WorldWideScience

Sample records for simulated nuclear core

  1. 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.)

  2. 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)

  3. 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)

  4. 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

  5. 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

  6. 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)

  7. 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.

  8. 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

  9. 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.

  10. 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)

  11. 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)

  12. 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.

  13. 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)

  14. 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

  15. 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

  16. 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.

  17. 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)

  18. 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)

  19. 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

  20. 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.

  1. 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.

  2. 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.)

  3. 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

  4. 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)

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. Simulation methods for nuclear production scheduling

    International Nuclear Information System (INIS)

    Miles, W.T.; Markel, L.C.

    1975-01-01

    Recent developments and applications of simulation methods for use in nuclear production scheduling and fuel management are reviewed. The unique characteristics of the nuclear fuel cycle as they relate to the overall optimization of a mixed nuclear-fossil system in both the short-and mid-range time frame are described. Emphasis is placed on the various formulations and approaches to the mid-range planning problem, whose objective is the determination of an optimal (least cost) system operation strategy over a multi-year planning horizon. The decomposition of the mid-range problem into power system simulation, reactor core simulation and nuclear fuel management optimization, and system integration models is discussed. Present utility practices, requirements, and research trends are described. 37 references

  10. 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

  11. 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

  12. Special nuclear material simulation device

    Science.gov (United States)

    Leckey, John H.; DeMint, Amy; Gooch, Jack; Hawk, Todd; Pickett, Chris A.; Blessinger, Chris; York, Robbie L.

    2014-08-12

    An apparatus for simulating special nuclear material is provided. The apparatus typically contains a small quantity of special nuclear material (SNM) in a configuration that simulates a much larger quantity of SNM. Generally the apparatus includes a spherical shell that is formed from an alloy containing a small quantity of highly enriched uranium. Also typically provided is a core of depleted uranium. A spacer, typically aluminum, may be used to separate the depleted uranium from the shell of uranium alloy. A cladding, typically made of titanium, is provided to seal the source. Methods are provided to simulate SNM for testing radiation monitoring portals. Typically the methods use at least one primary SNM spectral line and exclude at least one secondary SNM spectral line.

  13. 48{sup th} Annual meeting on nuclear technology (AMNT 2017). Key topic / Enhanced safety and operation excellence. Focus session: Uncertainty analyses in reactor core simulations

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, Winfried [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Forschungszentrum

    2017-12-15

    The supplementation of reactor simulations by uncertainty analyses is becoming increasingly important internationally due to the fact that the reliability of simulation calculations can be significantly increased by the quantification of uncertainties in comparison to the use of so-called conservative methods (BEPU- ''Best-Estimate plus Uncertainties''). While systematic uncertainty analyses for thermo-hydraulic calculations have been performed routinely for a long time, methods for taking into account uncertainties in nuclear data, which are the basis for neutron transport calculations, are under development. The Focus Session Uncertainty Analyses in Reactor Core Simulations was intended to provide an overview of international research and development with respect to supplementing reactor core simulations with uncertainty and sensitivity analyses, in research institutes as well as within the nuclear industry. The presented analyses not only focused on light water reactors, but also on advanced reactor systems. Particular emphasis was put on international benchmarks in the field. The session was chaired by Winfried Zwermann (Gesellschaft fuer Anlagen- und Reaktorsicherheit).

  14. 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

  15. 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)

  16. 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)

  17. 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.)

  18. 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

  19. 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.

  20. Application of Nuclear Power Plant Simulator for High School Student Training

    Energy Technology Data Exchange (ETDEWEB)

    Kong, Chi Dong; Choi, Soo Young; Park, Min Young; Lee, Duck Jung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-10-15

    In this context, two lectures on nuclear power plant simulator and practical training were provided to high school students in 2014. The education contents were composed of two parts: the micro-physics simulator and the macro-physics simulator. The micro-physics simulator treats only in-core phenomena, whereas the macro-physics simulator describes whole system of a nuclear power plant but it considers a reactor core as a point. The high school students showed strong interests caused by the fact that they operated the simulation by themselves. This abstract reports the training detail and evaluation of the effectiveness of the training. Lectures on nuclear power plant simulator and practical exercises were performed at Ulsan Energy High School and Ulsan Meister High School. Two simulators were used: the macro- and micro-physics simulator. Using the macro-physics simulator, the following five simulations were performed: reactor power increase/decrease, reactor trip, single reactor coolant pump trip, large break loss of coolant accident, and station black-out with D.C. power loss. Using the micro-physics simulator, the following three analyses were performed: the transient analysis, fuel rod performance analysis, and thermal-hydraulics analysis. The students at both high schools showed interest and strong support for the simulator-based training. After the training, the students showed passionate responses that the education was of help for them to get interest in a nuclear power plant.

  1. Application of Nuclear Power Plant Simulator for High School Student Training

    International Nuclear Information System (INIS)

    Kong, Chi Dong; Choi, Soo Young; Park, Min Young; Lee, Duck Jung

    2014-01-01

    In this context, two lectures on nuclear power plant simulator and practical training were provided to high school students in 2014. The education contents were composed of two parts: the micro-physics simulator and the macro-physics simulator. The micro-physics simulator treats only in-core phenomena, whereas the macro-physics simulator describes whole system of a nuclear power plant but it considers a reactor core as a point. The high school students showed strong interests caused by the fact that they operated the simulation by themselves. This abstract reports the training detail and evaluation of the effectiveness of the training. Lectures on nuclear power plant simulator and practical exercises were performed at Ulsan Energy High School and Ulsan Meister High School. Two simulators were used: the macro- and micro-physics simulator. Using the macro-physics simulator, the following five simulations were performed: reactor power increase/decrease, reactor trip, single reactor coolant pump trip, large break loss of coolant accident, and station black-out with D.C. power loss. Using the micro-physics simulator, the following three analyses were performed: the transient analysis, fuel rod performance analysis, and thermal-hydraulics analysis. The students at both high schools showed interest and strong support for the simulator-based training. After the training, the students showed passionate responses that the education was of help for them to get interest in a nuclear power plant

  2. 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

  3. 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.

  4. 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

  5. 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

  6. Nuclear engine system simulation (NESS) program update

    International Nuclear Information System (INIS)

    Scheil, C.M.; Pelaccio, D.G.; Petrosky, L.J.

    1993-01-01

    The second phase of development of a Nuclear Thermal Propulsion (NTP) engine system design analysis code has been completed. The standalone, versatile Nuclear Engine System Simulation (NESS) code provides an accurate, detailed assessment of engine system operating performance, weight, and sizes. The critical information is required to support ongoing and future engine system and stage design study efforts. This recent development effort included incorporation of an updated solid-core nuclear thermal reactor model that yields a reduced core weight and higher fuel power density when compared to a NERVA type reactor. NESS can now analyze expander, gas generator, and bleed cycles, along with multi-redundant propellant pump feed systems. Performance and weight of efficient multi-stage axial turbopump can now be determined, in addition to the traditional centrifugal pump

  7. INTERPLAY OF NEUTRINO OPACITIES IN CORE-COLLAPSE SUPERNOVA SIMULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Lentz, Eric J. [Department of Physics and Astronomy, University of Tennessee, Knoxville, TN 37996-1200 (United States); Mezzacappa, Anthony; Hix, W. Raphael [Physics Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6354 (United States); Messer, O. E. Bronson [National Center for Computational Sciences, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6164 (United States); Bruenn, Stephen W., E-mail: elentz@utk.edu [Department of Physics, Florida Atlantic University, 777 Glades Road, Boca Raton, FL 33431-0991 (United States)

    2012-11-20

    We have conducted a series of numerical experiments using spherically symmetric, general relativistic, neutrino radiation hydrodynamics with the code Agile-BOLTZTRAN to examine the effects of modern neutrino opacities on the development of supernova simulations. We test the effects of opacities by removing opacities or by undoing opacity improvements for individual opacities and groups of opacities. We find that improvements to electron capture (EC) on nuclei, namely EC on an ensemble of nuclei using modern nuclear structure models rather than the simpler independent-particle approximation (IPA) for EC on a mean nucleus, plays the most important role during core collapse of all tested neutrino opacities. Low-energy neutrinos emitted by modern nuclear EC preferentially escape during collapse without the energy downscattering on electrons required to enhance neutrino escape and deleptonization for the models with IPA nuclear EC. During shock breakout the primary influence on the emergent neutrinos arises from non-isoenergetic scattering (NIS) on electrons. For the accretion phase, NIS on free nucleons and pair emission by e {sup +} e {sup -} annihilation have the largest impact on the neutrino emission and shock evolution. Other opacities evaluated, including nucleon-nucleon bremsstrahlung and especially neutrino-positron scattering, have little measurable impact on neutrino emission or shock dynamics. Modern treatments of nuclear EC, e {sup +} e {sup -}-annihilation pair emission, and NIS on electrons and free nucleons are critical elements of core-collapse simulations of all dimensionality.

  8. 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

  9. Monte Carlo simulation of nuclear energy study (II). Annual report on Nuclear Code Evaluation Committee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-01-01

    In the report, research results discussed in 1999 fiscal year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. Present status of Monte Carlo simulation on nuclear energy study was described. Especially, besides of criticality, shielding and core analyses, present status of applications to risk and radiation damage analyses, high energy transport and nuclear theory calculations of Monte Carlo Method was described. The 18 papers are indexed individually. (J.P.N.)

  10. 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

  11. 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

  12. Simulated annealing algorithm for reactor in-core design optimizations

    International Nuclear Information System (INIS)

    Zhong Wenfa; Zhou Quan; Zhong Zhaopeng

    2001-01-01

    A nuclear reactor must be optimized for in core fuel management to make full use of the fuel, to reduce the operation cost and to flatten the power distribution reasonably. The author presents a simulated annealing algorithm. The optimized objective function and the punishment function were provided for optimizing the reactor physics design. The punishment function was used to practice the simulated annealing algorithm. The practical design of the NHR-200 was calculated. The results show that the K eff can be increased by 2.5% and the power distribution can be flattened

  13. 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.

  14. 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

  15. 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

  16. 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.

  17. Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Gilberto Espinosa-Paredes

    2012-01-01

    Full Text Available The loss-of-coolant accident (LOCA simulation in the boiling water reactor (BWR of Laguna Verde Nuclear Power Plant (LVNPP at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

  18. fissioncore: A desktop-computer simulation of a fission-bomb core

    Science.gov (United States)

    Cameron Reed, B.; Rohe, Klaus

    2014-10-01

    A computer program, fissioncore, has been developed to deterministically simulate the growth of the number of neutrons within an exploding fission-bomb core. The program allows users to explore the dependence of criticality conditions on parameters such as nuclear cross-sections, core radius, number of secondary neutrons liberated per fission, and the distance between nuclei. Simulations clearly illustrate the existence of a critical radius given a particular set of parameter values, as well as how the exponential growth of the neutron population (the condition that characterizes criticality) depends on these parameters. No understanding of neutron diffusion theory is necessary to appreciate the logic of the program or the results. The code is freely available in FORTRAN, C, and Java and is configured so that modifications to accommodate more refined physical conditions are possible.

  19. 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.

  20. Uncertainty analysis for the assembly and core simulation of BEAVRS at the HZP conditions

    Energy Technology Data Exchange (ETDEWEB)

    Wan, Chenghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Cao, Liangzhi, E-mail: caolz@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Shen, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2017-04-15

    Highlights: • Uncertainty analysis has been completed based on the “two-step” scheme. • Uncertainty analysis has been performed to BEAVRS at HZP. • For lattice calculations, the few-group constant’s uncertainty was quantified. • For core simulation, uncertainties of k{sub eff} and power distributions were quantified. - Abstract: Based on the “two-step” scheme for the reactor-physics calculations, the capability of uncertainty analysis for the core simulations has been implemented in the UNICORN code, an in-house code for the sensitivity and uncertainty analysis of the reactor-physics calculations. Applying the statistical sampling method, the nuclear-data uncertainties can be propagated to the important predictions of the core simulations. The uncertainties of the few-group constants introduced by the uncertainties of the multigroup microscopic cross sections are quantified first for the lattice calculations; the uncertainties of the few-group constants are then propagated to the core multiplication factor and core power distributions for the core simulations. Up to now, our in-house lattice code NECP-CACTI and the neutron-diffusion solver NECP-VIOLET have been implemented in UNICORN for the steady-state core simulations based on the “two-step” scheme. With NECP-CACTI and NECP-VIOLET, the modeling and simulation of the steady-state BEAVRS benchmark problem at the HZP conditions was performed, and the results were compared with those obtained by CASMO-4E. Based on the modeling and simulation, the UNICORN code has been applied to perform the uncertainty analysis for BAEVRS at HZP. The uncertainty results of the eigenvalues and two-group constants for the lattice calculations and the multiplication factor and the power distributions for the steady-state core simulations are obtained and analyzed in detail.

  1. Uncertainty analysis for the assembly and core simulation of BEAVRS at the HZP conditions

    International Nuclear Information System (INIS)

    Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei

    2017-01-01

    Highlights: • Uncertainty analysis has been completed based on the “two-step” scheme. • Uncertainty analysis has been performed to BEAVRS at HZP. • For lattice calculations, the few-group constant’s uncertainty was quantified. • For core simulation, uncertainties of k_e_f_f and power distributions were quantified. - Abstract: Based on the “two-step” scheme for the reactor-physics calculations, the capability of uncertainty analysis for the core simulations has been implemented in the UNICORN code, an in-house code for the sensitivity and uncertainty analysis of the reactor-physics calculations. Applying the statistical sampling method, the nuclear-data uncertainties can be propagated to the important predictions of the core simulations. The uncertainties of the few-group constants introduced by the uncertainties of the multigroup microscopic cross sections are quantified first for the lattice calculations; the uncertainties of the few-group constants are then propagated to the core multiplication factor and core power distributions for the core simulations. Up to now, our in-house lattice code NECP-CACTI and the neutron-diffusion solver NECP-VIOLET have been implemented in UNICORN for the steady-state core simulations based on the “two-step” scheme. With NECP-CACTI and NECP-VIOLET, the modeling and simulation of the steady-state BEAVRS benchmark problem at the HZP conditions was performed, and the results were compared with those obtained by CASMO-4E. Based on the modeling and simulation, the UNICORN code has been applied to perform the uncertainty analysis for BAEVRS at HZP. The uncertainty results of the eigenvalues and two-group constants for the lattice calculations and the multiplication factor and the power distributions for the steady-state core simulations are obtained and analyzed in detail.

  2. 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

  3. Groundwater-stream-simulation experiments for the evaluation of the safety of proposed nuclear waste repositories

    International Nuclear Information System (INIS)

    Seitz, M.G.

    1981-01-01

    A bench-scale experimental design which integrates repository components to simulate a groundwater stream infiltrating a breached repository is described in this paper. An experiment performed with a nuclear waste solid and one rock core is briefly summarized. The nuclear waste solid consists of borosilicate glass containing formulated nuclear waste and is the source of the leached radionuclides. The rock core used is of granite and serves as the adsorption medium for the leached radionuclides

  4. 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

  5. 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

  6. Monte Carlo simulation of a TRIGA source driven core configuration: Preliminary results

    International Nuclear Information System (INIS)

    Burgio, N.; Ciavola, C.; Santagata, A.

    2002-01-01

    The different core configurations with a k eff ranging from 0.93 to 0.98, and their response when driven by a pulsed neutron source were simulated with MCNP4C3 (Los Alamos - Monte Carlo N Particles). Simulation results could be considered both as preliminary check for nuclear data and a conceptual design for 'source jerk' experiments on the frame of TRIGA Accelerator Driven Experiment (TRADE) on the reactor facility of Casaccia research center. (author)

  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. 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

  9. Fundamental experiment on simulated molten core/concrete interaction

    International Nuclear Information System (INIS)

    Toda, S.; Katsumura, Y.

    1994-01-01

    If a complete and prolonged failure of coolant flow were to occur in a LWR or FBR, fission product decay heat would cause the fuel to overheat. If no available action to cool the fuel were taken, it would eventually melt. Ibis could lead to slumping of the molten core material and to the failure of the reactor pressure vessel and deposition of these materials into the concrete reactor cavity. Consequently, the molten core could melt and decompose the concrete. Vigorous agitation of the molten core pool by concrete decomposition gases is expected to enhance the convective heat transfer process. Besides the decomposition gases, melting concrete (slag) generated under the molten core pool will be buoyed up, and will also affect the downward heat transfer. Though, in this way, the heat transfer process across the interface is complicated by the slag and the gases evolved from the decomposed concrete, it is very important to make its process clear for the safety evaluation of nuclear reactors. Therefore, in this study, fundamental experiments were performed using simulated materials to observe the behaviors of the hot pool, slag and gases at the interface. Moreover, from the experimental observation, a correlation without empirical constants was proposed to calculate the interface heat transfer. The heat transfer across the interface would depend on thermo-physical interactions between the pool, slag and concrete which are changed by their thermal properties and interface temperature and so on. For example, the molten concrete is miscible in molten oxidic core debris, but is immiscible in metallic core debris. If a contact temperature between the molten core pool and the concrete falls below the solidus of the pool, solidification of the pool will occur. In this study, the case of immiscible slag in the pool is treated and solidification of the pool does not occur. Thus, water, paraffin and air were selected as the simulated molten core pool, concrete, and decomposition

  10. Palo Verde nuclear generating station EASEplus SIMULATE model

    International Nuclear Information System (INIS)

    McDonald, W.F.; Reed, M.L.; Fauste, J.L.

    1992-01-01

    The Palo Verde on-site reactor engineers have an extremely powerful and accurate tool for quickly predicting the effects of reactor power maneuvers on core axial shape index (ASI) and xenon worth. They can analyze postulated future power maneuvers quickly and supply the reactor operators with valuable predictions without having to consult with the off-site nuclear analysis group. The tool developed by the nuclear analysis group was an advanced nodal code with a graphic user interface (GUI) driver for ease of use. The advanced nodal code used was the Studsvik of America SIMULATE-3 Version 2.20-DSI. This SIMULATE version was compiled for use on a personal computer (PC) with a Definicon Systems' 50-MHz coprocessor board. The GUI face used was Expert-EASE Systems' EASE+SIM3 Version 3.0 pre-/postprocessor. The system was installed on Compaq Deskpro 386/20e PCs located in the control room of each of the three units, in the reactor engineering office, in the nuclear analysis office, and in the control room of the training simulator

  11. Application of Nuclear Application Programs to APR1400 Simulator

    International Nuclear Information System (INIS)

    Hwang, Do Hyun; Lee, Myeong Soo; Hong, Jin Hyuk

    2012-01-01

    Advanced Power Reactor 1400MWe (APR1400) simulator has been developed and installed at Kori Training Center for operators of ShinKori no.3, 4 nuclear power plant by Korea Hydro and Nuclear Power,s Central Research Institute (KHNP CRI). NAPS (Nuclear Application Programs) is a computerbased system which provides operators with past and real-time information for monitoring and controlling NSSS (Nuclear Steam Supply System), BOP (Balance Of Plant) and Electric system. NAPS consists of several programs such as COLSS (Core Operating Limit Supervisory System), SPADES+ (Safety Parameter Display and Evaluation System), CEA (Control Element Assembly) Application Program, and so on. Each program makes calculations based on its own algorithm and provides information available for operation. In order to use NAPS programs with a simulator even though they are being used in a real plant, they should be modified to add several simulation functions such as reset, snap, run/freeze and backtrack required by ANSI/ANS-3.5 to the original NAPS functionality. On top of that, interfacing programs should be developed for the data communication between respective NAPS programs and simulator sever. The purpose of this paper is to provide the overall architecture of the communication system between NAPS and simulator model, and to describe the method to apply NAPS to APR1400 simulator

  12. Fast core prediction simulator for load follow control

    International Nuclear Information System (INIS)

    Yim, Man Sung; Lee, Sang Hoon; Lee, Un Chul

    1990-01-01

    An operator-assisting system for the reactor core control under power changing operating condition was developed. The system is consisted of core simulator routine and Xenon and Iodine initial condition generation routine. The initial condition generation routine, without exactly knowing the core status, is capable of providing accurate number densities and axial offset conditions of Xenon and Iodine after several hours of predictor- corrector calculations using the plant instrumentation signals of power level and power axial offset. The core simulator routine, even with the two node core model, gives equivalently accurate results as the one-dimensional model for the core behaviour simulation under power changing condition and can provide proper control strategies for load follow operation. The core simulator can also be used by the operator to develop remedial actions to restore the distorted power distribution by using its prediction capability

  13. 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)

  14. 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)

  15. 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.

  16. Annual meeting on nuclear technology '85. Technical session on nuclear power plant simulators. Jahrestagung Kerntechnik '85. Fachsitzung Simulatoren bei Kernkraftwerken

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    The ten lectures read at this Technical Session deal with nuclear power plant simulators and their functions for the training of control room personnel, with test facilities for analysis and simulation of incidents and the relevant plant behaviour, and with the computer codes for fast simulation of reactor processes. The papers explain in particular the HAMMLAB project, the COSIMA fast core simulator, and the OPAL simulator for the KWU convoy process information system named PRINS.

  17. 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

  18. 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

  19. Nuclear ship engineering simulator

    International Nuclear Information System (INIS)

    Itoh, Yasuyoshi; Kusunoki, Tsuyoshi; Hashidate, Koji

    1991-01-01

    The nuclear ship engineering simulator, which analyzes overall system response of nuclear ship numerically, is now being developed by JAERI as an advanced design tool with the latest computer technology in software and hardware. The development of the nuclear ship engineering simulator aims at grasping characteristics of a reactor plant under the situation generated by the combination of ocean, a ship hull and a reactor. The data from various tests with the nuclear ship 'MUTSU' will be used for this simulator to modulate and verify its functions of reproducing realistic response of nuclear ship, and then the simulator will be utilized for the research and development of advanced marine reactors. (author)

  20. Simulation of the Long period Core Design for WH type of KHNP

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ji-Eun; Moon, Sang-Rae [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve.

  1. Simulation of the Long period Core Design for WH type of KHNP

    International Nuclear Information System (INIS)

    Jung, Ji-Eun; Moon, Sang-Rae

    2016-01-01

    The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve

  2. 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

  3. 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.)

  4. Adaptive Core Simulation Employing Discrete Inverse Theory - Part I: Theory

    International Nuclear Information System (INIS)

    Abdel-Khalik, Hany S.; Turinsky, Paul J.

    2005-01-01

    Use of adaptive simulation is intended to improve the fidelity and robustness of important core attribute predictions such as core power distribution, thermal margins, and core reactivity. Adaptive simulation utilizes a selected set of past and current reactor measurements of reactor observables, i.e., in-core instrumentation readings, to adapt the simulation in a meaningful way. A meaningful adaption will result in high-fidelity and robust adapted core simulator models. To perform adaption, we propose an inverse theory approach in which the multitudes of input data to core simulators, i.e., reactor physics and thermal-hydraulic data, are to be adjusted to improve agreement with measured observables while keeping core simulator models unadapted. At first glance, devising such adaption for typical core simulators with millions of input and observables data would spawn not only several prohibitive challenges but also numerous disparaging concerns. The challenges include the computational burdens of the sensitivity-type calculations required to construct Jacobian operators for the core simulator models. Also, the computational burdens of the uncertainty-type calculations required to estimate the uncertainty information of core simulator input data present a demanding challenge. The concerns however are mainly related to the reliability of the adjusted input data. The methodologies of adaptive simulation are well established in the literature of data adjustment. We adopt the same general framework for data adjustment; however, we refrain from solving the fundamental adjustment equations in a conventional manner. We demonstrate the use of our so-called Efficient Subspace Methods (ESMs) to overcome the computational and storage burdens associated with the core adaption problem. We illustrate the successful use of ESM-based adaptive techniques for a typical boiling water reactor core simulator adaption problem

  5. GPU based numerical simulation of core shooting process

    Directory of Open Access Journals (Sweden)

    Yi-zhong Zhang

    2017-11-01

    Full Text Available Core shooting process is the most widely used technique to make sand cores and it plays an important role in the quality of sand cores. Although numerical simulation can hopefully optimize the core shooting process, research on numerical simulation of the core shooting process is very limited. Based on a two-fluid model (TFM and a kinetic-friction constitutive correlation, a program for 3D numerical simulation of the core shooting process has been developed and achieved good agreements with in-situ experiments. To match the needs of engineering applications, a graphics processing unit (GPU has also been used to improve the calculation efficiency. The parallel algorithm based on the Compute Unified Device Architecture (CUDA platform can significantly decrease computing time by multi-threaded GPU. In this work, the program accelerated by CUDA parallelization method was developed and the accuracy of the calculations was ensured by comparing with in-situ experimental results photographed by a high-speed camera. The design and optimization of the parallel algorithm were discussed. The simulation result of a sand core test-piece indicated the improvement of the calculation efficiency by GPU. The developed program has also been validated by in-situ experiments with a transparent core-box, a high-speed camera, and a pressure measuring system. The computing time of the parallel program was reduced by nearly 95% while the simulation result was still quite consistent with experimental data. The GPU parallelization method can successfully solve the problem of low computational efficiency of the 3D sand shooting simulation program, and thus the developed GPU program is appropriate for engineering applications.

  6. 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.

  7. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Vargas O, D.; Chavez M, C.

    2012-10-01

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  8. 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

  9. Apparatus for simulating a reactor core

    International Nuclear Information System (INIS)

    Yokomizo, Osamu; Kiguchi, Takashi; Motoda, Hiroshi; Takeda, Renzo.

    1975-01-01

    Object: To facilitate searching of input and output of information and to efficiently perform trial-and-error in a short time. Structure: Kinds of necessary input information are stored in an input information converter and are displayed by an image display through an image control. An operator operates an information input device to input information. This input information is converted by the input information converter into a form used in a reactor core simulation counter. The reactor core simulation counter simulates a state of the core to the input information converted, and outputs it as an output information. An output information converter converts output information into a form that may be displayed as an image and feeds it to the image control. The operator may correct the input information while viewing the output information displayed on the image display to immediately perform succeeding calculation. (Kamimura, M.)

  10. Adaption of core simulations to detector readings

    International Nuclear Information System (INIS)

    Lindahl, S.Oe.

