element method. A validation of the model, on three experimental device, is proposed. The first one presents two fuel assemblies subjected to axial flow. One of the two fuel assemblies is deviated from its position of equilibrium and released, while the other is at rest. The second one presents a six assemblies row, immersed in water, placed on a shaking table that can simulate seismic loading. Finally, the last one presents nine fuel assemblies network, arranged in a three by three, subject to an axial flow. The displacement of the central fuel assembly is imposed. The simulations are in agreement with the experiments, the model reproduces the influence of the flow of fluid on the dynamics and coupling of the fuel assemblies. (author)

  15. First in-core simultaneous measurements of nuclear heating and thermal neutron flux obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lepeltier, Valerie; Bubendorff, Jacques; Carcreff, Hubert [Nuclear studies and reactor irradiation Service, CEA Saclay 91191 Gif sur Yvette (France); Salmon, Laurent [Thermalhydraulics and Fluid Mechanics Section, CEA Saclay 91191 Gif sur Yvette, (France)

    2015-07-01

    Nuclear heating inside a MTR reactor has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. This development required preliminary modelling and irradiation of mock-ups of the calorimetric probe in the ex-core area, where nuclear heating rate does not exceed 2 W.g{sup -1}. The calorimeter working modes, the different measurement procedures allowed with such a new probe, the main modeling and experimental results and expected advantages of this new technique have been already presented. However, these first in-core measurements were not performed beyond 6 W.g{sup -1}, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at the 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 10{sup 14} n.cm{sup -2}.s{sup -1} and nuclear heating up to 12 W.g{sup -1}. A comprehensive measurement campaign carried out from 2013 to 2015 inside all accessible irradiation locations of the core, allowed to qualify definitively this new device, not only in terms of measurement ability but also in terms of reliability. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a systematic comparison is made between results obtained by

  16. The APR1400 Core Design by Using APA Code System

    International Nuclear Information System (INIS)

    Choi, Yu Sun; Koh, Byung Marn

    2008-01-01

    The nuclear design for APR1400 has been performed to prepare the core model for Automatic Load Follow Operation Simulation. APA (ALPHA/ PHOENIXP/ ANC) code system is a tool for the multi-cycle depletion calculations for APR1400. Its detail versions for ALPHA, PHOENIX-P and ANC are 8.9.3, 8.6.1 and 8.10.5, respectively. The first and equilibrium core depletion calculations for APR1400 have been performed to assure the target cycle length and confirm the safety parameters. The parameters are satisfied within limitation about nuclear design criteria. This APR1400 core models will be based on the design parameters for APR1400 Simulator

  17. Modeling of molten core-concrete interactions and fission-product release

    International Nuclear Information System (INIS)

    Norkus, J.K.; Corradini, M.L.

    1991-09-01

    The study of molten core-concrete interaction is important in estimating the possible consequences of a severe nuclear reactor accident. CORCON-Mod2 is a computer program which models the thermal, chemical, and physical phenomena associated with molten core-concrete interactions. Models have been added to extend and improve the modeling of these phenomena. An ideal solution chemical equilibrium methodology is presented to predict the fission-product vaporization release. Additional chemical species have been added, and the calculation of chemical equilibrium has been expanded to the oxidic layer and to the mixed layer configuration. Recent experiments performed at Argonne National Laboratory are compared to CORCON predictions of melt temperature, erosion depth, and release fraction of fission products. The results consistently underpredicted the melt temperatures and erosion rates. However, the predictions of release of Te, Ba, Sr, and U were good. A sensitivity study of the effects of initial temperature, concrete type, use of the mixing option, degree of zirconium oxidation, cavity size, and amount of control material on erosion, gas production, and release of radioactive materials was performed for a PWR and a BWR. The initial melt temperature had the greatest effect on the results of interest. Concrete type and cavity size also had important effects. 78 refs., 35 figs., 40 tabs

  18. Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

    International Nuclear Information System (INIS)

    Tanabe, Fumiya

    2011-01-01

    In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

  19. Geodynamo Modeling of Core-Mantle Interactions

    Science.gov (United States)

    Kuang, Wei-Jia; Chao, Benjamin F.; Smith, David E. (Technical Monitor)

    2001-01-01

    Angular momentum exchange between the Earth's mantle and core influences the Earth's rotation on time scales of decades and longer, in particular in the length of day (LOD) which have been measured with progressively increasing accuracy for the last two centuries. There are four possible coupling mechanisms for transferring the axial angular momentum across the core-mantle boundary (CMB): viscous, magnetic, topography, and gravitational torques. Here we use our scalable, modularized, fully dynamic geodynamo model for the core to assess the importance of these torques. This numerical model, as an extension of the Kuang-Bloxham model that has successfully simulated the generation of the Earth's magnetic field, is used to obtain numerical results in various physical conditions in terms of specific parameterization consistent with the dynamical processes in the fluid outer core. The results show that depending on the electrical conductivity of the lower mantle and the amplitude of the boundary topography at CMB, both magnetic and topographic couplings can contribute significantly to the angular momentum exchange. This implies that the core-mantle interactions are far more complex than has been assumed and that there is unlikely a single dominant coupling mechanism for the observed decadal LOD variation.

  20. On Input Vector Representation for the SVR model of Reactor Core Loading Pattern Critical Parameters

    International Nuclear Information System (INIS)

    Trontl, K.; Pevec, D.; Smuc, T.

    2008-01-01

    Determination and optimization of reactor core loading pattern is an important factor in nuclear power plant operation. The goal is to minimize the amount of enriched uranium (fresh fuel) and burnable absorbers placed in the core, while maintaining nuclear power plant operational and safety characteristics. The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. Recently, we proposed a new method for fast loading pattern evaluation based on general robust regression model relying on the state of the art research in the field of machine learning. We employed Support Vector Regression (SVR) technique. SVR is a supervised learning method in which model parameters are automatically determined by solving a quadratic optimization problem. The preliminary tests revealed a good potential of the SVR method application for fast and accurate reactor core loading pattern evaluation. However, some aspects of model development are still unresolved. The main objective of the work reported in this paper was to conduct additional tests and analyses required for full clarification of the SVR applicability for loading pattern evaluation. We focused our attention on the parameters defining input vector, primarily its structure and complexity, and parameters defining kernel functions. All the tests were conducted on the NPP Krsko reactor core, using MCRAC code for the calculation of reactor core loading pattern critical parameters. The tested input vector structures did not influence the accuracy of the models suggesting that the initially tested input vector, consisted of the number of IFBAs and the k-inf at the beginning of the cycle, is adequate. The influence of kernel function specific parameters (σ for RBF kernel

  1. Coupling of the 3D neutron kinetic core model DYN3D with the CFD software ANSYS-CFX

    International Nuclear Information System (INIS)

    Grahn, Alexander; Kliem, Sören; Rohde, Ulrich

    2015-01-01

    Highlights: • Improved thermal hydraulic description of nuclear reactor cores. • Possibility of three-dimensional flow phenomena in the core, such as cross flow, flow reversal, flow around obstacles. • Simulation at higher spatial resolution as compared to system codes. - Abstract: This article presents the implementation of a coupling between the 3D neutron kinetic core model DYN3D and the commercial, general purpose computational fluid dynamics (CFD) software ANSYS-CFX. In the coupling approach, parts of the thermal hydraulic calculation are transferred to CFX for its better ability to simulate the three-dimensional coolant redistribution in the reactor core region. The calculation of the heat transfer from the fuel into the coolant remains with DYN3D, which incorporates well tested and validated heat transfer models for rod-type fuel elements. On the CFX side, the core region is modeled based on the porous body approach. The implementation of the code coupling is verified by comparing test case results with reference solutions of the DYN3D standalone version. Test cases cover mini and full core geometries, control rod movement and partial overcooling transients

  2. Burst shield for a pressurized nuclear-reactor core and method of operating same

    International Nuclear Information System (INIS)

    Beine, B.; Schilling, F.

    1976-01-01

    A pressurized nuclear-reactor core stands on a base up from which extends a cylindrical side wall formed of a plurality of hollow iron castings held together by circumferential and longitudinal prestressed elements. A cylindrical space between this shield and the core serves for inspection of the core and is normally filled with cast-iron segmental slabs so that if the core bursts pieces thrown out do not acquire any dangerous kinetic energy before engaging the burst shield. The top of the shield is removably secured to the side so that it can be moved out of the way periodically for removal of the filler slabs and inspection of the core. An anchor on the upper end of each longitudinal prestressing element bears against a sleeve pressing against the uppermost side element, and a nut engageable with this anchor is engageable down over the top to hold it in place, removal of this nut leaving the element prestressed in the side wall. 11 claims, 16 drawing figures

  3. Development of NUFREQ-N, an analytical model for the stability analysis of nuclear coupled density-wave oscillations in boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Park, G.C.

    1983-01-01

    A state-of-the-art one-dimensional thermal-hydraulic model has been developed to be used for the linear analysis of nuclear-coupled density-wave oscillations in a boiling water nuclear reactor (BWR). The model accounts for phasic slip, distributed spacers, subcooled boiling, space/time-dependent power distributions and distributed heated wall dynamics. In addition to a parallel channel stability analysis, a detailed model was derived for the BWR loop analysis of both the natural and forced circulation modes of operation. In its final form, this model constitutes a multi-input, multi-output (MIMO) linear system, which features a general nodal neutron kinetics model. Kinetics parameters for use in the kinetics model have been obtained by utilizing self-consistent nodal data and power distributions. The stability characteristics of a typical BWR/4 has been investigated with the Nyquist criterion. The computer implementation of this mode, NUFREQ-N, was used for the parametric study of a typical BWR/4 and comparison were made with existing in-core and out-of-core data. Also, NUFREQ-N was used to analyze the expected stability characteristics of a typical BWR/4. The parametric results revealed important factors influencing BWR stability margin. It was found that NUFREQ-N generally agreed well with out-of-core data. This was especially true for the predicted power-to-flow transfer function, which is the most important transfer function in thermal-hydraulic stability analysis

  4. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    International Nuclear Information System (INIS)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun

    2016-01-01

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes

  5. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes.

  6. Possible generation of heat from nuclear fusion in Earth's inner core.

    Science.gov (United States)

    Fukuhara, Mikio

    2016-11-23

    The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2 D +  2 D +  2 D → 2 1 H +  4 He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10 12  J/m 3 , based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

  7. Core fusion accidents in nuclear power reactors. Knowledge review

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

  8. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    International Nuclear Information System (INIS)

    Nigg, David W.; Steuhm, Devin A.

    2011-01-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V and V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V and V, within the next 3-4 years via the ATR Core Modeling and Simulation and V and V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose

  9. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Devin A. Steuhm

    2011-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore

  10. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach; Simulacao termohidraulica do nucleo do reator nuclear HTR-10 com o uso da abordagem realistica CFD

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S.; Dominguez, Dany S., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil); Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas, La Habana (Cuba); Lira, Carlos Alberto Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  11. Nuclear spins, magnetic moments and quadrupole moments of Cu isotopes from N = 28 to N = 46: probes for core polarization effects

    CERN Document Server

    Vingerhoets, P; Avgoulea, M; Billowes, J; Bissell, M L; Blaum, K; Brown, B A; Cheal, B; De Rydt, M; Forest, D H; Geppert, Ch; Honma, M; Kowalska, M; Kramer, J; Krieger, A; Mane, E; Neugart, R; Neyens, G; Nortershauser, W; Otsuka, T; Schug, M; Stroke, H H; Tungate, G; Yordanov, D T

    2010-01-01

    Measurements of the ground-state nuclear spins, magnetic and quadrupole moments of the copper isotopes from 61Cu up to 75Cu are reported. The experiments were performed at the ISOLDE facility, using the technique of collinear laser spectroscopy. The trend in the magnetic moments between the N=28 and N=50 shell closures is reasonably reproduced by large-scale shell-model calculations starting from a 56Ni core. The quadrupole moments reveal a strong polarization of the underlying Ni core when the neutron shell is opened, which is however strongly reduced at N=40 due to the parity change between the $pf$ and $g$ orbits. No enhanced core polarization is seen beyond N=40. Deviations between measured and calculated moments are attributed to the softness of the 56Ni core and weakening of the Z=28 and N=28 shell gaps.

  12. Materials behaviour in PWRs core

    International Nuclear Information System (INIS)

    Barbu, A.; Massoud, J.P.

    2008-01-01

    Like in any industrial facility, the materials of PWR reactors are submitted to mechanical, thermal or chemical stresses during particularly long durations of operation: 40 years, and even 60 years. Materials closer to the nuclear fuel are submitted to intense bombardment of particles (mainly neutrons) coming from the nuclear reactions inside the core. In such conditions, the damages can be numerous and various: irradiation aging, thermal aging, friction wear, generalized corrosion, stress corrosion etc.. The understanding of the materials behaviour inside the cores of reactors in operation is a major concern for the nuclear industry and its long term forecast is a necessity. This article describes the main ways of materials degradation without and under irradiation, with the means used to foresee their behaviour using physics-based models. Content: 1 - structures, components and materials: structure materials, nuclear materials; 2 - main ways of degradation without irradiation: thermal aging, stress corrosion, wear; 3 - main ways of degradation under irradiation: microscopic damaging - point defects, dimensional alterations, evolution of mechanical characteristics under irradiation, irradiation-assisted stress corrosion cracking (IASCC), synergies; 4 - forecast of materials evolution under irradiation using physics-based models: primary damage - fast dynamics, primary damage annealing - slow kinetics microstructural evolution, impact of microstructural changes on the macroscopic behaviour, insight on modeling methods; 5 - materials change characterization techniques: microscopic techniques - direct defects observation, nuclear techniques using a particle beam, global measurements, mechanical characterizations; 6 - perspectives. (J.S.)

  13. First In-Core Simultaneous Measurements of Nuclear Heating and Thermal Neutron Flux Obtained With the Innovative Mobile Calorimeter CALMOS Inside the OSIRIS Reactor

    Science.gov (United States)

    Carcreff, Hubert; Salmon, Laurent; Bubendorff, Jacques; Lepeltier, Valérie

    2016-10-01

    Nuclear heating inside a MTR reactor has to be known in order to design and run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. Calorimeter working modes, measurement procedures, main modeling and experimental results and expected advantages of this new technique have been already presented in previous papers. However, these first in-core measurements were not performed beyond 6 W · g-1, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 1014 n · cm-2 · s-1 and nuclear heating up to 12 W · g-1. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a comparison is made between results obtained by the probe calibration coefficient and the zero methods. Thermal neutron flux evaluation from SPND signal processing required a specific TRIPOLI-4 Monte Carlo calculation which has been performed with the precise CALMOS cell geometry. In addition, the Finite Element model for temperatures map prediction inside the calorimetric cell has been upgraded with recent experimental data obtained up to 12 W · g-1. Finally, the experience feedback led us to improvement perspectives. A second device is

  14. Integrated core-edge-divertor modeling studies

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    An integrated calculation model for simulating the interaction of physics phenomena taking place in the plasma core, in the plasma edge and in the SOL and divertor of tokamaks has been developed and applied to study such interactions. The model synthesises a combination of numerical calculations (1) the power and particle balances for the core plasma, using empirical confinement scaling laws and taking into account radiation losses (2), the particle, momentum and power balances in the SOL and divertor, taking into account the effects of radiation and recycling neutrals, (3) the transport of feeling and recycling neutrals, explicitly representing divertor and pumping geometry, and (4) edge pedestal gradient scale lengths and widths, evaluation of theoretical predictions (5) confinement degradation due to thermal instabilities in the edge pedestals, (6) detachment and divertor MARFE onset, (7) core MARFE onsets leading to a H-L transition, and (8) radiative collapse leading to a disruption and evaluation of empirical fits (9) power thresholds for the L-H and H-L transitions and (10) the width of the edge pedestals. The various components of the calculation model are coupled and must be iterated to a self-consistent convergence. The model was developed over several years for the purpose of interpreting various edge phenomena observed in DIII-D experiments and thereby, to some extent, has been benchmarked against experiment. Because the model treats the interactions of various phenomena in the core, edge and divertor, yet is computationally efficient, it lends itself to the investigation of the effects of different choices of various edge plasma operating conditions on overall divertor and core plasma performance. Studies of the effect of feeling location and rate, divertor geometry, plasma shape, pumping and over 'edge parameters' on core plasma properties (line average density, confinement, density limit, etc.) have been performed for DIII-D model problems. A

  15. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  16. How did Fukushima-Dai-ichi core meltdown change the probability of nuclear accidents?

    International Nuclear Information System (INIS)

    Escobar Rangel, Lina; Leveque, Francois

    2012-10-01

    How to predict the probability of a nuclear accident using past observations? What increase in probability the Fukushima Dai-ichi event does entail? Many models and approaches can be used to answer these questions. Poisson regression as well as Bayesian updating are good candidates. However, they fail to address these issues properly because the independence assumption in which they are based on is violated. We propose a Poisson Exponentially Weighted Moving Average (PEWMA) based in a state-space time series approach to overcome this critical drawback. We find an increase in the risk of a core meltdown accident for the next year in the world by a factor of ten owing to the new major accident that took place in Japan in 2011. (authors)

  17. Evaluation of the need for stochastic optimization of out-of-core nuclear fuel management decisions

    International Nuclear Information System (INIS)

    Thomas, R.L. Jr.

    1989-01-01

    Work has been completed on utilizing mathematical optimization techniques to optimize out-of-core nuclear fuel management decisions. The objective of such optimization is to minimize the levelized fuel cycle cost over some planning horizon. Typical decision variables include feed enrichments and number of assemblies, burnable poison requirements, and burned fuel to reinsert for every cycle in the planning horizon. Engineering constraints imposed consist of such items as discharge burnup limits, maximum enrichment limit, and target cycle energy productions. Earlier the authors reported on the development of the OCEON code, which employs the integer Monte Carlo Programming method as the mathematical optimization method. The discharge burnpups, and feed enrichment and burnable poison requirements are evaluated, initially employing a linear reactivity core physics model and refined using a coarse mesh nodal model. The economic evaluation is completed using a modification of the CINCAS methodology. Interest now is to assess the need for stochastic optimization, which will account for cost components and cycle energy production uncertainties. The implication of the present studies is that stochastic optimization in regard to cost component uncertainties need not be completed since deterministic optimization will identify nearly the same family of near-optimum cycling schemes

  18. Review of the Shoreham Nuclear Power Station Probabilistic Risk Assessment: internal events and core damage frequency

    International Nuclear Information System (INIS)

    Ilberg, D.; Shiu, K.; Hanan, N.; Anavim, E.

    1985-11-01

    A review of the Probabilistic Risk Assessment of the Shoreham Nuclear Power Station was conducted with the broad objective of evaluating its risks in relation to those identified in the Reactor Safety Study (WASH-1400). The scope of the review was limited to the ''front end'' part, i.e., to the evaluation of the frequencies of states in which core damage may occur. Furthermore, the review considered only internally generated accidents, consistent with the scope of the PRA. The review included an assessment of the assumptions and methods used in the Shoreham study. It also encompassed a reevaluation of the main results within the scope and general methodological framework of the Shoreham PRA, including both qualitative and quantitative analyses of accident initiators, data bases, and accident sequences which result in initiation of core damage. Specific comparisons are given between the Shoreham study, the results of the present review, and the WASH-1400 BWR, for the core damage frequency. The effect of modeling uncertainties was considered by a limited sensitivity study so as to show how the results would change if other assumptions were made. This review provides an independently assessed point value estimate of core damage frequency and describes the major contributors, by frontline systems and by accident sequences. 17 figs., 81 tabs

  19. Development of conceptual nuclear design of 10MWt research reactor core

    International Nuclear Information System (INIS)

    Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

    2008-03-01

    KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

  20. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    Energy Technology Data Exchange (ETDEWEB)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  1. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  2. Nuclear reactor with several cores

    International Nuclear Information System (INIS)

    Swars, H.

    1977-01-01

    Several sodium-cooled cores in separate vessels with removable closures are placed in a common reactor tank. Each individual vessel is protected against the consequences of an accident in the relevant core. Maintenance devices and inlet and outlet pipes for the coolant are also arranged within the reactor tank. The individual vessels are all enclosed by coolant in a way that in case of emergency cooling or refuelling each core can be continued to be cooled by means of the coolant loops of the other cores. (HP) [de

  3. Shock absorber in combination with a nuclear reactor core structure

    International Nuclear Information System (INIS)

    Housman, J.J.

    1976-01-01

    This invention relates to the provision of shock absorbers for use in blind control rod passages of a nuclear reactor core structure which are not subject to degradation. The shock absorber elements are made of a porous brittle carbonaceous material, a porous brittle ceramic material, or a porous brittle refractory oxide and have a void volume of between 30% and 70% of the total volume of the element for energy absorption by fracturing due to impact loading by a control rod. (UK)

  4. Pairing gaps from nuclear mean-field models

    International Nuclear Information System (INIS)

    Bender, M.; Rutz, K.; Maruhn, J.A.

    2000-01-01

    We discuss the pairing gap, a measure for nuclear pairing correlations, in chains of spherical, semi-magic nuclei in the framework of self-consistent nuclear mean-field models. The equations for the conventional BCS model and the approximate projection-before-variation Lipkin-Nogami method are formulated in terms of local density functionals for the effective interaction. We calculate the Lipkin-Nogami corrections of both the mean-field energy and the pairing energy. Various definitions of the pairing gap are discussed as three-point, four-point and five-point mass-difference formulae, averaged matrix elements of the pairing potential, and single-quasiparticle energies. Experimental values for the pairing gap are compared with calculations employing both a delta pairing force and a density-dependent delta interaction in the BCS and Lipkin-Nogami model. Odd-mass nuclei are calculated in the spherical blocking approximation which neglects part of the the core polarization in the odd nucleus. We find that the five-point mass difference formula gives a very robust description of the odd-even staggering, other approximations for the gap may differ from that up to 30% for certain nuclei. (orig.)

  5. Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

    International Nuclear Information System (INIS)

    1977-09-01

    After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities [de

  6. Modelling characteristics of ferromagnetic cores with the influence of temperature

    International Nuclear Information System (INIS)

    Górecki, K; Rogalska, M; Zarȩbski, J; Detka, K

    2014-01-01

    The paper is devoted to modelling characteristics of ferromagnetic cores with the use of SPICE software. Some disadvantages of the selected literature models of such cores are discussed. A modified model of ferromagnetic cores taking into account the influence of temperature on the magnetizing characteristics and the core losses is proposed. The form of the elaborated model is presented and discussed. The correctness of this model is verified by comparing the calculated and the measured characteristics of the selected ferromagnetic cores.

  7. Code package {open_quotes}SVECHA{close_quotes}: Modeling of core degradation phenomena at severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others

    1995-09-01

    The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.

  8. A study on Monte Carlo analysis of Pebble-type VHTR core for hydrogen production

    International Nuclear Information System (INIS)

    Kim, Hong Chul

    2005-02-01

    In order to pursue exact the core analysis for VHTR core which will be developed in future, a study on Monte Carol method was carried out. In Korea, pebble and prism type core are under investigation for VHTR core analysis. In this study, pebble-type core was investigated because it was known that it should not only maintain the nuclear fuel integrity but also have the advantage in economical efficiency and safety. The pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model. After the detailed MCNP modeling of the whole facility, calculations of nuclear characteristics were performed. The two core configurations, Core 4.3 and Core 5 (reference state no. 3), among the 10 configurations of the HTR-PROTEUS cores were chosen to be analyzed in order to treat different fuel loading pattern and modeled. The former is a random packing core and the latter deterministic packing core. Based on the experimental data and the benchmark result of other research groups for the two different cores, some nuclear characteristics were calculated. Firstly, keff was calculated for these cores. The effect for TRIO homogeneity model was investigated. Control rod and shutdown rod worths also were calculated and the sensitivity analysis on cross-section library and reflector thickness was pursued. Lastly, neutron flux profiles were investigated in reflector regions. It is noted that Monte Carlo analysis of pebble-type VHTR core was firstly carried out in Korea. Also, this study should not only provide the basic data for pebble-type VHTR core analysis for hydrogen production but also be utilized as the verified data to validate a computer code for VHTR core analysis which will be developed in future

  9. Strategic plan for the development of core technologies for the Korean advanced nuclear power reactor for export

    International Nuclear Information System (INIS)

    Moon, Joo Hyun; Cho, Young Ho

    2010-01-01

    With the soaring oil price and worsening global warming, nuclear power has attracted considerable attention on a global scale and a new large market of nuclear power plants (NPPs) is expected. The Korean government aims to export up to 10 NPPs by 2012, based on the successful export of 2 NPPs to the UAE in 2009. It is also going to develop a follow-up model of the Advanced Power Reactor (APR) 1400, and join the world's NPP market under the banner of Korea's original reactor type. For this, it promulgated the strategic plan, NuTech 2012, a technology development plan intended for the early acquisition of core technologies for the Korean advanced NPP design and domestic production of the main components in NPP. This paper introduces the strategic plan of NuTech 2012. (orig.)