    1985-05-01

    The shortcomings of the conventional core supervision methods are briefly discussed. A new strategy for core surveillance is proposed The strategy is based on a combination of analytical evaluation of detailed core power and adaption of these to detector measurements. The adaption is carried out 1) each time the simulator is executed by use of averaged detector readings and 2) once a year (approximately) in which case the coefficients of the simulator's equations are overviewed. In the yearly overview, calculations are tuned to measurements (TIP, γ-scannings, k-eff) by parameter optimization or by inversion of the diffusion equation. The proposed strategy is believed to increase the accuracy of the core surveillance, to yield improved thermal margins, to increase the accuracy of core predictions and design calculations, and to lessen the dependence of core surveillance on the detector equipment. (author)

  11. Analysis of core and core barrel heat-up under conditions simulating severe reactor accidents

    International Nuclear Information System (INIS)

    Chellaiah, S.; Viskanta, R.; Ranganathan, P.; Anand, N.K.

    1987-01-01

    This paper reports on the development of a model for estimating the temperature distributions in the reactor core, core barrel, thermal shield and reactor pressure vessel of a PWR during an undercooling transient. A number of numerical calculations simulating the core uncovering of the TMI-2 reactor and the subsequent heat-up of the core have been performed. The results of the calculations show that the exothermic heat release due to Zircaloy oxidation contributes to the sharp heat-up of the core. However, the core barrel temperature rise which is driven by the temperature increase of the edge of the core (e.g., the core baffle) is very modest. The maximum temperature of the core barrel never exceeded 610 K (at a system pressure of 68 bar) after a 75 minute simulation following the start of core uncovering

  12. 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.

  13. Neutronics simulations on hypothetical power excursion and possible core melt scenarios in CANDU6

    International Nuclear Information System (INIS)

    Kim, Yonghee

    2015-01-01

    LOCA (Loss of coolant accident) is an outstanding safety issue in the CANDU reactor system since the coolant void reactivity is strongly positive. To deal with the LOCA, the CANDU systems are equipped with specially designed quickly-acting secondary shutdown system. Nevertheless, the so-called design-extended conditions are requested to be taken into account in the safety analysis for nuclear reactor systems after the Fukushima accident. As a DEC scenario, the worst accident situation in a CANDU reactor system is a unprotected LOCA, which is supposed to lead to a power excursion and possibly a core melt-down. In this work, the hypothetical unprotected LOCA scenario is simulated in view of the power excursion and fuel temperature changes by using a simplified point-kinetics (PK) model accounting for the fuel temperature change. In the PK model, the core reactivity is assumed to be affected by a large break LOCA and the fuel temperature is simulated to account for the Doppler effect. In addition, unlike the conventional PK simulation, we have also considered the Xe-I model to evaluate the impact of Xe during the LOCA. Also, we tried to simulate the fuel and core melt-down scenario in terms of the reactivity through a series of neutronics calculations for hypothetical core conditions. In case of a power excursion and possible fuel melt-down situation, the reactor system behavior is very uncertain. In this work, we tried to understand the impacts of fuel melt and relocation within the pressure vessel on the core reactivity and failure of pressure and calandria tubes. (author)

  14. 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)

  15. Adaptive Core Simulation Employing Discrete Inverse Theory - Part II: Numerical Experiments

    International Nuclear Information System (INIS)

    Abdel-Khalik, Hany S.; Turinsky, Paul J.

    2005-01-01

    Use of adaptive simulation is intended to improve the fidelity and robustness of important core attribute predictions such as core power distribution, thermal margins, and core reactivity. Adaptive simulation utilizes a selected set of past and current reactor measurements of reactor observables, i.e., in-core instrumentation readings, to adapt the simulation in a meaningful way. The companion paper, ''Adaptive Core Simulation Employing Discrete Inverse Theory - Part I: Theory,'' describes in detail the theoretical background of the proposed adaptive techniques. This paper, Part II, demonstrates several computational experiments conducted to assess the fidelity and robustness of the proposed techniques. The intent is to check the ability of the adapted core simulator model to predict future core observables that are not included in the adaption or core observables that are recorded at core conditions that differ from those at which adaption is completed. Also, this paper demonstrates successful utilization of an efficient sensitivity analysis approach to calculate the sensitivity information required to perform the adaption for millions of input core parameters. Finally, this paper illustrates a useful application for adaptive simulation - reducing the inconsistencies between two different core simulator code systems, where the multitudes of input data to one code are adjusted to enhance the agreement between both codes for important core attributes, i.e., core reactivity and power distribution. Also demonstrated is the robustness of such an application

  16. Static core performance simulator SCOPERS-2 for light water reactors and its application

    International Nuclear Information System (INIS)

    Shimooke, Takanori; Itagaki, Masafumi; Osanai, Masao.

    1979-05-01

    SCOPERS-2 is a generalized FLARE-type computer program simulating both PWR and BWR. Features and the calculation model (generalized FLARE-type nodal equation, migration kernel, etc.) are first described. A calculation is then given of the core of nuclear ship MUTSU (PWR) for an example of the code application. The power distribution calculated by SCOPERS-2 and by CITATION (3-dimensional diffusion code) are in good agreement. (author)

  17. Multidimensional simulations of core-collapse supernovae with CHIMERA

    Science.gov (United States)

    Lentz, Eric J.; Bruenn, S. W.; Yakunin, K.; Endeve, E.; Blondin, J. M.; Harris, J. A.; Hix, W. R.; Marronetti, P.; Messer, O. B.; Mezzacappa, A.

    2014-01-01

    Core-collapse supernovae are driven by a multidimensional neutrino radiation hydrodynamic (RHD) engine, and full simulation requires at least axisymmetric (2D) and ultimately symmetry-free 3D RHD simulation. We present recent and ongoing work with our multidimensional RHD supernova code CHIMERA to understand the nature of the core-collapse explosion mechanism and its consequences. Recently completed simulations of 12-25 solar mass progenitors(Woosley & Heger 2007) in well resolved (0.7 degrees in latitude) 2D simulations exhibit robust explosions meeting the observationally expected explosion energy. We examine the role of hydrodynamic instabilities (standing accretion shock instability, neutrino driven convection, etc.) on the explosion dynamics and the development of the explosion energy. Ongoing 3D and 2D simulations examine the role that simulation resolution and the removal of the imposed axisymmetry have in the triggering and development of an explosion from stellar core collapse. Companion posters will explore the gravitational wave signals (Yakunin et al.) and nucleosynthesis (Harris et al.) of our simulations.

  18. 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

  19. 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

  20. 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)

  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. 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

  3. Right Size Determining the Staff Necessary to Sustain Simulation and Computing Capabilities for Nuclear Security

    Energy Technology Data Exchange (ETDEWEB)

    Nikkel, Daniel J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meisner, Robert [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2010-09-10

    The Advanced Simulation and Computing Campaign, herein referred to as the ASC Program, is a core element of the science-based Stockpile Stewardship Program (SSP), which enables assessment, certification, and maintenance of the safety, security, and reliability of the U.S. nuclear stockpile without the need to resume nuclear testing. The use of advanced parallel computing has transitioned from proof-of-principle to become a critical element for assessing and certifying the stockpile. As the initiative phase of the ASC Program came to an end in the mid-2000s, the National Nuclear Security Administration redirected resources to other urgent priorities, and resulting staff reductions in ASC occurred without the benefit of analysis of the impact on modern stockpile stewardship that is dependent on these new simulation capabilities. Consequently, in mid-2008 the ASC Program management commissioned a study to estimate the essential size and balance needed to sustain advanced simulation as a core component of stockpile stewardship. The ASC Program requires a minimum base staff size of 930 (which includes the number of staff necessary to maintain critical technical disciplines as well as to execute required programmatic tasks) to sustain its essential ongoing role in stockpile stewardship.

  4. Software for simulation of nuclear simulation of nuclear installations

    International Nuclear Information System (INIS)

    Castaneda, J.O.; Ramos, L.M.; Arjona, O.; Rodriguez, L.

    1993-01-01

    The software is an instrument to build conceptual-type simulators of low, medium and full scale for used in nuclear installations. The system is composed by composed by two basic modules: one for the edition and the other for the simulation. The first one allows to prepare the information to simulate: mathematical model, technological design (fundamentally, operation board or mnemotechnical design), parameters to be shown, failures to be simulated

  5. 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.

  6. 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.

  7. Current status of relativistic core collapse simulations

    Energy Technology Data Exchange (ETDEWEB)

    Font, Jose A [Departamento de Astronomia y Astrofisica, Universidad de Valencia, Dr. Moliner 50, 46100 Burjassot (Valencia) (Spain)

    2007-05-15

    With the first generation of ground-based gravitational wave laser interferometers already taking data, the availability of reliable waveform templates from astrophysical sources, which may help extract the signal from the anticipated noisy data, is urgently required. Gravitational stellar core collapse supernova has traditionally been considered among the most important astrophysical sources of potentially detectable gravitational radiation. Only very recently the first multidimensional simulations of relativistic rotational core collapse have been possible (albeit for models with simplified input physics), thanks to the use of conservative formulations of the hydrodynamics equations and advanced numerical methodology, as well as stable formulations of Einstein's equations. In this paper, the current status of relativistic core collapse simulations is discussed, with the emphasis given to the modelling of the collapse dynamics and to the computation of the gravitational radiation in the existing numerical approaches. Work employing the conformally-flat approximation (CFC) of the 3+1 Einstein's equations is reported, as well as extensions of this approximation (CFC+) and investigations within the framework of the so-called BSSN formulation of the 3+1 gravitational field equations (with no approximation for the spacetime dynamics). On the other hand, the incorporation of magnetic fields and the MHD equations in numerical codes to improve the realism of core collapse simulations in general relativity, is currently an emerging field where significant progress is bound to be soon achieved. The paper also contains a brief discussion of magneto-rotational simulations of core collapse, aiming at addressing the effects of magnetic fields on the collapse dynamics and on the gravitational waveforms.

  8. Current status of relativistic core collapse simulations

    International Nuclear Information System (INIS)

    Font, Jose A

    2007-01-01

    With the first generation of ground-based gravitational wave laser interferometers already taking data, the availability of reliable waveform templates from astrophysical sources, which may help extract the signal from the anticipated noisy data, is urgently required. Gravitational stellar core collapse supernova has traditionally been considered among the most important astrophysical sources of potentially detectable gravitational radiation. Only very recently the first multidimensional simulations of relativistic rotational core collapse have been possible (albeit for models with simplified input physics), thanks to the use of conservative formulations of the hydrodynamics equations and advanced numerical methodology, as well as stable formulations of Einstein's equations. In this paper, the current status of relativistic core collapse simulations is discussed, with the emphasis given to the modelling of the collapse dynamics and to the computation of the gravitational radiation in the existing numerical approaches. Work employing the conformally-flat approximation (CFC) of the 3+1 Einstein's equations is reported, as well as extensions of this approximation (CFC+) and investigations within the framework of the so-called BSSN formulation of the 3+1 gravitational field equations (with no approximation for the spacetime dynamics). On the other hand, the incorporation of magnetic fields and the MHD equations in numerical codes to improve the realism of core collapse simulations in general relativity, is currently an emerging field where significant progress is bound to be soon achieved. The paper also contains a brief discussion of magneto-rotational simulations of core collapse, aiming at addressing the effects of magnetic fields on the collapse dynamics and on the gravitational waveforms

  9. 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

  10. A simulator-independent optimization tool based on genetic algorithm applied to nuclear reactor design

    International Nuclear Information System (INIS)

    Abreu Pereira, Claudio Marcio Nascimento do; Schirru, Roberto; Martinez, Aquilino Senra

    1999-01-01

    Here is presented an engineering optimization tool based on a genetic algorithm, implemented according to the method proposed in recent work that has demonstrated the feasibility of the use of this technique in nuclear reactor core designs. The tool is simulator-independent in the sense that it can be customized to use most of the simulators which have the input parameters read from formatted text files and the outputs also written from a text file. As the nuclear reactor simulators generally use such kind of interface, the proposed tool plays an important role in nuclear reactor designs. Research reactors may often use non-conventional design approaches, causing different situations that may lead the nuclear engineer to face new optimization problems. In this case, a good optimization technique, together with its customizing facility and a friendly man-machine interface could be very interesting. Here, the tool is described and some advantages are outlined. (author)

  11. 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

  12. Reactor core simulations in Canada

    International Nuclear Information System (INIS)

    Roy, R.; Koclas, J.; Shen, W.; Jenkins, D. A.; Altiparmakov, D.; Rouben, B.

    2004-01-01

    This review will address the current simulation flow-chart currently used for reactor-physics simulations in the Canadian industry. The neutron behaviour in heavy-water moderated power reactors is quite different from that in other power reactors, thus the core physics approximations are somewhat different Some codes used are particular to the context of heavy-water reactors, and the paper focuses on this aspect. The paper also shows simulations involving new design features of the Advanced Candu Reactor TM (ACR TM), and provides insight into future development, expected in the coming years. (authors)

  13. Three-dimensional discrete element method simulation of core disking

    Science.gov (United States)

    Wu, Shunchuan; Wu, Haoyan; Kemeny, John

    2018-04-01

    The phenomenon of core disking is commonly seen in deep drilling of highly stressed regions in the Earth's crust. Given its close relationship with the in situ stress state, the presence and features of core disking can be used to interpret the stresses when traditional in situ stress measuring techniques are not available. The core disking process was simulated in this paper using the three-dimensional discrete element method software PFC3D (particle flow code). In particular, PFC3D is used to examine the evolution of fracture initiation, propagation and coalescence associated with core disking under various stress states. In this paper, four unresolved problems concerning core disking are investigated with a series of numerical simulations. These simulations also provide some verification of existing results by other researchers: (1) Core disking occurs when the maximum principal stress is about 6.5 times the tensile strength. (2) For most stress situations, core disking occurs from the outer surface, except for the thrust faulting stress regime, where the fractures were found to initiate from the inner part. (3) The anisotropy of the two horizontal principal stresses has an effect on the core disking morphology. (4) The thickness of core disk has a positive relationship with radial stress and a negative relationship with axial stresses.

  14. A Personal Computer-Based Simulator for Nuclear-Heating Reactors

    International Nuclear Information System (INIS)

    Liu Jie; Zhang Zuoyi; Lu Dongsen; Shi Zhengang; Chen Xiaoming; Dong Yujie

    2000-01-01

    A personal computer (PC)-based simulator for nuclear-heating reactors (NHRs), PC-NHR, has been developed to provide an educational tool for understanding the design and operational characteristics of an NHR system. A general description of the reactor system as well as the technical basis for the design and operation of the heating reactor is provided. The basic models and equations for the NHR simulation are then given, which include models of the reactor core, the reactor coolant system, the containment, and the control system. The graphical user interface is described in detail to provide a manual for the user to operate the simulator properly. Steady state and several transients have been simulated. The results of PC-NHR are in good agreement with design data and the results of RETRAN-02. The real-time capability is also confirmed

  15. 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

  16. 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.)

  17. Geoelectrical monitoring of simulated subsurface leakage to support high-hazard nuclear decommissioning at the Sellafield Site, UK

    Energy Technology Data Exchange (ETDEWEB)

    Kuras, Oliver, E-mail: oku@bgs.ac.uk [British Geological Survey, Keyworth, Nottingham NG12 5GG (United Kingdom); Wilkinson, Paul B.; Meldrum, Philip I.; Oxby, Lucy S. [British Geological Survey, Keyworth, Nottingham NG12 5GG (United Kingdom); Uhlemann, Sebastian [British Geological Survey, Keyworth, Nottingham NG12 5GG (United Kingdom); ETH-Swiss Federal Institute of Technology, Institute of Geophysics, Sonneggstr. 5, 8092 Zurich (Switzerland); Chambers, Jonathan E. [British Geological Survey, Keyworth, Nottingham NG12 5GG (United Kingdom); Binley, Andrew [Lancaster Environment Centre, Lancaster University, Lancaster LA1 4YQ (United Kingdom); Graham, James [National Nuclear Laboratory, Central Laboratory, Sellafield, Seascale, Cumbria CA20 1PG (United Kingdom); Smith, Nicholas T. [National Nuclear Laboratory, Central Laboratory, Sellafield, Seascale, Cumbria CA20 1PG (United Kingdom); School of Earth, Atmospheric and Environmental Sciences, Williamson Building, University of Manchester, Oxford Road, Manchester M13 9PL (United Kingdom); Atherton, Nick [Sellafield Ltd, Albion Square, Swingpump Lane, Whitehaven CA28 7NE (United Kingdom)

    2016-10-01

    A full-scale field experiment applying 4D (3D time-lapse) cross-borehole Electrical Resistivity Tomography (ERT) to the monitoring of simulated subsurface leakage was undertaken at a legacy nuclear waste silo at the Sellafield Site, UK. The experiment constituted the first application of geoelectrical monitoring in support of decommissioning work at a UK nuclear licensed site. Images of resistivity changes occurring since a baseline date prior to the simulated leaks revealed likely preferential pathways of silo liquor simulant flow in the vadose zone and upper groundwater system. Geophysical evidence was found to be compatible with historic contamination detected in permeable facies in sediment cores retrieved from the ERT boreholes. Results indicate that laterally discontinuous till units forming localized hydraulic barriers substantially affect flow patterns and contaminant transport in the shallow subsurface at Sellafield. We conclude that only geophysical imaging of the kind presented here has the potential to provide the detailed spatial and temporal information at the (sub-)meter scale needed to reduce the uncertainty in models of subsurface processes at nuclear sites. - Graphical abstract: 3D fractional resistivity change (resistivity change Δρ divided by baseline resistivity ρ{sub 0}) image showing results of Stage 1 silo liquor simulant injection. The black line delineates the preferential flow path; green cylinders show regions of historic contamination found in sediment cores from ERT boreholes. - Highlights: • 4D geoelectrical monitoring at Sellafield detected and tracked simulated silo leaks. • ERT revealed likely pathways of silo liquor simulant flow in the subsurface. • The method can reduce uncertainty in subsurface process models at nuclear sites. • Has been applied in this form at a UK nuclear licensed site for the first time • Study demonstrates value of 4D geophysics for nuclear decommissioning.

  18. 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.

  19. 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

  20. An approach to evaluate the cutting time for the nuclear dismantling simulation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jonghwan; Hyun, Dongjun; Kang, Sinyoung; Kim, Ikjune; Jeong, Kwan-Seong; Choi, Byung-Seon; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Nuclear power plant (NPP) decommissioning involves various processes and technologies. Decommissioning should be performed after a comprehensive review of the information related to these processes and technologies. There are various means of prior examination and evaluation to ensure the feasibility and safety of the decommissioning process plan. Our dismantling simulation system aims to simulate and evaluate whole processes related to the dismantlement of core equipment of NPP such as the device preparation, cutting operation, waste transfer, and so on. This paper introduces the estimation methodology of the time required for the cutting processes based on real cutting conditions in order to provide effective economic evaluation functionalities used for the system. The methodology to estimate the time required for the remote cutting process in the nuclear dismantling simulation system was proposed. Among the factors which mainly determine the time, the cutting trace was directly calculated from the simulation system and the continuous cutting speed was obtained by proper order of the spline fitting with constraint conditions.

  1. Accident sequences simulated at the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1998-01-01

    Different hypothetical accident sequences have been simulated at Unit 1 of the Juragua nuclear power plant in Cuba, a plant with two VVER-440 V213 units under construction. The computer code MELCOR was employed for these simulations. The sequences simulated are: (1) a design-basis accident (DBA) large loss of coolant accident (LOCA) with the emergency core coolant system (ECCS) on, (2) a station blackout (SBO), (3) a small LOCA (SLOCA) concurrent with SBO, (4) a large LOCA (LLOCA) concurrent with SBO, and (5) a LLOCA concurrent with SBO and with the containment breached at time zero. Timings of important events and source term releases have been calculated for the different sequences analyzed. Under certain weather conditions, the fission products released from the severe accident sequences may travel to southern Florida

  2. Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M.

    2013-01-01

    In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

  3. Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M., E-mail: amir@cdtn.br, E-mail: hcr@cdtn.br, E-mail: aacs@cdtn.br, E-mail: vitors@cdtn.br, E-mail: campolina@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/UFMG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel Artur P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

  4. LWR-core behaviour project

    International Nuclear Information System (INIS)

    Paratte, J.M.

    1982-07-01

    The LWR-Core behaviour project concerns the mathematical simulation of a light water reactor in normal operation (emergency situations excluded). Computational tools are assembled, i.e. programs and libraries of data. These computational tools can likewise be used in nuclear power applications, industry and control applications. The project is divided into three parts: the development and application of calculation methods for quantisation determination of LWR physics; investigation of the behaviour of nuclear fuels under radiation with special attention to higher burnup; simulation of the operating transients of nuclear power stations. (A.N.K.)

  5. Simulation of a postulated 2% cold leg break in Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Palmieiri, Elcio Tadeu; Azevedo, Carlos Vicente Goulart de; Aronne, Ivan Dionysio

    2007-01-01

    This paper presents the simulation of a 2% break in the cold leg pipe of Angra 2 nuclear power plant, with the computer code RELAP5/Mod3.3. The main boundary conditions specified for this simulation were: no injection from high pressure injection system; enhanced depressurization of the primary system by opening the pressure operated relief valve (PORV) and the safety relief valve (SRV) when core temperature reaches circa 100 K above saturation; and accumulator injection starting at 2.7 MPa. The specific objectives to be addressed with this simulation are: the core boil-off and dryout at relatively high pressure in the primary system; the phenomena during enhanced primary depressurization; the effectiveness of hot leg accumulator injection into the partially uncovered rod bundle; and the core rewetting. The results obtained were compared with the Lobi A1-93 test, which was performed under the same boundary conditions. This activity was executed in the scope of IAEA research project Evaluation of Uncertainties in the Simulation of Accidents in Angra 2 using RELAP5/MOD3 Code Applying CIAU Methodology (author)

  6. 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

  7. Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

    International Nuclear Information System (INIS)

    Guzina, Bojan; Kunerth, Dennis

    2014-01-01

    A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

  8. Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Guzina, Bojan [Univ. of Minnesota, Minneapolis, MN (United States); Kunerth, Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-30

    A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

  9. Influence of core sand properties on flow dynamics of core shooting process based on experiment and multiphase simulation

    Directory of Open Access Journals (Sweden)

    Chang-jiang Ni

    2017-03-01

    Full Text Available The influence of core sand properties on flow dynamics was investigated synchronously with various core sands, transparent core-box and high-speed camera. To confirm whether the core shooting process has significant turbulence, the flow pattern of sand particles in the shooting head and core box was reproduced with colored core sands. By incorporating the kinetic theory of granular flow (KTGF, kinetic-frictional constitutive correlation and turbulence model, a two-fluid model (TFM was established to study the flow dynamics of the core shooting process. Two-fluid model (TFM simulations were then performed and a areasonable agreement was achieved between the simulation and experimental results. Based on the experimental and simulation results, the effects of turbulence, sand density, sand diameter and binder ratio were analyzed in terms of filling process, sand volume fraction (αs and sand velocity (Vs.

  10. Simulation and Comparison of the Calorimeters Measuring the Nuclear Heating in the OSIRIS Reactor, with the TRIPOLI-4R Monte-Carlo Code

    International Nuclear Information System (INIS)

    Peron, A.; Malouch, F.; Diop, C.M.

    2013-06-01

    Two calorimeter devices are used in the OSIRIS MTR reactor (CEA-Saclay center) for the nuclear heating measurements. The first one is a fixed five-stage calorimeter device. The second one is an innovative mobile probe called 'CALMOS'. The design of these devices is different (in particular their geometry), implying modifications on the local neutron and photon fluxes and hence on nuclear heating measured values. The measurements performed by the two calorimeter devices cannot directly be compared; this requires perfect irradiation conditions in the reactor core, especially for the core loading and the control element positions. Simulation is here a good help to perform a fully relevant comparison. In this paper, differences between calorimeter devices in terms of nuclear heating and particle fluxes are evaluated using the TRIPOLI-4 Monte-Carlo code. After a description of the OSIRIS reactor and the design of the two calorimeter devices, the nuclear heating calculation scheme used for simulation will be introduced. Different simulations and results will be detailed and analyzed to determine the calorimeter geometry impact on the measured nuclear heating. (authors)

  11. AGN Heating in Simulated Cool-core Clusters

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yuan; Ruszkowski, Mateusz [Department of Astronomy, University of Michigan, 1085 S. University Avenue, Ann Arbor, MI 48109 (United States); Bryan, Greg L., E-mail: yuanlium@umich.edu [Department of Astronomy, Columbia University, Pupin Physics Laboratories, New York, NY 10027 (United States)

    2017-10-01

    We analyze heating and cooling processes in an idealized simulation of a cool-core cluster, where momentum-driven AGN feedback balances radiative cooling in a time-averaged sense. We find that, on average, energy dissipation via shock waves is almost an order of magnitude higher than via turbulence. Most of the shock waves in the simulation are very weak shocks with Mach numbers smaller than 1.5, but the stronger shocks, although rare, dissipate energy more effectively. We find that shock dissipation is a steep function of radius, with most of the energy dissipated within 30 kpc, more spatially concentrated than radiative cooling loss. However, adiabatic processes and mixing (of post-shock materials and the surrounding gas) are able to redistribute the heat throughout the core. A considerable fraction of the AGN energy also escapes the core region. The cluster goes through cycles of AGN outbursts accompanied by periods of enhanced precipitation and star formation, over gigayear timescales. The cluster core is under-heated at the end of each cycle, but over-heated at the peak of the AGN outburst. During the heating-dominant phase, turbulent dissipation alone is often able to balance radiative cooling at every radius but, when this is occurs, shock waves inevitably dissipate even more energy. Our simulation explains why some clusters, such as Abell 2029, are cooling dominated, while in some other clusters, such as Perseus, various heating mechanisms including shock heating, turbulent dissipation and bubble mixing can all individually balance cooling, and together, over-heat the core.

  12. Nuclear Thermal Rocket Simulation in NPSS

    Science.gov (United States)

    Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

    2013-01-01

    Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

  13. Simulation of the aspersion system of the core low pressure (LPCS) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Membrillo G, O. E.; Chavez M, C.

    2012-10-01

    The present work presents the modeling and simulation of the aspersion system to low pressure of reactor of the nuclear power plant of Laguna Verde using the nuclear code RELAP/SCDAP. The objective of the emergency systems inside a nuclear reactor is the cooling of the core, nor caring the performance of any other emergency system in the case of an accident design base for coolant loss. To obtain a simulation of the system is necessary to have a model based on their main components, pipes, pumps, valves, etc. This article describes the model for the simulation of the main line and the test line for the HPCS. At the moment we have the simulation of the reactor vessel and their systems associated to the nuclear power plant of Laguna Verde, this work will allow to associate the emergency system model LPCS to the vessel model. The simulation of the vessel and the emergency systems will allow knowing the behavior of the reactor in the stage of the coolant loos, giving the possibility to analyze diverse scenarios. The general model will provide an auxiliary tool for the training in classroom and at distance in the operation of nuclear power plants. (Author)

  14. 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.