  10. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant

    International Nuclear Information System (INIS)

    Zamora R, L.; Medina F, A.

    1999-01-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

  11. APROS 3-D core models for simulators and plant analyzers

    International Nuclear Information System (INIS)

    Puska, E.K.

    1999-01-01

    The 3-D core models of APROS simulation environment can be used in simulator and plant analyzer applications, as well as in safety analysis. The key feature of APROS models is that the same physical models can be used in all applications. For three-dimensional reactor cores the APROS models cover both quadratic BWR and PWR cores and the hexagonal lattice VVER-type cores. In APROS environment the user can select the number of flow channels in the core and either five- or six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the channel description have a decisive effect on the calculation time of the 3-D core model and thus just these selection make at present the major difference between a safety analysis model and a training simulator model. The paper presents examples of various types of 3-D LWR-type core descriptions for simulator and plant analyzer use and discusses the differences of calculation speed and physical results between a typical safety analysis model description and a real-time simulator model description in transients. (author)

  12. In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept

    International Nuclear Information System (INIS)

    Gauntt, R.O.; Slutz, S.A.; Harms, G.A.; Latham, T.S.; Roman, W.C.; Rodgers, R.J.

    1993-01-01

    The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>1800 s) and thrust (>445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (∼4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept

  13. Evaluation of the Angra-2 nuclear power plant using a RELAP5-PARCS coupled model

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patrícia A.L.; Hamers, Adolfo R.; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria A.F.; Verdú, Gumersindo [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Miró, Rafael, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Departamento de Ingenieria Quimica y Nuclear, Universidad Politécnica de Valencia (Spain)

    2017-07-01

    Studies of complex transients in nuclear reactors have been improved by the use of thermal hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes to possibility simulation of transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronic and reactor thermal hydraulics. In this work, steady state results using the verified model of TH RELAP5 code and the NK PARCS code to predict the Angra 2 reactor behavior have been presented. (author)

  14. Methodology for Identification of the Coolant Thermalhydraulic Regimes in the Core of Nuclear Reactors

    International Nuclear Information System (INIS)

    Sharaevsky, L.G.; Sharaevskaya, E.I.; Domashev, E.D.; Arkhypov, A.P.; Kolochko, V.N.

    2002-01-01

    The paper deals with one of the acute for the nuclear energy problem of accident regimes of NPPs recognition diagnostics using noise signal diagnostics methodology. The methodology intends transformation of the random noise signals of the main technological parameters at the exit of a nuclear facility (neutron flow, dynamic pressure etc.) which contain the important information about the technical status of the equipment. The effective algorithms for identification of random processes wore developed. After proper transformation its were considered as multidimensional random vectors. Automatic classification of these vectors in the developed algorithms is realized on the basis of the probability function in particular Bayes classifier and decision functions. Till now there no mathematical models for thermalhydraulic regimes of fuel assemblies recognition on the acoustic and neutron noises parameters in the core of nuclear facilities. The two mathematical models for analysis of the random processes submitted to the automatic classification is proposed, i.e. statistical (using Bayes classifier of acoustic spectral density diagnosis signals) and geometrical (on the basis of formation in the featured space of dividing hyper-plane). The theoretical basis of the bubble boiling regimes in the fuel assemblies is formulated as identification of these regimes on the basis of random parameters of auto spectral density of acoustic noise (ASD) measured in the fuel assemblies (dynamic pressure in the upper plenum in the paper). The elaborated algorithms allow recognize realistic status of the fuel assemblies. For verification of the proposed mathematical models the analysis of experimental measurements was carried out. The research of the boiling onset and definition of the local values of the flow parameters in the seven-beam fuel assembly (length of 1.3 m, diameter of 6 mm) have shown the correct identification of the bubble boiling regimes. The experimental measurements on

  15. Flow distribution in ET-RR-1 core

    International Nuclear Information System (INIS)

    Khattab, M.; Mina, A.R.

    1989-01-01

    In nuclear reactors the flow may be arranged through individual bundles by orifices to achieve better thermal performance. A model based on constant pressure drop across different core regions is developed to determine the flow distribution in reactor core. The friction and grids in the bundles as well as the orifices diameters have an influence on modifying the flow distribution. The application of the proposed model on ET-RR-1 gives reasonable prediction of flow distribution

  16. Nuclear many-body problem with repulsive hard core interactions

    Energy Technology Data Exchange (ETDEWEB)

    Haddad, L M

    1965-07-01

    The nuclear many-body problem is considered using the perturbation-theoretic approach of Brueckner and collaborators. This approach is outlined with particular attention paid to the graphical representation of the terms in the perturbation expansion. The problem is transformed to centre-of-mass coordinates in configuration space and difficulties involved in ordinary methods of solution of the resulting equation are discussed. A new technique, the 'reference spectrum method', devised by Bethe, Brandow and Petschek in an attempt to simplify the numerical work in presented. The basic equations are derived in this approximation and considering the repulsive hard core part of the interaction only, the effective mass is calculated at high momentum (using the same energy spectrum for both 'particle' and 'hole' states). The result of 0.87m is in agreement with that of Bethe et al. A more complete treatment using the reference spectrum method in introduced and a self-consistent set of equations is established for the reference spectrum parameters again for the case of hard core repulsions. (author)

  17. Neutronics calculation of RTP core

    Science.gov (United States)

    Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

    2017-01-01

    Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

  18. Spent nuclear fuel application of CORE reg-sign systems engineering software

    International Nuclear Information System (INIS)

    Grimm, R.J.

    1996-01-01

    The DOE has adopted a systems engineering approach for the successful completion of the Spent Nuclear Fuel (SNF) Program mission. The DOE has utilized systems engineering principles to develop the SNF program guidance documents and has held several systems engineering workshops to develop the functional hierarchies of both the programmatic and technical side of the SNF program. The sheer size and complexity of the SNF program has led to problems that the Westinghouse Savannah River Company (WSRC) is working to manage through the use of systems engineering software. WSRC began using CORE reg-sign, an off the shelf PC based software package, to assist DOE in management of the SNF program. This paper details the successful use of the CORE reg-sign systems engineering software to date and the proposed future activities

  19. A Core Language for Separate Variability Modeling

    DEFF Research Database (Denmark)

    Iosif-Lazăr, Alexandru Florin; Wasowski, Andrzej; Schaefer, Ina

    2014-01-01

    Separate variability modeling adds variability to a modeling language without requiring modifications of the language or the supporting tools. We define a core language for separate variability modeling using a single kind of variation point to define transformations of software artifacts in object...... hierarchical dependencies between variation points via copying and flattening. Thus, we reduce a model with intricate dependencies to a flat executable model transformation consisting of simple unconditional local variation points. The core semantics is extremely concise: it boils down to two operational rules...

  20. Modeling Transients and Designing a Passive Safety System for a Nuclear Thermal Rocket Using Relap5

    Science.gov (United States)

    Khatry, Jivan

    Long-term high payload missions necessitate the need for nuclear space propulsion. Several nuclear reactor types were investigated by the Nuclear Engine for Rocket Vehicle Application (NERVA) program of National Aeronautics and Space Administration (NASA). Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. A NERVA design known as the Pewee I was selected for this purpose. The following transients were run: (i) modeling of corrosion-induced blockages on the peripheral fuel element coolant channels and their impact on radiation heat transfer in the core, and (ii) modeling of loss-of-flow-accidents (LOFAs) and their impact on radiation heat transfer in the core. For part (i), the radiation heat transfer rate of blocked channels increases while their neighbors' decreases. For part (ii), the core radiation heat transfer rate increases while the flow rate through the rocket system is decreased. However, the radiation heat transfer decreased while there was a complete LOFA. In this situation, the peripheral fuel element coolant channels handle the majority of the radiation heat transfer. Recognizing the LOFA as the most severe design basis accident, a passive safety system was designed in order to respond to such a transient. This design utilizes the already existing tie rod tubes and connects them to a radiator in a closed loop. Hence, this is basically a secondary loop. The size of the core is unchanged. During normal steady-state operation, this secondary loop keeps the moderator cool. Results show that the safety system is able to remove the decay heat and prevent the fuel elements from melting, in response to a LOFA and subsequent SCRAM.

  1. Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jae; Kim, Do Gyeum [Korea Institute of Civil Engineering and Building Technology, Goyang (Korea, Republic of); Cho, Jae Leon [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Myung Suk [Korea Hydro and Nuclear Power Co., Central Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

  2. Investigation of EAS cores

    Directory of Open Access Journals (Sweden)

    Shaulov S.B.

    2017-01-01

    Full Text Available The development of nuclear-electromagnetic cascade models in air in the late forties have shown informational content of the study of cores of extensive air showers (EAS. These investigations were the main goal in different experiments which were carried out over many years by a variety of methods. Outcomes of such investigations obtained in the HADRON experiment using an X-ray emulsion chamber (XREC as a core detector are considered. The Ne spectrum of EAS associated with γ-ray families, spectra of γ-rays (hadrons in EAS cores and the Ne dependence of the muon number, ⟨Nμ⟩, in EAS with γ-ray families are obtained for the first time at energies of 1015–1017 eV with this method. A number of new effects were observed, namely, an abnormal scaling violation in hadron spectra which are fundamentally different from model predictions, an excess of muon number in EAS associated with γ-ray families, and the penetrating component in EAS cores. It is supposed that the abnormal behavior of γ-ray spectra and Ne dependence of the muon number are explained by the emergence of a penetrating component in the 1st PCR spectrum ‘knee’ range. Nuclear and astrophysical explanations of the origin of the penetrating component are discussed. The necessity of considering the contribution of a single close cosmic-ray source to explain the PCR spectrum in the knee range is noted.

  3. Final Report, Nuclear Energy Research Initiative (NERI) Project: An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model

    International Nuclear Information System (INIS)

    Anistratov, Dmitriy Y.; Adams, Marvin L.; Palmer, Todd S.; Smith, Kord S.; Clarno, Kevin; Hikaru Hiruta; Razvan Nes

    2003-01-01

    OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations

  4. Modification of Core Model for KNTC 2 Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y.K.; Lee, J.G.; Park, J.E.; Bae, S.N.; Chin, H.C. [Korea Electric Power Research Institute, Taejeon (Korea, Republic of)

    1997-12-31

    KNTC 2 simulator was developed in 1986 referencing YGN 1. Since the YGN 1 has changed its fuel cycle to long term cycle(cycle 9), the data such as rod worth, boron worth, moderator temperature coefficient, and etc. of the simulator and those of the YGN 1 became different. To incorporate these changes into the simulator and make the simulator more close to the reference plant, core model upgrade became a necessity. During this research, core data for the simulator was newly generated using APA of the WH. And to make it easy tuning and verification of the key characteristics of the reactor model, PC-Based tool was also developed. And to facilitate later core model upgrade, two procedures-`the Procedures for core characteristic generation` and `the Procedures for core characteristic modification`-were also developed. (author). 16 refs., 22 figs., 1 tab.

  5. Modelling the core magnetic field of the earth

    Science.gov (United States)

    Harrison, C. G. A.; Carle, H. M.

    1982-01-01

    It is suggested that radial off-center dipoles located within the core of the earth be used instead of spherical harmonics of the magnetic potential in modeling the core magnetic field. The off-center dipoles, in addition to more realistically modeling the physical current systems within the core, are if located deep within the core more effective at removing long wavelength signals of either potential or field. Their disadvantage is that their positions and strengths are more difficult to compute, and such effects as upward and downward continuation are more difficult to manipulate. It is nevertheless agreed with Cox (1975) and Alldredge and Hurwitz (1964) that physical realism in models is more important than mathematical convenience. A radial dipole model is presented which agrees with observations of secular variation and excursions.

  6. Development of the RSAC Automation System for Reload Core of WH NPP

    International Nuclear Information System (INIS)

    Choi, Yu Sun; Bae, Sung Man; Koh, Byung Marn; Hong, Sun Kwan

    2006-01-01

    The Nuclear Design for Reload Core of Westinghouse Nuclear Power Plant consists of 'Reload Core Model Search', 'Safety Analysis(RSAC)', 'NDR(Nuclear Design Report) and OCAP(Operational Core Analysis Package Generation)' phases. Since scores of calculations for various accidents are required to confirm that the safety analysis assumptions are valid, the Safety Analysis(RSAC) is the most important and time and effort consuming phase of reload core design sequence. The Safety Analysis Automation System supports core designer by the automation of safety analysis calculations in 'Safety Analysis' phase(about 20 calculations). More than 10 kinds of codes, APA(ALPHA/PHOENIX/ANC), APOLLO, VENUS, PHIRE XEFIT, INCORE, etc. are being used for Safety Analysis calculations. Westinghouse code system needs numerous inputs and outputs, so the possibility of human errors could not be ignored during Safety Analysis calculations. To remove these inefficiencies, all input files for Safety Analysis calculations are automatically generated and executed by this Safety Analysis Automation System. All calculation notes are generated and the calculation results are summarized in RSAC (Reload Safety Analysis Checklist) by this system. Therefore, The Safety Analysis Automation System helps the reload core designer to perform safety analysis of the reload core model instantly and correctly

  7. Quark-Meson-Coupling (QMC) model for finite nuclei, nuclear matter and beyond

    Science.gov (United States)

    Guichon, P. A. M.; Stone, J. R.; Thomas, A. W.

    2018-05-01

    The Quark-Meson-Coupling model, which self-consistently relates the dynamics of the internal quark structure of a hadron to the relativistic mean fields arising in nuclear matter, provides a natural explanation to many open questions in low energy nuclear physics, including the origin of many-body nuclear forces and their saturation, the spin-orbit interaction and properties of hadronic matter at a wide range of densities up to those occurring in the cores of neutron stars. Here we focus on four aspects of the model (i) a full comprehensive survey of the theory, including the latest developments, (ii) extensive application of the model to ground state properties of finite nuclei and hypernuclei, with a discussion of similarities and differences between the QMC and Skyrme energy density functionals, (iii) equilibrium conditions and composition of hadronic matter in cold and warm neutron stars and their comparison with the outcome of relativistic mean-field theories and, (iv) tests of the fundamental idea that hadron structure changes in-medium.

  8. MONJU experimental data analysis and its feasibility evaluation to build up the standard data base for large FBR nuclear core design

    International Nuclear Information System (INIS)

    Sugino, K.; Iwai, T.

    2006-01-01

    MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores developed in Japan. Subsequently, feasibility of the MONJU integral data was evaluated by the cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that the MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to considerably improve the prediction accuracy of neutronic parameters for MONJU. (authors)

  9. Enhanced Core Noise Modeling for Turbofan Engines

    Science.gov (United States)

    Stone, James R.; Krejsa, Eugene A.; Clark, Bruce J.

    2011-01-01

    This report describes work performed by MTC Technologies (MTCT) for NASA Glenn Research Center (GRC) under Contract NAS3-00178, Task Order No. 15. MTCT previously developed a first-generation empirical model that correlates the core/combustion noise of four GE engines, the CF6, CF34, CFM56, and GE90 for General Electric (GE) under Contract No. 200-1X-14W53048, in support of GRC Contract NAS3-01135. MTCT has demonstrated in earlier noise modeling efforts that the improvement of predictive modeling is greatly enhanced by an iterative approach, so in support of NASA's Quiet Aircraft Technology Project, GRC sponsored this effort to improve the model. Since the noise data available for correlation are total engine noise spectra, it is total engine noise that must be predicted. Since the scope of this effort was not sufficient to explore fan and turbine noise, the most meaningful comparisons must be restricted to frequencies below the blade passage frequency. Below the blade passage frequency and at relatively high power settings jet noise is expected to be the dominant source, and comparisons are shown that demonstrate the accuracy of the jet noise model recently developed by MTCT for NASA under Contract NAS3-00178, Task Order No. 10. At lower power settings the core noise became most apparent, and these data corrected for the contribution of jet noise were then used to establish the characteristics of core noise. There is clearly more than one spectral range where core noise is evident, so the spectral approach developed by von Glahn and Krejsa in 1982 wherein four spectral regions overlap, was used in the GE effort. Further analysis indicates that the two higher frequency components, which are often somewhat masked by turbomachinery noise, can be treated as one component, and it is on that basis that the current model is formulated. The frequency scaling relationships are improved and are now based on combustor and core nozzle geometries. In conjunction with the Task

  10. Modelling Pressurized Water Reactor cores in terms of porous media

    International Nuclear Information System (INIS)

    Ricciardi, G.; Collard, B.; Ricciardi, G.; Bellizzi, S.; Cochelin, B.

    2009-01-01

    The aim of this study is to develop a tractable model of a nuclear reactor core taking the complexity of the structure (including its nonlinear behaviour) and fluid flow coupling into account. The mechanical behaviour modelling includes the dynamics of both the fuel assemblies and the fluid. Each rod bundle is modelled in the form of a deformable porous medium; then, the velocity field of the fluid and the displacement field of the structure are defined over the whole domain. The fluid and the structure are first modelled separately, before being linked together. The equations of motion for the structure are obtained using a Lagrangian approach and, to be able to link up the fluid and the structure, the equations of motion for the fluid are obtained using an arbitrary Lagrangian Eulerian approach. The finite element method is applied to spatially discretize the equations. Simulations are performed to analyse the effects of the characteristics of the fluid and of the structure. Finally, the model is validated with a test involving two fuel assemblies, showing good agreement with the experimental data. (authors)

  11. Comparison of serpent and triton generated FEW group constants for APR1400 nuclear reactor core

    International Nuclear Information System (INIS)

    Elsawi, Mohamed A.; Alnoamani, Zainab

    2015-01-01

    The accuracy of full-core reactor power calculations using diffusion codes is strongly dependent on the quality of the homogenized cross sections and other few-group constants generated by lattice codes. For many years, deterministic lattice codes have been used to generate these constants using different techniques: the discrete ordinates, collision probability or the method of characteristics, just to name a few. These codes, however, show some limitations, for example, on complex geometries or near heavy absorbers as in modern pressurized water reactor (PWR) designs like the Korean Advanced Power Reactor 1400 (APR1400) core. The use of continuous-energy Monte Carlo (MC) codes to produce nuclear constants can be seen as an attractive option when dealing with fuel or reactor types that lie beyond the capabilities of conventional deterministic lattice transport codes. In this paper, the few-group constants generated by two of the state-of-the-art reactor physics codes, SERPENT and SCALE/TRITON, will be critically studied and their reliability for being used in subsequent diffusion calculations will be evaluated. SERPENT is a 3D, continuous-energy, Monte Carlo reactor physics code which has a built-in burn-up calculation capability. It has been developed at the Technical Research Center of Finland (VTT) since 2004. SCALE/TRITON, on the other hand, is a control module developed within the framework of SCALE package that enables performing deterministic 2-D transport calculations on nuclear reactor core lattices. The approach followed in this paper is as follows. First, the few-group nuclear constants for the APR1400 reactor core were generated using SERPENT (version 2.1.22) and NEWT (in SCALE version 6.1.2) codes. For both codes, the critical spectrum, calculated using the B1 method, was used as a weighting function. Second, 2-D diffusion calculations were performed using the US NRC core simulator PARCS employing the two few-group constant sets generated in the first

  12. Collapsing stellar cores and supernovae

    Energy Technology Data Exchange (ETDEWEB)

    Epstein, R J [Nordisk Inst. for Teoretisk Atomfysik, Copenhagen (Denmark); Noorgaard, H [Nordisk Inst. for Teoretisk Atomfysik, Copenhagen (Denmark); Chicago Univ., IL (USA). Enrico Fermi Inst.); Bond, J R [Niels Bohr Institutet, Copenhagen (Denmark); California Inst. of Tech., Pasadena (USA). W.K. Kellogg Radiation Lab.)

    1979-05-01

    The evolution of a stellar core is studied during its final quasi-hydrostatic contraction. The core structure and the (poorly known) properties of neutron rich matter are parametrized to include most plausible cases. It is found that the density-temperature trajectory of the material in the central part of the core (the core-center) is insensitive to nearly all reasonable parameter variations. The central density at the onset of the dynamic phase of the collapse (when the core-center begins to fall away from the rest of the star) and the fraction of the emitted neutrinos which are trapped in the collapsing core-center depend quite sensitively on the properties of neutron rich matter. We estimate that the amount of energy Ecm which is imparted to the core-mantle by the neutrinos which escape from the imploded core-center can span a large range of values. For plausible choices of nuclear and model parameters Ecm can be large enough to yield a supernova event.

  13. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    2013-01-01

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO 2 and MOX fuel rods, (3) analysis of isotopic composition data for UO 2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  14. Validation study of core analysis methods for full MOX BWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO{sub 2} and MOX fuel rods, (3) analysis of isotopic composition data for UO{sub 2} and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  15. Development of a multi-objective PBIL evolutionary algorithm applied to a nuclear reactor core reload optimization problem

    International Nuclear Information System (INIS)

    Machado, Marcelo D.; Dchirru, Roberto

    2005-01-01

    The nuclear reactor core reload optimization problem consists in finding a pattern of partially burned-up and fresh fuels that optimizes the plant's next operation cycle. This optimization problem has been traditionally solved using an expert's knowledge, but recently artificial intelligence techniques have also been applied successfully. The artificial intelligence optimization techniques generally have a single objective. However, most real-world engineering problems, including nuclear core reload optimization, have more than one objective (multi-objective) and these objectives are usually conflicting. The aim of this work is to develop a tool to solve multi-objective problems based on the Population-Based Incremental Learning (PBIL) algorithm. The new tool is applied to solve the Angra 1 PWR core reload optimization problem with the purpose of creating a Pareto surface, so that a pattern selected from this surface can be applied for the plant's next operation cycle. (author)

  16. Thermal hydraulic model validation for HOR mixed core fuel management

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Vries, J.W. de; Leege, P.F.A. de

    1997-01-01

    A thermal-hydraulic core management model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type university research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and for in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, and model validation are discussed. (author)

  17. Analysis of Random-Loading HTR-PROTEUS Cores with Continuous Energy Monte Carlo Code Based on A Statistical Geometry Model

    International Nuclear Information System (INIS)

    Murata, Isao; Miyamaru, Hiroyuki

    2008-01-01

    Spherical elements have remarkable features in various applications in the nuclear engineering field. In 1990's, by the project of HTR-PROTEUS at PSI various pebble bed reactor experiments were conducted including cores with a lot of spherical fuel elements loaded randomly. In this study, criticality experiments of the random-loading HTR-PROTEUS cores were analyzed by MCNP-BALL, which could deal with a random arrangement of spherical fuel elements exactly with a statistical geometry model. As a result of analysis, the calculated effective multiplication factors were in fairly good agreement with the measurements within about 0.5%Δk/k. In comparison with other numerical analysis, our effective multiplication factors were between the experimental values and the VSOP calculations. To investigate the discrepancy of the effective multiplication factors between the experiments and calculations, sensitivity analyses were performed. As the result, the sensitivity of impurity boron concentration was fairly large. The reason of the present slight overestimation was not made clear at present. However, the presently existing difference was thought to be related to the impurity boron concentration, not to the modelling of the reactor and the used nuclear data. From the present study, it was confirmed that MCNP-BALL would have an advantage to conventional transport codes by comparing with their numerical results and the experimental values. As for the criticality experiment of PROTEUS, we would conclude that the two cores of Core 4.2 and 4.3 could be regarded as an equivalent experiment of a reference critical core, which was packed in the packing fraction of RLP. (authors)

  18. Analysis of Random-Loading HTR-PROTEUS Cores with Continuous Energy Monte Carlo Code Based on A Statistical Geometry Model

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Isao; Miyamaru, Hiroyuki [Division of Electrical, Electronic and Information Engineering, Osaka University, Yamada-oka 2-1, Suita, Osaka, 565-0871 (Japan)

    2008-07-01

    Spherical elements have remarkable features in various applications in the nuclear engineering field. In 1990's, by the project of HTR-PROTEUS at PSI various pebble bed reactor experiments were conducted including cores with a lot of spherical fuel elements loaded randomly. In this study, criticality experiments of the random-loading HTR-PROTEUS cores were analyzed by MCNP-BALL, which could deal with a random arrangement of spherical fuel elements exactly with a statistical geometry model. As a result of analysis, the calculated effective multiplication factors were in fairly good agreement with the measurements within about 0.5%DELTAk/k. In comparison with other numerical analysis, our effective multiplication factors were between the experimental values and the VSOP calculations. To investigate the discrepancy of the effective multiplication factors between the experiments and calculations, sensitivity analyses were performed. As the result, the sensitivity of impurity boron concentration was fairly large. The reason of the present slight overestimation was not made clear at present. However, the presently existing difference was thought to be related to the impurity boron concentration, not to the modelling of the reactor and the used nuclear data. From the present study, it was confirmed that MCNP-BALL would have an advantage to conventional transport codes by comparing with their numerical results and the experimental values. As for the criticality experiment of PROTEUS, we would conclude that the two cores of Core 4.2 and 4.3 could be regarded as an equivalent experiment of a reference critical core, which was packed in the packing fraction of RLP. (authors)

  19. Spent nuclear fuel application of CORE reg-sign systems engineering software

    International Nuclear Information System (INIS)

    Grimm, R.J.

    1996-01-01

    The Department of Energy (DOE) has adopted a systems engineering approach for the successful completion of the Spent Nuclear Fuel (SNF) Program mission. The DOE has utilized systems engineering principles to develop the SNF Program guidance documents and has held several systems engineering workshops to develop the functional hierarchies of both the programmatic and technical side of the SNF Program. The sheer size and complexity of the SNF Program, however, has led to problems that the Westinghouse Savannah River Company (WSRC) is working to manage through the use of systems engineering software. WSRC began using CORE reg-sign, an off-the-shelf PC based software package, to assist the DOE in management of the SNF program. This paper details the successful use of the CORE reg-sign systems engineering software to date and the proposed future activities

  20. Restraint system for core elements of a reactor core

    International Nuclear Information System (INIS)

    Class, G.

    1975-01-01

    In a nuclear reactor, a core element bundle formed of a plurality of side-by-side arranged core elements is surrounded by restraining elements that exert a radially inwardly directly restraining force generating friction forces between the core elements in a restraining plane that is transverse to the core element axes. The adjoining core elements are in rolling contact with one another in the restraining plane by virtue of rolling-type bearing elements supported in the core elements. (Official Gazette)

  1. Core seismic behaviour: linear and non-linear models

    International Nuclear Information System (INIS)

    Bernard, M.; Van Dorsselaere, M.; Gauvain, M.; Jenapierre-Gantenbein, M.

    1981-08-01

    The usual methodology for the core seismic behaviour analysis leads to a double complementary approach: to define a core model to be included in the reactor-block seismic response analysis, simple enough but representative of basic movements (diagrid or slab), to define a finer core model, with basic data issued from the first model. This paper presents the history of the different models of both kinds. The inert mass model (IMM) yielded a first rough diagrid movement. The direct linear model (DLM), without shocks and with sodium as an added mass, let to two different ones: DLM 1 with independent movements of the fuel and radial blanket subassemblies, and DLM 2 with a core combined movement. The non-linear (NLM) ''CORALIE'' uses the same basic modelization (Finite Element Beams) but accounts for shocks. It studies the response of a diameter on flats and takes into account the fluid coupling and the wrapper tube flexibility at the pad level. Damping consists of one modal part of 2% and one part due to shocks. Finally, ''CORALIE'' yields the time-history of the displacements and efforts on the supports, but damping (probably greater than 2%) and fluid-structures interaction are still to be precised. The validation experiments were performed on a RAPSODIE core mock-up on scale 1, in similitude of 1/3 as to SPX 1. The equivalent linear model (ELM) was developed for the SPX 1 reactor-block response analysis and a specified seismic level (SB or SM). It is composed of several oscillators fixed to the diagrid and yields the same maximum displacements and efforts than the NLM. The SPX 1 core seismic analysis with a diagrid input spectrum which corresponds to a 0,1 g group acceleration, has been carried out with these models: some aspects of these calculations are presented here

  2. Nonlinear Dynamic Model of PMBLDC Motor Considering Core Losses

    DEFF Research Database (Denmark)

    Fasil, Muhammed; Mijatovic, Nenad; Jensen, Bogi Bech

    2017-01-01

    The phase variable model is used commonly when simulating a motor drive system with a three-phase permanent magnet brushless DC (PMBLDC) motor. The phase variable model neglects core losses and this affects its accuracy when modelling fractional-slot machines. The inaccuracy of phase variable mod...... on the detailed analysis of the flux path and the variation of flux in different components of the machine. A prototype of fractional slot axial flux PMBLDC in-wheel motor is used to assess the proposed nonlinear dynamic model....... of fractional-slot machines can be attributed to considerable armature flux harmonics, which causes an increased core loss. This study proposes a nonlinear phase variable model of PMBLDC motor that considers the core losses induced in the stator and the rotor. The core loss model is developed based...

  3. Development of a standard data base for FBR core nuclear design (XIII). Analysis of small sample reactivity experiments at ZPPR-9

    International Nuclear Information System (INIS)

    Sato, Wakaei; Fukushima, Manabu; Ishikawa, Makoto

    2000-09-01

    A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data library JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design. (author)

  4. Modeling the atmospheric dispersion of radioactive effluents in a nuclear accident situation

    International Nuclear Information System (INIS)

    Margeanu, Sorin

    2002-01-01

    In case of a nuclear accident, which could lead to release of radioactive contaminants, fastest countermeasures are needed related to sheltering, iodine distribution, evacuation and interdiction of food and water consumption. All these decisions should be based either on estimation of inhaled dose and the dose due to external exposure for public, or on the estimation of radioactive concentration in food (which will depend on the radioactive concentration in air and ground deposition). The dispersion model used, was a Gaussian 'puff' model. The vertical dispersion was considered not dependent on the release high. The used meteorological data are specific for the SCN - Pitesti site, collected every hour for one year. The meteorological data file contains: the wind speed (in m/s), wind direction (degrees clockwise from north), atmospheric stability category, precipitation rate (in mm/h) and the high of the mixing layer (in m). A hypothetical major nuclear accident at TRIGA - SSR of INR - Pitesti, due to a serious damage of the reactor core leading, to a large release of radioactive contaminants was examined. The release was considered as a single phase with of one hour duration. The release factors for the considered isotopic mixture are 100% noble gases (of the reactor core inventory), 40% iodine (of the reactor core inventory) and 40% particulate, i.e., 40% of the fission products of core fission products inventory, released as particles. The accuracy of the model could be increased by implementation of the code on a real-time system, where the acquisition of the parameters done is on-line, namely, the data are introduced as soon as the modification of meteorological and dosimetric conditions are produced. In this case, the parameters used in formulas can be adjusted according with the field situation. Unfortunately the real-time systems need more powerful resources: monitoring stations which can measure and send on-line the data and which can cover a large area

  5. Determination of the neutron activation profile of core drill samples by gamma-ray spectrometry.

    Science.gov (United States)

    Gurau, D; Boden, S; Sima, O; Stanga, D

    2018-04-01

    This paper provides guidance for determining the neutron activation profile of core drill samples taken from the biological shield of nuclear reactors using gamma spectrometry measurements. Thus, it provides guidance for selecting a model of the right form to fit data and using least squares methods for model fitting. The activity profiles of two core samples taken from the biological shield of a nuclear reactor were determined. The effective activation depth and the total activity of core samples along with their uncertainties were computed by Monte Carlo simulation. Copyright © 2017 Elsevier Ltd. All rights reserved.

  6. Sub-saturation matter in compact stars: Nuclear modelling in the framework of the extended Thomas-Fermi theory

    Energy Technology Data Exchange (ETDEWEB)

    Aymard, François; Gulminelli, Francesca [CNRS and ENSICAEN, UMR6534, LPC, 14050 Caen cédex (France); Margueron, Jérôme [Institut de Physique Nucléaire de Lyon, Université Claude Bernard Lyon 1, IN2P3-CNRS, F-69622 Villeurbanne Cedex (France)

    2015-02-24

    A recently introduced analytical model for the nuclear density profile [1] is implemented in the Extended Thomas-Fermi (ETF) energy density functional. This allows to (i) shed a new light on the issue of the sign of surface symmetry energy in nuclear mass formulas, as well as to (ii) show the importance of the in-medium corrections to the nuclear cluster energies in thermodynamic conditions relevant for the description of core-collapse supernovae and (proto)-neutron star crust.

  7. Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

    Energy Technology Data Exchange (ETDEWEB)

    Fok, Alex

    2013-10-30

    The proposed work addresses the key research need for the development of constitutive models and overall failure models for graphite and high temperature structural materials, with the long-term goal being to maximize the design life of the Next Generation Nuclear Plant (NGNP). To this end, the capability of a Continuum Damage Mechanics (CDM) model, which has been used successfully for modeling fracture of virgin graphite, will be extended as a predictive and design tool for the core components of the very high- temperature reactor (VHTR). Specifically, irradiation and environmental effects pertinent to the VHTR will be incorporated into the model to allow fracture of graphite and ceramic components under in-reactor conditions to be modeled explicitly using the finite element method. The model uses a combined stress-based and fracture mechanics-based failure criterion, so it can simulate both the initiation and propagation of cracks. Modern imaging techniques, such as x-ray computed tomography and digital image correlation, will be used during material testing to help define the baseline material damage parameters. Monte Carlo analysis will be performed to address inherent variations in material properties, the aim being to reduce the arbitrariness and uncertainties associated with the current statistical approach. The results can potentially contribute to the current development of American Society of Mechanical Engineers (ASME) codes for the design and construction of VHTR core components.

  8. Automated in-core image generation from video to aid visual inspection of nuclear power plant cores

    Energy Technology Data Exchange (ETDEWEB)

    Murray, Paul, E-mail: paul.murray@strath.ac.uk [Department of Electronic and Electrical Engineering, University of Strathclyde, Technology and Innovation Centre, 99 George Street, Glasgow, G1 1RD (United Kingdom); West, Graeme; Marshall, Stephen; McArthur, Stephen [Dept. Electronic and Electrical Engineering, University of Strathclyde, Royal College Building, 204 George Street, Glasgow G1 1XW (United Kingdom)

    2016-04-15

    Highlights: • A method is presented which improves visual inspection of reactor cores. • Significant time savings are made to activities on the critical outage path. • New information is extracted from existing data sources without additional overhead. • Examples from industrial case studies across the UK fleet of AGR stations. - Abstract: Inspection and monitoring of key components of nuclear power plant reactors is an essential activity for understanding the current health of the power plant and ensuring that they continue to remain safe to operate. As the power plants age, and the components degrade from their initial start-of-life conditions, the requirement for more and more detailed inspection and monitoring information increases. Deployment of new monitoring and inspection equipment on existing operational plant is complex and expensive, as the effect of introducing new sensing and imaging equipment to the existing operational functions needs to be fully understood. Where existing sources of data can be leveraged, the need for new equipment development and installation can be offset by the development of advanced data processing techniques. This paper introduces a novel technique for creating full 360° panoramic images of the inside surface of fuel channels from in-core inspection footage. Through the development of this technique, a number of technical challenges associated with the constraints of using existing equipment have been addressed. These include: the inability to calibrate the camera specifically for image stitching; dealing with additional data not relevant to the panorama construction; dealing with noisy images; and generalising the approach to work with two different capture devices deployed at seven different Advanced Gas Cooled Reactor nuclear power plants. The resulting data processing system is currently under formal assessment with a view to replacing the existing manual assembly of in-core defect montages. Deployment of the

  9. Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

    International Nuclear Information System (INIS)

    Mitchell, Jonathan; Fordham, Edmund J.

    2014-01-01

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general

  10. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

    2012-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core

  11. Analytical model for relativistic corrections to the nuclear magnetic shielding constant in atoms

    International Nuclear Information System (INIS)

    Romero, Rodolfo H.; Gomez, Sergio S.

    2006-01-01

    We present a simple analytical model for calculating and rationalizing the main relativistic corrections to the nuclear magnetic shielding constant in atoms. It provides good estimates for those corrections and their trends, in reasonable agreement with accurate four-component calculations and perturbation methods. The origin of the effects in deep core atomic orbitals is manifestly shown

  12. Analytical model for relativistic corrections to the nuclear magnetic shielding constant in atoms

    Energy Technology Data Exchange (ETDEWEB)

    Romero, Rodolfo H. [Facultad de Ciencias Exactas, Universidad Nacional del Nordeste, Avenida Libertad 5500 (3400), Corrientes (Argentina)]. E-mail: rhromero@exa.unne.edu.ar; Gomez, Sergio S. [Facultad de Ciencias Exactas, Universidad Nacional del Nordeste, Avenida Libertad 5500 (3400), Corrientes (Argentina)

    2006-04-24

    We present a simple analytical model for calculating and rationalizing the main relativistic corrections to the nuclear magnetic shielding constant in atoms. It provides good estimates for those corrections and their trends, in reasonable agreement with accurate four-component calculations and perturbation methods. The origin of the effects in deep core atomic orbitals is manifestly shown.

  13. U(6)-phonon model of nuclear collective motion

    International Nuclear Information System (INIS)

    Ganev, H.G.

    2015-01-01

    The U(6)-phonon model of nuclear collective motion with the semi-direct product structure [HW(21)]U(6) is obtained as a hydrodynamic (macroscopic) limit of the fully microscopic proton–neutron symplectic model (PNSM) with Sp(12, R) dynamical group. The phonon structure of the [HW(21)]U(6) model enables it to simultaneously include the giant monopole and quadrupole, as well as dipole resonances and their coupling to the low-lying collective states. The U(6) intrinsic structure of the [HW(21)]U(6) model, from the other side, gives a framework for the simultaneous shell-model interpretation of the ground state band and the other excited low-lying collective bands. It follows then that the states of the whole nuclear Hilbert space which can be put into one-to-one correspondence with those of a 21-dimensional oscillator with an intrinsic (base) U(6) structure. The latter can be determined in such a way that it is compatible with the proton–neutron structure of the nucleus. The macroscopic limit of the Sp(12, R) algebra, therefore, provides a rigorous mechanism for implementing the unified model ideas of coupling the valence particles to the core collective degrees of freedom within a fully microscopic framework without introducing redundant variables or violating the Pauli principle. (author)

  14. Review of coaxial flow gas core nuclear rocket fluid mechanics

    International Nuclear Information System (INIS)

    Weinstein, H.

    1976-01-01

    In a prematurely aborted attempt to demonstrate the feasibility of using a gas core nuclear reactor as a rocket engine, NASA initiated a number of studies on the relevant fluid mechanics problems. These studies were carried out at NASA laboratories, universities and industrial research laboratories. Because of the relatively sudden termination of most of this work, a unified overview was never presented which demonstrated the accomplishments of the program and pointed out the areas where additional work was required for a full understanding of the cavity flow. This review attempts to fulfill a part of this need in two important areas

  15. In-core assembly configuration having a dual-wall pressure boundary for nuclear reactor

    International Nuclear Information System (INIS)

    Todt, W.H. Sr.; Playfoot, K.C.

    1988-01-01

    This patent describes an in-core detector assembly of the type having an in-core part and an out-of-core part and having an elongated outer hollow housing tube with a wall thickness, an inner hollow calibration tube with a wall thickness and disposed concentrically within the outer tube to define an annular space therewith, and a plurality of discrete, circular, rod-like elements extending through the annular space, the improvement comprising: the elements having outer diameters and being of a number to substantially occupy the entire annular space of both the incore and out-of-core parts without significant voids between elements; each of the elements including at least an outer sheath and interior highly compacted mineral insulation for the entire length of the element; a first number of the elements also including center lead means connected to condition responsive element means in the in-core part of the length of the assembly and a second, remaining number of the elements being non-operating elements. The wall thickness of the housing tube and the wall thickness of the calibration tube, taken together with the diameter of the elements, provide a thickness dimension adequate to meet code primary pressure requirements for normal nuclear reactor in-core conditions, while the wall thickness of the calibration tube alone provides a thickness dimension less than adequate to meet such requirements

  16. Nuclear Power Reactor Core Melt Accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus- FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day

  17. Testing the HTA core model: experiences from two pilot projects

    DEFF Research Database (Denmark)

    Pasternack, Iris; Anttila, Heidi; Mäkelä, Marjukka

    2009-01-01

    OBJECTIVES: The aim of this study was to analyze and describe process and outcomes of two pilot assessments based on the HTA Core Model, discuss the applicability of the model, and explore areas of development. METHODS: Data were gathered from HTA Core Model and pilot Core HTA documents, their va...

  18. Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wehe, D K [Oak Ridge National Laboratory, Oak Ridge, TN (United States); King, J S; Lee, J C; Martin, W R [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    1985-07-01

    The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from meV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional {sup 238}U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above MeV. (author)

  19. A new nodal kinetics method for analyzing fast control rod motions in nuclear reactor cores

    International Nuclear Information System (INIS)

    Kaya, S.; Yavuz, H.

    2001-01-01

    A new nodal kinetics approach is developed for analyzing large reactivity accidents in nuclear reactor cores. This method shows promising that it has capability of inspecting promt criticality transients and it gives comparable results with respect to those of other techniques. (orig.)

  20. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y H; Cha, K H; Lee, S H [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  1. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  2. Applications for coupled core neutronics and thermal-hydraulic models

    International Nuclear Information System (INIS)

    Eller, J.

    1996-01-01

    The unprecedented increases in computing capacity that have occurred during the last decade have affected our sciences, and thus our lives, to an extent that is difficult to overstate. All indications are that this trend will continue for years to come. Nuclear reactor systems analysis is one of many areas of engineering that has changed dramatically as a result of this evolution. Our ability to model the various mechanical and physical systems in greater and greater detail has allowed significant improvements in operational efficiency in spite of increasing regulatory requirements. Many of these efficiencies result from the use of more complex and geometrically detailed computer modeling, which is used to justify a reduction or elimination of some of the conservatisms required by earlier, less sophisticated analyses. And more recently, as our industries open-quotes downsize,close quotes efforts are being made to find ways to use the ever-increasing computing capacity to design systems that accomplish more work, in less time, and with fewer people. The balance of this paper discusses some of the visions that Duke Power Company feels would most benefit their particular methodologies. One of the concepts receiving a lot of attention involves an automated coupling of a thermal-hydraulic plant systems analysis model to a three-dimensional core neutronics program. The thermal-hydraulic analysis of several postulated system transients incorporates large conservatisms because of limited ability to model complex time-dependent asymmetric heat sources in adequate geometric detail. For these transients, the core behavior is closely coupled with the thermal-hydraulic behavior of the total plant system and vice versa. Steam-line break, uncontrolled rod withdrawal, and rod drop anayses are likely to benefit most from this type of linked process

  3. Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

    2015-01-01

    As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

  4. Interaction between core analysis methodology and nuclear design: some PWR examples

    International Nuclear Information System (INIS)

    Rothleder, B.M.; Eich, W.J.

    1982-01-01

    The interaction between core analysis methodology and nuclear design is exemplified by PSEUDAX, a major improvement related to the Advanced Recycle methodology program (ARMP) computer code system, still undergoing development by the Electric Power Research Institute. The mechanism of this interaction is explored by relating several specific nulcear design changes to the demands placed by these changes on the ARMP system, and by examining the meeting of these demands, first within the standard ARMP methodology and then through augmentation of the standard methodology by development of PSEUDAX

  5. Theoretical analysis and numerical modelling of heat transfer and fuel migration in underlying soils and constructive elements of nuclear plants during an accident release from the core

    International Nuclear Information System (INIS)

    Arutunjan, R.V.; Bolshov, L.A.; Vitukov, V.V.; Goloviznin, V.M.; Dykhne, A.M.; Kiselev, V.P.; Klementova, S.V.; Krayushkin, I.E.; Moskovchenko, A.V.; Pismennii, V.D.; Popkov, A.G.; Chernov, S.Y.; Chudanov, V.V.; Khoruzhii, O.V.; Yudin, A.I.

    1990-01-01

    Migration of fuel fragments and core fission products during severe accidents on nuclear plants is studied analytically and numerically. The problems of heat transfer and migration of volume heat sources in construction materials and underlying soils are considered

  6. Control and balance of nuclear matters used for core fabrication of Super Phenix

    International Nuclear Information System (INIS)

    Beche, M.; Guillet, H.; Heyraud, H.; Levrard, J.; Pajot, J.

    1987-05-01

    The fabrication of the core of the fast breeder reactor set up at Creys Malville ended in March 1984. It started in 1978 and it required, for the fabrication of the 410 assemblies, the utilization of 7438 kg of plutonium. To satisfy national and international regulations, DPFER/SFER has used a methodology to follow and to control the movements of the nuclear materials. These controls are achieved by physical methods, chemical methods and empiric methods. Euratom has conducted a succession of inspections during the 5.5 years of that campaign. The inventory difference, in the fabrication of that core, represents about 0.1% of the total mass of the plutonium handled [fr

  7. Modelling perspectives on radiation chemistry in BWR reactor core

    International Nuclear Information System (INIS)

    Ibe, Eishi

    1991-01-01

    Development of a full-system boiling water reactor core model started in 1982. The model included a two-region reactor core, one with and one without boiling. Key design parameters consider variable dose rates in a three-layer liquid downcomer. Dose rates in the core and downcomer include both generation and recombination reactions of species. Agreement is good between calculations and experimental data of oxygen concentration as a function of hydrogen concentration for different bubble sizes. Oxygen concentration is reduced in the reactor pressure vessel (RPV) by increasing bubble size. The multilayer model follows the oxygen data better than a single-layered model at high concentrations of hydrogen. Key reactions are reduced to five radiolysis reactions and four decomposition reactions for hydrogen peroxide. Calculations by the DOT 3 code showed dose rates from neutrons and gamma rays in various parts of the core. Concentrations of oxygen, hydrogen peroxide, and hydrogen were calculated by the model as a function of time from core inlet. Similar calculations for NWC and HWC were made as a function of height from core inlet both in the boiling channel an the bypass channel. Finally the model was applied to calculate the oxygen plus half the hydrogen peroxide concentrations as a function of hydrogen concentration to compare with data from five plants. Power density distribution with core height was given for an early stage and an end stage of a cycle. Increases of dose rates in the turbine for seven plants were shown as a function of increased hydrogen concentration in the reactor water

  8. System modeling for the advanced thermionic initiative single cell thermionic space nuclear reactor

    International Nuclear Information System (INIS)

    Lee, H.H.; Lewis, B.R.; Klein, A.C.; Pawlowski, R.A.

    1993-01-01

    Incore thermionic space reactor design concepts which operate in a nominal power output range of 20 to 40 kWe are described. Details of the neutronics, thermionic, shielding, and heat rejection performance are presented. Two different designs, ATI-Driven and ATI-Driverless, are considered. Comparison of the core overall performance of these two configurations are described. The comparison of these two cores includes the overall conversion efficiency, reactor mass, shield mass, and heat rejection mass. An overall system design has been developed to model the advanced incore thermionic energy conversion based nuclear reactor systems for space applications in this power range

  9. Contribution of Anticipated Transients Without Scram (ATWS) to core melt at United States nuclear power plants

    International Nuclear Information System (INIS)

    Giachetti, R.T.

    1989-09-01

    This report looks at WASH-1400 and several other Probabilistic Risk Assessments (PRAs) and Probabilistic Safety Studies (PSSs) to determine the contribution of Anticipated Transients Without Scram (ATWS) events to the total core melt probability at eight nuclear power plants in the United States. After considering each plant individually, the results are compared from plant to plant to see if any generic conclusions regarding ATWS, or core melt in general, can be made. 8 refs., 34 tabs

  10. NEW EQUATIONS OF STATE IN SIMULATIONS OF CORE-COLLAPSE SUPERNOVAE

    International Nuclear Information System (INIS)

    Hempel, M.; Liebendörfer, M.; Fischer, T.; Schaffner-Bielich, J.

    2012-01-01

    We discuss three new equations of state (EOS) in core-collapse supernova simulations. The new EOS are based on the nuclear statistical equilibrium model of Hempel and Schaffner-Bielich (HS), which includes excluded volume effects and relativistic mean-field (RMF) interactions. We consider the RMF parameterizations TM1, TMA, and FSUgold. These EOS are implemented into our spherically symmetric core-collapse supernova model, which is based on general relativistic radiation hydrodynamics and three-flavor Boltzmann neutrino transport. The results obtained for the new EOS are compared with the widely used EOS of H. Shen et al. and Lattimer and Swesty. The systematic comparison shows that the model description of inhomogeneous nuclear matter is as important as the parameterization of the nuclear interactions for the supernova dynamics and the neutrino signal. Furthermore, several new aspects of nuclear physics are investigated: the HS EOS contains distributions of nuclei, including nuclear shell effects. The appearance of light nuclei, e.g., deuterium and tritium, is also explored, which can become as abundant as alphas and free protons. In addition, we investigate the black hole formation in failed core-collapse supernovae, which is mainly determined by the high-density EOS. We find that temperature effects lead to a systematically faster collapse for the non-relativistic LS EOS in comparison with the RMF EOS. We deduce a new correlation for the time until black hole formation, which allows the determination of the maximum mass of proto-neutron stars, if the neutrino signal from such a failed supernova would be measured in the future. This would give a constraint for the nuclear EOS at finite entropy, complementary to observations of cold neutron stars.

  11. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given

  12. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  13. Nuclear power investment risk economic model

    International Nuclear Information System (INIS)

    Houghton, W.J.; Postula, F.D.