  15. 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

  16. Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW

    International Nuclear Information System (INIS)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim; Ashoub, Nagieb

    2015-01-01

    This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

  17. Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Center

    2015-11-15

    This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

  18. 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

  19. 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

  20. Simulation and control of the site-dependent neutron density in a nuclear reactor

    International Nuclear Information System (INIS)

    Stark, K.

    1974-01-01

    The present work deals with the simulation and control of a pressurized-water reactor such as is used in nuclear power plants today. In the first part of the work, the mathematical model equations of the reactor are set up. They take into consideration the local distribution of the various reactor parameters as far as seems necessary for further investigations. Taking the given approximations, the mathematical model is locally one-dimensional; it is valid for the period of time in which a power control of the reactor must work. The model equations set up are calculated on an analog/hybride computer according to the modal simulation method in true time. The method is distinguished in the present problem here through good convergence and enables the observation of the simulation results as a stationary picture on an oscillograph screen. For this reason, a simulation of this type seems particularly suitable for the training of operational personnel. The aim of the second part of the work is the development of a simple control concept which enables the control of the total power of the reactor as well as of the distribution of the power density in the reactor core. The fundamentals of the control design are the non-linear system equations of the nuclear reactor. The developed control is based on the controlling of eigenfunctions; it controls the total power of the reactor as well as the distribution of the power density in the reactor core where a uniform burn-up of the nuclear fuel is seen to. Part-absorbing control rods amongst others are used as actuators like they are already used in that type of reactors. (orig./LH) [de

  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. PC-Reactor-core transient simulation code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    PC-REATOR, a reactor core transient simulation code has been developed for the real-time operator training on a IBM-PC microcomputer. The program presents capabilities for on-line exchange of the operating parameters during the transient simulation, by friendly keyboard instructions. The model is based on the point-kinetics approximation, with 2 delayed neutron percursors and up to 11 decay power generating groups. (author) [pt

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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)

  8. 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)

  9. 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)

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

    Science.gov (United States)

    Huddar, Lakshana Ravindranath

    With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by

  11. 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

  12. 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

  13. Influence of Concrete Properties on Molten Core-Concrete Interaction: A Simulation Study

    Directory of Open Access Journals (Sweden)

    Jin-yang Jiang

    2016-01-01

    Full Text Available In a severe nuclear power plant accident, the molten core can be released into the reactor pit and interact with sacrificial concrete. In this paper, a simulation study is presented that aims to address the influence of sacrificial concrete properties on molten core-concrete interaction (MCCI. In particular, based on the MELCOR Code, the ferrosiliceous concrete used in European Pressurized Water Reactor (EPR is taken into account with respect to the different ablation enthalpy and Fe2O3 and H2O contents. Results indicate that the concrete ablation rate as well as the hydrogen generation rate depends much on the concrete ablation enthalpy and Fe2O3 and H2O contents. In practice, the ablation enthalpy of sacrificial concrete is the higher the better, while the Fe2O3 and H2O content of sacrificial concrete is the lower the better.

  14. 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)

  15. Monte Carlo simulation in nuclear medicine

    International Nuclear Information System (INIS)

    Morel, Ch.

    2007-01-01

    The Monte Carlo method allows for simulating random processes by using series of pseudo-random numbers. It became an important tool in nuclear medicine to assist in the design of new medical imaging devices, optimise their use and analyse their data. Presently, the sophistication of the simulation tools allows the introduction of Monte Carlo predictions in data correction and image reconstruction processes. The availability to simulate time dependent processes opens up new horizons for Monte Carlo simulation in nuclear medicine. In a near future, these developments will allow to tackle simultaneously imaging and dosimetry issues and soon, case system Monte Carlo simulations may become part of the nuclear medicine diagnostic process. This paper describes some Monte Carlo method basics and the sampling methods that were developed for it. It gives a referenced list of different simulation software used in nuclear medicine and enumerates some of their present and prospective applications. (author)

  16. 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

  17. 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.

  18. 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.

  19. 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)

  20. Power-Energy Simulation for Multi-Core Processors in Bench-marking

    Directory of Open Access Journals (Sweden)

    Mona A. Abou-Of

    2017-01-01

    Full Text Available At Microarchitectural level, multi-core processor, as a complex System on Chip, has sophisticated on-chip components including cores, shared caches, interconnects and system controllers such as memory and ethernet controllers. At technological level, architects should consider the device types forecast in the International Technology Roadmap for Semiconductors (ITRS. Energy simulation enables architects to study two important metrics simultaneously. Timing is a key element of the CPU performance that imposes constraints on the CPU target clock frequency. Power and the resulting heat impose more severe design constraints, such as core clustering, while semiconductor industry is providing more transistors in the die area in pace with Moore’s law. Energy simulators provide a solution for such serious challenge. Energy is modelled either by combining performance benchmarking tool with a power simulator or by an integrated framework of both performance simulator and power profiling system. This article presents and asses trade-offs between different architectures using four cores battery-powered mobile systems by running a custom-made and a standard benchmark tools. The experimental results assure the Energy/ Frequency convexity rule over a range of frequency settings on different number of enabled cores. The reported results show that increasing the number of cores has a great effect on increasing the power consumption. However, a minimum energy dissipation will occur at a lower frequency which reduces the power consumption. Despite that, increasing the number of cores will also increase the effective cores value which will reflect a better processor performance.

  1. 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

  2. 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

  3. 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. .

  4. 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)

  5. Development of neutron own codes for the simulation of PWR reactor core; Desarrollo de codigos neutronicos propios para la simulacion del nucleo de reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.; Cabellos, O.; Garcia-Herranz, N.; Cuervo, D.; Herrero, J. J.; Jimenez, J.; Ochoa, R.

    2011-07-01

    The core physic simulation is enough complex to need computers and ad-hoc software, and its evolution is to best-estimate methodologies, in order to improve availability and safety margins in the power plant operation. the Nuclear Engineering Department (UPM) has developed the SEANAP System in use in several power plants in Spain, with simulation in 3D and at the pin level detail, of the nominal and actual core burnup, with the on-line surveillance, and operational maneuvers optimization. (Author) 8 refs.

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. Development of a research nuclear reactor simulator using LABVIEW®

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)

  11. Development of a research nuclear reactor simulator using LABVIEW®

    International Nuclear Information System (INIS)

    Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade

    2015-01-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)

  12. Axial power deviation control strategy and computer simulation for Daya Bay Nuclear Power Station

    International Nuclear Information System (INIS)

    Liao Yehong; Zhou Xiaoling, Xiao Min

    2004-01-01

    Daya Bay Nuclear Power Station has very tight operation diagram especially at its right side. Therefore the successful control of axial power deviation for PWR is crucial to nuclear safety. After analyzing various core characters' effect on axial power distribution, several axial power deviation control strategies has been proposed to comply with different power varying operation scenario. Application and computer simulation of the strategies has shown that our prediction of axial power deviation evolution are comparable to the measurement values, and that our control strategies are effective. Engineering experience shows that the application of our methodology can predict accurately the transient of axial power deviation, and therefore has become a useful tool for reactor operation and safety control. This paper presents the axial power control characteristics, reactor operation strategy research, computer simulation, and comparison to measurement results in Daya Bay Nuclear Power Station. (author)

  13. 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

  14. 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

  15. 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%

  16. 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.)

  17. Full Core Multiphysics Simulation with Offline Mesh Deformation

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Obabko, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Jain, Rajeev [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay [Argonne National Lab. (ANL), Argonne, IL (United States); Solberg, Jerome [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferencz, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-21

    In this report, building on previous reports issued in FY13 we describe our continued efforts to integrate thermal/hydraulics, neutronics, and structural mechanics modeling codes to perform coupled analysis of a representative fast sodium-cooled reactor core. The focus of the present report is a full core simulation with off-line mesh deformation.

  18. 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

  19. Simulation language of DSNP: dynamic simulator for nuclear power-plants

    International Nuclear Information System (INIS)

    Saphier, D.

    1978-09-01

    The Dynamic Simulator for Nuclear Power-plants (DSNP) is a system of programs and data sets by which a nuclear power plant or part thereof can be simulated at different levels of sophistication. The acronym DSNP is used interchangeably for the DSNP language, for the DSNP precompiler, for the DSNP libraries, and for the DSNP document generator. The DSNP language is a set of simple block oriented statements, which together with the appropriate data, comprise a simulation of a nuclear power plant. The majority of the DSNP statements will result in the inclusion of a simulated physical module into the program. FORTRAN statements can be inserted with no restrictions among DSNP statements

  20. Modelling of reactor control and protection systems in the core simulator program GARLIC

    International Nuclear Information System (INIS)

    Beraha, D.; Lupas, O.; Ploegert, K.

    1984-01-01

    For analysis of the interaction between control and limitation systems and the power distribution in the reactor core, a valuable tool is provided by the joint simulation of the core and the interacting systems. To this purpose, the core simulator GARLIC has been enhanced by models of the systems for controlling and limiting the reactor power and the power distribution in the core as well as by modules for calculating safety related core parameters. The computer-based core protection system, first installed in the Grafenrheinfeld NPP, has been included in the simulation. In order to evaluate the accuracy of GARLIC-simulations, the code has been compared with a design code in the train of a verification phase. The report describes the program extensions and the results of the verification. (orig.) [de

  1. 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

  2. Concept for Multi-cycle Nuclear Fuel Optimization Based On Parallel Simulated Annealing With Mixing of States

    International Nuclear Information System (INIS)

    Kropaczek, David J.

    2008-01-01

    A new concept for performing nuclear fuel optimization over a multi-cycle planning horizon is presented. The method provides for an implicit coupling between traditionally separate in-core and out-of-core fuel management decisions including determination of: fresh fuel batch size, enrichment and bundle design; exposed fuel reuse; and core loading pattern. The algorithm uses simulated annealing optimization, modified with a technique called mixing of states that allows for deployment in a scalable parallel environment. Analysis of algorithm performance for a transition cycle design (i.e. a PWR 6 month cycle length extension) demonstrates the feasibility of the approach as a production tool for fuel procurement and multi-cycle core design. (authors)

  3. GTOROTO: a simulation system for HTGR core seismic behavior

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Nakamura, Yasuhiro; Onuma, Yoshio

    1980-07-01

    One of the most important design of HTGR core is its aseismic structure. Therefore, it is necessary to predict the forces and motion of the core blocks. To meet the requirement, many efforts to develop analytical methods and computer programs are made. A graphic simulation system GTOROTO with a CRT graphic display and lightpen was developed to analyze the HTGR core behavior in seismic excitation. Feature of the GTOROTO are as follows: (1) Behavior of the block-type HTGR core during earthquake can be shown on the CRT-display. (2) Parameters of the computing scheme can be changed with the lightpen. (3) Routines of the computing scheme can be changed with the lightpen and an alteration switch. (4) Simulation pictures are shown automatically. Hardcopies are available by plotter in stopping the progress of simulation pictures. Graphic representation can be re-start with the predetermined program. (5) Graphic representation informations can be stored in assembly language on a disk for rapid representation. (6) A computer-generated cinema can be made by COM (Computer Output Microfilming) or filming directly the CRT pictures. These features in the GTOROTO are provided in on-line conversational mode. (author)

  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

    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)

  5. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    Baptista, Vinicius Damas

    1996-01-01

    The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

  6. 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

  7. 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.

  8. Simulation and Representation of Body, Emotion, and Core Consciousness

    NARCIS (Netherlands)

    Bosse, T.; Jonker, C.M.; Treur, J.; Henderson-Sellers, B.; Winikoff, M.

    2005-01-01

    This paper contributes an analysis and formalisation of Damasio's theory on core consciousness. Three important concepts in this theory are 'emotion', 'feeling', and 'feeling a feeling' (or core consciousness). In particular, a simulation model is described of the neural dynamics leading via emotion

  9. PUMA code simulation of recovery power transients after a short shutdown of the Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Villar, Javier; Pomerantz, Marcelo E.

    2003-01-01

    A simulation of recovery power transients after a short shutdown on Embalse nuclear power plant equilibrium core with slightly enriched uranium fuel was performed in order to know the response of the reactor under such conditions. Also, comparison against the same event in a natural uranium core were done. No significant restrictions were found in operating with enriched fuel in the conditions of the analyzed event and in fact, slightly differences arose with natural uranium fuels. (author)

  10. The compact simulator for Tihange nuclear plant

    International Nuclear Information System (INIS)

    Gueben, M.

    1982-01-01

    After an introduction about the simulators for nuclear plants, a description is given of the compact simulator for the Tihange nuclear power plant as well as the simulated circuits and equipments such as the primary and secondary coolant circuits. The extent of simulation, the functions used by the instructor, the use of the simulator, the formation programme and construction planning are described. (AF)

  11. Monte Carlo simulation of magnetic multi-core nanoparticles

    International Nuclear Information System (INIS)

    Schaller, Vincent; Wahnstroem, Goeran; Sanz-Velasco, Anke; Enoksson, Peter; Johansson, Christer

    2009-01-01

    In this paper, a Monte Carlo simulation is carried out to evaluate the equilibrium magnetization of magnetic multi-core nanoparticles in a liquid and subjected to a static magnetic field. The particles contain a magnetic multi-core consisting of a cluster of magnetic single-domains of magnetite. We show that the magnetization of multi-core nanoparticles cannot be fully described by a Langevin model. Inter-domain dipolar interactions and domain magnetic anisotropy contribute to decrease the magnetization of the particles, whereas the single-domain size distribution yields an increase in magnetization. Also, we show that the interactions affect the effective magnetic moment of the multi-core nanoparticles.

  12. Atucha II nuclear power plant digital simulation

    International Nuclear Information System (INIS)

    Santome, D.; Rovere, L.A.T.

    1987-01-01

    This paper describes the start-up of a digital simulation code apt to be performed in real time of Atucha II nuclear power plant, foreseeing its subsequent usage in a Basic Principles Simulator. Adaptability and modification of existing routines and development of modules in order to incorporate the necessary variables dynamics to couple the different modes, were the main tasks. The mathematical model used allows the representation of the following sub-systems: a) a reactor's core point model, which comprehends the neutronic kinetics, fission and decaying powers, thermal transfer and Xe-poisoning calculation; b) pressurizer, which considers two sub-systems that may or may not be in thermodynamic equilibrium, both in two phases; c) coolants and moderators bonds considering separate moderator loops with the aim of introducing asymmetric perturbations; d) secondary sub-subsystem, which includes the feed water loop, pumps, steam generators and control valves; e) steam generators; f) control and safety systems, including power control, steam generators levels, moderator's temperature primary loop system, limitations and protection. (Author)

  13. Nuclear system test simulator

    International Nuclear Information System (INIS)

    Sawyer, S.D.; Hill, W.D.; Wilson, P.A.; Steiner, W.M.

    1987-01-01

    A transportable test simulator is described for a nuclear power plant. The nuclear power plant includes a control panel, a reactor having actuated rods for moving into and out of a reactor for causing the plant to operate, and a control rod network extending between the control panel and the reactor rods. The network serially transmits command words between the panel and rods, and has connecting interfaces at preselected points remote from the control panel between the control panel and rods. The test simulator comprises: a test simulator input for transport to and connection into the network at at least one interface for receiving the serial command words from the network. Each serial command includes an identifier portion and a command portion; means for processing interior of the simulator for the serial command words for identifying that portion of the power plant designated in the identifier portion and processing the word responsive to the command portion of the word after the identification; means for generating a response word responsive to the command portion; and output means for sending and transmitting the response word to the nuclear power plant at the interface whereby the control panel responds to the response word

  14. 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).

  15. Simulators for nuclear power plants

    International Nuclear Information System (INIS)

    Ancarani, A.; Zanobetti, D.

    1983-01-01

    The different types of simulator for nuclear power plants depend on the kind of programme and the degree of representation to be achieved, which in turn determines the functions to duplicate. Different degrees correspond to different simulators and hence to different choices in the functions. Training of nuclear power plant operators takes advantage of the contribution of simulators of various degrees of complexity and fidelity. Reduced scope simulators are best for understanding basic phenomena; replica simulators are best used for formal qualification and requalification of personnel, while modular mini simulators of single parts of a plant are best for replay and assessment of malfunctions. Another category consists of simulators for the development of assistance during operation, with the inclusion of disturbance and alarm analysis. The only existing standard on simulators is, at present, the one adopted in the United States. This is too stringent and is never complied with by present simulators. A description of possible advantages of a European standard is therefore offered: it rests on methods of measurement of basic simulator characteristics such as fidelity in values and time. (author)

  16. Combined core/boundary layer plasma transport simulations in tokamaks

    International Nuclear Information System (INIS)

    Prinja, A.K.; Schafer, R.F. Jr.; Conn, R.W.; Howe, H.C.

    1987-01-01

    Significant new numerical results are presented from self-consistent core and boundary or scrape-off layer plasma simulations with 3-D neutral transport calculations. For a symmetric belt limiter it is shown that, for plasma conditions considered here, the pump limiter collection efficiency increases from 11% to 18% of the core efflux as a result of local reionization of blade deflected neutrals. This hitherto unobserved effect causes a significant amplification of upstream ion flux entering the pump limiter. Results from coupling of an earlier developed two-zone edge plasma model ODESSA to the PROCTR core plasma simulation code indicates that intense recycling divertor operation may not be possible because of stagnation of upstream flow velocity. This results in a self-consistent reduction of density gradient in an intermediate region between the central plasma and separatrix, and a concomitant reduction of core-efflux. There is also evidence of increased recycling at the first wall. (orig.)

  17. 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

  18. Primary loop simulation of the SP-100 space nuclear reactor

    International Nuclear Information System (INIS)

    Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F.

    2011-01-01

    Between 1983 and 1992 the SP-100 space nuclear reactor development project for electric power generation in a range of 100 to 1000 kWh was conducted in the USA. Several configurations were studied to satisfy different mission objectives and power systems. In this reactor the heat is generated in a compact core and refrigerated by liquid lithium, the primary loops flow are controlled by thermoelectric electromagnetic pumps (EMTE), and thermoelectric converters produce direct current energy. To define the system operation point for an operating nominal power, it is necessary the simulation of the thermal-hydraulic components of the space nuclear reactor. In this paper the BEMTE-3 computer code is used to EMTE pump design performance evaluation to a thermalhydraulic primary loop configuration, and comparison of the system operation points of SP-100 reactor to two thermal powers, with satisfactory results. (author)

  19. 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.

  20. Development of new Micro-Physics Nuclear Reactor Simulator™ and its possibility for introductory education of nuclear engineering

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Tsujita, Kosuke; Tamari, Yohei

    2015-01-01

    This paper describes recent activity on development of the Micro-Physics Nuclear Reactor Simulator™ and its application to introductory educations of nuclear engineering at high schools and university. The simulator has been continuously improved with active feedbacks from existing and potential users through its applications to exercises in classes/seminars. A newly developed reactor core transient analysis code, RAMBO-T has been adopted in the simulator along with SIMULATE-3K by Studsvik Scandpower Inc. (Borkowski, 1994) The internal data structure has been revised so that any combinations of the target reactor type, the core transient analysis code and the display language can be established. A new graphical user interface was implemented to realize the intuitive and easy-to-understand operations by novice users. The improved version of the Micro-Physics Nuclear Reactor Simulator has been practically used at educational institutions. In order to contribute to the activities on human resource development in the field of nuclear engineering, it is planned to donate the Micro-Physics Simulator™ Lite, a variation of the simulator that supports the only transient core analysis with RAMBO-T, to IAEA, the International Atomic Energy Agency. It will be included into the “NPP Simulators suite for Education” where complimentary copies are distributed to the member states countries. (author)

  1. 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

  2. Ultrafast vortex core dynamics investigated by finite-element micromagnetic simulations

    Energy Technology Data Exchange (ETDEWEB)

    Gliga, Sebastian

    2010-07-01

    The investigations carried out in this thesis concern the ultrafast dynamics of a fundamental micromagnetic configuration: the vortex. Over the past decade, a detailed understanding of the dynamic and static properties of such magnetic nanostructures has been achieved as a result of close interplay between experiments, theory and numeric simulations. Here, micromagnetic simulations were performed based on the finite-element method. The vortex structure arises in laterally-confined ferromagnets, in particular in thin-film elements, and is characterized by an in-plane curling of the magnetic moments around a very stable and narrow core. In the present study, a novel process in micromagnetism was found: the ultrafast reversal of the vortex core. The possibility of easily switching the core orientation by means of short in-plane field pulses is surprising in view of the very high stability of the core. Moreover, the simulations presented here showed that this reversal process unfolds on a time scale of only a few tens of picoseconds, which leads to the prediction of the fastest and most complex micromagnetic reversal process known to date. Indeed, the vortex core is not merely switched: it is destroyed and recreated in the immediate vicinity with an opposite direction. This is mediated by a rapid sequence of vortex-antivortex pair creation and annihilation subprocesses and results in a sudden burst-like emission of spin waves. Equally fascinating is the ultrafast dynamics of an isolated magnetic antivortex, the topological counterpart of the vortex. The simulations performed here showed that the static complementarity between vortices and antivortices is equally reflected in their ultrafast dynamics, which leads to the reversal of the antivortex core. A promising means for the control of the magnetization on the nanoscale consists in exploiting the spin-transfer torque effect. The study of the current-induced dynamics of vortices showed that the core reversal can be

  3. Multilevel criticality computations in AREVA NP'S core simulation code artemis - 195

    International Nuclear Information System (INIS)

    Van Geemert, R.

    2010-01-01

    This paper discusses the multi-level critical boron iteration approach that is applied per default in AREVA NP's whole-core neutronics and thermal hydraulics core simulation program ARTEMIS. This multi-level approach is characterized by the projection of variational boron concentration adjustments to the coarser mesh levels in a multi-level re-balancing hierarchy that is associated with the nodal flux equations to be solved in steady-state core simulation. At each individual re-balancing mesh level, optimized variational criticality tuning formulas are applied. The latter drive the core model to a numerically highly accurate self-sustaining state (i.e. with the neutronic eigenvalue being 1 up to a very high numerical precision) by continuous adjustment of the boron concentration as a system-wide scalar criticality parameter. Due to the default application of this approach in ARTEMIS reactor cycle simulations, an accuracy of all critical boron concentration estimates better than 0.001 ppm is enabled for all burnup time steps in a computationally efficient way. This high accuracy is relevant for precision optimization in industrial core simulation as well as for enabling accurate reactivity perturbation assessments. The developed approach is presented from a numerical methodology point of view with an emphasis on the multi-grid aspect of the concept. Furthermore, an application-relevant verification is presented in terms of achieved coupled iteration convergence efficiency for an application-representative industrial core cycle computation. (authors)

  4. 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)

  5. CLASS: Core Library for Advanced Scenario Simulations

    International Nuclear Information System (INIS)

    Mouginot, B.; Thiolliere, N.

    2015-01-01

    The nuclear reactor simulation community has to perform complex electronuclear scenario simulations. To avoid constraints coming from the existing powerful scenario software such as COSI, VISION or FAMILY, the open source Core Library for Advanced Scenario Simulation (CLASS) has been developed. The main asset of CLASS is its ability to include any type of reactor, whether the system is innovative or standard. A reactor is fully described by its evolution database which should contain a set of different validated fuel compositions in order to simulate transitional scenarios. CLASS aims to be a useful tool to study scenarios involving Generation-IV reactors as well as innovative fuel cycles, like the thorium cycle. In addition to all standard key objects required by an electronuclear scenario simulation (the isotopic vector, the reactor, the fuel storage and the fabrication units), CLASS also integrates two new specific modules: fresh fuel evolution and recycled fuel fabrication. The first module, dealing with fresh fuel evolution, is implemented in CLASS by solving Bateman equations built from a database induced cross-sections. The second module, which incorporates the fabrication of recycled fuel to CLASS, can be defined by user priorities and/or algorithms. By default, it uses a linear Pu equivalent-method, which allows predicting, from the isotopic composition, the maximum burn-up accessible for a set type of fuel. This paper presents the basis of the CLASS scenario, the fuel method applied to a MOX fuel and an evolution module benchmark based on the French electronuclear fleet from 1977 to 2012. Results of the CLASS calculation were compared with the inventory made and published by the ANDRA organisation in 2012. For UOX used fuels, the ANDRA reported 12006 tonnes of heavy metal in stock, including cooling, versus 18500 tonnes of heavy metal predicted by CLASS. The large difference is easily explained by the presence of 56 tonnes of plutonium already separated

  6. IEA-R1 reactor core simulation with RELAP5 code

    International Nuclear Information System (INIS)

    Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

    2005-01-01

    This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

  7. 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.

  8. Applications of simulation experiments in LMFBR core materials technology

    International Nuclear Information System (INIS)

    Appleby, W.K.

    1976-01-01

    The development of charged particle bombardment experiments to simulate neutron irradiation induced swelling in austenitic alloys is briefly described. The applications of these techniques in LMFBR core materials technology are discussed. It is shown that use of the techniques to study the behavior of cold-worked Type-316 was instrumental in demonstrating at an early date the need for advanced materials. The simulation techniques then were used to identify alloying elements which can markedly decrease swelling and thus a focused reactor irradiation program is now in place to allow the future use of a lower swelling alloy for LMFBR core components

  9. Simulation in CFD of a Pebble Bed: Advanced high temperature reactor core using OpenFOAM

    International Nuclear Information System (INIS)

    Dahl, Pamela M.; Su, Jian

    2017-01-01

    Numerical simulations of a Pebble Bed nuclear reactor core are presented using the multi-physics tool-kit OpenFOAM. The HTR-PM is modeled using the porous media approach, accounting both for viscous and inertial effects through the Darcy and Forchheimer model. Initially, cylindrical 2D and 3D simulations are compared, in order to evaluate their differences and decide if the 2D simulations carry enough of the sought information, considering the savings in computational costs. The porous medium is considered to be isotropic, with the whole length of the packed bed occupied homogeneously with the spherical fuel elements. Steady-state simulations for normal equilibrium operation are performed, using a semi sine function of the power density along the vertical axis as the source term for the energy balance equation.Total pressure drop is calculated and compared with that obtained from literature for a similar case. At a second stage, transient simulations are performed, where relevant parameters are calculated and compared to those of the literature. (author)

  10. Simulation in CFD of a Pebble Bed: Advanced high temperature reactor core using OpenFOAM

    Energy Technology Data Exchange (ETDEWEB)

    Dahl, Pamela M.; Su, Jian, E-mail: sujian@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    Numerical simulations of a Pebble Bed nuclear reactor core are presented using the multi-physics tool-kit OpenFOAM. The HTR-PM is modeled using the porous media approach, accounting both for viscous and inertial effects through the Darcy and Forchheimer model. Initially, cylindrical 2D and 3D simulations are compared, in order to evaluate their differences and decide if the 2D simulations carry enough of the sought information, considering the savings in computational costs. The porous medium is considered to be isotropic, with the whole length of the packed bed occupied homogeneously with the spherical fuel elements. Steady-state simulations for normal equilibrium operation are performed, using a semi sine function of the power density along the vertical axis as the source term for the energy balance equation.Total pressure drop is calculated and compared with that obtained from literature for a similar case. At a second stage, transient simulations are performed, where relevant parameters are calculated and compared to those of the literature. (author)

  11. Nuclear fuel cycle simulation system (VISTA)

    International Nuclear Information System (INIS)

    2007-02-01

    The Nuclear Fuel Cycle Simulation System (VISTA) is a simulation system which estimates long term nuclear fuel cycle material and service requirements as well as the material arising from the operation of nuclear fuel cycle facilities and nuclear power reactors. The VISTA model needs isotopic composition of spent nuclear fuel in order to make estimations of the material arisings from the nuclear reactor operation. For this purpose, in accordance with the requirements of the VISTA code, a new module called Calculating Actinide Inventory (CAIN) was developed. CAIN is a simple fuel depletion model which requires a small number of input parameters and gives results in a very short time. VISTA has been used internally by the IAEA for the estimation of: spent fuel discharge from the reactors worldwide, Pu accumulation in the discharged spent fuel, minor actinides (MA) accumulation in the spent fuel, and in the high level waste (HLW) since its development. The IAEA decided to disseminate the VISTA tool to Member States using internet capabilities in 2003. The improvement and expansion of the simulation code and the development of the internet version was started in 2004. A website was developed to introduce the simulation system to the visitors providing a simple nuclear material flow calculation tool. This website has been made available to Member States in 2005. The development work for the full internet version is expected to be fully available to the interested parties from IAEA Member States in 2007 on its website. This publication is the accompanying text which gives details of the modelling and an example scenario

  12. Laser-heating and Radiance Spectrometry for the Study of Nuclear Materials in Conditions Simulating a Nuclear Power Plant Accident.

    Science.gov (United States)

    Manara, Dario; Soldi, Luca; Mastromarino, Sara; Boboridis, Kostantinos; Robba, Davide; Vlahovic, Luka; Konings, Rudy

    2017-12-14

    Major and severe accidents have occurred three times in nuclear power plants (NPPs), at Three Mile Island (USA, 1979), Chernobyl (former USSR, 1986) and Fukushima (Japan, 2011). Research on the causes, dynamics, and consequences of these mishaps has been performed in a few laboratories worldwide in the last three decades. Common goals of such research activities are: the prevention of these kinds of accidents, both in existing and potential new nuclear power plants; the minimization of their eventual consequences; and ultimately, a full understanding of the real risks connected with NPPs. At the European Commission Joint Research Centre's Institute for Transuranium Elements, a laser-heating and fast radiance spectro-pyrometry facility is used for the laboratory simulation, on a small scale, of NPP core meltdown, the most common type of severe accident (SA) that can occur in a nuclear reactor as a consequence of a failure of the cooling system. This simulation tool permits fast and effective high-temperature measurements on real nuclear materials, such as plutonium and minor actinide-containing fission fuel samples. In this respect, and in its capability to produce large amount of data concerning materials under extreme conditions, the current experimental approach is certainly unique. For current and future concepts of NPP, example results are presented on the melting behavior of some different types of nuclear fuels: uranium-plutonium oxides, carbides, and nitrides. Results on the high-temperature interaction of oxide fuels with containment materials are also briefly shown.

  13. 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.)

  14. Simulation of hemp fibre bundle and cores using discrete element method

    Energy Technology Data Exchange (ETDEWEB)

    Al-Amin Sadek, M.; Chen, Y. [Manitoba Univ., Winnipeg, MB (Canada). Dept. of Biosystems Engineering; Lague, C. [Ottawa Univ., Ottawa, ON (Canada). Faculty of Engineering; Landry, H. [Prairie Agricultural Machinery Inst., Humboldt, SK (Canada); Peng, Q. [Manitoba Univ., Winnipeg, MB (Canada). Dept. of Mechanical and Manufacturing Engineering; Zhong, W. [Manitoba Univ., Winnipeg, MB (Canada). Dept. of Textile Sciences

    2010-07-01

    The mechanical behaviour of hemp fibre and core must be well understood in order to obtain high-grade hemp fibre that is currently in high demand for various industrial applications. Modelling by discrete element method can simulate the mechanical behaviour of such materials. A commercial discrete element software called Particle Flow Code was used in this study. In particular, the 3-dimension (PFC3D) was used to simulate hemp fibre and core. Since the basic PFC3D particles are spherical, the individual virtual hemp fibres were defined as strings of balls held together by PFC3D parallel bonds. The study showed that the virtual fibre is flexible and can bend and break by forces. This reflects the characteristics of hemp fibre. Using the clump logic of PFC3D, the virtual hemp core was defined as a rigid and unbreakable body, which reflect the characteristics of the core. The virtual fibre and core were defined with several microproperties, some of which were previously calibrated. The PFC3D bond properties were calibrated in this study. They included normal and shear stiffness; pb{sub k}n and pb{sub k}s; normal and shear strength; and bond disk radius, R of the virtual fibre. The calibration started with developing a PFC3D model to simulate fibre tensile test. The microproperties of virtual fibre and core were calibrated by running the PFC3D model. Literature data from fibre tensile tests was compared with simulation results.

  15. Hydrodynamical simulation of the core helium flash with two-dimensional convection

    International Nuclear Information System (INIS)

    Cole, P.W.

    1981-01-01

    The thermonuclear runaway of helium reactions under the condition of electron degeneracy in the hot, dense central regions of a low mass Population II red giant is investigated. A two-dimensional finite difference approach to time dependent convection has been applied to a peak energy production model of this phenomenon called the core helium flash. The dynamical conservation equations are integrated in two spatial dimensions and time which allow the horizontal variations of the dynamical variables to be followed explicitly. The unbalanced bouyancy forces in convectively unstable regions lead to mass flow (i.e., convective energy transport) by calculation of the velocity flow patterns produced by the conservation laws of mass, momentum, and energy without recourse to any phenomenological theory of convection. The initial phase of this hydrodynamical simulation is characterized by a thermal readjustment via downward convective energy transport into the neutrino cooled core in a series of convection modulated thermal pulses. Each of these pulses is driven by the thermal runaway and quenched by the convective energy transport when the actual temperature gradient in the flash region becomes sufficiently superadiabatic. These convection modulated thermal pulses are observed throughout 95% of the calculation, the duration of which is approximately 570,000 cycles or nearly 96,000 seconds of evolution. After this initial thermal restructuring, there ensues in the simulation a dynamic phase in which the thermonuclear runaway becomes violent. The degree of violence, the final composition, and the peak temperature depend sensitively on the nuclear energy generation rates of those reactions involving alpha particle captures

  16. Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

    International Nuclear Information System (INIS)

    Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

    2009-01-01

    The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

  17. Flow with boiling in four-cusp channels simulating damaged core in PWR type reactors

    International Nuclear Information System (INIS)

    Esteves, M.M.

    1985-01-01

    The study of subcooled nucleate flow boiling in non-circular channels is of great importance to engineering applications in particular to Nuclear Engineering. In the present work, an experimental apparatus, consisting basically of a refrigeration system, running on refrigerant-12, has been developed. Preliminary tests were made with a circular tube. The main objective has been to analyse subcooled flow boiling in four-cusp channels simulating the flow conditions in a PWR core degraded by accident. Correlations were developed for the forced convection film coefficient for both single-phase and subcooled flow boiling. The incipience of boiling in such geometry has also been studied. (author) [pt

  18. SIMODIS - a software package for simulating nuclear reactor components

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Borges, Eduardo M.

    2000-01-01

    In this paper it is presented the initial development effort in building a nuclear reactor component simulation package. This package was developed to be used in the MATLAB simulation environment. It uses the graphical capabilities from MATLAB and the advantages of compiled languages, as for instance FORTRAN and C ++ . From the MATLAB it takes the facilities for better displaying the calculated results. From the compiled languages it takes processing speed. So far models from reactor core, UTSG and OTSG have been developed. Also, a series a user-friendly graphical interfaces have been developed for the above models. As a by product a set of water and sodium thermal and physical properties have been developed and may be used directly as a function from MATLAB, or by being called from a model, as part of its calculation process. The whole set was named SIMODIS, which stands for SIstema MODular Integrado de Simulacao. (author)

  19. Operator training simulator for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Watanabe, Tadasu

    1988-01-01

    For the operation management of nuclear power stations with high reliability and safety, the role played by operators is very important. The effort of improving the man-machine interface in the central control rooms of nuclear power stations is energetically advanced, but the importance of the role of operators does not change. For the training of the operators of nuclear power stations, simulators have been used from the early stage. As the simulator facilities for operator training, there are the full scope simulator simulating faithfully the central control room of an actual plant and the small simulator mainly aiming at learning the plant functions. For BWR nuclear power stations, two full scope simulators are installed in the BWR Operator Training Center, and the training has been carried out since 1974. The plant function learning simulators have been installed in respective electric power companies as the education and training facilities in the companies. The role of simulators in operator training, the BTC No.1 simulator of a BWR-4 of 780 MWe and the BTC No.2 simulator of a BWR-5 of 1,100 MWe, plant function learning simulators, and the design of the BTC No.2 simulator and plant function learning simulators are reported. (K.I.)

  20. Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide. Final Report

    International Nuclear Information System (INIS)

    Pelaccio, D.G.; Scheil, C.M.; Petrosky, L.J.

    1993-03-01

    A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner

  1. 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)

  2. 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

  3. Simulators of nuclear power stations

    International Nuclear Information System (INIS)

    Zanobetti, D.

    1984-01-01

    The report deals with the simulators of nuclear power stations used for the training of operators and for the analysis of operations. It reviews the development of analogical, hybrid and digital simulators up to the present, indicating the impact resulting from the TMI-2 accident. It indicates, the components of simulators and the present accepted terminology for a classification of the various types of simulators. It reviews the present state of the art of the technology: how a basic mathematical model of a nuclear power system is worked out and what are the technical problems associated with more accurate models. Examples of elaborate models are given: for a PWR pressurizer, for an AGR steam generator. It also discusses certain problems of hardware technology. Characteristics of present replica simulators are given with certain details: simulated transient evolutions and malfunctions, accuracy of simulation. The work concerning the assessment of the validity of certain simulators is reported. A list of simulator manufacturers and a survey of the principal simulators in operation in the countries of the European Community, in the United States, and in certain other countries are presented. Problem associated with the use of simulators as training facilities, and their use as operational devices are discussed. Studies and research in progress for the expected future development of simulators are reviewed

  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. ROSA full-core and DNBR capabilities

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

    2013-01-01

    The latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter are presented. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost 2 decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

  6. ROSA full-core and DNBR capabilities

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

    2012-01-01

    This paper presents the latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost two decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

  7. CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Cleveland, J.C.

    1977-01-01

    CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP

  8. Multidimensional, multiphysics simulations of core-collapse supernovae

    Energy Technology Data Exchange (ETDEWEB)

    Messer, O E B [National Center for Computational Sciences, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6008 (United States); Bruenn, S W [Department of Physics, Florida Atlantic University, Boca Raton, FL 33431-0991 (United States); Blondin, J M [Department of Physics, North Carolina State University, Raleigh, NC 27695-8202 (United States); Hix, W R; Mezzacappa, A [Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6354 (United States)

    2008-07-15

    CHIMERA is a multi-dimensional radiation hydrodynamics code designed to study core-collapse supernovae. The code is made up of three essentially independent parts: a hydrodynamics module, a nuclear burning module, and a neutrino transport solver combined within an operator-split approach. We review the code's architecture and some recently improved implementations used in the code. We also briefly discuss preliminary results obtained with the code in three spatial dimensions.

  9. 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

  10. 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)

  11. 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)

  12. 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.

  13. 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.

  14. 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.

  15. Simulation of an MSLB scenario using the 3D neutron kinetic core model DYN3D coupled with the CFD software Trio-U

    Energy Technology Data Exchange (ETDEWEB)

    Grahn, Alexander, E-mail: a.grahn@hzdr.de; Gommlich, André; Kliem, Sören; Bilodid, Yurii; Kozmenkov, Yaroslav

    2017-04-15

    Highlights: • Improved thermal-hydraulic description of nuclear reactor cores. • Providing reactor dynamics code with realistic thermal-hydraulic boundary conditions. • Possibility of three-dimensional flow phenomena in the core, such as cross flow, flow reversal. • Simulation at higher spatial resolution as compared to system codes. - Abstract: In the framework of the European project NURESAFE, the reactor dynamics code DYN3D, developed at Helmholtz-Zentrum Dresden-Rossendorf (HZDR), was coupled with the Computational Fluid Dynamics (CFD) solver Trio-U, developed at CEA France, in order to replace DYN3D’s one-dimensional hydraulic part with a full three-dimensional description of the coolant flow in the reactor core at higher spatial resolution. The present document gives an introduction into the coupling method and shows results of its application to the simulation of a Main Steamline Break (MSLB) accident of a Pressurised Water Reactor (PWR).

  16. Development of Core Simulator (CoSi) for APR1400 And Analysis of LPPT Result using APR1400-CoSi

    International Nuclear Information System (INIS)

    Moon, Sang-Rae; Kim, Yong-Bae; Lee, Eun-Ki

    2014-01-01

    According to NRC guidelines, Low Power Physics Test (LPPT) is required to be performed in low temperature/low pressure (160 .deg. C/42.2 kg/cm 2 ) as well as NOT/NOP (291.3 .deg. C/158.2 kg/cm 2 ) because Shin-Kori Unit 3 is FOAK nuclear power plant. Low Power Physics Test (LPPT) is essential to verify the nuclear design and robustness of reactor safety. LPPT consists of initial criticality, Point of Adding Heat (POAH), All Rod Out (ARO) Boron Concentration, Isothermal Temperature Coefficient (ITC), Control Rod Worth measurements and so on. KHNP-CRI has developed the Core Simulator for APR1400 (APR1400-CoSi) in order to improve the ability performing the LPPT. Especially, APR1400-CoSi has enhanced capability to calculate the full Core neutronic parameters by revising RAST-K that is three dimensional real time core kinetic program. APR1400-CoSi has been developed for education-training of Low Power Physics Test(LPPT). Particularly, APR1400-CoSi has an enhanced capability to calculate the full core neutronic parameters by revising RAST-K which is a three dimensional real time core kinetics program. Low Power Physics Test (LPPT) was performed using APR1400-CoSi and the results showed very similar values with the predicted ones. In other words, the initial core model of Shin-Kori Unit 3 in APR1400-CoSi system has been verified to be appropriate enough. Also, it was confirmed that the test range of Low Power Physics Test (LPPT) reamains effective even though the largest incremental bank reactivity is inserted in core by analyzing the power change during the rod SWAP test

  17. 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)

  18. Analysis of core uncovery time in Kuosheng station blackout transient with MELCOR

    International Nuclear Information System (INIS)

    Wang, S.J.; Chien, C.S.

    1996-01-01

    The MELCOR code, developed by the Sandia National Laboratories, is capable of simulating severe accident phenomena of nuclear power plants. Core uncovery time is an important parameter in the probabilistic risk assessment. However, many MELCOR users do not generate the initial conditions in a station blackout (SBO) transient analysis. Thus, achieving reliable core uncovery time is difficult. The core uncovery time for the Kuosheng nuclear power plant during an SBO transient is analyzed. First, full-power steady-state conditions are generated with the application of a developed self-initialization algorithm. Then the response of the SBO transient up to core uncovery is simulated. The effects of key parameters including the initialization process and the reactor feed pump (RFP) coastdown time on the core uncovery time are analyzed. The initialization process is the most important parameter that affects the core uncovery time. Because SBO transient analysis, the correct initial conditions must be generated to achieve a reliable core uncovery time. The core uncovery time is also sensitive to the RFP coastdown time. A correct time constant is required

  19. 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.)

  20. Development of a fiber-coupled laser-induced breakdown spectroscopy instrument for analysis of underwater debris in a nuclear reactor core

    International Nuclear Information System (INIS)

    Saeki, Morihisa; Iwanade, Akio; Ohba, Hironori; Ito, Chikara; Wakaida, Ikuo; Thornton, Blair; Sakka, Tetsuo

    2014-01-01

    To inspect the post-accident nuclear core reactor of the TEPCO Fukushima Daiichi nuclear power plant (F1-NPP), a transportable fiber-coupled laser-induced breakdown spectroscopy (LIBS) instrument has been developed. The developed LIBS instrument was designed to analyze underwater samples in a high-radiation field by single-pulse breakdown with gas flow or double-pulse breakdown. To check the feasibility of the assembled fiber-coupled LIBS instrument for the analysis of debris material (mixture of the fuel core, fuel cladding, construction material and so on) in the F1-NPP, we investigated the influence of the radiation dose on the optical transmittance of the laser delivery fiber, compared data quality among various LIBS techniques for an underwater sample and studied the feasibility of the fiber-coupled LIBS system in an analysis of the underwater sample of the simulated debris in F1-NPP. In a feasible study conducted by using simulated debris, which was a mixture of CeO 2 (surrogate of UO 2 ), ZrO 2 and Fe, we selected atomic lines suitable for the analysis of materials, and prepared calibration curves for the component elements. The feasible study has guaranteed that the developed fiber-coupled LIBS system is applicable for analyzing the debris materials in the F1-NPP. (author)

  1. Computer code for simulating pressurized water reactor core

    International Nuclear Information System (INIS)

    Serrano, A.M.B.

    1978-01-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)

  2. 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.)

  3. Rethinking Sensitivity Analysis of Nuclear Simulations with Topology

    Energy Technology Data Exchange (ETDEWEB)

    Dan Maljovec; Bei Wang; Paul Rosen; Andrea Alfonsi; Giovanni Pastore; Cristian Rabiti; Valerio Pascucci

    2016-01-01

    In nuclear engineering, understanding the safety margins of the nuclear reactor via simulations is arguably of paramount importance in predicting and preventing nuclear accidents. It is therefore crucial to perform sensitivity analysis to understand how changes in the model inputs affect the outputs. Modern nuclear simulation tools rely on numerical representations of the sensitivity information -- inherently lacking in visual encodings -- offering limited effectiveness in communicating and exploring the generated data. In this paper, we design a framework for sensitivity analysis and visualization of multidimensional nuclear simulation data using partition-based, topology-inspired regression models and report on its efficacy. We rely on the established Morse-Smale regression technique, which allows us to partition the domain into monotonic regions where easily interpretable linear models can be used to assess the influence of inputs on the output variability. The underlying computation is augmented with an intuitive and interactive visual design to effectively communicate sensitivity information to the nuclear scientists. Our framework is being deployed into the multi-purpose probabilistic risk assessment and uncertainty quantification framework RAVEN (Reactor Analysis and Virtual Control Environment). We evaluate our framework using an simulation dataset studying nuclear fuel performance.

  4. Numerical simulation on coolant flow and heat transfer in core

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    To simulate the coolant flow and the heat transfer characteristics of a core, a computer code, THAPMA (Thermal Hydraulic Analysis Porous Medium Analysis) has been developed. In THAPMA code, conservation equations are based on a porous-medium formulation, which uses four parameters, i.e, volume porosity, directional surface porosity, distributed resistance, and distributed heat source (sink), to model the effects of fuel rods and other internal solid structures on flow and heat transfer. Because the scheme and the solution are very important in accuracy and speed of calculation, a new difference scheme (WSUC) has been used in the energy equation, and a modified PISO solution method have been employed to simulate the steady/transient states. The code has been proved reliable and can effectively solve the transient state problem by several numerical tests. According to the design of Qinshan NPP-II, the flow and heat transfer phenomena in reactor core have been numerically simulated. The distributions of the velocity and the temperature can provide a theoretical basis for core design and safety analysis

  5. The innovative simulator for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kurosawa, A [The Inst. of Applied Energy, Tokyo (Japan); Ohashi, H; Akiyama, M [Univ. of Tokyo (Japan). Dept. of Nuclear Engineering

    1994-12-31

    Nuclear power simulators are becoming more and more important tools for ensuring the safety and the reliability during the whole cycle of plants from design to operation. Recently, there has been remarkable progress in computer science such as increase of computing speed, refinement of mathematical models and emergence of various AI technologies. By fully exploiting this progress to nuclear plant simulators, it becomes possible to achieve much faster, more extensive and more realistic simulation than ever. The Institute of Applied Energy (IAE) has organized a feasibility study on the advanced simulator since 1990, to develop the concept of nuclear power plant simulators in future. In this study, several academic organizations make fundamental researches on parallelization of transient analyses, large-scale parallel computing, thermal-hydraulic analysis using cellular automata, code development methodology by module-integration and task scheduling methods for parallel compilers. The concept and impact of the innovative simulator, as a multipurpose simulator complex, are summarized from the viewpoints of wide range scenarios including severe accidents, 3D multi-media interface, much faster than real-time simulation, and innovative algorithms for analyses of thermal-hydraulics, structure, neutronkinetics and their coupled phenomena. (orig.) (2 refs., 2 figs.).

  6. The innovative simulator for nuclear power plants

    International Nuclear Information System (INIS)

    Kurosawa, A.; Ohashi, H.; Akiyama, M.

    1994-01-01

    Nuclear power simulators are becoming more and more important tools for ensuring the safety and the reliability during the whole cycle of plants from design to operation. Recently, there has been remarkable progress in computer science such as increase of computing speed, refinement of mathematical models and emergence of various AI technologies. By fully exploiting this progress to nuclear plant simulators, it becomes possible to achieve much faster, more extensive and more realistic simulation than ever. The Institute of Applied Energy (IAE) has organized a feasibility study on the advanced simulator since 1990, to develop the concept of nuclear power plant simulators in future. In this study, several academic organizations make fundamental researches on parallelization of transient analyses, large-scale parallel computing, thermal-hydraulic analysis using cellular automata, code development methodology by module-integration and task scheduling methods for parallel compilers. The concept and impact of the innovative simulator, as a multipurpose simulator complex, are summarized from the viewpoints of wide range scenarios including severe accidents, 3D multi-media interface, much faster than real-time simulation, and innovative algorithms for analyses of thermal-hydraulics, structure, neutronkinetics and their coupled phenomena. (orig.) (2 refs., 2 figs.)

  7. Simulation of the nuclear power economy

    International Nuclear Information System (INIS)

    Triplett, M.B.

    1977-01-01

    Evaluation of nuclear power development policies requires the ability to forecast the economic and resource impacts attributable to a given policy. A computer simulation has been used in several recent evaluations of alternate nuclear power growth scenarios for the U.S. By using a discrete event modeling approach, a flexible tool has been developed that can simulate most planned reactor systems in terms of their overall economics and their impacts upon fuel cycle industries

  8. 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

  9. The role of computer simulation in nuclear technologies development

    International Nuclear Information System (INIS)

    Tikhonchev, M.Yu.; Shimansky, G.A.; Lebedeva, E.E.; Lichadeev, V. V.; Ryazanov, D.K.; Tellin, A.I.

    2001-01-01

    In the report the role and purposes of computer simulation in nuclear technologies development is discussed. The authors consider such applications of computer simulation as nuclear safety researches, optimization of technical and economic parameters of acting nuclear plant, planning and support of reactor experiments, research and design new devices and technologies, design and development of 'simulators' for operating personnel training. Among marked applications the following aspects of computer simulation are discussed in the report: neutron-physical, thermal and hydrodynamics models, simulation of isotope structure change and damage dose accumulation for materials under irradiation, simulation of reactor control structures. (authors)

  10. Development of nuclear power plant simulators for Soviet-designed nuclear reactors

    International Nuclear Information System (INIS)

    Kohut, P.; Tutu, N.K.; Cleary, E.J.; Erickson, K.G.; Yoder, J.; Kroshilin, A.

    2001-01-01

    The US Department of Energy (US DOE), under the US government's International Nuclear Safety Program (INSP), is implementing a program of developing and providing simulators for many of the Russian and Ukrainian Nuclear Power Plants (NPPs). Pacific Northwest National Laboratory (PNNL) and Brookhaven National Laboratory (BNL) manage and provide technical oversight of the various INSP simulator projects for DOE. The program also includes a simulator technology transfer process to simulator design organizations in Russia and Ukraine. Training programs, installation of new simulators, and enhancements in existing simulators, are viewed as providing a relatively fast and cost-effective technology transfer that will result in measurable improvement in the safety culture and operation of NPPs. A review of this program, its present status, and its accomplishments are provided in this paper

  11. The ab initio simulation of the Earth's core.

    Science.gov (United States)

    Alfè, D; Gillan, M J; Vocadlo, L; Brodholt, J; Price, G D

    2002-06-15

    The Earth has a liquid outer and solid inner core. It is predominantly composed of Fe, alloyed with small amounts of light elements, such as S, O and Si. The detailed chemical and thermal structure of the core is poorly constrained, and it is difficult to perform experiments to establish the properties of core-forming phases at the pressures (ca. 300 GPa) and temperatures (ca. 5000-6000 K) to be found in the core. Here we present some major advances that have been made in using quantum mechanical methods to simulate the high-P/T properties of Fe alloys, which have been made possible by recent developments in high-performance computing. Specifically, we outline how we have calculated the Gibbs free energies of the crystalline and liquid forms of Fe alloys, and so conclude that the inner core of the Earth is composed of hexagonal close packed Fe containing ca. 8.5% S (or Si) and 0.2% O in equilibrium at 5600 K at the boundary between the inner and outer cores with a liquid Fe containing ca. 10% S (or Si) and 8% O.

  12. Computer Simulation To Assess The Feasibility Of Coring Magma

    Science.gov (United States)

    Su, J.; Eichelberger, J. C.