    1985-12-01

    This paper describes an economic model which was developed to evaluate the net costs incurred by a utility due to an accident induced outage at a nuclear power plant. During such an outage the portion of the plant operating costs associated with power production are saved; however, the owning utility faces a sizable expense as fossil fuels are burned as a substitute for the incapacitated nuclear power. Additional expenses are incurred by the utility for plant repair and if necessary, decontamination costs. The model makes provision for mitigating these costs by sales of power, property damage insurance payments, tax write-offs and increased rates. Over 60 economic variables contribute to the net cost uncertainty. The values of these variables are treated as uncertainty distributions and are used in a Monte carlo computer program to evaluate the cost uncertainty (investment risk) associated with damage which could occur from various categories of initiating accidents. As an example, results of computations for various levels of damage associated with a loss of coolant accident are shown as a range of consequential plant downtime and unrecovered cost. A typical investment risk profile is shown for these types of accidents. Cost/revenue values for each economic factor are presented for a Three Mile Island - II type accident, e.g., uncontrolled core heatup. 4 refs., 6 figs., 3 tabs

  14. Modeling the effect in of criticality from changes in key parameters for small High Temperature Nuclear Reactor (U-BatteryTM) using MCNP4C

    International Nuclear Information System (INIS)

    Pauzi, A M

    2013-01-01

    The neutron transport code, Monte Carlo N-Particle (MCNP) which was wellkown as the gold standard in predicting nuclear reaction was used to model the small nuclear reactor core called U -battery TM, which was develop by the University of Manchester and Delft Institute of Technology. The paper introduces on the concept of modeling the small reactor core, a high temperature reactor (HTR) type with small coated TRISO fuel particle in graphite matrix using the MCNPv4C software. The criticality of the core were calculated using the software and analysed by changing key parameters such coolant type, fuel type and enrichment levels, cladding materials, and control rod type. The criticality results from the simulation were validated using the SCALE 5.1 software by [1] M Ding and J L Kloosterman, 2010. The data produced from these analyses would be used as part of the process of proposing initial core layout and a provisional list of materials for newly design reactor core. In the future, the criticality study would be continued with different core configurations and geometries.

  15. A computational technique to identify the optimal stiffness matrix for a discrete nuclear fuel assembly model

    International Nuclear Information System (INIS)

    Park, Nam-Gyu; Kim, Kyoung-Joo; Kim, Kyoung-Hong; Suh, Jung-Min

    2013-01-01

    Highlights: ► An identification method of the optimal stiffness matrix for a fuel assembly structure is discussed. ► The least squares optimization method is introduced, and a closed form solution of the problem is derived. ► The method can be expanded to the system with the limited number of modes. ► Identification error due to the perturbed mode shape matrix is analyzed. ► Verification examples show that the proposed procedure leads to a reliable solution. -- Abstract: A reactor core structural model which is used to evaluate the structural integrity of the core contains nuclear fuel assembly models. Since the reactor core consists of many nuclear fuel assemblies, the use of a refined fuel assembly model leads to a considerable amount of computing time for performing nonlinear analyses such as the prediction of seismic induced vibration behaviors. The computational time could be reduced by replacing the detailed fuel assembly model with a simplified model that has fewer degrees of freedom, but the dynamic characteristics of the detailed model must be maintained in the simplified model. Such a model based on an optimal design method is proposed in this paper. That is, when a mass matrix and a mode shape matrix are given, the optimal stiffness matrix of a discrete fuel assembly model can be estimated by applying the least squares minimization method. The verification of the method is completed by comparing test results and simulation results. This paper shows that the simplified model's dynamic behaviors are quite similar to experimental results and that the suggested method is suitable for identifying reliable mathematical model for fuel assemblies

  16. A calculation model for a HTR core seismic response

    International Nuclear Information System (INIS)

    Buland, P.; Berriaud, C.; Cebe, E.; Livolant, M.

    1975-01-01

    The paper presents the experimental results obtained at Saclay on a HTGR core model and comparisons with analytical results. Two series of horizontal tests have been performed on the shaking table VESUVE: sinusoidal test and time history response. Acceleration of graphite blocks, forces on the boundaries, relative displacement of the core and PCRB model, impact velocity of the blocks on the boundaries were recorded. These tests have shown the strongly non-linear dynamic behaviour of the core. The resonant frequency of the core is dependent on the level of the excitation. These phenomena have been explained by a computer code, which is a lumped mass non-linear model. Good correlation between experimental and analytical results was obtained for impact velocities and forces on the boundaries. This comparison has shown that the damping of the core is a critical parameter for the estimation of forces and velocities. Time history displacement at the level of PCRV was reproduced on the shaking table. The analytical model was applied to this excitation and good agreement was obtained for forces and velocities. (orig./HP) [de

  17. Localization of Vibrating Noise Sources in Nuclear Reactor Cores

    International Nuclear Information System (INIS)

    Hultqvist, Pontus

    2004-09-01

    In this thesis the possibility of locating vibrating noise sources in a nuclear reactor core from the neutron noise has been investigated using different localization methods. The influence of the vibrating noise source has been considered to be a small perturbation of the neutron flux inside the reactor. Linear perturbation theory has been used to construct the theoretical framework upon which the localization methods are based. Two different cases have been considered: one where a one-dimensional one-group model has been used and another where a two-dimensional two-energy group noise simulator has been used. In the first case only one localization method is able to determine the position with good accuracy. This localization method is based on finding roots of an equation and is sensitive to other perturbations of the neutron flux. It will therefore work better with the assistance of approximative methods that reconstruct the noise source to determine if the results are reliable or not. In the two-dimensional case the results are more promising. There are several different localization techniques that reproduce both the vibrating noise source position and the direction of vibration with enough precision. The approximate methods that reconstruct the noise source are substantially better and are able to support the root finding method in a more constructive way. By combining the methods, the results will be more reliable

  18. Power systems with nuclear-electric generators - Modelling methods

    International Nuclear Information System (INIS)

    Valeca, Serban Constantin

    2002-01-01

    This is a vast analysis on the issue of sustainable nuclear power development with direct conclusions regarding the Nuclear Programme of Romania. The work is targeting specialists and decision making boards. Specific to the nuclear power development is its public implication, the public being most often misinformed by non-professional media. The following problems are debated thoroughly: - safety, nuclear risk, respectively, is treated in chapter 1 and 7 aiming at highlighting the quality of nuclear power and consequently paving the way to public acceptance; - the environment considered both as resource of raw materials and medium essential for life continuation, which should be appropriately protected to ensure healthy and sustainable development of human society; its analysis is also presented in chapter 1 and 7, where the problem of safe management of radioactive waste is addressed too; - investigation methods based on information science of nuclear systems, applied in carrying out the nuclear strategy and planning are widely analyzed in the chapter 2, 3 and 6; - optimizing the processes by following up the structure of investment and operation costs, and, generally, the management of nuclear units is treated in the chapter 5 and 7; - nuclear weapon proliferation as a possible consequence of nuclear power generation is treated as a legal issue. The development of Romanian NPP at Cernavoda, practically, the core of the National Nuclear Programme, is described in chapter 8. Actually, the originality of the present work consists in the selection and adaptation from a multitude of mathematical models applicable to the local and specific conditions of nuclear power plant at Cernavoda. The Romanian economy development and power development oriented towards reduction of fossil fuel consumption and protection of environment, most reliably ensured by the nuclear power, is discussed in the frame of the world trends of the energy production. Various scenarios are

  19. Thermal control of high energy nuclear waste, space option. [mathematical models

    Science.gov (United States)

    Peoples, J. A.

    1979-01-01

    Problems related to the temperature and packaging of nuclear waste material for disposal in space are explored. An approach is suggested for solving both problems with emphasis on high energy density waste material. A passive cooling concept is presented which utilized conduction rods that penetrate the inner core. Data are presented to illustrate the effectiveness of the rods and the limit of their capability. A computerized thermal model is discussed and developed for the cooling concept.

  20. Core of a liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Wright, J.R.; McFall, A.

    1975-01-01

    The core of a liquid-cooled nuclear reactor, e.g. of a sodium-cooled fast reactor, is protected in such a way that the recoil wave resulting from loss of coolant in a cooling channel and caused by released gas is limited to a coolant inlet chamber of this cooling channel. The channels essentially consist of the coolant inlet chamber and a fuel chamber - with a fission gas storage plenum - through which the coolant flows. Between the two chambers, a locking device within a tube is provided offering a much larger flow resistance to the backflow of gas or coolant than in flow direction. The locking device may be a hydraulic countertorque control system, e.g. a valvular line. Other locking devices have got radially helical vanes running around an annular flow space. Furthermore, the locking device may consist of a number of needles running parallel to each other and forming a circular grid. Though it can be expanded by the forward flow - the needles are spreading - , it acts as a solid barrier for backflows. (TK) [de

  1. Nuclear physics, neutron physics and nuclear energy. Proceedings

    International Nuclear Information System (INIS)

    Andrejtscheff, W.; Elenkov, D.

    1994-01-01

    The book contains of proceedings of XI International School on Nuclear Physics, Neutron Physics and Nuclear Energy organized traditionally every two years by Bulgarian Academy of Sciences and the Physics Department of Sofia University held near the city of Varna. It provides a good insight to the large range of theoretical and experimental results, prospects, problems, difficulties and challenges which are at the core of nuclear physics today. The efforts and achievements of scientists to search for new phenomena in nuclei at extreme circumstances as superdeformation and band crossing in nuclear structure understanding are widely covered. From this point of view the achievements and future in the field of high-precision γ-spectroscopy are included. Nuclear structure models and methods, models for strong interaction, particle production and properties, resonance theory and its application in reactor physics are comprised also. (V.T.)

  2. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Directory of Open Access Journals (Sweden)

    Gholam Reza Ansarifar

    2015-12-01

    Full Text Available One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC, which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  3. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  4. Nuclear Neutrino Spectra in Late Stellar Evolution

    Science.gov (United States)

    Misch, G. Wendell; Sun, Yang; Fuller, George

    2018-05-01

    Neutrinos are the principle carriers of energy in massive stars, beginning from core carbon burning and continuing through core collapse and after the core bounce. In fact, it may be possible to detect neutrinos from nearby pre-supernova stars. Therefore, it is of great interest to understand the neutrino energy spectra from these stars. Leading up to core collapse, beginning around core silicon burning, nuclei become dominant producers of neutrinos, particularly at high neutrino energy, so a systematic study of nuclear neutrino spectra is desirable. We have done such a study, and we present our sd-shell model calculations of nuclear neutrino energy spectra for nuclei in the mass number range A = 21 - 35. Our study includes neutrinos produced by charged lepton capture, charged lepton emission, and neutral current nuclear deexcitation. Previous authors have tabulated the rates of charged current nuclear weak interactions in astrophysical conditions, but the present work expands on this not only by providing neutrino energy spectra, but also by including the heretofore untabulated neutral current de-excitation neutrino pairs.

  5. Verification Results of Safety-grade Optical Modem for Core Protection Calculator (CPC) in Korea Standard Nuclear Power Plant (KSNP)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jangyeol; Son, Kwangseop; Lee, Youngjun; Cheon, Sewoo; Cha, Kyoungho; Lee, Jangsoo; Kwon, Keechoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    We confirmed that the coverage criteria for a safety-grade optical modem of a Core Protection Calculator is satisfactory using a traceability analysis matrix between high-level requirements and lower-level system test case data set. This paper describes the test environment, test components and items, a traceability analysis, and system tests as a result of system verification and validation based on Software Requirement Specifications (SRS) for a safety-grade optical modem of a Core Protection Calculator (CPC) in a Korea Standard Nuclear Power Plant (KSNP), and Software Design Specifications (SDS) for a safety-grade optical modem of a CPC in a KSNP. All tests were performed according to the test plan and test procedures. Functional testing, performance testing, event testing, and scenario based testing for a safety-grade optical modem of a Core Protection Calculator in a Korea Standard Nuclear Power Plant as a thirty-party verifier were successfully performed.

  6. Development of cutting technique of reactor core internals by CO laser

    International Nuclear Information System (INIS)

    Takano, G.; Beppu, S.; Matsumoto, O.; Sakamoto, N.; Onozawa, T.; Sugihara, M.; Miya, K.

    1995-01-01

    The CO laser is superior in the absorption characteristic to materials to the CO 2 laser due to its shorter wavelength. In consideration of this characteristic Nuclear Power Engineering Corporation is studying this applicability sponsored by the Ministry of International Trade Industry of Japan to cutting of reactor core internals of commercial nuclear power plant. In decommissioning of reactor core internals it is necessary to cut stainless steel plates of 305 mm thick. The authors cut stainless steel plates of up to 310mm thick in air and those of up to 150 mm thick underwater with a 20kW class laser. Further, models simulating key structural elements of PWR core internals were cut and secondary products to clarify the applicability of the CO laser cutting to reactor core internals were evaluated. (author)

  7. Summary of multi-core hardware and programming model investigations

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Suzanne Marie; Pedretti, Kevin Thomas Tauke; Levenhagen, Michael J.

    2008-05-01

    This report summarizes our investigations into multi-core processors and programming models for parallel scientific applications. The motivation for this study was to better understand the landscape of multi-core hardware, future trends, and the implications on system software for capability supercomputers. The results of this study are being used as input into the design of a new open-source light-weight kernel operating system being targeted at future capability supercomputers made up of multi-core processors. A goal of this effort is to create an agile system that is able to adapt to and efficiently support whatever multi-core hardware and programming models gain acceptance by the community.

  8. Core-electron binding energies from self-consistent field molecular orbital theory using a mixture of all-electron real atoms and valence-electron model atoms

    International Nuclear Information System (INIS)

    Quinn, C.M.; Schwartz, M.E.

    1981-01-01

    The chemistry of large systems such as clusters may be readily investigated by valence-electron theories based on model potentials, but such an approach does not allow for the examination of core-electron binding energies which are commonly measured experimentally for such systems. Here we merge our previously developed Gaussian based valence-electron model potential theory with all-electron ab initio theory to allow for the calculation of core orbital binding energies when desired. For the atoms whose cores are to be examined, we use the real nuclear changes, all of the electrons, and the appropriate many-electron basis sets. For the rest of the system we use reduced nuclear charges, the Gaussian based model potentials, only the valence electrons, and appropriate valence-electron basis sets. Detailed results for neutral Al 2 are presented for the cases of all-electron, mixed real--model, and model--model SCF--MO calculations. Several different all-electron and valence electron calculations have been done to test the use of the model potential per se, as well as the effect of basis set choice. The results are in all cases in excellent agreement with one another. Based on these studies, a set of ''double-zeta'' valence and all-electron basis functions have been used for further SCF--MO studies on Al 3 , Al 4 , AlNO, and OAl 3 . For a variety of difference combinations of real and model atoms we find excellent agreement for relative total energies, orbital energies (both core and valence), and Mulliken atomic populations. Finally, direct core-hole-state ionic calculations are reported in detail for Al 2 and AlNO, and noted for Al 3 and Al 4 . Results for corresponding frozen-orbital energy differences, relaxed SCF--MO energy differences, and relaxation energies are in all cases in excellent agreement (never differing by more than 0.07 eV, usually by somewhat less). The study clearly demonstrates the accuracy of the mixed real--model theory

  9. Improvement of core monitoring code cecor by the virtual segmentation of the self powered neutron detector loaded at Korean Standard Nuclear Plant

    International Nuclear Information System (INIS)

    Choi, T.; Jung, Y.S.

    2006-01-01

    Full text: Full text: Korean Standard Nuclear Plant uses Self Powered Neutron Detectors (SPNDs) to measure the neutron flux in the reactor core. The SPND's height is 40 cm and is located axially at the five different positions and 45 radial places. The design code simulated a reactor core is calculated by segmentation of the core. The segmentation is called as 'node', of which size is normally 20 cm. The axial height of the detector is larger than that of the node, and the larger detector's height maybe product some error on the axially complex shape. The analysis with the detector's signals showed some errors at the non-cosine axial flux shape. In order to reduce the errors for the shape, we tried to divide the detector by introducing the virtual boundary in the detector. Then, each axially 5 detectors had two virtual segmentations respectively and the detector's signal was divided by the inputs. So the more virtual detector's signals were gotten, the more accurate axial shape was produced. The result with virtual segmentations in a detector gave less deviation than the case without virtual segmentation (the current model). After the middle of cycle at the initial core specially, the axial neutron flux shape is changed to the saddle type one. The current model gave some error in Root Mean Square (RMS) between the measured value and the calculated one. The virtual segmentation model gave the better agreement at that time

  10. observer-based diagnostics and monitoring of vibrations in nuclear reactor core cooling system

    International Nuclear Information System (INIS)

    Siry, S.A K.

    2007-01-01

    analysis and diagnostics of vibration in industrial systems play a significant rule to prevent severe severe damages . drive shaft vibration is a complicated phenomenon composed of two independent forms of vibrations, translational and torsional. translational vibration measurements in case of the reactor core cooling system are introduced. the system under study consists of the three phase induction motor, flywheel, centrifugal pump, and two coupling between motor-flywheel, and flywheel-pump. this system structure is considered to be one where the blades are pegged into the discs fitting into the shafts. a non-linear model to simulate vibration in the reactor core cooling system will be introduced. simulation results of an operating reactor core cooling system using the actual parameters will be presented to validate the accuracy and reliability of the proposed analytical method the accuracy in analyzing the results depends on the system model. the shortcomings of the conventional model will be avoided through the use of that accurate nonlinear model which improve the simulation of the reactor core cooling system

  11. Three-particle forces and nuclear models

    International Nuclear Information System (INIS)

    Krutov, V.A.

    1980-01-01

    Different nuclear models accounting and unaccounting for three-particle internucleon forces (TIF) are reviewed. At present only two nuclear models use manifestly TIP: the Vautherin-Brink-Skyrme (VBS) model and the model proposed by the author of the review and called the semiphenomenological (SP) nuclear model. There is a short discussion of major drawbacks of models unaccounting for TIF: multiparticle shell model, ''superfluid model'', Harty-Fock calculations with two-particle forces, Bruckner-Hartry-Fock calculations, the relativistic self-consistent nuclear model. The VBS and SP models are discussed in detail. It is concluded, that the employment of TIF even in a very simplified form (extremely short-range) puts away a lot of problems characteristic to models limited by two-particle forces (collapse at iteratious in Hartry-Fock, simultaneous fitting of the binding energy of a nucleus and the binding energy of a nucleon, etc.) and makes it possible to obtain in a rather simple way such nuclear characteristics as nuclear binding energy, nuclear mean square root radii, nucleon density of a nucleus

  12. Evaluation of core modeling effect on transients for multi-flow zone design of SFR

    International Nuclear Information System (INIS)

    Shin, Andong; Choi, Yong Won

    2016-01-01

    SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

  13. Evaluation of core modeling effect on transients for multi-flow zone design of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Choi, Yong Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

  14. Scaling of Core Material in Rubble Mound Breakwater Model Tests

    DEFF Research Database (Denmark)

    Burcharth, H. F.; Liu, Z.; Troch, P.

    1999-01-01

    The permeability of the core material influences armour stability, wave run-up and wave overtopping. The main problem related to the scaling of core materials in models is that the hydraulic gradient and the pore velocity are varying in space and time. This makes it impossible to arrive at a fully...... correct scaling. The paper presents an empirical formula for the estimation of the wave induced pressure gradient in the core, based on measurements in models and a prototype. The formula, together with the Forchheimer equation can be used for the estimation of pore velocities in cores. The paper proposes...... that the diameter of the core material in models is chosen in such a way that the Froude scale law holds for a characteristic pore velocity. The characteristic pore velocity is chosen as the average velocity of a most critical area in the core with respect to porous flow. Finally the method is demonstrated...

  15. Nuclear reactor core

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo; Ishibashi, Yoko; Mochida, Takaaki; Haikawa, Katsumasa; Yamanaka, Akihiro.

    1995-01-01

    A reactor core is radially divided into an inner region, an outer region and an outermost region. As a fuel, three kinds of fuels, namely, a high enrichment degree fuel at 3.4%, a middle enrichment degree fuel at 2.3% and a low enrichment degree at 1.1% of a fuel average enrichment degree of fission product are used. Each of the fuels is bisected to upper and lower portions at an axial center thereof. The difference of average enrichment degrees between upper and lower portions is 0.1% for the high enrichment degree fuel, 0.3% for the middle enrichment degree fuel and 0.2% for the low enrichment degree fuel. In addition, the composition of fuels in each of radial regions comprises 100% of the low enrichment degree fuels in the outermost region, 91% of the higher enrichment degree fuels and 9% of the middle enrichment degree fuels in the outer region, and 34% of the high enrichment degree fuels and 30% of the middle enrichment degree fuels in the inner region. With such a constitution, fuel economy can be improved while maintaining the thermal margin in an initially loaded reactor core of a BWR type reactor. (I.N.)

  16. Diffusion induced nuclear reactions in metals: a possible source of heat in the core

    International Nuclear Information System (INIS)

    Hamza, V.M.; Iyer, S.S.S.

    1989-01-01

    It has recently been proposed that diffusion of light nuclei in metals can give rise to unusual electrical charge distributions in their lattice structures, inducing thereby certain nuclear reactions that are otherwise uncommon. In the light of these results we advance the hypothesis that such nuclear reactions take place in the metal rich core of the earth, based on following observations: 1 - The solubility of hydrogen in metals is relatively high compared to that in silicates. 2 - Studies of rare gas samples in intraplate volcanos and diamonds show that 3 He/ He ratio increases with depth in the mantle. 3 - There are indications that He is positively correlated with enrichment of metals in lavas. We propose that hydrogen incorporated into metallic phases at the time of planetary accretion was carried to the core by downward migration of metal rich melts during the early states of proto-earth. Preliminary estimates suggest that cold fusion reactions can give rise to an average rate of heat generation of 8.2x10 12 W and may thus serve as a supplementary source of energy for the geomagnetic dynamo. (author)

  17. Fluid structure interaction in LMFBR cores modelling by an homogenization method

    International Nuclear Information System (INIS)

    Brochard, D.

    1988-01-01

    The upper plenum of the internals of PWR, the steam generator bundle, the nuclear reactor core, may be schematically represented by a beam bundle immersed in a fluid. The dynamical study of such a system needs to take into account fluid structure interaction. A refined model at the scale of the tubes can be used but leads to a very difficult problem to solve even on the largest computers. The homogenization method allows to have an approximation of the fluid structure interaction for the global behaviour of the bundle. It consists of replacing the heterogeneous physical medium (tubes and fluid) by an equivalent homogeneous medium whose characteristics are determined from the resolution of a set of problems on the elementary cell. The aim of this paper is to present the main steps of the determination of this equivalent medium in the case of small displacements (acoustic behaviour of the fluid). Then an application to LMFBR core geometry has been realised, which shows the lowering effect on eigenfrequencies due to the fluid. Some comparisons with test results will be presented. 6 refs, 7 figs, 2 tabs

  18. LAMBDA-hyperon superfluidity in neutron star cores

    CERN Document Server

    Takatsuka, T

    2000-01-01

    Superfluidity of LAMBDA hyperons in neutron star cores is investigated by a realistic approach to use reliable LAMBDA LAMBDA interactions and the effective mass of LAMBDA based on the G-matrix calculations. It is found that LAMBDA superfluid can exist at rho approx = (rho sub t approx rho sub d) with rho sub t approx = 2 rho sub 0 (rho sub 0 being the nuclear density) and rho sub d approx = (3 - 4.5)rho sub 0 , depending on hyperon core models.

  19. Test model of WWER core

    International Nuclear Information System (INIS)

    Tikhomirov, A. V.; Gorokhov, A. K.

    2007-01-01

    The objective of this paper is creation of precision test model for WWER RP neutron-physics calculations. The model is considered as a tool for verification of deterministic computer codes that enables to reduce conservatism of design calculations and enhance WWER RP competitiveness. Precision calculations were performed using code MCNP5/1/ (Monte Carlo method). Engineering computer package Sapfir 9 5andRC V VER/2/ is used in comparative analysis of the results, it was certified for design calculations of WWER RU neutron-physics characteristic. The object of simulation is the first fuel loading of Volgodon NPP RP. Peculiarities of transition in calculation using MCNP5 from 2D geometry to 3D geometry are shown on the full-scale model. All core components as well as radial and face reflectors, automatic regulation in control and protection system control rod are represented in detail description according to the design. The first stage of application of the model is assessment of accuracy of calculation of the core power. At the second stage control and protection system control rod worth was assessed. Full scale RP representation in calculation using code MCNP5 is time consuming that calls for parallelization of computational problem on multiprocessing computer (Authors)

  20. Studies on WWER core diagnostics

    International Nuclear Information System (INIS)

    Lunin, G.L.; Mitin, V.I.; Bulavin, V.V.

    1987-01-01

    The reliability and safety of nuclear power plants have decisive meaning under the situation that nuclear power generation steadily increases, and among various measures aiming at ensuring the reliability and safety in the operation of nuclear power plants, the countermeasures for protecting reactor core, main process equipment and high pressure circuits from damage have the important role, and the monitoring of condition and the organization of forecast, which are carried out continuously or periodically during the operation of nuclear power stations using the diagnostic expert system specially developed for the purpose, are included in them. Such monitoring enables the early detection of mechanical damage, increase of vibration, defects caused during operation and so on in reactor cores and primary and secondary circuits, and the continuous watching of defect developments. Also boiling in a core is detected, the place of abnormality occurrence is identified, and the intensity and characteristics of boiling are determined, thus the occurrence of dangerous condition is prevented. The developments of an in-core monitoring system and noise diagnostic systems are reported. (Kako, I.)

  1. Radionuclide release and aerosol generation during core debris interactions with concrete

    International Nuclear Information System (INIS)

    Powers, D.A.

    1986-01-01

    During severe accidents at nuclear power plants, it is possible for the reactor fuel to melt and penetrate the reactor vessel. This can lead to vigorous interaction of core materials (UO 2 , ZrO 2 , Zr, and stainless steel) with structural concrete. Sparging of the molten core debris by gases (H 2 O and CO 2 ) liberated from the concrete can lead to rapid release of radionuclides from the core debris. A theoretical description of this release process has been developed and is called the VANESA model. The treatments in the VANESA model of the thermodynamics of radionuclide vaporization and the kinetic barriers to vaporization will be described. Predictions obtained from the model will be compared to the results of tests of core debris/concrete interactions

  2. Gaseous core nuclear-driven engines featuring a self-shutoff mechanism to provide nuclear safety

    International Nuclear Information System (INIS)

    Heidrich, J.; Pettibone, J.; Chow, Tze-Show; Condit, R.; Zimmerman, G.

    1991-11-01

    Nuclear driven engines are described that could be run in either pulsed or steady state modes. In the pulsed mode nuclear energy is released by fissioning of uranium or plutonium in a supercritical assembly of fuel and working gas. In a steady state mode a fuel-gas mixture is injected into a magnetic nozzle where it is compressed into a critical state and produces energy. Engine performance is modeled using a code that calculates hydrodynamics, fission energy production, and neutron transport self-consistently. Results are given demonstrating a large negative temperature coefficient that produces self-shutoff or control of energy production. Reduced fission product inventory and the self-shutoff provide inherent nuclear safety. It is expected that nuclear engine reactor units could be scaled up from about 100 MW e

  3. 78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2013-09-12

    ... 52 [Docket No. PRM-50-105; NRC-2012-0056] In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission. ACTION: Petition for rulemaking; denial...-core thermocouples at different elevations and radial positions throughout the reactor core to enable...

  4. 3D CAD model of the subcritical nuclear reactor of IPN

    International Nuclear Information System (INIS)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

    2016-09-01

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  5. Modeling nuclear processes by Simulink

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Nahrul Khair Alang Md, E-mail: nahrul@iium.edu.my [Faculty of Engineering, International Islamic University Malaysia, Jalan Gombak, Selangor (Malaysia)

    2015-04-29

    Modelling and simulation are essential parts in the study of dynamic systems behaviours. In nuclear engineering, modelling and simulation are important to assess the expected results of an experiment before the actual experiment is conducted or in the design of nuclear facilities. In education, modelling can give insight into the dynamic of systems and processes. Most nuclear processes can be described by ordinary or partial differential equations. Efforts expended to solve the equations using analytical or numerical solutions consume time and distract attention from the objectives of modelling itself. This paper presents the use of Simulink, a MATLAB toolbox software that is widely used in control engineering, as a modelling platform for the study of nuclear processes including nuclear reactor behaviours. Starting from the describing equations, Simulink models for heat transfer, radionuclide decay process, delayed neutrons effect, reactor point kinetic equations with delayed neutron groups, and the effect of temperature feedback are used as examples.

  6. Modeling nuclear processes by Simulink

    International Nuclear Information System (INIS)

    Rashid, Nahrul Khair Alang Md

    2015-01-01

    Modelling and simulation are essential parts in the study of dynamic systems behaviours. In nuclear engineering, modelling and simulation are important to assess the expected results of an experiment before the actual experiment is conducted or in the design of nuclear facilities. In education, modelling can give insight into the dynamic of systems and processes. Most nuclear processes can be described by ordinary or partial differential equations. Efforts expended to solve the equations using analytical or numerical solutions consume time and distract attention from the objectives of modelling itself. This paper presents the use of Simulink, a MATLAB toolbox software that is widely used in control engineering, as a modelling platform for the study of nuclear processes including nuclear reactor behaviours. Starting from the describing equations, Simulink models for heat transfer, radionuclide decay process, delayed neutrons effect, reactor point kinetic equations with delayed neutron groups, and the effect of temperature feedback are used as examples

  7. Analytical one-dimensional frequency response and stability model for PWR nuclear power plants

    International Nuclear Information System (INIS)

    Hoeld, A.

    1975-01-01

    A dynamic model for PWR nuclear power plants is presented. The plant is assumed to consist of one-dimensional single-channel core, a counterflow once-through steam generator (represented by two nodes according to the nonboiling and boiling region) and the necessary connection coolant lines. The model describes analytically the frequency response behaviour of important parameters of such a plant with respect to perturbations in reactivity, subcooling or mass flow (both at the entrances to the reactor core and/or the secondary steam generator side), the perturbations in steam load or system pressure (on the secondary side of the steam generator). From corresponding 'open' loop considerations it can then be concluded - by applying the Nyquist criterion - upon the degree of the stability behaviour of the underlying system. Based on this theoretical model, a computer code named ADYPMO has been established. From the knowledge of the frequency response behaviour of such a system, the corresponding transient behaviour with respect to a stepwise or any other perturbation signal can also be calculated by applying an appropriate retransformation method, e.g. by using digital code FRETI. To demonstrate this procedure, a transient experimental curve measured during the pre-operational test period at the PWR nuclear power plant KKS Stade was recalculated using the combination ADYPMO-FRETI. Good agreement between theoretical calculations and experimental results give an insight into the validity and efficiency of the underlying theoretical model and the applied retransformation method. (Auth.)

  8. About the application of MCNP4 code in nuclear reactor core design calculations

    International Nuclear Information System (INIS)

    Svarny, J.

    2000-01-01

    This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

  9. Integral Full Core Multi-Physics PWR Benchmark with Measured Data

    Energy Technology Data Exchange (ETDEWEB)

    Forget, Benoit; Smith, Kord; Kumar, Shikhar; Rathbun, Miriam; Liang, Jingang

    2018-04-11

    In recent years, the importance of modeling and simulation has been highlighted extensively in the DOE research portfolio with concrete examples in nuclear engineering with the CASL and NEAMS programs. These research efforts and similar efforts worldwide aim at the development of high-fidelity multi-physics analysis tools for the simulation of current and next-generation nuclear power reactors. Like all analysis tools, verification and validation is essential to guarantee proper functioning of the software and methods employed. The current approach relies mainly on the validation of single physic phenomena (e.g. critical experiment, flow loops, etc.) and there is a lack of relevant multiphysics benchmark measurements that are necessary to validate high-fidelity methods being developed today. This work introduces a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading and re-loading patterns. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from 58 instrumented assemblies. The benchmark description is now available online and has been used by many groups. However, much work remains to be done on the quantification of uncertainties and modeling sensitivities. This work aims to address these deficiencies and make this benchmark a true non-proprietary international benchmark for the validation of high-fidelity tools. This report details the BEAVRS uncertainty quantification for the first two cycle of operations and serves as the final report of the project.

  10. Modelling the carbonation of cementitious matrixes by means of the unreacted-core model, UR-CORE

    International Nuclear Information System (INIS)

    Castellote, M.; Andrade, C.

    2008-01-01

    This paper presents a model for the carbonation of cementitious matrixes (UR-CORE). The model is based on the principles of the 'unreacted-core' systems, typical of chemical engineering processes, in which the reacted product remains in the solid as a layer of inert ash, adapted for the specific case of carbonation. Development of the model has been undertaken in three steps: 1) Establishment of the controlling step in the global carbonation rate, by using data of fractional conversion of different phases of the cementitious matrixes, obtained by the authors through neutron diffraction data experiments, and reported in [M. Castellote, C. Andrade, X. Turrillas, J. Campo, G. Cuello, Accelerated carbonation of cement pastes in situ monitored by neutron diffraction, Cem. Concr. Res. (2008), doi:10.1016/j.cemconres.2008.07.002]. 2) Then, the model has been adapted and applied to the cementitious materials using different concentrations of CO 2 , with the introduction of the needed assumptions and factors. 3) Finally, the model has been validated with laboratory data at different concentrations (taken from literature) and for long term natural exposure of concretes. As a result, the model seems to be reliable enough to be applied to cementitious materials, being able to extrapolate the results from accelerated tests in any conditions to predict the rate of carbonation in natural exposure, being restricted, at present stage, to conditions with a constant relative humidity

  11. Core calculations of JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)

  12. Preliminary Nuclear Analysis for the HANARO Fuel Element with Burnable Absorber

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, So Young; In, Won Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Burnable absorber is used for reducing reactivity swing and power peaking in high performance research reactors. Development of the HANARO fuel element with burnable absorber was started in the U-Mo fuel development program at HANARO, but detailed full core analysis was not performed because the current HANARO fuel management system is uncertain to analysis the HANARO core with burnable absorber. A sophisticated reactor physics system is required to analysis the core. The McCARD code was selected and the detailed McCARD core models, in which the basic HANARO core model was developed by one of the McCARD developers, are used in this study. The development of nuclear fuel requires a long time and correct developing direction especially by the nuclear analysis. This paper presents a preliminary nuclear analysis to promote the fuel development. Based on the developed fuel, the further nuclear analysis will improve reactor performance and safety. Basic nuclear analysis for the HANARO and the AHR were performed for getting the proper fuel elements with burnable absorber. Addition of 0.3 - 0.4% Cd to the fuel meat is promising for the current HANARO fuel element. Small addition of burnable absorber may not change any fuel characteristics of the HANARO fuel element, but various basic tests and irradiation tests at the HANARO core are required.

  13. A nuclear heuristic for application to metaheuristics in-core fuel management optimization

    Energy Technology Data Exchange (ETDEWEB)

    Meneses, Anderson Alvarenga de Moura, E-mail: ameneses@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program; Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Gambardella, Luca Maria, E-mail: luca@idsia.c [Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Schirru, Roberto, E-mail: schirru@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program

    2009-07-01

    The In-Core Fuel Management Optimization (ICFMO) is a well-known problem of nuclear engineering whose features are complexity, high number of feasible solutions, and a complex evaluation process with high computational cost, thus it is prohibitive to have a great number of evaluations during an optimization process. Heuristics are criteria or principles for deciding which among several alternative courses of action are more effective with respect to some goal. In this paper, we propose a new approach for the use of relational heuristics for the search in the ICFMO. The Heuristic is based on the reactivity of the fuel assemblies and their position into the reactor core. It was applied to random search, resulting in less computational effort concerning the number of evaluations of loading patterns during the search. The experiments demonstrate that it is possible to achieve results comparable to results in the literature, for future application to metaheuristics in the ICFMO. (author)

  14. A nuclear heuristic for application to metaheuristics in-core fuel management optimization

    International Nuclear Information System (INIS)

    Meneses, Anderson Alvarenga de Moura; Gambardella, Luca Maria; Schirru, Roberto

    2009-01-01

    The In-Core Fuel Management Optimization (ICFMO) is a well-known problem of nuclear engineering whose features are complexity, high number of feasible solutions, and a complex evaluation process with high computational cost, thus it is prohibitive to have a great number of evaluations during an optimization process. Heuristics are criteria or principles for deciding which among several alternative courses of action are more effective with respect to some goal. In this paper, we propose a new approach for the use of relational heuristics for the search in the ICFMO. The Heuristic is based on the reactivity of the fuel assemblies and their position into the reactor core. It was applied to random search, resulting in less computational effort concerning the number of evaluations of loading patterns during the search. The experiments demonstrate that it is possible to achieve results comparable to results in the literature, for future application to metaheuristics in the ICFMO. (author)

  15. Assessment of water hammer effects on boiling water nuclear reactor core dynamics

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2007-01-01

    Full Text Available Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

  16. Core baffle for nuclear reactors

    International Nuclear Information System (INIS)

    Machado, O.J.; Berringer, R.T.

    1977-01-01

    The invention concerns the design of the core of a LWR with a large number of fuel assemblies formed by fuel rods and kept in position by spacer grids. According to the invention, at the level of the spacer grids match plates are mounted with openings so the flow of coolant directed upwards will not be obstructed and a parallel bypass will be obtained in the space between the core barrel and the baffle plates. In case of an accident, this configuration reduces or avoids damage from overpressure reactions. (HP) [de

  17. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  18. JSPS-CAS Core University Program seminar on summary of 10-year collaborations in plasma and nuclear fusion research area

    International Nuclear Information System (INIS)

    Toi, Kazuo; Wang Kongjia

    2011-07-01

    The JSPS-CAS Core University Program (CUP) seminar on “Summary of 10-year Collaborations in Plasma and Nuclear Fusion Research Area” was held from March 9 to March 11, 2011 in the Okinawa Prefectural Art Museum, Naha city, Okinawa, Japan. The collaboration program on plasma and nuclear fusion started from 2001 under the auspices of Japanese Society of Promotion of Science (JSPS) and Chinese Academy of Sciences (CAS). This year is the last year of the CUP. This seminar was organized in the framework of the CUP. In the seminar, 29 oral talks were presented, having 14 Chinese and 30 Japanese participants. These presentations covered key topics related to the collaboration categories: (1) improvement of core plasma properties, (2) basic research on fusion reactor technologies, and (3) theory and numerical simulation. This seminar aims at summarizing the results obtained through the collaborations for 10 years, and discussing future prospects of China-Japan collaboration in plasma and nuclear fusion research areas. (author)

  19. Nuclear force from string theory

    International Nuclear Information System (INIS)

    Hashimoto, Koji; Sakai, Tadakatsu; Sugimoto, Shigeki

    2009-01-01

    We compute the nuclear force in a holographic model of QCD on the basis of a D4-D8 brane configuration in type IIA string theory. The repulsive core of nucleons is important in nuclear physics, but its origin has not been well understood in strongly coupled QCD. We find that the string theory via gauge/string duality deduces this repulsive core at a short distance between nucleons. Since baryons in the model are realized as solitons given by Yang-Mills instanton configuration on flavor D8-branes, ADHM construction of two instantons probes well the nucleon interaction at short scale, which provides the nuclear force quantitatively. We obtain a central force, as well as a tensor force, which is strongly repulsive as suggested in experiments and lattice results. In particular, the nucleon-nucleon potential V(r) (as a function of the distance) scales as r -2 , which is peculiar to the holographic model. We compare our results with the one-boson exchange model using the nucleon-nucleon-meson coupling obtained in our previous paper. (author)

  20. Normal Mode Derived Models of the Physical Properties of Earth's Outer Core

    Science.gov (United States)

    Irving, J. C. E.; Cottaar, S.; Lekic, V.; Wu, W.

    2017-12-01

    Earth's outer core, the largest reservoir of metal in our planet, is comprised of an iron alloy of an uncertain composition. Its dynamical behaviour is responsible for the generation of Earth's magnetic field, with convection driven both by thermal and chemical buoyancy fluxes. Existing models of the seismic velocity and density of the outer core exhibit some variation, and there are only a small number of models which aim to represent the outer core's density.It is therefore important that we develop a better understanding of the physical properties of the outer core. Though most of the outer core is likely to be well mixed, it is possible that the uppermost outer core is stably stratified: it may be enriched in light elements released during the growth of the solid, iron enriched, inner core; by elements dissolved from the mantle into the outer core; or by exsolution of compounds previously dissolved in the liquid metal which will eventually be swept into the mantle. The stratified layer may host MAC or Rossby waves and it could impede communication between the chemically differentiated mantle and outer core, including screening out some of the geodynamo's signal. We use normal mode center frequencies to estimate the physical properties of the outer core in a Bayesian framework. We estimate the mineral physical parameters needed to best produce velocity and density models of the outer core which are consistent with the normal mode observations. We require that our models satisfy realistic physical constraints. We create models of the outer core with and without a distinct uppermost layer and assess the importance of this region.Our normal mode-derived models are compared with observations of body waves which travel through the outer core. In particular, we consider SmKS waves which are especially sensitive to the uppermost outer core and are therefore an important way to understand the robustness of our models.

  1. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  2. Effect of core burnup on the dynamic behavior of fast reactors

    International Nuclear Information System (INIS)

    Ilberg, D.; Saphier, D.; Yiftah, S.

    1977-01-01

    Performance of a dynamic analysis, taking burnup changes into account, requires fission-product nuclear data of relatively small uncertainty, suitable burnup calculation models, and dynamic computer programs. These were prepared and used with the following results: (1) Significant changes in static and dynamic parameters were observed when investigating the effect of burnup. These changes were found to be larger than differences introduced by the uncertainty of the fission-product nuclear data. (2) A one-dimensional burnup computer program was prepared. It was found that a burnup model based on the generalized radioactive decay scheme is suitable for accurate fast reactor calculations. (3) Space-time dynamic calculations of fast reactors having different burnup levels were performed. The stability difference between ''clean'' and high burnup cores is greater when local rather than uniform perturbations are inserted along the entire core length. The magnitude by which the ''end-of-life'' core increases the transient excursion over that of the clean core depends on the particular region in which the perturbation is inserted. The end-of-life core will magnify the transient excursion more than the clean core whenever the perturbation is inserted into a region having a higher adjoint flux level than that of the clean core. However, when a reactor safety system operates successfully, the difference in the temperature transient of the clean and end-of-life cores will be relatively small. It is suggested that only the analysis of large local perturbations be performed for end-of-life cores as well as for clean cores in the safety evaluation of fast reactors

  3. Uncertainties in Nuclear Proliferation Modeling

    International Nuclear Information System (INIS)

    Kim, Chul Min; Yim, Man-Sung; Park, Hyeon Seok

    2015-01-01

    There have been various efforts in the research community to understand the determinants of nuclear proliferation and develop quantitative tools to predict nuclear proliferation events. Such systematic approaches have shown the possibility to provide warning for the international community to prevent nuclear proliferation activities. However, there are still large debates for the robustness of the actual effect of determinants and projection results. Some studies have shown that several factors can cause uncertainties in previous quantitative nuclear proliferation modeling works. This paper analyzes the uncertainties in the past approaches and suggests future works in the view of proliferation history, analysis methods, and variable selection. The research community still lacks the knowledge for the source of uncertainty in current models. Fundamental problems in modeling will remain even other advanced modeling method is developed. Before starting to develop fancy model based on the time dependent proliferation determinants' hypothesis, using graph theory, etc., it is important to analyze the uncertainty of current model to solve the fundamental problems of nuclear proliferation modeling. The uncertainty from different proliferation history coding is small. Serious problems are from limited analysis methods and correlation among the variables. Problems in regression analysis and survival analysis cause huge uncertainties when using the same dataset, which decreases the robustness of the result. Inaccurate variables for nuclear proliferation also increase the uncertainty. To overcome these problems, further quantitative research should focus on analyzing the knowledge suggested on the qualitative nuclear proliferation studies

  4. Collective models of transition nuclei Pt. 2

    International Nuclear Information System (INIS)

    Dombradi, Zs.

    1982-01-01

    The models describing the even-odd and odd-odd transition nuclei (nuclei of moderate ground state deformation) are reviewed. The nuclear core is described by models of even-even nuclei, and the interaction of a single particle and the core is added. Different models of particle-core coupling (phenomenological models, collective models, nuclear field theory, interacting boson-fermion model, vibration nucleon cluster model) and their results are discussed. New developments like dynamical supersymmetry and new research trends are summarized. (D.Gy.)

  5. Thermal hydraulics model for Sandia's annular core research reactor

    International Nuclear Information System (INIS)

    Rao, Dasari V.; El-Genk, Mohamed S.; Rubio, Reuben A.; Bryson, James W.; Foushee, Fabian C.

    1988-01-01

    A thermal hydraulics model was developed for the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. The coupled mass, momentum and energy equations for the core were solved simultaneously using an explicit forward marching numerical technique. The model predictions of the temperature rise across the central channel of the ACRR core were within ± 10 percent agreement with the in-core temperature measurements. The model was then used to estimate the coolant mass flow rate and the axial distribution of the cladding surface temperature in the central and average channels as functions of the operating power and the water inlet subcooling. Results indicated that subcooled boiling occurs at the cladding surface in the central channels of the ACRR at power levels in excess of 0.5 MW. However, the high heat transfer coefficient due to subcooled boiling causes the cladding temperature along most of the active fuel rod region to be quite uniform and to increase very little with the reactor power. (author)

  6. Simulation an Accelerator driven Subcritical Reactor core with thorium fuel

    International Nuclear Information System (INIS)

    Shirmohammadi, L.; Pazirandeh, A.

    2011-01-01

    The main purpose of this work is simulation An Accelerator driven Subcritical core with Thorium as a new generation nuclear fuel. In this design core , A subcritical core coupled to an accelerator with proton beam (E p =1 GeV) is simulated by MCNPX code .Although the main purpose of ADS systems are transmutation and use MA (Minor Actinides) as a nuclear fuel but another use of these systems are use thorium fuel. This simulated core has two fuel assembly type : (Th-U) and (U-Pu) . Consequence , Neutronic parameters related to ADS core are calculated. It has shown that Thorium fuel is use able in this core and less nuclear waste ,Although Iran has not Thorium reserves but study on Thorium fuel cycle can open a new horizontal in use nuclear energy as a clean energy and without nuclear waste

  7. GARLIC-B. A digital code for real-time calculation of the transient behaviour of nodal and global core and plant parameters of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.; Hoeld, A.; Lupas, O.

    1982-04-01

    A program description of the code GARLIC-B is given. The code is based on a nonlinear transient model for BWR nuclear power plants which consist of a 3D-core, a top plenum, steam removal and feed water systems and a downcomer with main coolant recirculation pumps. The core is subdivided into a number of superboxes and flow channels with different coolant mass flow rates. Subcooled boiling within these channels has an important reactivity feed back effect and has to be taken also into account. The code computes the local and global core and plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. (orig./GL) [de

  8. Sensitivity of control times in function of core parameters and oscillations control in thermal nuclear systems

    International Nuclear Information System (INIS)

    Amorim, E.S. do; D'Oliveira, A.B.; Galvao, O.B.; Oyama, K.

    1981-03-01

    Sensitivity of control times to variation of a thermal reactor core parameters is defined by suitable changes in the power coefficient, core size and fuel enrichment. A control strategy is developed based on control theory concepts and on considerations of the physics of the problem. Digital diffusion theory simulation is described which tends to verify the control concepts considered, face dumped oscillations introduced in one thermal nuclear power system. The effectivity of the control actions, in terms of eliminating oscillations, provided guidelines for the working-group engaged in the analysis of the control rods and its optimal performance. (Author) [pt

  9. A core concept for the self-consistent nuclear energy system based on the promising future technology

    International Nuclear Information System (INIS)

    Arie, K.; Suzuki, M.; Kawashima, M.; Igashira, M.; Shimizu, A.; Fujii-e, Y.

    1995-01-01

    Feasibility of FP burning while maintaining fuel breeding capability for the Self-Consistent Nuclear Energy System is evaluated through neutron balance and a fast reactor core. It is shown that all radioactive FPs produced by itself can be burnt by a fast reactor while maintaining breeding capability, assuming separation of radioactive FP and stable FP isotopes. Assuming that the recovery system of fuel and FPs to be burnt is based on a pyro-chemical process, the major long-lived FPs of I, Pd, Tc, Sn, Se can be burnt with keeping breeding capability by suitability arranging materials in the fast reactor core. (Author)

  10. Three-dimensional transport coefficient model and prediction-correction numerical method for thermal margin analysis of PWR cores

    International Nuclear Information System (INIS)

    Chiu, C.

    1981-01-01

    Combustion Engineering Inc. designs its modern PWR reactor cores using open-core thermal-hydraulic methods where the mass, momentum and energy equations are solved in three dimensions (one axial and two lateral directions). The resultant fluid properties are used to compute the minimum Departure from Nuclear Boiling Ratio (DNBR) which ultimately sets the power capability of the core. The on-line digital monitoring and protection systems require a small fast-running algorithm of the design code. This paper presents two techniques used in the development of the on-line DNB algorithm. First, a three-dimensional transport coefficient model is introduced to radially group the flow subchannel into channels for the thermal-hydraulic fluid properties calculation. Conservation equations of mass, momentum and energy for this channels are derived using transport coefficients to modify the calculation of the radial transport of enthalpy and momentum. Second, a simplified, non-iterative numerical method, called the prediction-correction method, is applied together with the transport coefficient model to reduce the computer execution time in the determination of fluid properties. Comparison of the algorithm and the design thermal-hydraulic code shows agreement to within 0.65% equivalent power at a 95/95 confidence/probability level for all normal operating conditions of the PWR core. This algorithm accuracy is achieved with 1/800th of the computer processing time of its parent design code. (orig.)