    2017-12-01

    Lava lakes on Kilauea Volcano, Hawaii have been successfully cored many times, often with nearly complete recovery and at temperatures exceeding 1100oC. Water exiting nozzles on the diamond core bit face quenches melt to glass just ahead of the advancing bit. The bit readily cuts a clean annulus and the core, fully quenched lava, passes smoothly into the core barrel. The core remains intact after recovery, even when there are comparable amounts of glass and crystals with different coefficients of thermal expansion. The unique resulting data reveal the rate and sequence of crystal growth in cooling basaltic lava and the continuous liquid line of descent as a function of temperature from basalt to rhyolite. Now that magma bodies, rather than lava pooled at the surface, have been penetrated by geothermal drilling, the question arises as to whether similar coring could be conducted at depth, providing fundamentally new insights into behavior of magma. This situation is considerably more complex because the coring would be conducted at depths exceeding 2 km and drilling fluid pressures of 20 MPa or more. Criteria that must be satisfied include: 1) melt is quenched ahead of the bit and the core itself must be quenched before it enters the barrel; 2) circulating drilling fluid must keep the temperature of the coring assembling cooled to within operational limits; 3) the drilling fluid column must nowhere exceed the local boiling point. A fluid flow simulation was conducted to estimate the process parameters necessary to maintain workable temperatures during the coring operation. SolidWorks Flow Simulation was used to estimate the effect of process parameters on the temperature distribution of the magma immediately surrounding the borehole and of drilling fluid within the bottom-hole assembly (BHA). A solid model of the BHA was created in SolidWorks to capture the flow behavior around the BHA components. Process parameters used in the model include the fluid properties and

  13. Nuclear lattice simulations

    Directory of Open Access Journals (Sweden)

    Epelbaum E.

    2010-04-01

    Full Text Available We review recent progress on nuclear lattice simulations using chiral effective field theory. We discuss lattice results for dilute neutron matter at next-to-leading order, three-body forces at next-to-next-toleading order, isospin-breaking and Coulomb effects, and the binding energy of light nuclei.

  14. NMTC/JAM, Simulates High Energy Nuclear Reactions and Nuclear-Meson Transport Processes

    International Nuclear Information System (INIS)

    Furihata, Shiori

    2002-01-01

    1 - Description of program or function: NMTC/JAM is an upgraded version of the code system NMTC/JAERI97. NMTC/JAERI97 simulates high energy nuclear reactions and nucleon-meson transport processes. It implements an intra-nuclear cascade model taking account of the in-medium nuclear effects and the pre-equilibrium calculation model based on the exciton one. For treating the nucleon transport process, the nucleon-nucleus cross sections are revised to those derived by the systematics of Pearlstein. Moreover, the level density parameter derived by Ignatyuk is included as a new option for particle evaporation calculation. A geometry package based on the Combinatorial Geometry with multi-array system and the importance sampling technique is implemented in the code. Tally function is also employed for obtaining such physical quantities as neutron energy spectra, heat deposition and nuclide yield without editing a history file. The code can simulate both the primary spallation reaction and the secondary particle transport in the intermediate energy region from 20 MeV to 3.5 GeV by the use of the Monte Carlo technique. The code has been employed in combination with the neutron-photon transport codes available to the energy region below 20 MeV for neutronics calculation of accelerator-based subcritical reactors, analyses of thick target spallation experimented and so on. 2 - Methods: High energy nuclear reactions induced by incident high energy protons, neutrons and pions are simulated with the Monte Carlo Method by the intra-nuclear nucleon-nucleon reaction probabilities based on an intra-nuclear nucleon cascade model followed by the particle evaporation including high energy fission process. Jet-Aa Microscopic transport model (JAM) is employed to simulate high energy nuclear reactions in the energy range of GeV. All reaction channels are taken into account in the JAM calculation. An intra-nuclear cascade model (ISOBAR code) taking account of the in-medium nuclear effects

  15. Advanced gadolinia core and Toshiba advanced reactor management system

    International Nuclear Information System (INIS)

    Miyamoto, Toshiki; Yoshioka, Ritsuo; Ebisuya, Mitsuo

    1988-01-01

    At the Hamaoka Nuclear Power Station, Unit No. 3, advanced core design and core management technology have been adopted, significantly improving plant availability, operability and reliability. The outstanding technologies are the advanced gadolinia core (AGC) which utilizes gadolinium for the axial power distribution control, and Toshiba advanced reactor management system (TARMS) which uses a three-dimensional core physics simulator to calculate the power distribution. Presented here are the effects of these advanced technologies as observed during field testing. (author)

  16. Nuclear reactor multi-physics simulations with coupled MCNP5 and STAR-CCM+

    International Nuclear Information System (INIS)

    Cardoni, Jeffrey Neil; Rizwan-uddin

    2011-01-01

    The MCNP5 Monte Carlo particle transport code has been coupled to the computational fluid dynamics code, STAR-CCM+, to provide a high fidelity multi-physics simulation tool for pressurized water nuclear reactors. The codes are executed separately and coupled externally through a Perl script. The Perl script automates the exchange of temperature, density, and volumetric heating information between the codes using ASCII text data files. Fortran90 and Java utility programs assist job automation with data post-processing and file management. The MCNP5 utility code, MAKXSF, pre-generates temperature dependent cross section libraries for the thermal feedback calculations. The MCNP5–STAR-CCM+ coupled simulation tool, dubbed MULTINUKE, was applied to a steady state, PWR cell model to demonstrate its usage and capabilities. The demonstration calculation showed reasonable results that agree with PWR values typically reported in literature. Temperature and fission reaction rate distributions were realistic and intuitive. Reactivity coefficients were also deemed reasonable in comparison to historically reported data. The demonstration problem consisted of 9,984 CFD cells and 7,489 neutronic cells. MCNP5 tallied fission energy deposition over 3,328 UO_2 cells. The coupled solution converged within eight hours and in three MULTINUKE iterations. The simulation was carried out on a 64 bit, quad core, Intel 2.8 GHz microprocessor with 1 GB RAM. The simulations on a quad core machine indicated that a massively parallelized implementation of MULTINUKE can be used to assess larger multi-million cell models. (author)

  17. 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.

  18. Relativistic MHD simulations of stellar core collapse and magnetars

    Energy Technology Data Exchange (ETDEWEB)

    Font, Jose A; Gabler, Michael [Departamento de AstronomIa y Astrofisica, Universitat de Valencia, 46100 Burjassot (Valencia) (Spain); Cerda-Duran, Pablo; Mueller, Ewald [Max-Planck-Institut fuer Astrophysik, Karl-Schwarzschild-Str. 1, 85741 Garching (Germany); Stergioulas, Nikolaos, E-mail: j.antonio.font@uv.es [Department of Physics, Aristotle University of Thessaloniki, Thessaloniki 54124 (Greece)

    2011-02-01

    We present results from simulations of magneto-rotational stellar core collapse along with Alfven oscillations in magnetars. These simulations are performed with the CoCoA/CoCoNuT code, which is able to handle ideal MHD flows in dynamical spacetimes in general relativity. Our core collapse simulations highlight the importance of genuine magnetic effects, like the magneto-rotational instability, for the dynamics of the flow. For the modelling of magnetars we use the anelastic approximation to general relativistic MHD, which allows for an effective suppression of fluid modes and an accurate description of Alfven waves. We further compute Alfven oscillation frequencies along individual magnetic field lines with a semi-analytic approach. Our work confirms previous results based on perturbative approaches regarding the existence of two families of quasi-periodic oscillations (QPOs), with harmonics at integer multiples of the fundamental frequency. Additional material is presented in the accompanying contribution by Gabler et al (2010b) in these proceedings.

  19. 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

  20. Simulators in the training program for nuclear power plants

    International Nuclear Information System (INIS)

    Grimm, E.

    1988-01-01

    The principle simulator of the reactor school of the Paul Scherrer Institute is described. A compact simulator at the nuclear power plant Beznau is used for beginners as well as for refresher courses. Full simulator training cannot be taken in Switzerland. The Swiss nuclear power plants take advantage of the services of foreign nuclear power plants or training centers. The role of the instructor is discussed

  1. The role of computer simulation in nuclear technology development

    International Nuclear Information System (INIS)

    Tikhonchev, M.Yu.; Shimansky, G.A.; Lebedeva, E.E.; Lichadeev, VV.; Ryazanov, D.K.; Tellin, A.I.

    2000-01-01

    In the report, the role and purpose of computer simulation in nuclear technology development is discussed. The authors consider such applications of computer simulation as: (a) Nuclear safety research; (b) Optimization of technical and economic parameters of acting nuclear plant; (c) Planning and support of reactor experiments; (d) Research and design new devices and technologies; (f) Design and development of 'simulators' for operating personnel training. Among marked applications, the following aspects of computer simulation are discussed in the report: (g) Neutron-physical, thermal and hydrodynamics models; (h) Simulation of isotope structure change and dam- age dose accumulation for materials under irradiation; (i) Simulation of reactor control structures. (authors)

  2. Teaching simulator for divulgation of the nuclear energy

    International Nuclear Information System (INIS)

    Ortega B, M.G.; Gutierrez F, R.

    2003-01-01

    To solicitude of the authorities of the 'Universum' sciences museum of the UNAM, it develops a highly interactive computational system, to provide of information to the population in general about basic principles, uses and benefits of the nuclear energy. The objective is to achieve a better understanding and acceptance of the nuclear technology in our country. The system allows the visualization and simulation of nuclear processes as well as of its applications. The system is divided in three levels: basic, intermediate and simulation. In the basic level multimedia information is included on diverse basic concepts of the nuclear energy. The intermediate level includes the description and operation of some systems of the Laguna Verde nuclear power plant (CNLV). Finally the simulation level contains representative scenarios that the user can control by means of virtual control panels of the main systems of the CNLV. Inside the system a part of interactive games is included with the purpose that the user remembers with more easiness all the concepts and advantages of the nuclear energy mentioned during the previous levels. The system contributes, by means of the development of multimedia computational tools and of simulation, to the popularization of the use and applications of the nuclear energy in Mexico. (Author)

  3. Equipment and performance upgrade of compact nuclear simulator

    International Nuclear Information System (INIS)

    Park, J. C.; Kwon, K. C.; Lee, D. Y.; Hwang, I. K.; Park, W. M.; Cha, K. H.; Song, S. J.; Lee, J. W.; Kim, B. G.; Kim, H. J.

    1999-01-01

    The simulator at Nuclear Training Center in KAERI became old and has not been used effectively for nuclear-related training and researches due to the problems such as aging of the equipment, difficulties in obtaining consumables and their high cost, and less personnel available who can handle the old equipment. To solve the problems, this study was performed for recovering the functions of the simulator through the technical design and replacement of components with new ones. As results of this study, our test after the replacement showed the same simulation status as the previous one, and new graphic displays added to the simulator was effective for the training and easy for maintenance. This study is meaningful as demonstrating the way of upgrading nuclear training simulators that lost their functioning due to the obsolescence of simulators and the unavailability of components

  4. 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

  5. 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

  6. Measurement and simulation of thermal neutron flux distribution in the RTP core

    Science.gov (United States)

    Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

    2018-01-01

    The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

  7. 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

  8. 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

  9. Non Nuclear NTR Environmental Simulator

    International Nuclear Information System (INIS)

    Emrich, William J. Jr.

    2006-01-01

    Nuclear Thermal Rockets or NTR's have been suggested as a propulsion system option for vehicles traveling to the moon or Mars. These engines are capable of providing high thrust at specific impulses at least twice that of today's best chemical engines. The performance constraints on these engines are mainly the result of temperature limitations on the fuel coupled with a limited ability to withstand chemical attack by the hot hydrogen propellant. To operate at maximum efficiency, fuel forms are desired which can withstand the extremely hot, hostile environment characteristic of NTR operation for at least several hours. The simulation of such an environment would require an experimental device which could simultaneously approximate the power, flow, and temperature conditions which a nuclear fuel element (or partial element) would encounter during NTR operation. Such a simulation would allow detailed studies of the fuel behavior and hydrogen flow characteristics under reactor like conditions to be performed. The goal of these simulations would be directed toward expanding the performance envelope of NTR engines over that which was demonstrated during the Rover and NERVA nuclear rocket programs of the 1970's. Current planning calls for such a simulator to be constructed at the Marshall Space Flight Center over the coming year, and it is anticipated that it will be used in the future to evaluate a wide variety of fuel element designs and the materials of which they are constructed. This present work addresses the initial experimental objectives of the NTR simulator with regard to reproducing the fuel degradation patterns previously observed during the NERVA testing

  10. 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)

  11. An improved one-and-a-half group BWR core simulator for a new-generation core management system

    International Nuclear Information System (INIS)

    Iwamoto, Tatsuya; Yamamoto, Munenari

    2000-01-01

    An improved one-and-a-half group core simulator method for a next-generation BWR core management system is presented. In the improved method, intranodal spectral index (thermal to fast flux ratio) is expanded with analytic solutions to the diffusion equation, and the nodal power density and the interface net current are calculated, taking the intranodal flux shape into consideration. A unique method was developed for assembly heterogeneity correction. Thus eliminating the insufficiencies of the conventional one-and-a-half group method, we can have accurate power distributions as well as local peaking factors for cores having large spectral mismatch between fuel assemblies. The historical effects of spectral mismatch are also considered in both nodal power and local peaking calculations. Although reflectors are not solved explicitly, there is essentially no need for core dependent adjustable parameters, since boundary conditions are derived in the same manner as in the interior nodes. Calculation time for nodal solutions is comparable to that for the conventional method, and is less than 1/10 of a few-group nodal simulator. Verifications of the present method were made by comparing the results with those obtained by heterogeneous fine-mesh multi-group core depletion calculations, and the accuracy was shown to be fairly good. (author)

  12. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    International Nuclear Information System (INIS)

    Lioce, D.; Asztalos, M.; Alemberti, A.; Barucca, L.; Frogheri, M.; Saiu, G.

    2012-01-01

    Highlights: ► Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. ► The two tests have been simulated by means of the Relap5 computer code. ► Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000 ® plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two selected tests has been defined and, in order to perform the pre-operational tests simulations, a

  13. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Lioce, D., E-mail: donato.lioce@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Asztalos, M., E-mail: asztalmj@westinghouse.com [Westinghouse Electric Company, Cranberry Twp, PA 16066 (United States); Alemberti, A., E-mail: alessandro.alemberti@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Barucca, L. [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Frogheri, M., E-mail: monicalinda.frogheri@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Saiu, G., E-mail: gianfranco.saiu@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. Black-Right-Pointing-Pointer The two tests have been simulated by means of the Relap5 computer code. Black-Right-Pointing-Pointer Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000{sup Registered-Sign} plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two

  14. Transferring Nuclear Knowledge by NPP Simulators Developers

    International Nuclear Information System (INIS)

    Levchenko, A.; Duginov, O.; Levchenko, V.

    2016-01-01

    Full text: In relation to nuclear power, safety is closely linked with the human factor. Knowledge and skills of staff should more closely match the needs of the industry and employers. This can be achieved through more efficient training using simulators. Such simulators must be available at all stages of study and cover the needs of trainees with different levels of knowledge. Simulator developers can solve this problem by implement knowledge management in the using of simulators for practical training. Due to the nature of their activities, they have everything needed for this. This paper describes the solutions of Simulation Systems Ltd Company for nuclear power plants, universities and other parties. (author

  15. Framework Application for Core Edge Transport Simulation (FACETS)

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, Sergei; Pigarov, Alexander

    2011-10-15

    The FACETS (Framework Application for Core-Edge Transport Simulations) project of Scientific Discovery through Advanced Computing (SciDAC) Program was aimed at providing a high-fidelity whole-tokamak modeling for the U.S. magnetic fusion energy program and ITER through coupling separate components for each of the core region, edge region, and wall, with realistic plasma particles and power sources and turbulent transport simulation. The project also aimed at developing advanced numerical algorithms, efficient implicit coupling methods, and software tools utilizing the leadership class computing facilities under Advanced Scientific Computing Research (ASCR). The FACETS project was conducted by a multi-discipline, multi-institutional teams, the Lead PI was J.R. Cary (Tech-X Corp.). In the FACETS project, the Applied Plasma Theory Group at the MAE Department of UCSD developed the Wall and Plasma-Surface Interaction (WALLPSI) module, performed its validation against experimental data, and integrated it into the developed framework. WALLPSI is a one-dimensional, coarse grained, reaction/advection/diffusion code applied to each material boundary cell in the common modeling domain for a tokamak. It incorporates an advanced model for plasma particle transport and retention in the solid matter of plasma facing components, simulation of plasma heat power load handling, calculation of erosion/deposition, and simulation of synergistic effects in strong plasma-wall coupling.

  16. 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)

  17. Simulation of the turbine trip of Unit 1 of the Laguna Verde nuclear power plant using the code Simulate-3K

    International Nuclear Information System (INIS)

    Alegria A, A.; Filio L, C.; Ortiz V, J.

    2017-09-01

    In order to compare the results obtained from the model developed in the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) with the code Simulate-3K (S3K) with respect to those reported by the process computer of the Central (SIIP), the simulation of the turbine trip transient was carried out, caused by the firing of the main generator, the low differential pressure of oil of its seals and the automatic Scram of Unit 1 of the Laguna Verde nuclear power plant, at 87% of power nominal during the operation cycle 16. Since the reactor was brought to a safe stop due to Scram, was enough to simulate 20 seconds to observe the maximum increase in pressure with S3K. In this work, the following parameters are shown and compared: the neutron flux, the thermal power, the pressure in the dome, the flow at the entrance to the core, the steam flow that leaves the vessel and the minimal critical power ratio (MCPR). The neutron flux of the average power range monitors of the nuclear power plant was compared with the S3K detectors model. Finally, the MCPR was calculated with a different correlation to that of the fuel supplier and its deviation from its safety limit was determined. In conclusion, the results obtained show the current state of the model for the simulation of reactivity transients and the opportunity areas to consolidate this tool in support of the process of licensing refueling in the CNSNS. (Author)

  18. Nuclear Application Programs Development and Integration for a Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun-Joon; Lee, Tae-Woo [KEPCO Engineering and Construction Co., Deajeon (Korea, Republic of)

    2016-10-15

    KEPCO E and C participated in the NAPS (Nuclear Application Programs) development project for BNPP (Barakah Nuclear Power Plant) simulator. The 3KEY MASTER™ was adopted for this project, which is comprehensive simulation platform software developed by WSC (Western Services Corporation) for the development, and control of simulation software. The NAPS based on actual BNPP project was modified in order to meet specific requirements for nuclear power plant simulators. Considerations regarding software design for BNPP simulator and interfaces between the 3KM platform and application programs are discussed. The repeatability is one of functional requirements for nuclear power plant simulators. In order to migrate software from actual plants to simulators, software functions for storing and retrieving plant conditions and program variables should be implemented. In addition, software structures need to be redesigned to meet the repeatability, and source codes developed for actual plants would have to be optimized to reflect simulator’s characteristics as well. The synchronization is an important consideration to integrate external application programs into the 3KM simulator.

  19. 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

  20. 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)

  1. Development of a compact nuclear power station engineering simulator

    International Nuclear Information System (INIS)

    Jian Jianfeng; Yang Yanhua; Lin Meng; Hu Rui

    2003-01-01

    The compact nuclear power plant project simulator is developed based on the Chashma nuclear power plant. This simulator consists of simulation computation code, data communication module and human-machine interface. This paper discusses the design and implementation of the simulator from such aspect as computer system, hydrothermal model, programming language, human-machine interface and data communication in details

  2. Simulators in nuclear power sector

    International Nuclear Information System (INIS)

    Mathey, C.; Roux, J.

    1984-01-01

    The simulator has established itself as an indispensable tool for training nuclear power station operators. After summarizing the main advantages of this training method, the author examines different types of simulators (for training or planning) and their architecture. He then describes the mathematical models used to simulate operation of the various elements of the ''power station'' and guarantee accurate representativity of phenomena associated with the power station operating under normal and accident conditions [fr

  3. Evolution dynamics modeling and simulation of logistics enterprise's core competence based on service innovation

    Science.gov (United States)

    Yang, Bo; Tong, Yuting

    2017-04-01

    With the rapid development of economy, the development of logistics enterprises in China is also facing a huge challenge, especially the logistics enterprises generally lack of core competitiveness, and service innovation awareness is not strong. Scholars in the process of studying the core competitiveness of logistics enterprises are mainly from the perspective of static stability, not from the perspective of dynamic evolution to explore. So the author analyzes the influencing factors and the evolution process of the core competence of logistics enterprises, using the method of system dynamics to study the cause and effect of the evolution of the core competence of logistics enterprises, construct a system dynamics model of evolution of core competence logistics enterprises, which can be simulated by vensim PLE. The analysis for the effectiveness and sensitivity of simulation model indicates the model can be used as the fitting of the evolution process of the core competence of logistics enterprises and reveal the process and mechanism of the evolution of the core competence of logistics enterprises, and provide management strategies for improving the core competence of logistics enterprises. The construction and operation of computer simulation model offers a kind of effective method for studying the evolution of logistics enterprise core competence.

  4. Dynamic Simulator for Nuclear Power Plants (DSNP)

    International Nuclear Information System (INIS)

    Saphier, D.

    1976-01-01

    A new simulation language DSNP (Dynamic Simulator for Nuclear Power Plants) is being developed. It is a simple block oriented simulation language with an extensive library of component and auxiliary modules. Each module is a self-contained unit of a part of a physical component to be found in nuclear power plants. Each module will be available in four levels of sophistication, the fourth being a user supplied model. A module can be included in the simulation by a single statement. The precompiler translates DSNP statements into FORTRAN statements, takes care of the module parameters and the intermodular communication blocks, prepares proper data files and I/0 statements and searches the various libraries for the appropriate component modules. The documentation is computerized and all the necessary information for a particular module can be retrieved by a special document generator. The DSNP will be a flexible tool which will allow dynamic simulations to be performed on a large variety of nuclear power plants or specific components of these plants

  5. Design of compact nuclear power marine engineering simulator

    International Nuclear Information System (INIS)

    Gao Jinghui; Xing Hongchuan; Zhang Ronghua; Yang Yanhua; Xu Jijun

    2004-01-01

    The essentiality of compact nuclear power marine engineering simulator (NPMES) is discussed. The technology of nuclear power plant engineering simulator (NPPES) for NPMES development is introduced, and the function design, general design and model design are given in details. A compact NPMES based on the nuclear power marine of 'Mutsu' is developed. The design can help the development of NPMES, which will improve operation safety and management efficiency of marine. (authors)

  6. Performance Test of Core Protection and Monitoring Algorithm with DLL for SMART Simulator Implementation

    International Nuclear Information System (INIS)

    Koo, Bonseung; Hwang, Daehyun; Kim, Keungkoo

    2014-01-01

    A multi-purpose best-estimate simulator for SMART is being established, which is intended to be used as a tool to evaluate the impacts of design changes on the safety performance, and to improve and/or optimize the operating procedure of SMART. In keeping with these intentions, a real-time model of the digital core protection and monitoring systems was developed and the real-time performance of the models was verified for various simulation scenarios. In this paper, a performance test of the core protection and monitoring algorithm with a DLL file for the SMART simulator implementation was performed. A DLL file of the simulator application code was made and several real-time evaluation tests were conducted for the steady-state and transient conditions with simulated system variables. A performance test of the core protection and monitoring algorithms for the SMART simulator was performed. A DLL file of the simulator version code was made and several real-time evaluation tests were conducted for various scenarios with a DLL file and simulated system variables. The results of all test cases showed good agreement with the reference results and some features caused by algorithm change were properly reflected to the DLL results. Therefore, it was concluded that the SCOPS S SIM and SCOMS S SIM algorithms and calculational capabilities are appropriate for the core protection and monitoring program in the SMART simulator

  7. 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

  8. Transfer coefficients in a four-cusp duct simulating a typical nuclear reactor channel degraded by accident

    International Nuclear Information System (INIS)

    Souza Dutra, A. de.

    1985-01-01

    An experimental study on forced convection in a four-cusp duct simulating a typical nuclear reactor channel degraded by accident is presented. Transfer coefficients were obtained by using the analogy between heat and mass tranfer, with the naphtalene sublimation technique. The experiment consisted in forcing air past a four-cusp naphthalene moulded duct. Mass transfer coefficients were determined in nondimensional form as Sherwood number. Experimental curves correlating the Sherwood number with a nondimensional length, x + , were obtained for Reynolds number varying from 891 to 30.374. This range covers typical flow rates that are expected to exist in a degraded nuclear reactor core. (Author) [pt

  9. 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

  10. Simulation experiment on the flooding behaviour of core melts: KATS-9

    International Nuclear Information System (INIS)

    Fieg, G.; Massier, H.; Schuetz, W.; Stegmaier, U.; Stern, G.

    2000-11-01

    For future Light Water Reactors special devices (core catchers) are being developed to prevent containment failure by basement erosion after reactor pressure vessel meltthrough during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher devices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent water cooling. A KATS-experiment has been performed to investigate the flooding behaviour of high temperature melts using alumina-iron thermite melts as a simulant. The oxidic thermite melt is conditioned by adding other oxides to simulate a realistic corium melt as close as possible in terms of liquidus and solidus temperatures. Before flooding with water, spreading of the separate oxidic and metallic melts has been done in one-dimensional channels with a silicate concrete as the substrate. The flooding rate was, in relation to the melt surface, identical to the flooding rate in EPR. (orig.) [de

  11. Geoelectrical monitoring of simulated subsurface leakage to support high-hazard nuclear decommissioning at the Sellafield Site, UK.

    Science.gov (United States)

    Kuras, Oliver; Wilkinson, Paul B; Meldrum, Philip I; Oxby, Lucy S; Uhlemann, Sebastian; Chambers, Jonathan E; Binley, Andrew; Graham, James; Smith, Nicholas T; Atherton, Nick

    2016-10-01

    A full-scale field experiment applying 4D (3D time-lapse) cross-borehole Electrical Resistivity Tomography (ERT) to the monitoring of simulated subsurface leakage was undertaken at a legacy nuclear waste silo at the Sellafield Site, UK. The experiment constituted the first application of geoelectrical monitoring in support of decommissioning work at a UK nuclear licensed site. Images of resistivity changes occurring since a baseline date prior to the simulated leaks revealed likely preferential pathways of silo liquor simulant flow in the vadose zone and upper groundwater system. Geophysical evidence was found to be compatible with historic contamination detected in permeable facies in sediment cores retrieved from the ERT boreholes. Results indicate that laterally discontinuous till units forming localized hydraulic barriers substantially affect flow patterns and contaminant transport in the shallow subsurface at Sellafield. We conclude that only geophysical imaging of the kind presented here has the potential to provide the detailed spatial and temporal information at the (sub-)meter scale needed to reduce the uncertainty in models of subsurface processes at nuclear sites. Copyright © 2016 British Geological Survey, NERC. Published by Elsevier B.V. All rights reserved.