  11. FLICA-4 (version 1). A computer code for three dimensional thermal analysis of nuclear reactor cores

    International Nuclear Information System (INIS)

    Raymond, P.; Allaire, G.; Boudsocq, G.; Caruge, D.; Gramont, T. de; Toumi, I.

    1995-01-01

    FLICA-4 is a thermal-hydraulic computer code, developed at the French Atomic Energy Commission (CEA) for three-dimensional steady-state or transient two-phase flow, and aimed at design and safety thermal analysis of nuclear reactor cores. It is available for various UNIX workstations and CRAY computers under UNICOS.It is based on four balance equations which include three balance equations for the mixture and a mass balance equation for the less concentrated phase which allows for the calculation of non equilibrium flows such as sub-cooled boiling and superheated steam. A drift velocity model takes into account the velocity unbalance between phases. The equations are solved using a finite volume numerical scheme. Typical running time, specific features (coupling with other codes) and auxiliary programs are presented. 1 tab., 9 refs

  12. Modeling the Power Variability of Core Speed Scaling on Homogeneous Multicore Systems

    Directory of Open Access Journals (Sweden)

    Zhihui Du

    2017-01-01

    Full Text Available We describe a family of power models that can capture the nonuniform power effects of speed scaling among homogeneous cores on multicore processors. These models depart from traditional ones, which assume that individual cores contribute to power consumption as independent entities. In our approach, we remove this independence assumption and employ statistical variables of core speed (average speed and the dispersion of the core speeds to capture the comprehensive heterogeneous impact of subtle interactions among the underlying hardware. We systematically explore the model family, deriving basic and refined models that give progressively better fits, and analyze them in detail. The proposed methodology provides an easy way to build power models to reflect the realistic workings of current multicore processors more accurately. Moreover, unlike the existing lower-level power models that require knowledge of microarchitectural details of the CPU cores and the last level cache to capture core interdependency, ours are easier to use and scalable to emerging and future multicore architectures with more cores. These attributes make the models particularly useful to system users or algorithm designers who need a quick way to estimate power consumption. We evaluate the family of models on contemporary x86 multicore processors using the SPEC2006 benchmarks. Our best model yields an average predicted error as low as 5%.

  13. Development of real-time core monitoring system models with accuracy-enhanced neural network and its application

    International Nuclear Information System (INIS)

    Koo, Bon Hyun

    1994-02-01

    In a complicated system like pressurized water reactor, a number of key safety parameters need to be selected to represent the reactor systems safety. It could be more effective for the reactor safety to make the key safety parameters in real-time available directly to the reactor operator. Direct representation of key safety parameters is also desirable in the view of reactor core design since it could reduce unnecessary margins for various components of uncertainties. In this thesis, real-time core monitoring system models have been developed with use of artificial neural networks for the prediction of nuclear hot channel factor (HCF) and core departure from nucleate boiling ratio (DNBR) which are known to be the fundamental core safety parameters for pressurized water reactors. Artificial neural network algorithm, has been shown to be successful for the conservative and accurate prediction of the HCF and DNBR. For the development of system models, training patterns were generated using the FLAIR and COBRAIV-i computer codes for the HCF and DNBR. The selected input variables were the core power, reactor coolant flow, temperature, pressure, power distribution, boron concentration, and rod position. The developed system models could replace the existing core monitoring systems and then afford a better efficiency by using the additional margin which otherwise needs to be reserved for various unidentified uncertainties. Several variations of the neural network technique have been proposed and compared based on numerical experiments. The neural network can be augmented by use of a functional link to improve the performance of the network model. The functional link is found to be very effective especially when the relationship between the input parameters and the output parameters is overly complicated such as in the core HCF and DNBR. For the further enhancement of DNBR accuracy, two-fold weight sets were used. The coarse weight set can provide a quick and

  14. Toward full MOX core design

    International Nuclear Information System (INIS)

    Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

    1999-01-01

    This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

  15. Global nuclear material control model

    International Nuclear Information System (INIS)

    Dreicer, J.S.; Rutherford, D.A.

    1996-01-01

    The nuclear danger can be reduced by a system for global management, protection, control, and accounting as part of a disposition program for special nuclear materials. The development of an international fissile material management and control regime requires conceptual research supported by an analytical and modeling tool that treats the nuclear fuel cycle as a complete system. Such a tool must represent the fundamental data, information, and capabilities of the fuel cycle including an assessment of the global distribution of military and civilian fissile material inventories, a representation of the proliferation pertinent physical processes, and a framework supportive of national or international perspective. They have developed a prototype global nuclear material management and control systems analysis capability, the Global Nuclear Material Control (GNMC) model. The GNMC model establishes the framework for evaluating the global production, disposition, and safeguards and security requirements for fissile nuclear material

  16. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  17. A new thermal-hydraulic core module based on the drift-flux model for the DSNP

    International Nuclear Information System (INIS)

    Silverman, I.; Shapira, M.; Saphier, D.; Elias, E.

    1996-01-01

    As a part of expanding the capabilities of the reactor calculations group at Soreq - NRC a new core fuel channel module is under development. The module solves the energy equations inside the fuel rod and mass, momentum and energy equations in the coolant channel. The module uses an approximation to the drift-flux model for the solution of the coolant conditions. This module is a part of DSNP library of modules and is used in the transient simulation of nuclear power plants. Several test cases were executed simulating the AP600 PWR. Comparison of the channel model with COBRA-4I and RELAP-5 calculations have shown good agreement. It was found that the previous homogeneous equilibrium model produced adequate results for power plant simulation until boiling conditions appear in a fuel channel (authors)

  18. A new thermal-hydraulic core module based on the drift-flux model for the DSNP

    Energy Technology Data Exchange (ETDEWEB)

    Silverman, I; Shapira, M; Saphier, D [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center; Elias, E [Technion-Israel Inst. of Tech., Haifa (Israel). Dept. of Mechanical Engineering

    1996-12-01

    As a part of expanding the capabilities of the reactor calculations group at Soreq - NRC a new core fuel channel module is under development. The module solves the energy equations inside the fuel rod and mass, momentum and energy equations in the coolant channel. The module uses an approximation to the drift-flux model for the solution of the coolant conditions. This module is a part of DSNP library of modules and is used in the transient simulation of nuclear power plants. Several test cases were executed simulating the AP600 PWR. Comparison of the channel model with COBRA-4I and RELAP-5 calculations have shown good agreement. It was found that the previous homogeneous equilibrium model produced adequate results for power plant simulation until boiling conditions appear in a fuel channel (authors).

  19. A Practical Core Loss Model for Filter Inductors of Power Electronic Converters

    DEFF Research Database (Denmark)

    Matsumori, Hiroaki; Shimizu, Toshihisa; Wang, Xiongfei

    2018-01-01

    This paper proposes a core loss model for filter inductors of power electronic converters. The model allows a computationally efficient analysis on the core loss of the inductor under the square voltage excitation and the premagnetization condition. First, the core loss of the filter inductor under...... buck chopper excitation is evaluated with the proposed model and compared with the conventional methods. The comparison shows that the proposed method results in a better core loss prediction under the premagnetized condition than that of conventional alternatives. Then, the core loss of the filter...... inductor with the pulsewidth modulated inverter excitation is evaluated, which shows that the proposed model not only accurately predicts the core loss but also identifies the hysteresis loss part. These results demonstrate that the approach can further be used for the development of magnetic materials...

  20. Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

    International Nuclear Information System (INIS)

    Sevon, T.

    2005-11-01

    In a hypothetical severe accident in a nuclear power plant, the molten core of the reactor may flow onto the concrete floor of containment building. This would cause a molten core . concrete interaction (MCCI), in which the heat transfer from the hot melt to the concrete would cause melting of the concrete. In assessing the safety of nuclear reactors, it is important to know the consequences of such an interaction. As background to the subject, this publication includes a description of the core melt stabilization concept of the European Pressurized water Reactor (EPR), which is being built in Olkiluoto in Finland. The publication includes a description of the basic theory of the interaction and the process of spalling or cracking of concrete when it is heated rapidly. A literature survey and some calculations of the physical properties of concrete and corium. concrete mixtures at high temperatures have been conducted. In addition, an equation is derived for conservative calculation of the maximum possible concrete ablation depth. The publication also includes a literature survey of experimental research on the subject of the MCCI and discussion of the results and deficiencies of the experiments. The main result of this work is the general design of an experimental facility to examine the interaction of molten metals and concrete. The main objective of the experiments is to assess the probability of spalling, or cracking, of concrete under pouring of molten material. A program of five experiments has been designed, and pre-test calculations of the experiments have been conducted with MELCOR 1.8.5 accident analysis program and conservative analytic calculations. (orig.)

  1. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  2. NEW EQUATIONS OF STATE BASED ON THE LIQUID DROP MODEL OF HEAVY NUCLEI AND QUANTUM APPROACH TO LIGHT NUCLEI FOR CORE-COLLAPSE SUPERNOVA SIMULATIONS

    International Nuclear Information System (INIS)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2013-01-01

    We construct new equations of state for baryons at subnuclear densities for the use in core-collapse simulations of massive stars. The abundance of various nuclei is obtained together with thermodynamic quantities. A model free energy is constructed, based on the relativistic mean field theory for nucleons and the mass formula for nuclei with the proton number up to ∼1000. The formulation is an extension of the previous model, in which we adopted the liquid drop model to all nuclei under the nuclear statistical equilibrium. We reformulate the new liquid drop model so that the temperature dependences of bulk energies could be taken into account. Furthermore, we extend the region in the nuclear chart, in which shell effects are included, by using theoretical mass data in addition to experimental ones. We also adopt a quantum-theoretical mass evaluation of light nuclei, which incorporates the Pauli- and self-energy shifts that are not included in the ordinary liquid drop model. The pasta phases for heavy nuclei are taken into account in the same way as in the previous model. We find that the abundances of heavy nuclei are modified by the shell effects of nuclei and temperature dependence of bulk energies. These changes may have an important effect on the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. The abundances of light nuclei are also modified by the new mass evaluation, which may affect the heating and cooling rates of supernova cores and shocked envelopes

  3. New Equations of State Based on the Liquid Drop Model of Heavy Nuclei and Quantum Approach to Light Nuclei for Core-collapse Supernova Simulations

    Science.gov (United States)

    Furusawa, Shun; Sumiyoshi, Kohsuke; Yamada, Shoichi; Suzuki, Hideyuki

    2013-08-01

    We construct new equations of state for baryons at subnuclear densities for the use in core-collapse simulations of massive stars. The abundance of various nuclei is obtained together with thermodynamic quantities. A model free energy is constructed, based on the relativistic mean field theory for nucleons and the mass formula for nuclei with the proton number up to ~1000. The formulation is an extension of the previous model, in which we adopted the liquid drop model to all nuclei under the nuclear statistical equilibrium. We reformulate the new liquid drop model so that the temperature dependences of bulk energies could be taken into account. Furthermore, we extend the region in the nuclear chart, in which shell effects are included, by using theoretical mass data in addition to experimental ones. We also adopt a quantum-theoretical mass evaluation of light nuclei, which incorporates the Pauli- and self-energy shifts that are not included in the ordinary liquid drop model. The pasta phases for heavy nuclei are taken into account in the same way as in the previous model. We find that the abundances of heavy nuclei are modified by the shell effects of nuclei and temperature dependence of bulk energies. These changes may have an important effect on the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. The abundances of light nuclei are also modified by the new mass evaluation, which may affect the heating and cooling rates of supernova cores and shocked envelopes.

  4. NEW EQUATIONS OF STATE BASED ON THE LIQUID DROP MODEL OF HEAVY NUCLEI AND QUANTUM APPROACH TO LIGHT NUCLEI FOR CORE-COLLAPSE SUPERNOVA SIMULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Furusawa, Shun; Yamada, Shoichi [Advanced Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Sumiyoshi, Kohsuke [Numazu College of Technology, Ooka 3600, Numazu, Shizuoka 410-8501 (Japan); Suzuki, Hideyuki, E-mail: furusawa@heap.phys.waseda.ac.jp [Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan)

    2013-08-01

    We construct new equations of state for baryons at subnuclear densities for the use in core-collapse simulations of massive stars. The abundance of various nuclei is obtained together with thermodynamic quantities. A model free energy is constructed, based on the relativistic mean field theory for nucleons and the mass formula for nuclei with the proton number up to {approx}1000. The formulation is an extension of the previous model, in which we adopted the liquid drop model to all nuclei under the nuclear statistical equilibrium. We reformulate the new liquid drop model so that the temperature dependences of bulk energies could be taken into account. Furthermore, we extend the region in the nuclear chart, in which shell effects are included, by using theoretical mass data in addition to experimental ones. We also adopt a quantum-theoretical mass evaluation of light nuclei, which incorporates the Pauli- and self-energy shifts that are not included in the ordinary liquid drop model. The pasta phases for heavy nuclei are taken into account in the same way as in the previous model. We find that the abundances of heavy nuclei are modified by the shell effects of nuclei and temperature dependence of bulk energies. These changes may have an important effect on the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. The abundances of light nuclei are also modified by the new mass evaluation, which may affect the heating and cooling rates of supernova cores and shocked envelopes.

  5. Two-dimensional horizontal model seismic test and analysis for HTGR core

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1988-05-01

    The resistance against earthquakes of high-temperature gas-cooled reactor (HTGR) core with block-type fuels is not fully ascertained yet. Seismic studies must be made if such a reactor plant is to be installed in areas with frequent earthquakes. The paper presented the test results of seismic behavior of a half scale two-dimensional horizontal slice core model and analysis. The following is a summary of the more important results. (1) When the core is subjected to the single axis excitation and simultaneous two-axis excitations to the core across-corners, it has elliptical motion. The core stays lumped motion at the low excitation frequencies. (2) When the load is placed on side fixed reflector blocks from outside to the core center, the core displacement and reflector impact reaction force decrease. (3) The maximum displacement occurs at simultaneous two-axis excitations. The maximum displacement occurs at the single axis excitation to the core across-flats. (4) The results of two-dimensional horizontal slice core model was compared with the results of two-dimensional vertical one. It is clarified that the seismic response of actual core can be predicted from the results of two-dimensional vertical slice core model. (5) The maximum reflector impact reaction force for seismic waves was below 60 percent of that for sinusoidal waves. (6) Vibration behavior and impact response are in good agreement between test and analysis. (author)

  6. Rift Valley fever phlebovirus NSs protein core domain structure suggests molecular basis for nuclear filaments.

    Science.gov (United States)

    Barski, Michal; Brennan, Benjamin; Miller, Ona K; Potter, Jane A; Vijayakrishnan, Swetha; Bhella, David; Naismith, James H; Elliott, Richard M; Schwarz-Linek, Ulrich

    2017-09-15

    Rift Valley fever phlebovirus (RVFV) is a clinically and economically important pathogen increasingly likely to cause widespread epidemics. RVFV virulence depends on the interferon antagonist non-structural protein (NSs), which remains poorly characterized. We identified a stable core domain of RVFV NSs (residues 83-248), and solved its crystal structure, a novel all-helical fold organized into highly ordered fibrils. A hallmark of RVFV pathology is NSs filament formation in infected cell nuclei. Recombinant virus encoding the NSs core domain induced intranuclear filaments, suggesting it contains all essential determinants for nuclear translocation and filament formation. Mutations of key crystal fibril interface residues in viruses encoding full-length NSs completely abrogated intranuclear filament formation in infected cells. We propose the fibrillar arrangement of the NSs core domain in crystals reveals the molecular basis of assembly of this key virulence factor in cell nuclei. Our findings have important implications for fundamental understanding of RVFV virulence.

  7. Toward a mineral physics reference model for the Moon's core.

    Science.gov (United States)

    Antonangeli, Daniele; Morard, Guillaume; Schmerr, Nicholas C; Komabayashi, Tetsuya; Krisch, Michael; Fiquet, Guillaume; Fei, Yingwei

    2015-03-31

    The physical properties of iron (Fe) at high pressure and high temperature are crucial for understanding the chemical composition, evolution, and dynamics of planetary interiors. Indeed, the inner structures of the telluric planets all share a similar layered nature: a central metallic core composed mostly of iron, surrounded by a silicate mantle, and a thin, chemically differentiated crust. To date, most studies of iron have focused on the hexagonal closed packed (hcp, or ε) phase, as ε-Fe is likely stable across the pressure and temperature conditions of Earth's core. However, at the more moderate pressures characteristic of the cores of smaller planetary bodies, such as the Moon, Mercury, or Mars, iron takes on a face-centered cubic (fcc, or γ) structure. Here we present compressional and shear wave sound velocity and density measurements of γ-Fe at high pressures and high temperatures, which are needed to develop accurate seismic models of planetary interiors. Our results indicate that the seismic velocities proposed for the Moon's inner core by a recent reanalysis of Apollo seismic data are well below those of γ-Fe. Our dataset thus provides strong constraints to seismic models of the lunar core and cores of small telluric planets. This allows us to propose a direct compositional and velocity model for the Moon's core.

  8. Core clamping device for a nuclear reactor

    International Nuclear Information System (INIS)

    Guenther, R.W.

    1974-01-01

    The core clamping device for a fast neutron reactor includes clamps to support the fuel zone against the pressure vessel. The clamps are arranged around the circumference of the core. They consist of torsion bars arranged parallel at some distance around the core with lever arms attached to the ends whose force is directed in the opposite direction, pressing against the wall of the pressure vessel. The lever arms and pressure plates also actuated by the ends of the torsion bars transfer the stress, the pressure plates acting upon the fuel elements or fuel assemblies. Coupling between the ends of the torsion bars and the pressure plates is achieved by end carrier plates directly attached to the torsion bars and radially movable. This clamping device follows the thermal expansions of the core, allows specific elements to be disengaged in sections and saves space between the core and the neutron reflectors. (DG) [de

  9. Benchmark calculation for water reflected STACY cores containing low enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori; Yamamoto, Toshihiro; Nakamura, Takemi

    2001-01-01

    In order to validate the availability of criticality calculation codes and related nuclear data library, a series of fundamental benchmark experiments on low enriched uranyl nitrate solution have been performed with a Static Experiment Criticality Facility, STACY in JAERI. The basic core composed of a single tank with water reflector was used for accumulating the systematic data with well-known experimental uncertainties. This paper presents the outline of the core configurations of STACY, the standard calculation model, and calculation results with a Monte Carlo code and JENDL 3.2 nuclear data library. (author)

  10. The Great Deluge Algorithm applied to a nuclear reactor core design optimization problem

    International Nuclear Information System (INIS)

    Sacco, Wagner F.; Oliveira, Cassiano R.E. de

    2005-01-01

    The Great Deluge Algorithm (GDA) is a local search algorithm introduced by Dueck. It is an analogy with a flood: the 'water level' rises continuously and the proposed solution must lie above the 'surface' in order to survive. The crucial parameter is the 'rain speed', which controls convergence of the algorithm similarly to Simulated Annealing's annealing schedule. This algorithm is applied to the reactor core design optimization problem, which consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a 3-enrichment-zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. This problem was previously attacked by the canonical genetic algorithm (GA) and by a Niching Genetic Algorithm (NGA). NGAs were designed to force the genetic algorithm to maintain a heterogeneous population throughout the evolutionary process, avoiding the phenomenon known as genetic drift, where all the individuals converge to a single solution. The results obtained by the Great Deluge Algorithm are compared to those obtained by both algorithms mentioned above. The three algorithms are submitted to the same computational effort and GDA reaches the best results, showing its potential for other applications in the nuclear engineering field as, for instance, the nuclear core reload optimization problem. One of the great advantages of this algorithm over the GA is that it does not require special operators for discrete optimization. (author)

  11. CFD approach to modeling of core-concrete interaction

    International Nuclear Information System (INIS)

    Vladimir V Chudanov; Anna E Aksenova; Valerii A Pervichko

    2005-01-01

    Full text of publication follows: A large attention is given to research behavior of concrete structures at high mechanical and thermal loadings, which those suffer at the severe accidents on Nuclear Power Plants with core melting and falling of the molten corium mass into reactor shaft. There are enough programs for analysis of heat and mass transfer processes at interaction of the molten corium with concrete. Most known among them CORCON and WECHSL, which were developed more than twenty years ago, allow considering a quasi-stationary phase decomposition of concrete and the some transition regimes. In opposing to the mentioned codes a new more generalized mathematical model and software are developed for modeling of a wide range of the heat and mass transfer processes under study of the molten core-concrete interaction. The developed mathematical model is based on the Navier-Stokes equations with variable properties with taking into account of a density jump under melting of concrete together with a heat transfer equation. The offered numerical technique is based on modern algorithms with small scheme diffusion, whose discrete approximations are constructed with use of finite-volume methods and the fully staggered grids. The developed software corresponds to modern level of development of computers and takes into account all phenomenology, used by mentioned codes, and allows to simulate the such phenomena and processes as: multidimensional heat transfer in concrete for modeling of transients for an intermediate thermal flux to concrete; direct erosion of concrete at a quasi-stationary regime of interaction with molten fuel masses; heat and mass transfer in corium and convective intermixing in a melt of corium with taking into account of its stratification on two layers of the metal and oxide components and heat transfer by radiation in a cavity of the reactor shaft; change physical properties of corium at concrete decomposition and release in corium of its

  12. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hae Kyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-10-15

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was

  13. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    International Nuclear Information System (INIS)

    Park, Hae Kyun; Kim, Sang Nyung

    2013-01-01

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

  14. Adaptive control method for core power control in TRIGA Mark II reactor

    Science.gov (United States)

    Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

    2018-01-01

    The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  15. Modelling of magnetostriction of transformer magnetic core for vibration analysis

    Science.gov (United States)

    Marks, Janis; Vitolina, Sandra

    2017-12-01

    Magnetostriction is a phenomenon occurring in transformer core in normal operation mode. Yet in time, it can cause the delamination of magnetic core resulting in higher level of vibrations that are measured on the surface of transformer tank during diagnostic tests. The aim of this paper is to create a model for evaluating elastic deformations in magnetic core that can be used for power transformers with intensive vibrations in order to eliminate magnetostriction as a their cause. Description of the developed model in Matlab and COMSOL software is provided including restrictions concerning geometry and properties of materials, and the results of performed research on magnetic core anisotropy are provided. As a case study modelling of magnetostriction for 5-legged 200 MVA power transformer with the rated voltage of 13.8/137kV is conducted, based on which comparative analysis of vibration levels and elastic deformations is performed.

  16. Modelling of magnetostriction of transformer magnetic core for vibration analysis

    Directory of Open Access Journals (Sweden)

    Marks Janis

    2017-12-01

    Full Text Available Magnetostriction is a phenomenon occurring in transformer core in normal operation mode. Yet in time, it can cause the delamination of magnetic core resulting in higher level of vibrations that are measured on the surface of transformer tank during diagnostic tests. The aim of this paper is to create a model for evaluating elastic deformations in magnetic core that can be used for power transformers with intensive vibrations in order to eliminate magnetostriction as a their cause. Description of the developed model in Matlab and COMSOL software is provided including restrictions concerning geometry and properties of materials, and the results of performed research on magnetic core anisotropy are provided. As a case study modelling of magnetostriction for 5-legged 200 MVA power transformer with the rated voltage of 13.8/137kV is conducted, based on which comparative analysis of vibration levels and elastic deformations is performed.

  17. 3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

    Energy Technology Data Exchange (ETDEWEB)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

    2016-09-15

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  18. Covariance matrices for nuclear cross sections derived from nuclear model calculations

    International Nuclear Information System (INIS)

    Smith, D. L.

    2005-01-01

    The growing need for covariance information to accompany the evaluated cross section data libraries utilized in contemporary nuclear applications is spurring the development of new methods to provide this information. Many of the current general purpose libraries of evaluated nuclear data used in applications are derived either almost entirely from nuclear model calculations or from nuclear model calculations benchmarked by available experimental data. Consequently, a consistent method for generating covariance information under these circumstances is required. This report discusses a new approach to producing covariance matrices for cross sections calculated using nuclear models. The present method involves establishing uncertainty information for the underlying parameters of nuclear models used in the calculations and then propagating these uncertainties through to the derived cross sections and related nuclear quantities by means of a Monte Carlo technique rather than the more conventional matrix error propagation approach used in some alternative methods. The formalism to be used in such analyses is discussed in this report along with various issues and caveats that need to be considered in order to proceed with a practical implementation of the methodology

  19. Quantification of cost of margin associated with in-core nuclear fuel management for a PWR

    International Nuclear Information System (INIS)

    Kropaczek, D.J.; Turinsky, P.J.

    1989-01-01

    The problem of in-core nuclear fuel management optimization is discussed. The problem is to determine the location of core material, such as the fuel and burnable poisons, so as to minimize (maximize) a stated objective within engineering constraints. Typical objectives include maximization of cycle energy production or discharged fuel exposure, and minimization of power peaking factor or reactor vessel fluence. Constraints include discharge burnup limits and one or more of the possible objectives if not selected as the objective. The optimization problem can be characterized as a large combinatorial problem with nonlinear objective function and constraints, which are likely to be active. The authors have elected to employ the integer Monte Carlo programming method to address this optimization problem because of the just-noted problem characteristics. To evaluate the core physics characteristics as a function of fuel loading pattern, second-order accurate perturbation theory is employed with successive application to improve estimates of the optimum loading pattern. No constraints on fuel movement other than requiring quarter-core symmetry were imposed. In this paper the authors employed this methodology to address a related problem. The problem being addressed can be stated as What is the cost associated with margin? Specifically, they wish to assign some financial value in terms of increased levelized fuel cycle cost associated with an increase in core margin of some type, such as power peaking factor

  20. Nuclear fuel cycle modelling using MESSAGE

    International Nuclear Information System (INIS)

    Guiying Zhang; Dongsheng Niu; Guoliang Xu; Hui Zhang; Jue Li; Lei Cao; Zeqin Guo; Zhichao Wang; Yutong Qiu; Yanming Shi; Gaoliang Li

    2017-01-01

    In order to demonstrate the possibilities of application of MESSAGE tool for the modelling of a Nuclear Energy System at the national level, one of the possible open nuclear fuel cycle options based on thermal reactors has been modelled using MESSAGE. The steps of the front-end and back-end of nuclear fuel cycle and nuclear reactor operation are described. The optimal structure for Nuclear Power Development and optimal schedule for introducing various reactor technologies and fuel cycle options; infrastructure facilities, nuclear material flows and waste, investments and other costs are demonstrated. (author)

  1. Neutron flux and power in RTP core-15

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis; Bayar, Abi Muttaqin Jalal; Hamzah, Na’im Syauqi Bin [Nuclear and reactor Physics Section, Nuclear Technology Center, Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core with literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.

  2. Hypernuclear properties derived from the Nijmegen soft-core OBE potential

    International Nuclear Information System (INIS)

    Yamamoto, Yasuo; Bando, Hiroharu.

    1990-01-01

    The Nijmegen soft-core YN potential is applied to the G-matrix calculation in nuclear matter, characteristics of which are investigated in comparison with the hard-core models D and F. The ΛN G-matrix interaction is simulated in a three-range Gaussian form and applied to various hypernuclear calculations. Λ binding energies in ground and excited states are wholly reproduced from light to medium heavy hypernuclei observed in experiments. (author)

  3. Research on intelligent monitor for 3D power distribution of reactor core

    International Nuclear Information System (INIS)

    Xia, Hong; Li, Bin; Liu, Jianxin

    2014-01-01

    Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

  4. A simple model for induction core voltage distributions

    International Nuclear Information System (INIS)

    Briggs, Richard J.; Fawley, William M.

    2004-01-01

    In fall 2003 T. Hughes of MRC used a full EM simulation code (LSP) to show that the electric field stress distribution near the outer radius of the longitudinal gaps between the four Metglas induction cores is very nonuniform in the original design of the DARHT-2 accelerator cells. In this note we derive a simple model of the electric field distribution in the induction core region to provide physical insights into this result. The starting point in formulating our model is to recognize that the electromagnetic fields in the induction core region of the DARHT-2 accelerator cells should be accurately represented within a quasi-static approximation because the timescale for the fields to change is much longer than the EM wave propagation time. The difficulty one faces is the fact that the electric field is a mixture of both a ''quasi-magnetostatic field'' (having a nonzero curl, with Bdot the source) and a ''quasi-electrostatic field'' (the source being electric charges on the various metal surfaces). We first discuss the EM field structure on the ''micro-scale'' of individual tape windings in Section 2. The insights from that discussion are then used to formulate a ''macroscopic'' description of the fields inside an ''equivalent homogeneous tape wound core region'' in Section 3. This formulation explicitly separates the nonlinear core magnetics from the quasi-electrostatic components of the electric field. In Section 4 a physical interpretation of the radial dependence of the electrostatic component of the electric field derived from this model is presented in terms of distributed capacitances, and the voltage distribution from gap to gap is related to various ''equivalent'' lumped capacitances. Analytic solutions of several simple multi-core cases are presented in Sections 5 and 6 to help provide physical insight into the effect of various proposed changes in the geometrical parameters of the DARHT-2 accelerator cell. Our results show that over most of the gap

  5. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  6. The nuclear Thomas-Fermi model

    International Nuclear Information System (INIS)

    Myers, W.D.; Swiatecki, W.J.

    1994-08-01

    The statistical Thomas-Fermi model is applied to a comprehensive survey of macroscopic nuclear properties. The model uses a Seyler-Blanchard effective nucleon-nucleon interaction, generalized by the addition of one momentum-dependent and one density-dependent term. The adjustable parameters of the interaction were fitted to shell-corrected masses of 1654 nuclei, to the diffuseness of the nuclear surface and to the measured depths of the optical model potential. With these parameters nuclear sizes are well reproduced, and only relatively minor deviations between measured and calculated fission barriers of 36 nuclei are found. The model determines the principal bulk and surface properties of nuclear matter and provides estimates for the more subtle, Droplet Model, properties. The predicted energy vs density relation for neutron matter is in striking correspondence with the 1981 theoretical estimate of Friedman and Pandharipande. Other extreme situations to which the model is applied are a study of Sn isotopes from 82 Sn to 170 Sn, and the rupture into a bubble configuration of a nucleus (constrained to spherical symmetry) which takes place when Z 2 /A exceeds about 100

  7. Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

    2017-06-01

    Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

  8. Modeling in the Common Core State Standards

    Science.gov (United States)

    Tam, Kai Chung

    2011-01-01

    The inclusion of modeling and applications into the mathematics curriculum has proven to be a challenging task over the last fifty years. The Common Core State Standards (CCSS) has made mathematical modeling both one of its Standards for Mathematical Practice and one of its Conceptual Categories. This article discusses the need for mathematical…

  9. Geomagnetic core field models in the satellite era

    DEFF Research Database (Denmark)

    Lesur, Vincent; Olsen, Nils; Thomson, Alan W. P.

    2011-01-01

    After a brief review of the theoretical basis and difficulties that modelers are facing, we present three recent models of the geomagnetic field originating in the Earth’s core. All three modeling approaches are using recent observatory and near-Earth orbiting survey satellite data. In each case...

  10. Design and test of the borosilicate glass burnable poison rod for Qinshan nuclear power plant core

    International Nuclear Information System (INIS)

    Huang Jinhua; Sun Hanhong

    1988-08-01

    Material for the burnable poison of Qinshan Nuclear Power Plant core is GG-17 borosilicate glass. The chemical composition and physico-chemical properties of GG-17 is very close to Pyrex-7740 glass used by Westinghouse. It is expected from the results of the experiments that the borosilicate glass burnable poison rod can be successfully used in Qinshan Nuclear Power Plant due to good physical, mechanical, corrosion-resistant and irradiaton properties for both GG-17 glass and cold-worked stainless steel cladding. Change of material for burnable poison from boron-bearing stainless steel to borosilicate glass will bring about much more economic benefit to Qinshan Naclear Power Plant

  11. Complex models of nodal nuclear data

    International Nuclear Information System (INIS)

    Dufek, Jan

    2011-01-01

    During the core simulations, nuclear data are required at various nodal thermal-hydraulic and fuel burnup conditions. The nodal data are also partially affected by thermal-hydraulic and fuel burnup conditions in surrounding nodes as these change the neutron energy spectrum in the node. Therefore, the nodal data are functions of many parameters (state variables), and the more state variables are considered by the nodal data models the more accurate and flexible the models get. The existing table and polynomial regression models, however, cannot reflect the data dependences on many state variables. As for the table models, the number of mesh points (and necessary lattice calculations) grows exponentially with the number of variables. As for the polynomial regression models, the number of possible multivariate polynomials exceeds the limits of existing selection algorithms that should identify a few dozens of the most important polynomials. Also, the standard scheme of lattice calculations is not convenient for modelling the data dependences on various burnup conditions since it performs only a single or few burnup calculations at fixed nominal conditions. We suggest a new efficient algorithm for selecting the most important multivariate polynomials for the polynomial regression models so that dependences on many state variables can be considered. We also present a new scheme for lattice calculations where a large number of burnup histories are accomplished at varied nodal conditions. The number of lattice calculations being performed and the number of polynomials being analysed are controlled and minimised while building the nodal data models of a required accuracy. (author)

  12. Preliminary model for core/concrete interactions

    International Nuclear Information System (INIS)

    Murfin, W.B.

    1977-08-01

    A preliminary model is described for computing the rate of penetration of concrete by a molten LWR core. Among the phenomena included are convective stirring of the melt by evolved gases, admixture of concrete decomposition products to the melt, chemical reactions, radiative heat loss, and variation of heat transfer coefficients with local pressure. The model is most applicable to a two-phase melt (metallic plus oxidic) having a fairly high metallic content

  13. Interactive Real-time Simulation of a Nuclear Reactor Emergency Core Cooling System on a Desktop Computer

    International Nuclear Information System (INIS)

    Muncharoen, C.; Chanyotha, S.; Bereznai, G.

    1998-01-01

    The simulation of the Emergency Core Cooling System for a 900 MW nuclear power plant has been developed by using object oriented programming language. It is capable of generating code that executes in real-time on a PENTIUM 100 or equivalent personal computer. Graphical user interface ECCS screens have been developed using Lab VIEW to allow interactive control of ECCS. The usual simulator functions, such as freeze, run, iterate, have been provided, and a number of malfunctions may be activated. A large pipe break near the reactor inlet header has been simulated to verify the response of the ECCS model. LOCA detection, ECC initiation, injection and recovery phased are all modeled, and give results consistent with safety analysis data for a 100% break. With stand alone ECCS simulation, the changes of flow and pressure in ECCS can be observed. The operator can study operational procedures and get used to LOCA in case of the LOCA. Practicing with malfunction, the operator will improve problem solving skills and gain a deeper comprehension of ECCS

  14. Testing Numerical Models of Cool Core Galaxy Cluster Formation with X-Ray Observations

    Science.gov (United States)

    Henning, Jason W.; Gantner, Brennan; Burns, Jack O.; Hallman, Eric J.

    2009-12-01

    Using archival Chandra and ROSAT data along with numerical simulations, we compare the properties of cool core and non-cool core galaxy clusters, paying particular attention to the region beyond the cluster cores. With the use of single and double β-models, we demonstrate a statistically significant difference in the slopes of observed cluster surface brightness profiles while the cluster cores remain indistinguishable between the two cluster types. Additionally, through the use of hardness ratio profiles, we find evidence suggesting cool core clusters are cooler beyond their cores than non-cool core clusters of comparable mass and temperature, both in observed and simulated clusters. The similarities between real and simulated clusters supports a model presented in earlier work by the authors describing differing merger histories between cool core and non-cool core clusters. Discrepancies between real and simulated clusters will inform upcoming numerical models and simulations as to new ways to incorporate feedback in these systems.

  15. Training reactor deployment. Advanced experimental course on designing new reactor cores

    International Nuclear Information System (INIS)

    Skoda, Radek

    2009-01-01

    Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

  16. IAEA CRP on HTGR Uncertainties in Modeling: Assessment of Phase I Lattice to Core Model Uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Rouxelin, Pascal Nicolas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    II 1a. The steady state core calculations were simulated with the INL coupled-code system known as the Parallel and Highly Innovative Simulation for INL Code System (PHISICS) and the system thermal-hydraulics code known as the Reactor Excursion and Leak Analysis Program (RELAP) 5 3D using the nuclear data libraries previously generated with NEWT. It was observed that significant differences in terms of multiplication factor and neutron flux exist between the various permutations of the Phase I super-cell lattice calculations. The use of these cross section libraries only leads to minor changes in the Phase II core simulation results for fresh fuel but shows significantly larger discrepancies for spent fuel cores. Furthermore, large incongruities were found between the SCALE NEWT and KENO VI results for the super cells, and while some trends could be identified, a final conclusion on this issue could not yet be reached. This report will be revised in mid 2016 with more detailed analyses of the super-cell problems and their effects on the core models, using the latest version of SCALE (6.2). The super-cell models seem to show substantial improvements in terms of neutron flux as compared to single-block models, particularly at thermal energies.

  17. LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

    International Nuclear Information System (INIS)

    Khan, E.U.

    1980-04-01

    A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist

  18. Nuclear technology and materials science

    International Nuclear Information System (INIS)

    Olander, D.R.

    1992-01-01

    Current and expected problems in the materials of nuclear technology are reviewed. In the fuel elements of LWRs, cladding waterside corrosion, secondary hydriding and pellet-cladding interaction may be significant impediments to extended burnup. In the fuel, fission gas release remains a key issue. Materials issues in the structural alloys of the primary system include stress-corrosion cracking of steel, corrosion of steam generator tubing and pressurized thermal shock of the reactor vessel. Prediction of core behavior in severe accidents requires basic data and models for fuel liquefaction, aerosol formation, fission product transport and core-concrete interaction. Materials questions in nuclear waste management and fusion technology are briefly reviewed. (author)

  19. A nonlinear 3D real-time model for simulation of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.

    1982-02-01

    A nonlinear transient model for BWR nuclear power plants which consists of a 3D-core (subdivided into a number of superboxes, and with parallel flow and subcooled boiling), a top plenum, steam removal and feed water systems and main coolant recirculation pumps is given. The model describes the local core and global plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. The model which forms the basis for the digital code GARLIC-B (Er et al. 82) is aimed to be used on an on-site process computer in parallel to the actual reactor process (or even in predictive mode). Thus, special measures had to be taken into account in order to increase the computational speed and reduce the necessary computer storage. This could be achieved by - separating the neutron and power kinetics from the xenon-iodine dynamics, - treating the neutron kinetics and most of the thermodynamics and hydrodynamics in a pseudostationary way, - developing a special coupling coefficient concept to describe the neutron diffusion, calculating the coupling coefficients from a basic neutron kinetics code, - combining coarse mesh elements into superboxes, taking advantage of the symmetry properties of the core and - applying a sparse matrix technique for solving the resulting algebraic power equation system. (orig.) [de

  20. Design of 50 MWe HTR-PBMR reactor core and nuclear power plant fuel using SRAC2006 programme

    International Nuclear Information System (INIS)

    Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

    2014-01-01

    This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U 235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (k eff ) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U 235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.1 2% at BOL. The addition Gd 2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er 2O3 with an optimum value 290 ppm has an excess reactivity 1.24 % at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. (author)

  1. Nucleosynthesis in Core-Collapse Supernovae

    Science.gov (United States)

    Stevenson, Taylor Shannon; Viktoria Ohstrom, Eva; Harris, James Austin; Hix, William R.

    2018-01-01

    The nucleosynthesis which occurs in core-collapse supernovae (CCSN) is one of the most important sources of elements in the universe. Elements from Oxygen through Iron come predominantly from supernovae, and contributions of heavier elements are also possible through processes like the weak r-process, the gamma process and the light element primary process. The composition of the ejecta depends on the mechanism of the explosion, thus simulations of high physical fidelity are needed to explore what elements and isotopes CCSN can contribute to Galactic Chemical Evolution. We will analyze the nucleosynthesis results from self-consistent CCSN simulations performed with CHIMERA, a multi-dimensional neutrino radiation-hydrodynamics code. Much of our understanding of CCSN nucleosynthesis comes from parameterized models, but unlike CHIMERA these fail to address essential physics, including turbulent flow/instability and neutrino-matter interaction. We will present nucleosynthesis predictions for the explosion of a 9.6 solar mass first generation star, relying both on results of the 160 species nuclear reaction network used in CHIMERA within this model and on post-processing with a more extensive network. The lowest mass iron core-collapse supernovae, like this model, are distinct from their more massive brethren, with their explosion mechanism and nucleosynthesis being more like electron capture supernovae resulting from Oxygen-Neon white dwarves. We will highlight the differences between the nucleosynthesis in this model and more massive supernovae. The inline 160 species network is a feature unique to CHIMERA, making this the most sophisticated model to date for a star of this type. We will discuss the need and mechanism to extrapolate the post-processing to times post-simulation and analyze the uncertainties this introduces for supernova nucleosynthesis. We will also compare the results from the inline 160 species network to the post-processing results to study further

  2. Nuclear energy technology: theory and practice of commercial nuclear power

    International Nuclear Information System (INIS)

    Knief, R.A.

    1982-01-01

    Reviews Nuclear Energy Technology: Theory and Practice of Commercial Nuclear Power by Ronald Allen Knief, whose contents include an overview of the basic concepts of reactors and the nuclear fuel cycle; the basics of nuclear physics; reactor theory; heat removal; economics; current concerns at the front and back ends of the fuel cycle; design descriptions of domestic and foreign reactor systems; reactor safety and safeguards; Three Mile Island; and a brief overview of the basic concepts of nuclear fusion. Both magnetic and inertial confinement techniques are clearly outlined. Also reviews Nuclear Fuel Management by Harry W. Graves, Jr., consisting of introductory subjects (e.g. front end of fuel cycle); core physics methodology required for fuel depletion calculations; power capability evaluation (analyzes physical parameters that limit potential core power density); and fuel management topics (economics, loading arrangements and core operation strategies)

  3. Development of supplier evaluation model applying in nuclear power plants

    International Nuclear Information System (INIS)

    Wang Yonggang; Fang Chunfa

    2006-01-01

    It is essential for the safe and stable operations of Nuclear Power Plants that various resources in the supply chain are effectively managed. Supplier is a significant resource of nuclear entities serving as an extension of the operation process. Scientific and radiation evaluation of the performance of suppliers is of vital importance to an effective and high quality supply chain. This paper establishes an advance and practical supplier evaluation system that is applicable for the operational nuclear power plants, based on the analysis of the current operation status of Daya Bay Nuclear Power Station against its targeted objectives, the acquisition of relevant practices home and abroad and the benchmarking with advanced peers, in order to enhance the core competence of nuclear power plant. (authors)

  4. CHARACTERIZING AND MODELING FERRITE-CORE PROBES

    International Nuclear Information System (INIS)

    Sabbagh, Harold A.; Murphy, R. Kim; Sabbagh, Elias H.; Aldrin, John C.

    2010-01-01

    In this paper, we accurately and carefully characterize a ferrite-core probe that is widely used for aircraft inspections. The characterization starts with the development of a model that can be executed using the proprietary volume-integral code, VIC-3D(c), and then the model is fitted to measured multifrequency impedance data taken with the probe in freespace and over samples of a titanium alloy and aluminum. Excellent results are achieved, and will be discussed.

  5. Computer code validation study of PWR core design system, CASMO-3/MASTER-α

    International Nuclear Information System (INIS)

    Lee, K. H.; Kim, M. H.; Woo, S. W.

    1999-01-01

    In this paper, the feasibility of CASMO-3/MASTER-α nuclear design system was investigated for commercial PWR core. Validation calculation was performed as follows. Firstly, the accuracy of cross section generation from table set using linear feedback model was estimated. Secondly, the results of CASMO-3/MASTER-α was compared with CASMO-3/NESTLE 5.02 for a few benchmark problems. Microscopic cross sections computed from table set were almost the same with those from CASMO-3. There were small differences between calculated results of two code systems. Thirdly, the repetition of CASMO-3/MASTER-α calculation for Younggwang Unit-3, Cycle-1 core was done and their results were compared with nuclear design report(NDR) and uncertainty analysis results of KAERI. It was found that uncertainty analysis results were reliable enough because results were agreed each other. It was concluded that the use of nuclear design system CASMO-3/MASTER-α was validated for commercial PWR core

  6. Nuclear models relevant to evaluation

    International Nuclear Information System (INIS)

    Arthur, E.D.; Chadwick, M.B.; Hale, G.M.; Young, P.G.

    1991-01-01

    The widespread use of nuclear models continues in the creation of data evaluations. The reasons include extension of data evaluations to higher energies, creation of data libraries for isotopic components of natural materials, and production of evaluations for radiative target species. In these cases, experimental data are often sparse or nonexistent. As this trend continues, the nuclear models employed in evaluation work move towards more microscopically-based theoretical methods, prompted in part by the availability of increasingly powerful computational resources. Advances in nuclear models applicable to evaluation will be reviewed. These include advances in optical model theory, microscopic and phenomenological state and level density theory, unified models that consistently describe both equilibrium and nonequilibrium reaction mechanism, and improved methodologies for calculation of prompt radiation from fission. 84 refs., 8 figs

  7. HELIOS/DRAGON/NESTLE codes' simulation of void reactivity in a CANDU core

    International Nuclear Information System (INIS)

    Sarsour, H.N.; Rahnema, F.; Mosher, S.; Turinsky, P.J.; Serghiuta, D.; Marleau, G.; Courau, T.

    2002-01-01

    This paper presents results of simulation of void reactivity in a CANDU core using the NESTLE core simulator, cross sections from the HELIOS lattice physics code in conjunction with incremental cross sections from the DRAGON lattice physics code. First, a sub-region of a CANDU6 core is modeled using the NESTLE core simulator and predictions are contrasted with predictions by the MCNP Monte Carlo simulation code utilizing a continuous energy model. In addition, whole core modeling results are presented using the NESTLE finite difference method (FDM), NESTLE nodal method (NM) without assembly discontinuity factors (ADF), and NESTLE NM with ADF. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculational methods and codes developed independently from the CANDU industry. (author)

  8. The Nuclear Thomas-Fermi Model

    Science.gov (United States)

    Myers, W. D.; Swiatecki, W. J.

    1994-08-01

    The statistical Thomas-Fermi model is applied to a comprehensive survey of macroscopic nuclear properties. The model uses a Seyler-Blanchard effective nucleon-nucleon interaction, generalized by the addition of one momentum-dependent and one density-dependent term. The adjustable parameters of the interaction were fitted to shell-corrected masses of 1654 nuclei, to the diffuseness of the nuclear surface and to the measured depths of the optical model potential. With these parameters nuclear sizes are well reproduced, and only relatively minor deviations between measured and calculated fission barriers of 36 nuclei are found. The model determines the principal bulk and surface properties of nuclear matter and provides estimates for the more subtle, Droplet Model, properties. The predicted energy vs density relation for neutron matter is in striking correspondence with the 1981 theoretical estimate of Friedman and Pandharipande. Other extreme situations to which the model is applied are a study of Sn isotopes from {sup 82}Sn to {sup 170}Sn, and the rupture into a bubble configuration of a nucleus (constrained to spherical symmetry) which takes place when Z{sup 2}/A exceeds about 100.

  9. Study and analysis for the flow-induced vibration of the core barrel of a PWR

    International Nuclear Information System (INIS)

    Yao Weida; Shi Guolin; Jiang Nanyan

    1989-01-01

    The resemblance criteria are derived and a test model is designed by applying the flow-soild coupling theory. After having completed the model analysis of the pressurized water reactor (PWR) core barrel in an 1:10 model, the dynamic characteristics are obtained. In an 1:5 reactor model with a hydraulic closed loop, the hydraulic vibration tests of the core barrel are performed, and the relations between the flow rate and the flow-induced pulse pressure on core barrel, acceleration and strain signals have been measured. The corresponding responses and a group of computational equations for hydraulic vibration are derived from these two experiments. The computational hydraulic vibration responses for core barrel in Qinshan Nuclear Power Plant are in good agreement with the test results, and it shows that the core barrel is safe within its lifetime of 30 years

  10. Core/corona modeling of diode-imploded annular loads

    Science.gov (United States)

    Terry, R. E.; Guillory, J. U.

    1980-11-01

    The effects of a tenuous exterior plasma corona with anomalous resistivity on the compression and heating of a hollow, collisional aluminum z-pinch plasma are predicted by a one-dimensional code. As the interior ("core") plasma is imploded by its axial current, the energy exchange between core and corona determines the current partition. Under the conditions of rapid core heating and compression, the increase in coronal current provides a trade-off between radial acceleration and compression, which reduces the implosion forces and softens the pitch. Combined with a heuristic account of energy and momentum transport in the strongly coupled core plasma and an approximate radiative loss calculation including Al line, recombination and Bremsstrahlung emission, the current model can provide a reasonably accurate description of imploding annular plasma loads that remain azimuthally symmetric. The implications for optimization of generator load coupling are examined.