  12. Temperature-dependent errors in nuclear lattice simulations

    International Nuclear Information System (INIS)

    Lee, Dean; Thomson, Richard

    2007-01-01

    We study the temperature dependence of discretization errors in nuclear lattice simulations. We find that for systems with strong attractive interactions the predominant error arises from the breaking of Galilean invariance. We propose a local 'well-tempered' lattice action which eliminates much of this error. The well-tempered action can be readily implemented in lattice simulations for nuclear systems as well as cold atomic Fermi systems

  13. CANDU 9 nuclear power plant simulator

    International Nuclear Information System (INIS)

    Kattan, M.; MacBeth, M.J.; Lam, K.

    1995-01-01

    Simulators are playing, an important role in the design and operations of CANDU reactors. They are used to analyze operating procedures under standard and upset conditions. The CANDU 9 nuclear power plant simulator is a low fidelity, near full scope capability simulator. It is designed to play an integral part in the design and verification of the control centre mock-up located in the AECL design office. It will also provide CANDU plant process dynamic data to the plant display system (PDS), distributed control system (DCS) and to the mock-up panel devices. The simulator model employs dynamic mathematical models of the various process and control components that make up a nuclear power plant. It provides the flexibility to add, remove or update user supplied component models. A block oriented process input is provided with the simulator. Individual blocks which represent independent algorithms of the model are linked together to generate the required overall plant model. As a design tool the simulator will be used for control strategy development, human factors studies (information access, readability, graphical display design, operability), analysis of overall plant control performance, tuning estimates for major control loops and commissioning strategy development. As a design evaluation tool, the simulator will be used to perform routine and non-routine procedures, practice 'what if' scenarios for operational strategy development, practice malfunction recovery procedures and verify human factors activities. This paper will describe the CANDU 9 plant simulator and demonstrate its implementation and proposed utility as a tool in the control system and control centre design of a CANDU 9 nuclear power plant. (author). 2 figs

  14. Corrosion of simulated nuclear waste glass

    International Nuclear Information System (INIS)

    Music, S.; Ristic, M.; Gotic, M.; Foric, J.

    1988-01-01

    In this study the preparation and characterization of borosilicate glasses of different chemical composition were investigated. Borosilicate glasses were doped with simulated nuclear waste oxides. The chemical corrosion in water of these glasses was followed by measuring the leach rates as a function of time. It was found that a simulated nuclear waste glass with the chemical composition (weight %), 15.61% Na 2 O, 10.39% B 2 O 3 , 45.31% SiO 2 , 13.42% ZnO, 6.61% TiO 2 and 8.66% waste oxides, is characterized by low melting temperature and with good corrosion resistance in water. Influence of passive layers on the leaching behaviour of nuclear waste glasses is discussed. (author) 20 refs.; 7 figs.; 4 tabs

  15. Simulation of a marine nuclear reactor

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Kyouya, Masahiko; Kobayashi, Hideo; Ochiai, Masaaki

    1995-01-01

    A Nuclear-powered ship Engineering Simulation SYstem (NESSY) has been developed by the Japan Atomic Energy Research Institute as an advanced design tool for research and development of future marine reactors. A marine reactor must respond to changing loads and to the ship's motions because of the ship's maneuvering and its presence in a marine environment. The NESSY has combined programs for the reactor plant behavior calculations and the ship's motion calculations. Thus, it can simulate reactor power fluctuations caused by changing loads and the ship's motions. It can also simulate the behavior of water in the pressurizer and steam generators. This water sloshes in response to the ship's motions. The performance of NESSY has been verified by comparing the simulation calculations with the measured data obtained by experiments performed using the nuclear ship Mutsu. The effects of changing loads and the ship's motions on the reactor behavior can be accurately simulated by NESSY

  16. 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

  17. Nuclear Power Plant Simulators: Goals and Evolution

    International Nuclear Information System (INIS)

    Miettinen, J.

    2008-01-01

    OECD / CSNI organised two Specialists Meetings in the field of plant simulators. With the development of the digital computation technology the progress in this field has been fast and there was a distinct need for reviewing the current state-of-the-art and the future development. The meeting were hosted by the Technical Research Centre of Finland (VTT), and they were arranged in 1992 and 1997. The paper discusses the evolution of the simulator technology and especially modelling of the main systems in the simulators used for the operator training and evaluation of the plant characteristics. The model development in 1990 was rapid due the increased computer capability and an interest towards the severe transients and full 3 D core neutronic simulation. The analysis models without real time calculation capability included a detailed model for the primary circuit, core or containment. The simulator models tried to describe the entire plant, but the primary circuit, core and containment were described with a robust, fast running model. After rapid development in the computational speed the detailed analysis models and robust simulator models do not need to be considered separately anymore. But there are new features, which force the simulator technology following the development. The hard control panels are replaced by CRT's, and / or the programmable digital technology replaces the analogue automation. The future full-scope simulators are replica's of the Crt control room and programmed automation. Only the process itself is calculated by the codes. One could ask, how with the fast development the state-of-the-art in 1992 and 1997 is useful still in 2008 in this fast developing area. One argument may be that understanding the history may help in creating the new understanding. And in fact many visions for the future were seen in those meeting. Some of them have been realized, some other not.

  18. Experimental simulation of fragmentation and stratification of core debris on the core catcher of a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pillai, Dipin S.; Vignesh, R. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Sudha, A. Jasmin, E-mail: jasmin@igcar.gov.in [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Pushpavanam, S.; Sundararajan, T. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Nashine, B.K.; Selvaraj, P. [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2016-05-15

    Highlights: • Fragmentation of two simultaneous metals jets in a bulk coolant analysed. • Particle size from experiments compared with theoretical analysis. • Jet breakup modes explained using dimensionless numbers. • Settling aspects of aluminium and lead debris on collector plate studied. • Results analysed in light of core debris settling on core catcher in a FBR. - Abstract: The complex and coupled phenomena of two simultaneous molten metal jets fragmenting inside a quiescent liquid pool and settling on a collector plate are experimentally analysed in the context of safety analysis of a fast breeder reactor (FBR) in the post accident heat removal phase. Following a hypothetical core melt down accident in a FBR, a major portion of molten nuclear fuel and clad/structural material which are collectively termed as ‘corium’ undergoes fragmentation in the bulk coolant sodium in the lower plenum of the reactor main vessel and settles on the core catcher plate. The coolability of this decay heat generating debris bed is dependent on the particle size distribution and its layering i.e., stratification. Experiments have been conducted with two immiscible molten metals of different densities poured inside a coolant medium to understand their fragmentation behaviour and to assess the possibility of formation of a stratified debris bed. Molten aluminium and lead have been used as simulants in place of molten stainless steel and nuclear fuel to facilitate easy handling. This paper summarizes the major findings from these experiments. The fragmentation of the two molten metals are explained in the light of relevant dimensionless numbers such as Reynolds number and Weber Number. The mass median diameter of the fragmented debris is predicted from nonlinear stability analysis of slender jets for lead jet and using Rayleigh's classical theory of jet breakup for aluminium jet. The agreement of the predicted values with the experimental results is good. These

  19. A Two-Step Approach to Uncertainty Quantification of Core Simulators

    Directory of Open Access Journals (Sweden)

    Artem Yankov

    2012-01-01

    Full Text Available For the multiple sources of error introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily to quantify the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the “two-step” method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the two-step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor and in the core power distribution was examined in the framework of phase I-3 of the OECD Uncertainty Analysis in Modeling benchmark. With the Three Mile Island Unit 1 core as a base model for analysis, the XSUSA and two-step methods were applied with certain limitations, and the results were compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions were drawn as to the method that is currently more viable for computing uncertainties in burnup and transient calculations.

  20. Military nuclear activities. The simulation program

    International Nuclear Information System (INIS)

    Delpuech, A.

    2000-01-01

    The durability of the French nuclear weapon dissuasion has to integrate two kind of problems: the geopolitical situation with the comprehensive nuclear test ban treaty (CTBT) and the aging of weapons. The replacement of decayed weapons requires a complete safety and reliability validation of the new weapons which is performed using simulation. This paper gives a brief presentation of the simulation program and of the technical means developed by the military division of the French atomic energy commission (CEA-DAM): the Airix X-ray radiography installation and the 'megajoule' laser facility. (J.S.)

  1. Simulations. 3D nuclear reactions; Simulations. Reactions nucleaires en 3D

    Energy Technology Data Exchange (ETDEWEB)

    Deleurence, Guillaume

    2012-05-15

    At CEA Saclay, the research programme for the forth generation of nuclear reactors, named 'Astrid' (Advanced Sodium Technological Reactor for Industrial Demonstration) of the Direction of nuclear energy (DEN), uses 2D or 3D movies and a 16 m{sup 2} image wall for the display of simulation results. This digest paper focusses on the technological means used for the simulations: the Curie supercomputer designed by Bull and the images display device. (J.S.)

  2. 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)

  3. Impact of Neutrino Opacities on Core-collapse Supernova Simulations

    Science.gov (United States)

    Kotake, Kei; Takiwaki, Tomoya; Fischer, Tobias; Nakamura, Ko; Martínez-Pinedo, Gabriel

    2018-02-01

    The accurate description of neutrino opacities is central to both the core-collapse supernova (CCSN) phenomenon and the validity of the explosion mechanism itself. In this work, we study in a systematic fashion the role of a variety of well-selected neutrino opacities in CCSN simulations where the multi-energy, three-flavor neutrino transport is solved using the isotropic diffusion source approximation (IDSA) scheme. To verify our code, we first present results from one-dimensional (1D) simulations following the core collapse, bounce, and ∼250 ms postbounce of a 15 {M}ȯ star using a standard set of neutrino opacities by Bruenn. A detailed comparison with published results supports the reliability of our three-flavor IDSA scheme using the standard opacity set. We then investigate in 1D simulations how individual opacity updates lead to differences with the baseline run with the standard opacity set. Through detailed comparisons with previous work, we check the validity of our implementation of each update in a step-by-step manner. Individual neutrino opacities with the largest impact on the overall evolution in 1D simulations are selected for systematic comparisons in our two-dimensional (2D) simulations. Special attention is given to the criterion of explodability in the 2D models. We discuss the implications of these results as well as its limitations and the requirements for future, more elaborate CCSN modeling.

  4. Validation of Simulation Codes for Future Systems: Motivations, Approach and the Role of Nuclear Data

    International Nuclear Information System (INIS)

    G. Palmiotti; M. Salvatores; G. Aliberti

    2007-01-01

    The validation of advanced simulation tools will still play a very significant role in several areas of reactor system analysis. This is the case of reactor physics and neutronics, where nuclear data uncertainties still play a crucial role for many core and fuel cycle parameters. The present paper gives a summary of validation motivations, objectives and approach. A validation effort is in particular necessary in the frame of advanced (e.g. Generation-IV or GNEP) reactors and associated fuel cycles assessment and design

  5. Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code

    Directory of Open Access Journals (Sweden)

    Yutthapong Pinanroj

    2014-04-01

    Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.

  6. Track 3: growth of nuclear technology and research numerical and computational aspects of the coupled three-dimensional core/plant simulations: organization for economic cooperation and development/U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-I. 5. Analyses of the OECD MSLB Benchmark with the Codes DYN3D and DYN3D/ATHLET

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.

    2001-01-01

    The code DYN3D coupled with ATHLET was used for the analysis of the OECD Main-Steam-Line-Break (MSLB) Benchmark, which is based on real plant design and operational data of the TMI-1 pressurized water reactor (PWR). Like the codes RELAP or TRAC,ATHLET is a thermal-hydraulic system code with point or one-dimensional neutron kinetic models. ATHLET, developed by the Gesellschaft for Anlagen- und Reaktorsicherheit, is widely used in Germany for safety analyses of nuclear power plants. DYN3D consists of three-dimensional nodal kinetic models and a thermal-hydraulic part with parallel coolant channels of the reactor core. DYN3D was coupled with ATHLET for analyzing more complex transients with interactions between coolant flow conditions and core behavior. It can be applied to the whole spectrum of operational transients and accidents, from small and intermediate leaks to large breaks of coolant loops or steam lines at PWRs and boiling water reactors. The so-called external coupling is used for the benchmark, where the thermal hydraulics is split into two parts: DYN3D describes the thermal hydraulics of the core, while ATHLET models the coolant system. Three exercises of the benchmark were simulated: Exercise 1: point kinetics plant simulation (ATHLET) Exercise 2: coupled three-dimensional neutronics/core thermal-hydraulics evaluation of the core response for given core thermal-hydraulic boundary conditions (DYN3D) Exercise 3: best-estimate coupled core-plant transient analysis (DYN3D/ATHLET). Considering the best-estimate cases (scenarios 1 of exercises 2 and 3), the reactor does not reach criticality after the reactor trip. Defining more serious tests for the codes, the efficiency of the control rods was decreased (scenarios 2 of exercises 2 and 3) to obtain recriticality during the transient. Besides the standard simulation given by the specification, modifications are introduced for sensitivity studies. The results presented here show (a) the influence of a reduced

  7. Integrated core-SOL simulations of L-mode plasma in ITER and Indian demo

    International Nuclear Information System (INIS)

    Wisitsorasak, Apiwat; Onjun, Thawatchai; Kanjanaput, Wittawat

    2015-01-01

    Core-SOL simulations are carried out using 1.5D BALDUR integrated predictive modeling code to investigate tokamak plasma in ITER and Indian DEMO reactors operating in low confinement mode (L-Mode). In each simulation, the plasma current, temperature, and density profiles in both core and SOL region are evolved self-consistency. The SOL is simulated by integrating the fluid equations, including sources, along the field lines. The solutions in SOL subsequently provide as the boundary conditions of core plasma region on low-confinement mode. The core plasma transport model is described using a combination of anomalous transport by Multi-Mode-Model version 2001 (MMM2001) and neoclassical transport calculated by NCLASS module together with the toroidal velocity based on the torque due to Neoclassical Toroidal Viscosity (NTV). In addition, a sensitivity analysis is explored by varying plasma parameters, such as plasma density and auxiliary heating power. Furthermore, the ignition tests are conducted to observed plasma response in each design after shutting down an auxiliary heating. (author)

  8. An FPGA computing demo core for space charge simulation

    International Nuclear Information System (INIS)

    Wu, Jinyuan; Huang, Yifei

    2009-01-01

    In accelerator physics, space charge simulation requires large amount of computing power. In a particle system, each calculation requires time/resource consuming operations such as multiplications, divisions, and square roots. Because of the flexibility of field programmable gate arrays (FPGAs), we implemented this task with efficient use of the available computing resources and completely eliminated non-calculating operations that are indispensable in regular micro-processors (e.g. instruction fetch, instruction decoding, etc.). We designed and tested a 16-bit demo core for computing Coulomb's force in an Altera Cyclone II FPGA device. To save resources, the inverse square-root cube operation in our design is computed using a memory look-up table addressed with nine to ten most significant non-zero bits. At 200 MHz internal clock, our demo core reaches a throughput of 200 M pairs/s/core, faster than a typical 2 GHz micro-processor by about a factor of 10. Temperature and power consumption of FPGAs were also lower than those of micro-processors. Fast and convenient, FPGAs can serve as alternatives to time-consuming micro-processors for space charge simulation.

  9. An FPGA computing demo core for space charge simulation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Jinyuan; Huang, Yifei; /Fermilab

    2009-01-01

    In accelerator physics, space charge simulation requires large amount of computing power. In a particle system, each calculation requires time/resource consuming operations such as multiplications, divisions, and square roots. Because of the flexibility of field programmable gate arrays (FPGAs), we implemented this task with efficient use of the available computing resources and completely eliminated non-calculating operations that are indispensable in regular micro-processors (e.g. instruction fetch, instruction decoding, etc.). We designed and tested a 16-bit demo core for computing Coulomb's force in an Altera Cyclone II FPGA device. To save resources, the inverse square-root cube operation in our design is computed using a memory look-up table addressed with nine to ten most significant non-zero bits. At 200 MHz internal clock, our demo core reaches a throughput of 200 M pairs/s/core, faster than a typical 2 GHz micro-processor by about a factor of 10. Temperature and power consumption of FPGAs were also lower than those of micro-processors. Fast and convenient, FPGAs can serve as alternatives to time-consuming micro-processors for space charge simulation.

  10. 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.

  11. Simulator of nuclear power plant with WWER-440 units

    International Nuclear Information System (INIS)

    Krcek, V.

    1985-01-01

    The use is discussed of simulators in the training of qualified personnel for the construction and operation of nuclear power plants. Simulators are used for training all activities and thinking processes related to the control of a nuclear reactor in the course of quasi-steady and non-steady states. The development and implementation is summed up of the construction of such a simulator for WWER-440 nuclear power plants. The main parts of the simulator include the unit control room, the computer system, the teacher's workplace and the interface system. The possibility of simulating the functions of the unit for personnel training is based on the description of the behaviour of the simulated object in form of mathematical models of its basic technological subsystems and their interrelations within the range of operating patterns. (J.C.)

  12. Design of blast simulators for nuclear testing

    International Nuclear Information System (INIS)

    Mark, A.; Opalka, K.O.; Kitchens, C.W. Jr.

    1983-01-01

    A quasi-one-dimensional computational technique is used to model the flow of a large, complicated shock tube. The shock tube, or Large Blast Simulator, is used to simulate conventional or nuclear explosions by shaping the pressure history. Results from computations show favorable agreement when compared with data taken in the facility at Gramat, France. Such future shock tubes will include a thermal irradiation capability to better simulate a nuclear event. The computations point to the need for venting of the combustion products since the pressure history will be considerably altered as the shock propagates through these hot gases

  13. 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

  14. 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)

  15. 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.

  16. Simulation of the turbine trip of Unit 1 of the Laguna Verde nuclear power plant using the code Simulate-3K; Simulacion del disparo de turbina de la Unidad 1 de la central nuclear Laguna Verde empleando el codigo Simulate-3K

    Energy Technology Data Exchange (ETDEWEB)

    Alegria A, A. [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico); Filio L, C. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, 07738 Ciudad de Mexico (Mexico); Ortiz V, J., E-mail: aalegria@cnsns.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    In order to compare the results obtained from the model developed in the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) with the code Simulate-3K (S3K) with respect to those reported by the process computer of the Central (SIIP), the simulation of the turbine trip transient was carried out, caused by the firing of the main generator, the low differential pressure of oil of its seals and the automatic Scram of Unit 1 of the Laguna Verde nuclear power plant, at 87% of power nominal during the operation cycle 16. Since the reactor was brought to a safe stop due to Scram, was enough to simulate 20 seconds to observe the maximum increase in pressure with S3K. In this work, the following parameters are shown and compared: the neutron flux, the thermal power, the pressure in the dome, the flow at the entrance to the core, the steam flow that leaves the vessel and the minimal critical power ratio (MCPR). The neutron flux of the average power range monitors of the nuclear power plant was compared with the S3K detectors model. Finally, the MCPR was calculated with a different correlation to that of the fuel supplier and its deviation from its safety limit was determined. In conclusion, the results obtained show the current state of the model for the simulation of reactivity transients and the opportunity areas to consolidate this tool in support of the process of licensing refueling in the CNSNS. (Author)

  17. An on-line advanced plant simulator (OLAPS)

    International Nuclear Information System (INIS)

    Samuels, J.W.

    1989-01-01

    A PC based on-line advanced plant simulator (OLAPS) for high quality simulations of Portland General Electric's Trojan Nuclear Facility is presented. OLAPS is designed to simulate the thermal-hydraulics of the primary system including core, steam generators, pumps, piping and pressurizer. The simulations are based on a five equation model that has two mass equations, two energy equations, two energy equations, and one momentum equation with a drift flux model to provide closure. A regionwise point reactor kinetics model is used to model the neutron kinetics in the core. The conservation equations, constitutive models and the numerical methods used to solve them are described. OLAPS results are compared with data from chapter 15 of the Trojan Nuclear Facility's final safety analysis report

  18. Nuclear power plant training simulator fidelity assessment

    International Nuclear Information System (INIS)

    Carter, R.J.; Laughery, K.R.

    1985-01-01

    The fidelity assessment portion of a methodology for evaluating nuclear power plant simulation facilities in regard to their appropriateness for conducting the Nuclear Regulatory Commission's operating test was described. The need for fidelity assessment, data sources, and fidelity data to be collected are addressed. Fidelity data recording, collection, and analysis are discussed. The processes for drawing conclusions from the fidelity assessment and evaluating the adequacy of the simulator control-room layout were presented. 3 refs

  19. 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)

  20. Validation of simulation codes for future systems: motivations, approach, and the role of nuclear data

    International Nuclear Information System (INIS)

    Palmiotti, G.; Salvatores, M.; Aliberti, G.

    2007-01-01

    The validation of advanced simulation tools will still play a very significant role in several areas of reactor system analysis. This is the case of reactor physics and neutronics, where nuclear data uncertainties still play a crucial role for many core and fuel cycle parameters. The present paper gives a summary of validation motivations, objectives and approach. A validation effort is in particular necessary in the frame of advanced (e.g. Generation-IV or GNEP) reactors and associated fuel cycles assessment and design. Validation of simulation codes is complementary to the 'verification' process. In fact, 'verification' addresses the question 'are we solving the equations correctly' while validation addresses the question 'are we solving the correct equations with the correct parameters'. Verification implies comparisons with 'reference' equation solutions or with analytical solutions, when they exist. Most of what is called 'numerical validation' falls in this category. Validation strategies differ according to the relative weight of the methods and of the parameters that enter into the simulation tools. Most validation is based on experiments, and the field of neutronics where a 'robust' physics description model exists and which is function of 'input' parameters not fully known, will be the focus of this paper. In fact, in the case of reactor core, shielding and fuel cycle physics the model (theory) is well established (the Boltzmann and Bateman equations) and the parameters are the nuclear cross-sections, decay data etc. Two types of validation approaches can and have been used: (a) Mock-up experiments ('global' validation): need for a very close experimental simulation of a reference configuration. Bias factors cannot be extrapolated beyond reference configuration; (b) Use of 'clean', 'representative' integral experiments ('bias factor and adjustment' method). Allows to define bias factors, uncertainties and can be used for a wide range of applications. It

  1. 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

  2. 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

  3. 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

  4. KATS experiments to simulate corium spreading in the EPR core catcher concept

    International Nuclear Information System (INIS)

    Eppinger, B.; Fieg, G.; Schuetz, W.; Stegmaier, U.

    2001-01-01

    In future Light Water Reactors special devices (core catchers) might be required to prevent containment failure by basement erosion after reactor pressure vessel melt-through during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher de-vices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent cooling by flooding with water. Therefore a series of experiments to investigate high temperature melt spreading on flat surfaces has been carried out using alumina-iron thermite melts as a simulant. The oxidic thermite melt is conditioned by adding other oxides to simulate a realistic corium melt as close as possible. Spreading of oxidic and metallic melts have been performed in one- and two-dimensional geometry. Substrates were chemically inert ceramic layers, dry concrete and concrete with a shallow water layer on top. (authors)

  5. A comprehensive review on the methodologies to simulate the nuclear fuel bundle for the thermal hydraulic experiments

    International Nuclear Information System (INIS)

    Vishnoi, A.K.; Chandraker, D.K.; Pal, A.K.; Vijayan, P.K.; Saha, D.

    2011-01-01

    The designer of a nuclear reactor system has to ensure its safety during normal operation as well as accidental conditions. This requires, among other things, a proper understanding of the various thermal hydraulic phenomena occurring in the reactor core. In a nuclear reactor core the fuel elements are the heat source and highly loaded components of the reactor system. Therefore their behaviour under normal and accidental conditions must be extensively investigated. Data generation for Critical heat flux (CHF) in full scale bundle and parallel channel instability studies with at least two full size channels are required in order to evaluate the thermal margin and stability margin of the reactor. The complex nature of these phenomena calls for exhaustive experimental investigations. Fuel Rod Cluster Simulator (FRCS) is a very important component required for the experimental investigation of the thermal hydraulic behaviour of reactor fuel elements under normal and accidental conditions. This paper brings out a comprehensive review of the FRCS elaborating the challenges and important design aspects of the FRCS. Some of the main features and analysis results on the performance of the developed FRCS with respect to the actual nuclear fuel bundle will be presented in the paper. (author)

  6. DSNP, Program and Data Library System for Dynamic Simulation of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Saphier, D.; Madell, J.; Dean, E.

    1988-01-01

    1 - Description of problem or function: DSNP (Dynamic Simulator for Nuclear Power-Plants) is a system of programs and data files by which a nuclear power plant, or part thereof, can be simulated. The acronym DSNP is used interchangeably for the DSNP language, the DSNP libraries, the DSNP pre-compiler, and the DSNP document generator. The DSNP language is a special-purpose, block-oriented digital- simulation language developed to facilitate the preparation of dynamic simulations of a large variety of nuclear power plants. It is a user-oriented language that permits the user to prepare simulation programs directly from power plant block diagrams and flow charts by recognizing the symbolic DSNP statements for the appropriate physical components and listing these statements in a logical sequence according to the flow of physical properties in the simulated power plant. Physical components of nuclear power plants are represented by functional blocks, or modules. Many of the more complex components are represented by several modules. The nuclear reactor, for example, has a kinetic module, a power distribution module, a feedback module, a thermodynamic module, a hydraulic module, and a radioactive heat decay module. These modules are stored in DSNP libraries in the form of a DSNP subroutine or function, a block of statements, a macro, or a combination of the above. Basic functional blocks such as integrators, pipes, function generators, connectors, and many auxiliary functions representing properties of materials used in nuclear power plants are also available. The DSNP pre-compiler analyzes the DSNP simulation program, performs the appropriate translations, inserts the requested modules from the library, links these modules together, searches necessary data files, and produces a simulation program in FORTRAN. FORTRAN is considered to be a subset of DSNP and can be inserted anywhere in the simulation program without restriction. I/O statements can be located anywhere in

  7. 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

  8. Application of simulation techniques for accident management training in nuclear power plants

    International Nuclear Information System (INIS)

    2003-05-01

    core. These capabilities include the optimized use of design margins as well as complementary measures for the prevention of accident progression, its monitoring, and the mitigation of severe accidents. Finally, level 5 includes off-site emergency response measures, the objective of which is to mitigate the radiological consequences of significant releases of radioactive material. Accident management is defined in the IAEA Safety Report on Development and Implementation of Accident Management Programmes in Nuclear Power Plants. The IAEA definitions are in line with the definitions of severe accident management in OECD/NEA documents as given, for example. This report describes simulation techniques used in the training of personnel involved in accident management of NPPs. This concerns both the plant personnel and the persons involved in the management of off-site releases. The report pertains to light water reactors (LWRs) and pressurized heavy water reactors (PHWRs), but it can equally be applied to power reactors of other types. The report is intended for use by experts responsible for planning, developing, executing or supervising the training of personnel involved in the implementation of AMPs in NPPs. It concentrates on existing techniques, but future prospects are also discussed. Various simulation techniques are considered, from incorporating graphical interfaces into existing severe accident codes to full-scope replica simulators. Both preventive and mitigative accident management measures, different training levels and different target personnel groups are taken into account. Based on the available information compiled worldwide, present views on the applicability of simulation techniques for the training of personnel involved in accident management are provided in this report. Apart from the introduction, this report consists of four sections and three appendices. In Section 2, specific aspects of accident management are summarized. Basic approaches in the

  9. Code for the core simulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Serrano, M.A.B.

    1978-08-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numericaly. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistence added to the film coeficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (Author) [pt

  10. The MARS simulation of the nuclear weapons preparedness LOTTA scenario

    International Nuclear Information System (INIS)

    Tovedal, H.

    2001-03-01

    The simulation method MARS, Mathematical Radiac Simulation, is primarily intended for preparedness exercises in nuclear fallout areas and simulates the ionizing radiation dose rates from fission products deposited on the ground, i.e. fallout from a nuclear weapons explosion or from a release of radioactive material from a nuclear reactor. MARS gives at any time after the fictitious explosion or reactor release the dose rates at any position in the fallout area. MARS has been used for simulation of an exercise scenario called LOTTA, designed for training and test of a radiac preparedness group in the Dept. of Nuclear Protection at FOI. The group is a member of a national preparedness organisation under the Swedish Radiation Protection Institute, SSI. This MARS application was a simulation of the entire course of events following a fictitious nuclear weapons explosion, including the fission product deposition process and the ultimate activity and dose rate distribution in the fallout area. The simulation was based on deposition and fallout prognoses worked out by FOI, using the prognosis model PELLO. This report presents a short description of the simulation of the LOTTA scenario. A more detailed presentation of the general MARS method can be found in the report 'Mathematical Radiac Simulation, MARS'

  11. 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

  12. Operator training simulator for nuclear power plant

    International Nuclear Information System (INIS)

    Shiozuka, Hiromi

    1977-01-01

    In nuclear power plants, training of the operators is important. In Japan, presently there are two training centers, one is BWR operation training center at Okuma-cho, Fukushima Prefecture, and another the nuclear power generation training center in Tsuruga City, Fukui Prefecture, where the operators of PWR nuclear power plants are trained. This report describes the BWR operation training center briefly. Operation of a nuclear power plant is divided into three stages of start-up, steady state operation, and shut down. Start-up is divided into the cold-state start-up after the shut down for prolonged period due to periodical inspection or others and the hot-state start-up from stand-by condition after the shut down for a short time. In the cold-state start-up, the correction of reactivity change and the heating-up control to avoid excessive thermal stress to the primary system components are important. The BWR operation training center offers the next three courses, namely beginner's course, retraining course and specific training course. The training period is 12 weeks and the number of trainees is eight/course in the beginner's course. The simulator was manufactured by modeling No. 3 plant of Fukushima First Nuclear Power Station, Tokyo Electric Power Co. The simulator is composed of the mimic central control panel and the digital computer. The software system comprises the monitor to supervise the whole program execution, the logic model simulating the plant interlock system and the dynamic model simulating the plant physical phenomena. (Wakatsuki, Y.)

  13. Development of Nuclear ship Engineering Simulation SYstem (NESSY)

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Kyouya, Masahiko; Takahashi, Teruo; Kobayashi, Hideo; Ochiai, Masa-aki; Hashidate, Kouji.

    1993-11-01

    NESSY has been developed for design studies of advanced marine reactors as a part of nuclear ship research and development since 1987. Engineering simulation model of the Mutsu, which is the first nuclear ship in Japan, was completed in March of 1993. In this report we concentration on detail description of softwares for Mutsu modeling. The aims of development of NESSY are as follows; (1) Assessment and confirmation on plant performance of an advanced marine reactor in each step of nuclear ship design (2) Development of abnormality diagnosis system and operator support system as a part of enhanced automization study, and study of human interface with hardware The characteristics of NESSY are the followings. (1) Total engineering simulation system simulate simultaneously ship motions, propulsion system behavior, and nuclear plant behavior under given weather and sea conditions. (2) Models based on physical theory as far as possible. (3) The simulator has high extensibility and flexibility. It is able to apply to other reactors, as the simulation model consists of the part of basic model and the part of plant data which are easy to change. After completion of Mutsu modeling, we are planning to utilize this system as one of design tools for an advanced marine reactor. (author)

  14. Introducing FACETS, the Framework Application for Core-Edge Transport Simulations

    International Nuclear Information System (INIS)

    Cary, John R.; Candy, Jeff; Cohen, Ronald H.; Krasheninnikov, Sergei I.; McCune, Douglas C.; Estep, Donald J.; Larson, Jay W.; Malony, Allen; Worley, Patrick H.; Carlsson, Johann Anders; Hakim, A.H.; Hamill, P.; Kruger, Scott E.; Muzsala, S.; Pletzer, Alexander; Shasharina, Svetlana; Wade-Stein, D.; Wang, N.; McInnes, Lois C.; Wildey, T.; Casper, T.A.; Diachin, Lori A.; Epperly, Thomas; Rognlien, T.D.; Fahey, Mark R.; Kuehn, Jeffery A.; Morris, A.; Shende, Sameer; Feibush, E.; Hammett, Gregory W.; Indireshkumar, K.; Ludescher, C.; Randerson, L.; Stotler, D.; Pigarov, A.; Bonoli, P.; Chang, C.S.; D'Ippolito, D.A.; Colella, Philip; Keyes, David E.; Bramley, R.

    2007-01-01

    The FACETS (Framework Application for Core-Edge Transport Simulations) project began in January 2007 with the goal of providing core to wall transport modeling of a tokamak fusion reactor. This involves coupling previously separate computations for the core, edge, and wall regions. Such a coupling is primarily through connection regions of lower dimensionality. The project has started developing a component-based coupling framework to bring together models for each of these regions. In the first year, the core model will be a 1 dimensional model (1D transport across flux surfaces coupled to a 2D equilibrium) with fixed equilibrium. The initial edge model will be the fluid model, UEDGE, but inclusion of kinetic models is planned for the out years. The project also has an embedded Scientific Application Partnership that is examining embedding a full-scale turbulence model for obtaining the crosssurface fluxes into a core transport code.

  15. Nuclear Power Plant Simulation Game.

    Science.gov (United States)

    Weiss, Fran

    1979-01-01

    Presents a nuclear power plant simulation game which is designed to involve a class of 30 junior or senior high school students. Scientific, ecological, and social issues covered in the game are also presented. (HM)

  16. 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)

  17. 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

  18. Autonomous dynamic decision making in a nuclear fuel cycle simulator

    International Nuclear Information System (INIS)

    Pelakauskas, Martynas; Auzans, Aris; Schneider, Erich A.; Tkaczyk, Alan H.

    2013-01-01

    Highlights: • Objective criteria based decision making in a nuclear fuel cycle simulator. • Simulation driven by an evolving performance metric. • Implementation of the model in a nuclear fuel cycle simulator. • Verification of dynamic decision making based on uranium price evolution. -- Abstract: Growing energy demand and the push to move toward carbon-free ways of electricity generation have renewed the world's interest in nuclear energy. Due to the high technical and economic uncertainties related to nuclear energy, simulation tools have become a necessity in order to plan and evaluate possible nuclear fuel cycles (NFCs). Most of the NFC simulators today work by running the simulation with a user-defined set of facility build orders and preferences. While this allows for a simple way to change the simulation conditions, it may not always lead to optimal results and strongly relies on the user defining the correct parameters. This study looks into the possibility of using the expected cost of electricity (CoE) as the driving build decision variable instead of relying on user-defined build orders. This is a first step toward a more general decision making strategy in dynamic fuel cycle simulation. For this purpose, additional modules were implemented in an NFC simulator, VEGAS, with the consumption dependent price of uranium as a time-varying NFC cost component that drives the cost competitiveness of available NFC options. The model was demonstrated to verify the correct operation of a CoE-driven NFC simulator

  19. 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)

  20. 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

  1. Development and verification of a high performance multi-group SP3 transport capability in the ARTEMIS core simulator

    International Nuclear Information System (INIS)

    Van Geemert, Rene

    2008-01-01

    For satisfaction of future global customer needs, dedicated efforts are being coordinated internationally and pursued continuously at AREVA NP. The currently ongoing CONVERGENCE project is committed to the development of the ARCADIA R next generation core simulation software package. ARCADIA R will be put to global use by all AREVA NP business regions, for the entire spectrum of core design processes, licensing computations and safety studies. As part of the currently ongoing trend towards more sophisticated neutronics methodologies, an SP 3 nodal transport concept has been developed for ARTEMIS which is the steady-state and transient core simulation part of ARCADIA R . For enabling a high computational performance, the SP N calculations are accelerated by applying multi-level coarse mesh re-balancing. In the current implementation, SP 3 is about 1.4 times as expensive computationally as SP 1 (diffusion). The developed SP 3 solution concept is foreseen as the future computational workhorse for many-group 3D pin-by-pin full core computations by ARCADIA R . With the entire numerical workload being highly parallelizable through domain decomposition techniques, associated CPU-time requirements that adhere to the efficiency needs in the nuclear industry can be expected to become feasible in the near future. The accuracy enhancement obtainable by using SP 3 instead of SP 1 has been verified by a detailed comparison of ARTEMIS 16-group pin-by-pin SP N results with KAERI's DeCart reference results for the 2D pin-by-pin Purdue UO 2 /MOX benchmark. This article presents the accuracy enhancement verification and quantifies the achieved ARTEMIS-SP 3 computational performance for a number of 2D and 3D multi-group and multi-box (up to pin-by-pin) core computations. (authors)

  2. 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)

  3. NUCLEBRAS' experience in the implantation of a nuclear power plants simulator

    International Nuclear Information System (INIS)

    Spitalnik, J.; Fonseca, G.

    1985-01-01

    The experiences gained by NUCLEBRAS in the setting up of a Nuclear Power Plant Simulator Training Center, cover the design and manufacture of the simulator for the Angra-2 type nuclear power plants, the training of the simulator operation and maintenance personnel, in preparation for the training of the Brazilian nuclear power plant operators, the development of the simulator training programs and materials, the temporary installation and utilization of the simulator in the FRG, including the training of operators of the Trillo nuclear power plant, in Spain. The simulator shall be finally installed in Brazil, at the vicinity of the Angra-2 site, in 1985, when it is foreseen to start the training of the Angra-2 operators. (Author) [pt

  4. 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

  5. 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.

  6. 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.

  7. Validation of SIMULATE-3K for stability analysis of Laguna Verde nuclear plant

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, Rogelio, E-mail: rogelio.castillo@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Alonso, Gustavo, E-mail: gustavo.alonso@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Instituto Politecnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ramírez, J. Ramón, E-mail: ramon.ramirez@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico)

    2013-12-15

    Highlights: • Neutronic/thermal hydraulic event in Laguna Verde is modeled. • A good agreement is obtained between SIMULATE-3K results and data plant for frequency and DR. • Other noise analysis techniques are used for the same purpose with good agreement. • Validation of SIMULATE-3K for stability analysis of Laguna Verde is confirmed - Abstract: Boiling Water Reactors are two phase flow systems which are susceptible to different types of flow instabilities. Among these are the coupled neutronic/thermal-hydraulic instabilities, these may compromise established fuel safety limits. These instabilities are characterized by periodic core-power and hydraulic oscillations. SIMULATE-3K code has been tested for stability analysis for several benchmarks, however to qualify the SIMULATE-3K code for a particular power plant a specific reactor plant analysis must be done. In this paper, the plant model of Laguna Verde Nuclear Power Plant is built and SIMULATE-3K is tested against the 1995 coupled neutronic/thermal-hydraulic instability event of Laguna Verde. Results obtained show the adequacy of this code to specific Laguna Verde power plant stability analysis.

  8. An Embalse nuclear power plant basic principles simulator

    International Nuclear Information System (INIS)

    Fernandez, Oscar; Galdoz, Erwin; Flury, Celso; Fontanini, Horacio

    1989-01-01

    A nuclear power plant basic principles simulator is a package of programs that numerically solve the dynamic equations of the simulated plant. This kind of tools is mainly used in the first step of training of operational personnel, to allow mental representation of physical phenomena governing the plant. They are also used for students or professional training, and experienced operators can also improve there performance under abnormal operation situations using the simulator. For the Embalse nuclear power plant, mainly the thermohydraulic behaviour, is simulated. The mathematical model was adapted from MANUVR, a code developed at the Electric Systems and Control Department of the National Atomic Energy Commission (CNEA). (Author) [es

  9. 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.

  10. 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.

  11. 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)

  12. 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.)

  13. 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

  14. Advanced nuclear reactors and their simulators

    International Nuclear Information System (INIS)

    Chaushevski, Anton; Boshevski, Tome

    2003-01-01

    Population growth, economy development and improvement life standard impact on continually energy needs as well as electricity. Fossil fuels have limited reserves, instability market prices and destroying environmental impacts. The hydro energy capacities highly depend on geographic and climate conditions. The nuclear fission is significant factor for covering electricity needs in this century. Reasonable capital costs, low fuel and operating expenses, environmental acceptable are some of the facts that makes the nuclear energy an attractive option especially for the developing countries. The simulators for nuclear reactors are an additional software tool in order to understand, study research and analyze the processes in nuclear reactors. (Original)

  15. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    Science.gov (United States)

    Emrich, William J. Jr.; Moran, Robert P.; Pearson, J. Boise

    2012-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities

  16. 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)

  17. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo alta presion (HPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, D.; Chavez M, C., E-mail: danmirnyi@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  18. 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)

  19. A new coupling kernel for the three-dimensional simulation of a boiling water reactor core by the nodal coupling method

    International Nuclear Information System (INIS)

    Gupta, N.K.

    1981-01-01

    A new coupling kernel is developed for the three-dimensional (3-D) simulation of Boiling Water Reactors (BWR's) by the nodal coupling method. The new kernel depends not only on the properties of the node under consideration but also on the properties of its neighbouring nodes. This makes the kernel more useful in particular for fuel bundles lying in a surrounding of different nuclear characteristics, e.g. for a controlled bundle in the surrounding of uncontrolled bundles or vice-versa. The main parameter in the new kernel is a space-dependent factor obtained from the ratio of thermal-to-fast flux. The average value of the above ratio for each node is evaluated analytically. The kernel is incorporated in a 3-D BWR core simulation program MOGS. As an experimental verification of the model, the cycle-6 operations of the two units of the Tarapur Atomic Power Station (TAPS) are simulated and the result of the simulation are compared with Travelling Incore Probe (TIP) data. (orig.)

  20. 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.

  1. 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

  2. An approach for coupled-code multiphysics core simulations from a common input

    International Nuclear Information System (INIS)

    Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger; Clarno, Kevin; Simunovic, Srdjan; Slattery, Stuart; Turner, John; Palmtag, Scott

    2015-01-01

    Highlights: • We describe an approach for coupled-code multiphysics reactor core simulations. • The approach can enable tight coupling of distinct physics codes with a common input. • Multi-code multiphysics coupling and parallel data transfer issues are explained. • The common input approach and how the information is processed is described. • Capabilities are demonstrated on an eigenvalue and power distribution calculation. - Abstract: This paper describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak

  3. Teaching simulator for divulgation of the nuclear energy; Simulador docente para divulgacion de la energia nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Ortega B, M.G.; Gutierrez F, R. [FI-UNAM, DEPFI Campus Morelos (Mexico)] e-mail: cchavez2@cableonline.com.mx

    2003-07-01

    To solicitude of the authorities of the 'Universum' sciences museum of the UNAM, it develops a highly interactive computational system, to provide of information to the population in general about basic principles, uses and benefits of the nuclear energy. The objective is to achieve a better understanding and acceptance of the nuclear technology in our country. The system allows the visualization and simulation of nuclear processes as well as of its applications. The system is divided in three levels: basic, intermediate and simulation. In the basic level multimedia information is included on diverse basic concepts of the nuclear energy. The intermediate level includes the description and operation of some systems of the Laguna Verde nuclear power plant (CNLV). Finally the simulation level contains representative scenarios that the user can control by means of virtual control panels of the main systems of the CNLV. Inside the system a part of interactive games is included with the purpose that the user remembers with more easiness all the concepts and advantages of the nuclear energy mentioned during the previous levels. The system contributes, by means of the development of multimedia computational tools and of simulation, to the popularization of the use and applications of the nuclear energy in Mexico. (Author)

  4. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  5. Part-task simulator for a WWER-440 nuclear power plant subsystem

    International Nuclear Information System (INIS)

    Szabo, B.K.

    1988-07-01

    PC-based part-task simulators for simulating subsystems of nuclear power plants are low cost tools for operator training. In the Central Research Institute for Physics, Budapest, a simulator system has been developed to facilitate fast development of such simulators. The first application simulates the Neutron Flux Monitoring System of WWER-440 nuclear power plants. (author) 9 refs.; 2 figs

  6. Simulation analysis on accident at Fukushima Daiichi Nuclear Power Plant Unit 2 by SAMPSON code

    International Nuclear Information System (INIS)

    Takahashi, Atsuo; Pellegrini, Marco; Mizouchi, Hideo; Suzuki, Hiroaki; Naitoh, Masanori

    2015-01-01

    The accident occurred at the Fukushima Daiichi Nuclear Power Plant Unit 2 has been investigated by the severe accident analysis code, SAMPSON with more realistic boundary conditions and newly introduced models. In Unit 2, the Reactor Core Isolation Cooling system (RCIC) is thought to have worked for unexpectedly long time (about 70 hours) without batteries. It is thought to be due to balance between injected water from the RCIC pump and supplied mixture of steam and water to the RCIC turbine. To confirm the RCIC working condition and reproduce the measured plant properties, such as pressure and water level in the reactor pressure vessel (RPV), we introduced two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by RCIC was calculated through mass flow rate of steam included in extracted two-phase flow, steam generated from flashing of water included in extracted two-phase flow, and turbine efficiency degradation originated by the mixture of steam and water flowing to the RCIC turbine. To reproduce the dry well (DW) pressure, we assumed that torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Simulation results by SAMPSON basically agree with the measured values such as pressure in the RPV and in the DW until several days after the scram. However, some contradictions between the simulation results and the measured values, such as that inversion of the RPV pressure at 10 hours after scram in the measurement happened at 14 hours in the simulation and that the DW pressure showed different behavior between simulation and measurement when SRV started periodic operation at 71 hours, are still remain and are under consideration. In the current calculation, model for falling core to the lower plenum was modified so that debris is not retained at the core plate based on observation of the XR2-1 experiment. Additionally, model of the RPV failure by melting of the penetrating pipe

  7. Simulators for training nuclear power plant personnel

    International Nuclear Information System (INIS)

    1993-01-01

    Simulator training and retraining of operations personnel is essential for their acquiring the necessary knowledge, skills and qualification for operating a nuclear power plant, and for effective feedback of experience including human based operating errors. Simulator training is the most effective way by far of training operations personnel in co-operation and communication in a team, which also involves instilling attitudes and approaches for achieving excellence and individual responsibility and alertness. This technical document provides guidance to Member States on the procurement, setting up and utilization of a simulator training centre; it will also be useful for organizations with previous experience in the use of simulators for training. The document is the result of a series of advisory and consultants meetings held in the framework of the International Working Group on Nuclear Power Plant Control and Instrumentation in 1989-1992. 17 refs, 2 tabs

  8. Interactive real-time nuclear plant simulations on a UNIX based supercomputer

    International Nuclear Information System (INIS)

    Behling, S.R.

    1990-01-01

    Interactive real-time nuclear plant simulations are critically important to train nuclear power plant engineers and operators. In addition, real-time simulations can be used to test the validity and timing of plant technical specifications and operational procedures. To accurately and confidently simulate a nuclear power plant transient in real-time, sufficient computer resources must be available. Since some important transients cannot be simulated using preprogrammed responses or non-physical models, commonly used simulation techniques may not be adequate. However, the power of a supercomputer allows one to accurately calculate the behavior of nuclear power plants even during very complex transients. Many of these transients can be calculated in real-time or quicker on the fastest supercomputers. The concept of running interactive real-time nuclear power plant transients on a supercomputer has been tested. This paper describes the architecture of the simulation program, the techniques used to establish real-time synchronization, and other issues related to the use of supercomputers in a new and potentially very important area. (author)

  9. Memory bottlenecks and memory contention in multi-core Monte Carlo transport codes

    International Nuclear Information System (INIS)

    Tramm, J.R.; Siegel, A.R.

    2013-01-01

    The simulation of whole nuclear cores through the use of Monte Carlo codes requires an impracticably long time-to-solution. We have extracted a kernel that executes only the most computationally expensive steps of the Monte Carlo particle transport algorithm - the calculation of macroscopic cross sections - in an effort to expose bottlenecks within multi-core, shared memory architectures. (authors)

  10. 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.

  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. DSNP: a new approach to simulate nuclear power plants

    International Nuclear Information System (INIS)

    Saphier, D.

    1977-01-01

    The DSNP (Dynamic Simulator for Nuclear Power-plants) is a special purpose block oriented simulation language. It provides for simulations of a large variety of nuclear power plants or various parts of the power plant in a simple straightforward manner. The system is composed of five basic elements, namely, the DSNP language, the precompiler-or the DSNP language translator, the components library, the document generator, and the system data files. The DSNP library of modules includes the selfcontained models of components or physical processes found in a nuclear power plant, and various auxiliary modules such as material properties, control modules, integration schemes, various basic transfer functions etc. In its final form DSNP will have four libraries

  13. Common modelling approaches for training simulators for nuclear power plants

    International Nuclear Information System (INIS)

    1990-02-01

    Training simulators for nuclear power plant operating staff have gained increasing importance over the last twenty years. One of the recommendations of the 1983 IAEA Specialists' Meeting on Nuclear Power Plant Training Simulators in Helsinki was to organize a Co-ordinated Research Programme (CRP) on some aspects of training simulators. The goal statement was: ''To establish and maintain a common approach to modelling for nuclear training simulators based on defined training requirements''. Before adapting this goal statement, the participants considered many alternatives for defining the common aspects of training simulator models, such as the programming language used, the nature of the simulator computer system, the size of the simulation computers, the scope of simulation. The participants agreed that it was the training requirements that defined the need for a simulator, the scope of models and hence the type of computer complex that was required, the criteria for fidelity and verification, and was therefore the most appropriate basis for the commonality of modelling approaches. It should be noted that the Co-ordinated Research Programme was restricted, for a variety of reasons, to consider only a few aspects of training simulators. This report reflects these limitations, and covers only the topics considered within the scope of the programme. The information in this document is intended as an aid for operating organizations to identify possible modelling approaches for training simulators for nuclear power plants. 33 refs

  14. Scientific codes developed and used at GRS. Nuclear simulation chain

    Energy Technology Data Exchange (ETDEWEB)

    Schaffrath, Andreas; Sonnenkalb, Martin; Sievers, Juergen; Luther, Wolfgang; Velkov, Kiril [Gesellschaft fuer Anlagen und Reaktorsicherheit (GRS) gGmbH, Garching/Muenchen (Germany). Forschungszentrum

    2016-05-15

    Over 60 technical experts of the reactor safety research division of the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH are developing and validating reliable methods and computer codes - summarized under the term nuclear simulation chain - for the safety-related assessment for all types of nuclear power plants (NPP) and other nuclear facilities considering the current state of science and technology. This nuclear simulation chain has to be able to simulate and assess all relevant physical processes and phenomena for all operating states and (severe) accidents. In the present contribution, the nuclear simulation chain developed and applied by GRS as well as selected examples of its application are presented. The latter demonstrate impressively the width of its scope and its performance. The GRS codes can be passed on request to other (national as well as international) organizations. This contributes to a worldwide increase of the nuclear safety standards. The code transfer is especially important for developing and emerging countries lacking the financial means and/or the necessary know-how for this purpose. At the end of this contribution, the respective course of action is described.

  15. Dynamic model for the control system simulation and design of a 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Zhang Yuai; Liu Longzhi; Ma Changwen

    1999-01-01

    The author develops a nonlinear dynamic model used in a wide range control system simulation for a 200 MW Nuclear Heating Reactor (NHR-200). Besides a one-point neutron kinetics equation and temperature feedback based on the lumped fuel and coolant temperature, which are the usual methods used in modeling of PWR, two other factors are also considered in order to suit the wide range operation. The first consideration is the natural circulation in the primary loop because it affects the heat transfer coefficients in the core and in the primary heat exchanger (PHE). The second consideration is the flow rate variation in the secondary loop which leads to some nonlinear properties. The simulation results show that the model is accurate enough for control system simulation. Some model reduction basis can be obtained through the dynamic analysis

  16. 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)

  17. Experimental investigation of the vibration response of a flexible tube due to simulated reactor core, cross and annular exit flows

    International Nuclear Information System (INIS)

    Haslinger, K.H.; Martin, M.L.; Higgins, W.H.; Rossano, F.V.

    1989-01-01

    Instrumentation tubes in pressurized nuclear reactors have experienced wear due to excessive flow-induced vibrations. Experiments to identify the predominant flow excitation mechanism at a particular plant, and to develop a sleeve design to remedy the wear problem are reported. An instrumented flow visualization model enabled simulation of a wide range of individual or combined reactor core flow, cross flow and thimble flow conditions. The instrumentation scheme adopted for these experiments used proximity displacement transducers and a force transducer to measure respectively tube motion and contact/impact forces at the wear region. Extensive testing of the original, in-plant configuration identified the normal core flow as the primary source of excitation. Shielding the In-Core-Instrumentation thimble tube from the normal core flow curtailed vibration amplitudes; however, thimble flow excitation then became more pronounced. Various outlet nozzle configurations were investigated. An internal cavity combined with radial outlet slots became the optimum solution for the problem. The paper presents typical test data in the form of orbital tube motion, spectrum analysis and time history collages. The effectiveness of shielding the instrumentation tube from the flow is demonstrated. (author)

  18. Closure simulation of the MSIV of Unit 1 of the Laguna Verde nuclear power plant using the Simulate 3K code

    International Nuclear Information System (INIS)

    Alegria A, A.