  11. CORE DESIGNS OF ABWR FOR PROPOSED OF THE FIRST NUCLEAR POWER PLANT IN INDONESIA

    Directory of Open Access Journals (Sweden)

    Yohannes Sardjono

    2015-04-01

    Full Text Available Indonesia as an archipelago has been experiencing high growth industry and energy demand due to high population growth, dynamic economic activities. The total population is around 230 million people and 75 % to the total population is living in Java. The introduction of Nuclear Power Plant on Java Bali electricity grid will be possible in 2022 for 2 GWe, using proven technology reactor like ABWR or others light water reactor with nominal power 1000 MWe. In this case, the rated thermal power for the equilibrium cycles is 3926 MWt, the cycle length is 18 month and overall capacity factor is 87 %. The designs were performed for an 872-fuel bundles ABWR core using GE-11 fuel type in an 9×9 fuel rod arrays with 2 Large Central Water Rods (LCWR. The calculations were divided into two steps; the first is to generate bundle library and the other is to make the thermal and reactivity limits satisfied for the core designs. Toshiba General Electric Bundle lattice Analysis (TGBLA and PANACEA computer codes were used as designs tools. TGBLA is a General Electric proprietary computer code which is used to generate bundle lattice library for fuel designs. PANACEA is General Electric proprietary computer code which is used as thermal hydraulic and neutronic coupled BWR core simulator. This result of core designs describes reactivity and thermal margins i.e.; Maximum Linear Heat Generation rate (MLHGR is lower than 14.4 kW/ft, Minimum Critical Power Ratio (MCPR is upper than 1.25, Hot Excess Reactivity (HOTXS is upper than 1 %Dk at BOC and 0.8 %Dk at 200 MWD/ST and Cold Shutdown Margin Reactivity (CSDM is upper than 1 %Dk. It is concluded that the equilibrium core design using GE-11 fuel bundle type satisfies the core design objectives for the proposed of the firs Indonesia ABWR Nuclear Power Plant. Keywords: The first NPP in Indonesia, ABWR-1000 MWe, and core designs.   Indonesia adalah sebagai negara kepulauan yang laju pertumbuhan industri, energi, penduduk

  12. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant; Analisis de documentos de los materiales de la envolvente del nucleo del reactor nuclear de la CLV U-1

    Energy Technology Data Exchange (ETDEWEB)

    Zamora R, L.; Medina F, A. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

  13. The nuclear reaction model code MEDICUS

    International Nuclear Information System (INIS)

    Ibishia, A.I.

    2008-01-01

    The new computer code MEDICUS has been used to calculate cross sections of nuclear reactions. The code, implemented in MATLAB 6.5, Mathematica 5, and Fortran 95 programming languages, can be run in graphical and command line mode. Graphical User Interface (GUI) has been built that allows the user to perform calculations and to plot results just by mouse clicking. The MS Windows XP and Red Hat Linux platforms are supported. MEDICUS is a modern nuclear reaction code that can compute charged particle-, photon-, and neutron-induced reactions in the energy range from thresholds to about 200 MeV. The calculation of the cross sections of nuclear reactions are done in the framework of the Exact Many-Body Nuclear Cluster Model (EMBNCM), Direct Nuclear Reactions, Pre-equilibrium Reactions, Optical Model, DWBA, and Exciton Model with Cluster Emission. The code can be used also for the calculation of nuclear cluster structure of nuclei. We have calculated nuclear cluster models for some nuclei such as 177 Lu, 90 Y, and 27 Al. It has been found that nucleus 27 Al can be represented through the two different nuclear cluster models: 25 Mg + d and 24 Na + 3 He. Cross sections in function of energy for the reaction 27 Al( 3 He,x) 22 Na, established as a production method of 22 Na, are calculated by the code MEDICUS. Theoretical calculations of cross sections are in good agreement with experimental results. Reaction mechanisms are taken into account. (author)

  14. Technology of nuclear reactors

    International Nuclear Information System (INIS)

    Ravelet, F.

    2016-01-01

    This academic report for graduation in engineering first presents operation principles of a nuclear reactor core. It presents core components, atomic nuclei, the notions of transmutation and radioactivity, quantities used to characterize ionizing radiations, the nuclear fission, statistical aspects of fission and differences between fast and slow neutrons, a comparison between various heat transfer fluids, the uranium enrichment process, and different types of reactor (boiling water, natural uranium and heavy water, pressurized water, and fourth generation). Then, after having recalled the French installed power, the author proposes an analysis of a typical 900 MWe nuclear power plant: primary circuit, reactor, fuel, spent fuel, pressurizer and primary pump, secondary circuit, aspects related to control-command, regulation, safety and exploitation. The last part proposes a modelling of the thermodynamic cycle of a pressurized water plant by using an equivalent Carnot cycle, a Rankine cycle, and a two-phase expansion cycle with drying-overheating

  15. First in-core measurement results obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS material testing reactor

    International Nuclear Information System (INIS)

    Carcreff, Hubert; Salmon, Laurent; Courtaux, Cedric

    2014-01-01

    Nuclear heating rate inside an MTR has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. An innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. Development of the calorimetric probe required manufacturing and irradiation of mock-ups in the ex-core area, where nuclear heating rate does not exceed 2 W.g -1 . The calorimeter working mode, the different measurement procedures, main modeling and ex-core experimental results have been already presented in previous papers. In this paper, we present in-core results obtained from 2011 to 2013 with the final device. For the first time, this new experimental measurement system was operated in several experimental locations, with nominal in-core thermal hydraulic conditions, nominal neutron flux and nuclear heating rate up to 6 W.g -1 (in graphite). After a brief presentation of the displacement system specificities, first nuclear heating distributions are presented and discussed. The Finite Element model of the calorimeter was upgraded in order to match calculated temperatures with measured ones. This 'validated' model allowed to estimate a Kc factor which tends to correct small nonlinearities when heating rate is calculated from the 'calibration method'. A comparison is made between nuclear heating rates determined from 'calibration' and 'zero methods'. In addition, an evaluation of the global uncertainty associated to the measurements is detailed. Finally, a comparison is made with available measurements obtained from previous calorimeters. (authors)

  16. Adjustment of cast metal post/cores modeled with different acrylic resins

    OpenAIRE

    Gusmão, João Milton Rocha; Pereira, Renato Piai; Alves, Guilhermino Oliveira; Pithon, Matheus Melo; Moreira, David Costa

    2016-01-01

    Aim: Evaluate the performance of four commercially available chemically-activated acrylic resins (CAARs) by measuring the level of displacement of the cores following casting. Materials and Methods: Two devices were constructed to model the cores based on a natural tooth. Forty post/cores were modeled, 10 in each of the following CAARs: Duralay (Reliance Dental, Illinois, USA), Pattern Resin (GC, Tokyo, Japan), Dencrilay (Dencril, Sao Paulo, Brazil), and Jet (Clássico, Sao Paulo, Brazil). Two...

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

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    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 .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  18. An integrated expert system for optimum in core fuel management

    International Nuclear Information System (INIS)

    Abd Elmoatty, Mona S.; Nagy, M.S.; Aly, Mohamed N.; Shaat, M.K.

    2011-01-01

    Highlights: → An integrated expert system constructed for optimum in core fuel management. → Brief discussion of the ESOIFM Package modules, inputs and outputs. → Package was applied on the DALAT Nuclear Research Reactor (0.5 MW). → The Package verification showed good agreement. - Abstract: An integrated expert system called Efficient and Safe Optimum In-core Fuel Management (ESOIFM Package) has been constructed to achieve an optimum in core fuel management and automate the process of data analysis. The Package combines the constructed mathematical models with the adopted artificial intelligence techniques. The paper gives a brief discussion of the ESOIFM Package modules, inputs and outputs. The Package was applied on the DALAT Nuclear Research Reactor (0.5 MW). Moreover, the data of DNRR have been used as a case study for testing and evaluation of ESOIFM Package. This paper shows the comparison between the ESOIFM Package burn-up results, the DNRR experimental burn-up data, and other DNRR Codes burn-up results. The results showed good agreement.

  19. Apparatus for controlling nuclear core debris

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    Disclosed is an apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling

  20. Transient core characteristics of small molten salt reactor coupling problem between heat transfer/flow and nuclear fission reaction

    International Nuclear Information System (INIS)

    Yamamoto, Takahisa; Mitachi, Koshi

    2004-01-01

    This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k 0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k 0 , putting the fission reaction rate back on the rated state. (author)

  1. The QCD model of hadron cores of the meson theory

    International Nuclear Information System (INIS)

    Pokrovskii, Y.E.

    1985-01-01

    It was shown that in the previously proposed QCD model of hadron cores the exchange and self-energy contributions of the virtual quark-antiquark-gluon cloud on the outside of a bag which radius coincides with the hardon core radius of the meson theory (∼ 0.4 Fm) have been taken into account at the phenomenological level. Simulation of this cloud by the meson field results in realistic estimations of the nucleon's electroweak properties, moment fractions carried by gluons, quarks, antiquarks and hadron-hadron interaction cross-sections within a wide range of energies. The authors note that the QCD hadron core model proposed earlier not only realistically reflects the hadron masses, but reflects self-consistently main elements of the structure and interaction of hadrons at the quark-gluon bag radius (R - 0.4Fm) being close to the meson theory core radius

  2. Critical experiments on enriched uranium graphite moderated cores

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko; Akino, Fujiyoshi; Kitadate, Kenji; Kurokawa, Ryosuke

    1978-07-01

    A variety of 20 % enriched uranium loaded and graphite-moderated cores consisting of the different lattice cells in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments systematically. In the present report, the experimental results for homogeneously or heterogeneously fuel loaded cores and for simulation core of the experimental reactor for a multi-purpose high temperature reactor are filed so as to be utilized for evaluating the accuracy of core design calculation for the experimental reactor. The filed experimental data are composed of critical masses of uranium, kinetic parameters, reactivity worths of the experimental control rods and power distributions in the cores with those rods. Theoretical analyses are made for the experimental data by adopting a simple ''homogenized cylindrical core model'' using the nuclear data of ENDF/B-III, which treats the neutron behaviour after smearing the lattice cell structure. It is made clear from a comparison between the measurement and the calculation that the group constants and fundamental methods of calculations, based on this theoretical model, are valid for the homogeneously fuel loaded cores, but not for both of the heterogeneously fuel loaded cores and the core for simulation of the experimental reactor. Then, it is pointed out that consideration to semi-homogeneous property of the lattice cells for reactor neutrons is essential for high temperature graphite-moderated reactors using dispersion fuel elements of graphite and uranium. (author)

  3. Conceptual Models Core to Good Design

    CERN Document Server

    Johnson, Jeff

    2011-01-01

    People make use of software applications in their activities, applying them as tools in carrying out tasks. That this use should be good for people--easy, effective, efficient, and enjoyable--is a principal goal of design. In this book, we present the notion of Conceptual Models, and argue that Conceptual Models are core to achieving good design. From years of helping companies create software applications, we have come to believe that building applications without Conceptual Models is just asking for designs that will be confusing and difficult to learn, remember, and use. We show how Concept

  4. Development of Core Design Technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Il; Hong, S. G.; Jang, J. W. (and others)

    2007-06-15

    This report describes the contents of core design technology and computer code system development performed during 2005 and 2006 on the objects of nuclear proliferation resistant core and nuclear fuel basic key technology development security. Also, it is including the future application plans for the results and the developed methodology, important information and the materials acquired in this period. Two core designs with single enrichment were considered for the KALIMER-600 during the first year : 1) the first core uses the non-fuel rods such as B4C, ZrH1.8, and dummy rods, 2) the core using different cladding thickness for each core region (inner, middle, and outer cores) without non-fuel rods to flatten the power distribution. In particular, the latter design was intended to simplify the fuel assembly design by eliminating the heterogeneity. It was found that the proposed design satisfy all of the Gen IV SFR design goals on the cycle length longer than 18 EFPM, fuel discharge burnup larger than 80GWd/t, sodium void worth, conversion ratio, reactivity burnup swing and so on. For this object reactor, the structure integrity outside of reactor is confirmed for the radiation exposure during the plant life according to the result of shielding design and evaluation. The transmutation capability and the core characteristics of sodium cooled fast reactor was also evaluated according to the change of MA amount. The reactivity coefficients for the BN-600 reactor with MA fueled are calculated and the results are compared and evaluated with other participants results. Even though the discrepancies between the results of participants are somewhat large but the K-CORE results are close to the average within a standard deviation. To have the capability of 3-dimensional core dynamic analysis such as analyzing power distribution and reactivity variations according to the asymmetric insertion/withdrawal of control rods, the calculation module for core dynamic parameters was

  5. Core flow inversion tested with numerical dynamo models

    Science.gov (United States)

    Rau, Steffen; Christensen, Ulrich; Jackson, Andrew; Wicht, Johannes

    2000-05-01

    We test inversion methods of geomagnetic secular variation data for the pattern of fluid flow near the surface of the core with synthetic data. These are taken from self-consistent 3-D models of convection-driven magnetohydrodynamic dynamos in rotating spherical shells, which generate dipole-dominated magnetic fields with an Earth-like morphology. We find that the frozen-flux approximation, which is fundamental to all inversion schemes, is satisfied to a fair degree in the models. In order to alleviate the non-uniqueness of the inversion, usually a priori conditions are imposed on the flow; for example, it is required to be purely toroidal or geostrophic. Either condition is nearly satisfied by our model flows near the outer surface. However, most of the surface velocity field lies in the nullspace of the inversion problem. Nonetheless, the a priori constraints reduce the nullspace, and by inverting the magnetic data with either one of them we recover a significant part of the flow. With the geostrophic condition the correlation coefficient between the inverted and the true velocity field can reach values of up to 0.65, depending on the choice of the damping parameter. The correlation is significant at the 95 per cent level for most spherical harmonic degrees up to l=26. However, it degrades substantially, even at long wavelengths, when we truncate the magnetic data sets to l currents, similar to those seen in core-flow models derived from geomagnetic data, occur in the equatorial region. However, the true flow does not contain this flow component. The results suggest that some meaningful information on the core-flow pattern can be retrieved from secular variation data, but also that the limited resolution of the magnetic core field could produce serious artefacts.

  6. Performance modeling and analysis of parallel Gaussian elimination on multi-core computers

    Directory of Open Access Journals (Sweden)

    Fadi N. Sibai

    2014-01-01

    Full Text Available Gaussian elimination is used in many applications and in particular in the solution of systems of linear equations. This paper presents mathematical performance models and analysis of four parallel Gaussian Elimination methods (precisely the Original method and the new Meet in the Middle –MiM– algorithms and their variants with SIMD vectorization on multi-core systems. Analytical performance models of the four methods are formulated and presented followed by evaluations of these models with modern multi-core systems’ operation latencies. Our results reveal that the four methods generally exhibit good performance scaling with increasing matrix size and number of cores. SIMD vectorization only makes a large difference in performance for low number of cores. For a large matrix size (n ⩾ 16 K, the performance difference between the MiM and Original methods falls from 16× with four cores to 4× with 16 K cores. The efficiencies of all four methods are low with 1 K cores or more stressing a major problem of multi-core systems where the network-on-chip and memory latencies are too high in relation to basic arithmetic operations. Thus Gaussian Elimination can greatly benefit from the resources of multi-core systems, but higher performance gains can be achieved if multi-core systems can be designed with lower memory operation, synchronization, and interconnect communication latencies, requirements of utmost importance and challenge in the exascale computing age.

  7. In core reload design for cycle 4 of Daya Bay nuclear power station both units

    International Nuclear Information System (INIS)

    Zhang Zongyao; Liu Xudong; Xian Chunyu; Li Dongsheng; Zhang Hong; Liu Changwen; Rui Min; Wang Yingming; Zhao Ke; Zhang Hong; Xiao Min

    1998-01-01

    The basic principles and the contents of the reload design for Daya Bay nuclear power station are briefly introduced. The in core reload design results, and the comparison between the calculated values and the measured values of both units the fourth cycle are also given. The reload design results of the two units satisfy all the economic requirements and safety criteria. The experimented results shown that the predicated values are tally good with all the measurement values

  8. Core design studies on various forms of coolants and fuel materials. 2. Studies on liquid heavy metal and gas cooled cores, small cores and evaluation of 4-type cores

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Sakashita, Yoshiyuki; Naganuma, Masayuki; Takaki, Naoyuki; Mizuno, Tomoyasu; Ikegami, Tetsuo

    2001-01-01

    Alternative concepts to sodium cooled fast reactors, such as heavy metal liquid cooled reactors and gas cooled fast reactors were studied in Phase-1 of the feasibility studies, aiming at simplification of the system, high thermal efficiency and enhancing safety. Fuel and core specifications and nuclear characteristics were surveyed to meet the targets for commercialization of fast reactor cycle. Nuclear characteristics of small fast reactor cores were also surveyed from the perspective of the possibility of multi-purpose use and dispersed power stations. The key points of the design study for each concept in Phase-2 were summarized from the aspect of the screening of the candidates for FR commercialization. (author)

  9. Development of an artificial neural network model for on-line thermal margin estimation of a nuclear reactor core

    International Nuclear Information System (INIS)

    Kim, Hyun Koon

    1992-02-01

    One of the key safety parameters related to thermal margin in a Pressurized Water Reactor (PWR) core, is Departure from Nucleate Boiling Ratio (DNBR), which is to be assessed and continuously monitored during operation via either an analog or a digital monitoring system. The digital monitoring system, in general, allows more thermal margin than the analog system through the on-line computation of DNBR using the measured parameters as inputs to a simplified, fast running computer code. The purpose of this thesis is to develop an advanced method for on-line DNBR estimation by introducing an artifactual neural network model for best-estimation of DNBR at the given reactor operating conditions. the neural network model, consisting of three layers with five operating parameters in the input layer, provides real-time prediction accuracy of DNBR by training the network against the detailed simulation results for various operating conditions. The overall training procedure is developed to learn the characteristics of DNBR behaviour in the reactor core. First, a set of random combination of input variables is generated by Latin Hypercube Sampling technique performed on a wide range of input parameters. Second, the target values of DNBR to be referenced for training are calculated using a detailed simulation code, COBRA-IV. Third, the optimized training input data are selected. Then, training is performed using an Error Back Propagation algorithm. After completion of training, the network is tested on the examining data set in order to investigate the generalization capability of the network responses for the steady state operating condition as well as for the transient situations where DNB is of a primary concern. The test results show that the values of DNBR predicted by the neural network are maintained at a high level of accuracy for the steady state condition, and are in good agreements with the transient situation, although slightly conservative as compared to those

  10. Organizational Models for Non-Core Processes Management: A Classification Framework

    Directory of Open Access Journals (Sweden)

    Alberto F. De Toni

    2012-12-01

    The framework enables the identification and the explanation of the main advantages and disadvantages of each strategy and to highlight how a company should coherently choose an organizational model on the basis of: (a the specialization/complexity of the non‐core processes, (b the focus on core processes, (c its inclination towards know‐how outsourcing, and (d the desired level of autonomy in the management of non‐core processes.

  11. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    International Nuclear Information System (INIS)

    Griffiths, M.; Coleman, C.E.; Holt, R.A.; Sagat, S.; Urbanic, V.F.; Chow, C.K.

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and β-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the α-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of 'breakaway' growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs

  12. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, M; Coleman, C E; Holt, R A; Sagat, S; Urbanic, V F [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.; Chow, C K [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Nuclear Research Establishment

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and {beta}-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the {alpha}-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of `breakaway` growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs.

  13. Development of advanced nuclear core analysis system applicable to various reactor types (II)

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2003-03-01

    A 900 group cross section library based on the specification determined last year was produced for 27 nuclei of the fast reactor benchmark problem evaluated in nuclear data file JENDL-3.2. In addition, the new SLAROM code, which has been developed as an advanced detail analysis system, was revised so as to make cell calculations effectively with the above 900 group library. Furthermore, new functions were added to the SLAROM so that the SLAROM evaluates assembly parameters using effective cross sections derived by the SLAROM and produces any condensed effective cross section set for core performance analysis. With the 900 group cross section library and the revised SALROM, three cell calculations for fast and medium neutron speed reactors having different neutron spectrum were performed, and the results were compared with those calculated by the continuos energy Monte Carlo code MVP. By the comparisons, it is concluded that the newly revised SLAROM and a 900 group cross section library give accuracy comparable to MVP for predicting core performances. (author)

  14. Mathematical Methodology for New Modeling of Water Hammer in Emergency Core Cooling System

    International Nuclear Information System (INIS)

    Lee, Seungchan; Yoon, Dukjoo; Ha, Sangjun

    2013-01-01

    In engineering insight, the water hammer study has carried out through the experimental work and the fluid mechanics. In this study, a new access methodology is introduced by Newton mechanics and a mathematical method. Also, NRC Generic Letter 2008-01 requires nuclear power plant operators to evaluate the effect of water-hammer for the protection of pipes of the Emergency Core Cooling System, which is related to the Residual Heat Removal System and the Containment Spray System. This paper includes modeling, the processes of derivation of the mathematical equations and the comparison with other experimental work. To analyze the effect of water-hammer, this mathematical methodology is carried out. This study is in good agreement with other experiment results as above. This method is very efficient to explain the water-hammer phenomena

  15. Mathematical Methodology for New Modeling of Water Hammer in Emergency Core Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seungchan; Yoon, Dukjoo; Ha, Sangjun [Korea Hydro Nuclear Power Co. Ltd, Daejeon (Korea, Republic of)

    2013-05-15

    In engineering insight, the water hammer study has carried out through the experimental work and the fluid mechanics. In this study, a new access methodology is introduced by Newton mechanics and a mathematical method. Also, NRC Generic Letter 2008-01 requires nuclear power plant operators to evaluate the effect of water-hammer for the protection of pipes of the Emergency Core Cooling System, which is related to the Residual Heat Removal System and the Containment Spray System. This paper includes modeling, the processes of derivation of the mathematical equations and the comparison with other experimental work. To analyze the effect of water-hammer, this mathematical methodology is carried out. This study is in good agreement with other experiment results as above. This method is very efficient to explain the water-hammer phenomena.

  16. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

  17. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: pnch@yonsei.ac.kr [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)

    2017-03-15

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  18. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    International Nuclear Information System (INIS)

    Park, Jong-beom; Park, No-Cheol; Lee, Sang-Jeong; Park, Young-Pil; Choi, Youngin

    2017-01-01

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  19. Quasi-exactly solvable relativistic soft-core Coulomb models

    Energy Technology Data Exchange (ETDEWEB)

    Agboola, Davids, E-mail: davagboola@gmail.com; Zhang, Yao-Zhong, E-mail: yzz@maths.uq.edu.au

    2012-09-15

    By considering a unified treatment, we present quasi exact polynomial solutions to both the Klein-Gordon and Dirac equations with the family of soft-core Coulomb potentials V{sub q}(r)=-Z/(r{sup q}+{beta}{sup q}){sup 1/q}, Z>0, {beta}>0, q{>=}1. We consider cases q=1 and q=2 and show that both cases are reducible to the same basic ordinary differential equation. A systematic and closed form solution to the basic equation is obtained using the Bethe ansatz method. For each case, the expressions for the energies and the allowed parameters are obtained analytically and the wavefunctions are derived in terms of the roots of a set of Bethe ansatz equations. - Highlights: Black-Right-Pointing-Pointer The relativistic bound-state solutions of the soft-core Coulomb models. Black-Right-Pointing-Pointer Quasi-exact treatments of the Dirac and Klein-Gordon equations for the soft-core Coulomb models. Black-Right-Pointing-Pointer Solutions obtained in terms of the roots to the Bethe ansatz equations. Black-Right-Pointing-Pointer The hidden Lie algebraic structure discussed for the models. Black-Right-Pointing-Pointer Results useful in describing mesonic atoms and interaction of intense laser fields with atom.

  20. Multi-objective genetic algorithm parameter estimation in a reduced nuclear reactor model

    Energy Technology Data Exchange (ETDEWEB)

    Marseguerra, M.; Zio, E.; Canetta, R. [Polytechnic of Milan, Dept. of Nuclear Engineering, Milano (Italy)

    2005-07-01

    The fast increase in computing power has rendered, and will continue to render, more and more feasible the incorporation of dynamics in the safety and reliability models of complex engineering systems. In particular, the Monte Carlo simulation framework offers a natural environment for estimating the reliability of systems with dynamic features. However, the time-integration of the dynamic processes may render the Monte Carlo simulation quite burdensome so that it becomes mandatory to resort to validated, simplified models of process evolution. Such models are typically based on lumped effective parameters whose values need to be suitably estimated so as to best fit to the available plant data. In this paper we propose a multi-objective genetic algorithm approach for the estimation of the effective parameters of a simplified model of nuclear reactor dynamics. The calibration of the effective parameters is achieved by best fitting the model responses of the quantities of interest to the actual evolution profiles. A case study is reported in which the real reactor is simulated by the QUAndry based Reactor Kinetics (Quark) code available from the Nuclear Energy Agency and the simplified model is based on the point kinetics approximation to describe the neutron balance in the core and on thermal equilibrium relations to describe the energy exchange between the different loops. (authors)