    2015-09-01

    In this paper the simulation of closure transient of all main steam isolation valves (MSIV) was performed with the Simulate-3K (S-3K) code for the Unit 1 of the Laguna Verde nuclear power plant (NPP-LV), which operates to thermal power of 2317 MWt, corresponding to the cycle 15 of operation. The set points for the performance of systems correspond to those set out in transient analysis: 3 seconds for the closure of all MSIV; the start of Scram when 121% of the neutron flux is reached, respect from baseline before the transient; the opening by peer of safety relief valves (SRV) in relief mode when the set point of the pressure is reached, the shoot of the feedwater flow seconds after the start of closing of the MSIV and the shoot of the recirculation water pumps when the pressure is reached in the dome of 1048 psig. The simulation time was of 57 seconds, with the top 50 to reach the steady state, from which the closure of all MSIV starts. In this paper the behavior of the pressure in the dome are analyzed, thermal power, neutron flux, the collapsed water level, the flow at the entrance of core, the steam flow coming out of vessel and the flow through of the SRV; the fuel temperature, the minimal critical power ratio, the readings in the instrumentation systems and reactivities. Instrumentation systems were implemented to analyze the neutron flux, these consist of 96 local power range monitors (LPRM) located in different radial and axial positions of the core and 4 channels of average power range monitors, which grouped at 24 LPRM each one. LPRM response to the change of neutron flux in the center of the core, at different axial positions is also shown. Finally, the results show that the safety limit MCPR is not exceeded. (Author)

  19. Programming for a nuclear reactor instrument simulation

    International Nuclear Information System (INIS)

    Cohn, C.

    1988-01-01

    This note discusses 8086/8087 machine-language programming for simulation of nuclear reactor instrument current inputs by means of a digital-analog converter (DAC) feeding a bank of series input resistors. It also shows FORTRAN programming for generating the parameter tales used in the simulation. These techniques would be generally useful for high-speed simulation of quantities varying over many orders of magnitude

  20. BWR nuclear plant maintenance simulation

    International Nuclear Information System (INIS)

    Stuart, I.F.

    1985-01-01

    As early as 1977, the General Electric Company, USA, Nuclear Energy Operation was making plans to construct a maintenance-type simulator to support Training and Services. The Company's pioneering experience with control room simulators started in 1968 with the Dresden simulator and showed clearly the benefits of having such facilities for training, checkout of procedures and, in the case of maintenance, match-up of equipment or tools as needed. Since the dedication of the facility, it has proved to be an invaluable resource in the training of refuelling and servicing crews. The facility has also been extensively used as developmental and test facility for in-vessel servicing equipment and procedures. (author)

  1. THREE-DIMENSIONAL CORE-COLLAPSE SUPERNOVA SIMULATED USING A 15 M{sub ⊙} PROGENITOR

    Energy Technology Data Exchange (ETDEWEB)

    Lentz, Eric J.; Mezzacappa, Anthony; Harris, J. Austin; Yakunin, Konstantin N. [Department of Physics and Astronomy, University of Tennessee, Knoxville, TN 37996-1200 (United States); Bruenn, Stephen W. [Department of Physics, Florida Atlantic University, 777 Glades Road, Boca Raton, FL 33431-0991 (United States); Hix, W. Raphael [Physics Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6354 (United States); Messer, O. E. Bronson [National Center for Computational Sciences, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6164 (United States); Endeve, Eirik [Computer Science and Mathematics Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6164 (United States); Blondin, John M. [Department of Physics, North Carolina State University, Raleigh, NC 27695-8202 (United States); Marronetti, Pedro, E-mail: elentz@utk.edu [Physics Division, National Science Foundation, Arlington, VA 22207 (United States)

    2015-07-10

    We have performed ab initio neutrino radiation hydrodynamics simulations in three and two spatial dimensions (3D and 2D) of core-collapse supernovae from the same 15 M{sub ☉} progenitor through 440 ms after core bounce. Both 3D and 2D models achieve explosions; however, the onset of explosion (shock revival) is delayed by ∼100 ms in 3D relative to the 2D counterpart and the growth of the diagnostic explosion energy is slower. This is consistent with previously reported 3D simulations utilizing iron-core progenitors with dense mantles. In the ∼100 ms before the onset of explosion, diagnostics of neutrino heating and turbulent kinetic energy favor earlier explosion in 2D. During the delay, the angular scale of convective plumes reaching the shock surface grows and explosion in 3D is ultimately lead by a single, large-angle plume, giving the expanding shock a directional orientation not dissimilar from those imposed by axial symmetry in 2D simulations. We posit that shock revival and explosion in the 3D simulation may be delayed until sufficiently large plumes form, whereas such plumes form more rapidly in 2D, permitting earlier explosions.

  2. 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

  3. 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)

  4. Parallelization and automatic data distribution for nuclear reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Liebrock, L.M. [Liebrock-Hicks Research, Calumet, MI (United States)

    1997-07-01

    Detailed attempts at realistic nuclear reactor simulations currently take many times real time to execute on high performance workstations. Even the fastest sequential machine can not run these simulations fast enough to ensure that the best corrective measure is used during a nuclear accident to prevent a minor malfunction from becoming a major catastrophe. Since sequential computers have nearly reached the speed of light barrier, these simulations will have to be run in parallel to make significant improvements in speed. In physical reactor plants, parallelism abounds. Fluids flow, controls change, and reactions occur in parallel with only adjacent components directly affecting each other. These do not occur in the sequentialized manner, with global instantaneous effects, that is often used in simulators. Development of parallel algorithms that more closely approximate the real-world operation of a reactor may, in addition to speeding up the simulations, actually improve the accuracy and reliability of the predictions generated. Three types of parallel architecture (shared memory machines, distributed memory multicomputers, and distributed networks) are briefly reviewed as targets for parallelization of nuclear reactor simulation. Various parallelization models (loop-based model, shared memory model, functional model, data parallel model, and a combined functional and data parallel model) are discussed along with their advantages and disadvantages for nuclear reactor simulation. A variety of tools are introduced for each of the models. Emphasis is placed on the data parallel model as the primary focus for two-phase flow simulation. Tools to support data parallel programming for multiple component applications and special parallelization considerations are also discussed.

  5. Parallelization and automatic data distribution for nuclear reactor simulations

    International Nuclear Information System (INIS)

    Liebrock, L.M.

    1997-01-01

    Detailed attempts at realistic nuclear reactor simulations currently take many times real time to execute on high performance workstations. Even the fastest sequential machine can not run these simulations fast enough to ensure that the best corrective measure is used during a nuclear accident to prevent a minor malfunction from becoming a major catastrophe. Since sequential computers have nearly reached the speed of light barrier, these simulations will have to be run in parallel to make significant improvements in speed. In physical reactor plants, parallelism abounds. Fluids flow, controls change, and reactions occur in parallel with only adjacent components directly affecting each other. These do not occur in the sequentialized manner, with global instantaneous effects, that is often used in simulators. Development of parallel algorithms that more closely approximate the real-world operation of a reactor may, in addition to speeding up the simulations, actually improve the accuracy and reliability of the predictions generated. Three types of parallel architecture (shared memory machines, distributed memory multicomputers, and distributed networks) are briefly reviewed as targets for parallelization of nuclear reactor simulation. Various parallelization models (loop-based model, shared memory model, functional model, data parallel model, and a combined functional and data parallel model) are discussed along with their advantages and disadvantages for nuclear reactor simulation. A variety of tools are introduced for each of the models. Emphasis is placed on the data parallel model as the primary focus for two-phase flow simulation. Tools to support data parallel programming for multiple component applications and special parallelization considerations are also discussed

  6. Constraining the supersaturation density equation of state from core-collapse supernova simulations? Excluded volume extension of the baryons

    International Nuclear Information System (INIS)

    Fischer, Tobias

    2016-01-01

    In this article the role of the supersaturation density equation of state (EOS) is explored in simulations of failed core-collapse supernova explosions. Therefore the nuclear EOS is extended via a one-parameter excluded-volume description for baryons, taking into account their finite and increasing volume with increasing density in excess of saturation density. Parameters are selected such that the resulting supernova EOS represent extreme cases, with high pressure variations at supersaturation density which feature extreme stiff and soft EOS variants of the reference case, i.e. without excluded-volume corrections. Unlike in the interior of neutron stars with central densities in excess of several times saturation density, central densities of core-collapse supernovae reach only slightly above saturation density. Hence, the impact of the supersaturation density EOS on the supernova dynamics as well as the neutrino signal is found to be negligible. It is mainly determined from the low- and intermediate-density domain, which is left unmodified within this generalized excluded volume approach. (orig.)

  7. Angra 1 nuclear power plant full scope simulator development project

    Energy Technology Data Exchange (ETDEWEB)

    Selvatici, Edmundo; Castanheira, Luiz Carlos C.; Silva Junior, Nilo Garcia da, E-mail: edsel@eletronuclear.gov.br, E-mail: lccast@eletronuclear.gov.br, E-mail: nilogar@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (SCO/ELETRONUCLEAR), Angra dos Reis, RJ (Brazil). Superintendencia de Coordenacao da Operacao; Zazo, Francisco Javier Lopez; Ruiz, Jose Antonio, E-mail: jlopez@tecnatom.es, E-mail: jaruiz@tecnatom.es [Tecnatom S.A., San Sebastian de los Reyes, Madrid (Spain)

    2015-07-01

    Specific Full Scope Simulators are an essential tool for training NPP control room operators, in the formation phase as well as for maintaining their qualifications. In the last years availability of a Plant specific simulator has also become a Regulator requirement for Nuclear Power Plant operation. By providing real-time practical training for the operators, the use of a simulator allows improving the operator's performance, reducing the number of unplanned shutdowns and more effective response to abnormal and emergency operating conditions. It can also be used, among other uses, to validate procedures, test proposed plant modifications, perform engineering studies and to provide operation training for the technical support staff of the plant. The NPP site, in Angra dos Reis-RJ, Brazil, comprises the two units in operation, Unit 1, 640 MWe, Westinghouse PWR and Unit 2, 1350 MWe, KWU/Areva PWR and one unit in construction, Unit 3, 1405 MWe, KWU/Areva PWR, of the same design of Angra 2. Angra 2 has had its full scope simulator from the beginning, however this was not the case of Angra 1, that had to train its operators abroad, due to lack of a specific simulator. Eletronuclear participated in all the phases of the project, from data supply to commissioning and validation. The Angra 1 full scope simulator encompasses more than 80 systems of the plant including the Primary system, reactor core and associated auxiliary systems, the secondary system and turbo generator as well as all the Plant operational and safety I and C. The Angra 1 Main Control Room panels were reproduced in the simulator control room as well as the remote shutdown panels that are outside the control room. This paper describes the project for development of the Angra 1 NPP Full Scope Simulator, supplied by Tecnatom S.A., in the period of Feb.2012 to Feb.2015. (author)

  8. Angra 1 nuclear power plant full scope simulator development project

    International Nuclear Information System (INIS)

    Selvatici, Edmundo; Castanheira, Luiz Carlos C.; Silva Junior, Nilo Garcia da

    2015-01-01

    Specific Full Scope Simulators are an essential tool for training NPP control room operators, in the formation phase as well as for maintaining their qualifications. In the last years availability of a Plant specific simulator has also become a Regulator requirement for Nuclear Power Plant operation. By providing real-time practical training for the operators, the use of a simulator allows improving the operator's performance, reducing the number of unplanned shutdowns and more effective response to abnormal and emergency operating conditions. It can also be used, among other uses, to validate procedures, test proposed plant modifications, perform engineering studies and to provide operation training for the technical support staff of the plant. The NPP site, in Angra dos Reis-RJ, Brazil, comprises the two units in operation, Unit 1, 640 MWe, Westinghouse PWR and Unit 2, 1350 MWe, KWU/Areva PWR and one unit in construction, Unit 3, 1405 MWe, KWU/Areva PWR, of the same design of Angra 2. Angra 2 has had its full scope simulator from the beginning, however this was not the case of Angra 1, that had to train its operators abroad, due to lack of a specific simulator. Eletronuclear participated in all the phases of the project, from data supply to commissioning and validation. The Angra 1 full scope simulator encompasses more than 80 systems of the plant including the Primary system, reactor core and associated auxiliary systems, the secondary system and turbo generator as well as all the Plant operational and safety I and C. The Angra 1 Main Control Room panels were reproduced in the simulator control room as well as the remote shutdown panels that are outside the control room. This paper describes the project for development of the Angra 1 NPP Full Scope Simulator, supplied by Tecnatom S.A., in the period of Feb.2012 to Feb.2015. (author)

  9. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Phase II Upgrade Activities

    Science.gov (United States)

    Emrich, William J.; Moran, Robert P.; Pearson, J. Bose

    2013-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities. Keywords: Nuclear Thermal Propulsion, Simulator

  10. Nuclear power plant simulation facility evaluation methodology

    International Nuclear Information System (INIS)

    Haas, P.M.; Carter, R.J.; Laughery, K.R. Jr.

    1985-01-01

    A methodology for evaluation of nuclear power plant simulation facilities with regard to their acceptability for use in the US Nuclear Regulatory Commission (NRC) operator licensing exam is described. The evaluation is based primarily on simulator fidelity, but incorporates some aspects of direct operator/trainee performance measurement. The panel presentation and paper discuss data requirements, data collection, data analysis and criteria for conclusions regarding the fidelity evaluation, and summarize the proposed use of direct performance measurment. While field testing and refinement of the methodology are recommended, this initial effort provides a firm basis for NRC to fully develop the necessary methodology

  11. Neutronic simulation of a research reactor core of (232Th, 235U)O2 fuel using MCNPX2.6 code

    International Nuclear Information System (INIS)

    Feghhi, Seyed Amir Hossein; Rezazadeh, Marzieh; Kadi, Yacine; ); Tenreiro, Claudio; Aref, Morteza; Gholamzadeh, Zohreh

    2013-01-01

    The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can he adopted because a high fissile production rate of 233 U converted from 232 Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and 2.2 % enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core. (author)

  12. Computational simulation of Argonauta/IEN nuclear reactor using MCNPX code

    International Nuclear Information System (INIS)

    Cunha, Victor Lusis Lassance; Silva Junior, Wilson F. Rebello da

    2011-01-01

    The study consisted of developing a computer simulation of a nuclear research reactor using the MCNPX. The reactor modeled is the Argonauta located at IEN (Rio de Janeiro) designed by Argonne National Laboratory (USA), which is primarily used for non-destructive testing with neutron beam and teaching purposes. It was entirely modeled with geometric fidelity, including detailed material description, shielding and irradiation channels. When available, the model was based on the as-built drawings. Four different simulations were made, the first set of two for criticality calculations and the other set for flux measurement. The first simulation set consisted of estimating the reactors reactivity. The second set consisted of placing detectors on specific places where the reactor is monitored and on the fuel axis covering the multiplicative and non-multiplicative media. Based on this data, the thermal neutron flux profile was plotted. All the outputs were compared with experimental data. Since it is a stochastic method, the statistical convergence was successfully checked for all simulations. The results were in good agreement with the experimental values. For the criticality calculations, the relative error was smaller then 1%. The flux measurements were also very well reproduced. The values were normalized for a reference point and the proportionality between the different spots was respected. The neutron flux profile along the core had the expected shape and values. Based on the good results, it can be said that the model is validated. (author)

  13. Understanding the core-halo relation of quantum wave dark matter from 3D simulations.

    Science.gov (United States)

    Schive, Hsi-Yu; Liao, Ming-Hsuan; Woo, Tak-Pong; Wong, Shing-Kwong; Chiueh, Tzihong; Broadhurst, Tom; Hwang, W-Y Pauchy

    2014-12-31

    We examine the nonlinear structure of gravitationally collapsed objects that form in our simulations of wavelike cold dark matter, described by the Schrödinger-Poisson (SP) equation with a particle mass ∼10(-22)  eV. A distinct gravitationally self-bound solitonic core is found at the center of every halo, with a profile quite different from cores modeled in the warm or self-interacting dark matter scenarios. Furthermore, we show that each solitonic core is surrounded by an extended halo composed of large fluctuating dark matter granules which modulate the halo density on a scale comparable to the diameter of the solitonic core. The scaling symmetry of the SP equation and the uncertainty principle tightly relate the core mass to the halo specific energy, which, in the context of cosmological structure formation, leads to a simple scaling between core mass (Mc) and halo mass (Mh), Mc∝a(-1/2)Mh(1/3), where a is the cosmic scale factor. We verify this scaling relation by (i) examining the internal structure of a statistical sample of virialized halos that form in our 3D cosmological simulations and by (ii) merging multiple solitons to create individual virialized objects. Sufficient simulation resolution is achieved by adaptive mesh refinement and graphic processing units acceleration. From this scaling relation, present dwarf satellite galaxies are predicted to have kiloparsec-sized cores and a minimum mass of ∼10(8)M⊙, capable of solving the small-scale controversies in the cold dark matter model. Moreover, galaxies of 2×10(12)M⊙ at z=8 should have massive solitonic cores of ∼2×10(9)M⊙ within ∼60  pc. Such cores can provide a favorable local environment for funneling the gas that leads to the prompt formation of early stellar spheroids and quasars.

  14. 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

  15. Turbulent geodynamo simulations: a leap towards Earth's core

    Science.gov (United States)

    Schaeffer, N.; Jault, D.; Nataf, H.-C.; Fournier, A.

    2017-10-01

    We present an attempt to reach realistic turbulent regime in direct numerical simulations of the geodynamo. We rely on a sequence of three convection-driven simulations in a rapidly rotating spherical shell. The most extreme case reaches towards the Earth's core regime by lowering viscosity (magnetic Prandtl number Pm = 0.1) while maintaining vigorous convection (magnetic Reynolds number Rm > 500) and rapid rotation (Ekman number E = 10-7) at the limit of what is feasible on today's supercomputers. A detailed and comprehensive analysis highlights several key features matching geomagnetic observations or dynamo theory predictions—all present together in the same simulation—but it also unveils interesting insights relevant for Earth's core dynamics. In this strong-field, dipole-dominated dynamo simulation, the magnetic energy is one order of magnitude larger than the kinetic energy. The spatial distribution of magnetic intensity is highly heterogeneous, and a stark dynamical contrast exists between the interior and the exterior of the tangent cylinder (the cylinder parallel to the axis of rotation that circumscribes the inner core). In the interior, the magnetic field is strongest, and is associated with a vigorous twisted polar vortex, whose dynamics may occasionally lead to the formation of a reverse polar flux patch at the surface of the shell. Furthermore, the strong magnetic field also allows accumulation of light material within the tangent cylinder, leading to stable stratification there. Torsional Alfvén waves are frequently triggered in the vicinity of the tangent cylinder and propagate towards the equator. Outside the tangent cylinder, the magnetic field inhibits the growth of zonal winds and the kinetic energy is mostly non-zonal. Spatio-temporal analysis indicates that the low-frequency, non-zonal flow is quite geostrophic (columnar) and predominantly large-scale: an m = 1 eddy spontaneously emerges in our most extreme simulations, without any

  16. Core followup studies of the Tarapur Reactors with the three dimensional BWR simulator COMTEG

    Energy Technology Data Exchange (ETDEWEB)

    Dwivedi, S. R.; Jagannathan, V.; Mohanakrishnan, P.; Srinivasan, K. R.; Rastogi, B. P.

    1976-07-01

    Both the units of the Tarapur Atomic Power Station started operation in the year 1969. Since then, these units have completed three cycles. For efficient operation and fuel management of these reactors, a three dimensional BWR simulator COMETG has been developed. The reactors are closely being followed using the simulator. The detailed analyses for cycle 3/4 operation of both the units are described in the paper. The results show very good agreement between calculated and measured values. It is concluded that reactor core behaviour could be predicted in a satisfactory manner with the core simulator COMETG.

  17. Development of a 3D-Multigroup program to simulate anomalous diffusion phenomena in the nuclear reactors

    International Nuclear Information System (INIS)

    Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto

    2015-01-01

    Highlights: • The new version of neutron diffusion equation for simulating anomalous diffusion is presented. • Application of fractional calculus in the nuclear reactor is revealed. • A 3D-Multigroup program is developed based on the fractional operators. • The super-diffusion and sub-diffusion phenomena are modeled in the nuclear reactors core. - Abstract: The diffusion process is categorized in three parts, normal diffusion, super-diffusion and sub-diffusion. The classical neutron diffusion equation is used to model normal diffusion. A new scheme of derivatives is required to model anomalous diffusion phenomena. The fractional space derivatives are employed to model anomalous diffusion processes where a plume of particles spreads at an inconsistent rate with the classical Brownian motion model. In the fractional diffusion equation, the fractional Laplacians are used; therefore the statistical jump length of neutrons is unrestricted. It is clear that the fractional Laplacians are capable to model the anomalous phenomena in nuclear reactors. We have developed a NFDE-3D (neutron fractional diffusion equation) as a core calculation code to model normal and anomalous diffusion phenomena. The NFDE-3D is validated against the LMW-LWR reactor. The results demonstrate that reactors exhibit complex behavior versus order of the fractional derivatives which depends on the competition between neutron absorption and super-diffusion phenomenon

  18. ac power control in the Core Flow Test Loop

    International Nuclear Information System (INIS)

    McDonald, D.W.

    1980-01-01

    This work represents a status report on a development effort to design an ac power controller for the Core Flow Test Loop. The Core Flow Test Loop will be an engineering test facility which will simulate the thermal environment of a gas-cooled fast-breeder reactor. The problems and limitations of using sinusoidal ac power to simulate the power generated within a nuclear reactor are addressed. The transformer-thyristor configuration chosen for the Core Flow Test Loop power supply is presented. The initial considerations, design, and analysis of a closed-loop controller prototype are detailed. The design is then analyzed for improved performance possibilities and failure modes are investigated at length. A summary of the work completed to date and a proposed outline for continued development completes the report

  19. Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    International Nuclear Information System (INIS)

    Emrich, William J. Jr.

    2008-01-01

    To support a potential future development of a nuclear thermal rocket engine, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts

  20. Recent advances in nuclear power plant simulation

    International Nuclear Information System (INIS)

    Zerbino, H.; Plisson, P.; Friant, J.Y.

    1997-01-01

    The field of industrial simulation has experienced very significant progress in recent years, and power plant simulation in particular has been an extremely active area. Improvements may be recorded in practically all simulator subsystems. In Europe, the construction of new full- or optimized-scope nuclear power plant simulators during the middle 1990's has been remarkable intense. In fact, it is possible to identify a distinct simulator generation, which constitutes a new de facto simulation standard. Thomson Training and Simulation has taken part in these developments by designing, building, and validation several of these new simulators for Dutch, German and French nuclear power plants. Their characteristics are discussed in this paper. The following main trends may be identified: Process modeling is clearly evolving towards obtaining engineering-grade performance, even under the added constraints of real-time operation and a very wide range of operating conditions to be covered; Massive use of modern graphic user interfaces (GUI) ensures an unprecedented flexibility and user-friendliness for the Instructor Station; The massive use of GUIs also allows the development of Trainee Stations (TS), which significantly enhance the in-depth training value of the simulators; The development of powerful Software Development Environments (SDE) enables the simulator maintenance teams to keep abreast of modifications carried out in the reference plants; Finally, simulator maintenance and its compliance with simulator fidelity requirements are greatly enhanced by integrated Configuration Management Systems (CMS). In conclusion, the power plant simulation field has attained a strong level of maturity, which benefits its approximately forty years of service to the power generation industry. (author)

  1. 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

  2. Nuclear plant simulation using the Nuclear Plant Analyzer

    International Nuclear Information System (INIS)

    Beelman, R.J.; Laats, E.T.; Wagner, R.J.

    1984-01-01

    The Nuclear Plant Analyzer (NPA), a state-of-the-art computerized safety analysis and engineering tool, was employed to simulate nuclear plant response to an abnormal transient during a training exercise at the US Nuclear Regulatory Commission (USNRC) in Washington, DC. Information relative to plant status was taken from a computer animated color graphics display depicting the course of the transient and was transmitted to the NRC Operations Center in a manner identical to that employed during an actual event. Recommendations from the Operations Center were implemented during on-line, interactive execution of the RELAP5 reactor systems code through the NPA allowing a degree of flexibility in training exercises not realized previously. When the debriefing was conducted, the RELAP5 calculations were replayed by way of the color graphics display, adding a new dimension to the debriefing and greatly enhancing the critique of the exercise

  3. Nuclear Fuel Cycle Analysis and Simulation Tool (FAST)

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kwon, Eun Ha; Kim, Ho Dong

    2005-06-15

    This paper describes the Nuclear Fuel Cycle Analysis and Simulation Tool (FAST) which has been developed by the Korea Atomic Energy Research Institute (KAERI). Categorizing various mix of nuclear reactors and fuel cycles into 11 scenario groups, the FAST calculates all the required quantities for each nuclear fuel cycle component, such as mining, conversion, enrichment and fuel fabrication for each scenario. A major advantage of the FAST is that the code employs a MS Excel spread sheet with the Visual Basic Application, allowing users to manipulate it with ease. The speed of the calculation is also quick enough to make comparisons among different options in a considerably short time. This user-friendly simulation code is expected to be beneficial to further studies on the nuclear fuel cycle to find best options for the future all proliferation risk, environmental impact and economic costs considered.

  4. 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...

  5. 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

  6. LOGOS. HX: a core simulator for high conversion boiling water reactors

    International Nuclear Information System (INIS)

    Tsuiki, Makoto; Sakurada, Koichi; Yoshida, Hiroyuki.

    1988-01-01

    A three-dimensional physics simulator 'LOGOS. HX' has been developed for the designing analysis of high conversion boiling water reactor (HCBWR) cores. Its functions, calculational methods, and verification results will briefly be discussed. (author)

  7. Final Report for the "Fusion Application for Core-Edge Transport Simulations (FACETS)"

    Energy Technology Data Exchange (ETDEWEB)

    Cary, John R; Kruger, Scott

    2014-10-02

    The FACETS project over its lifetime developed the first self-consistent core-edge coupled capabilities, a new transport solver for modeling core transport in tokamak cores, developed a new code for modeling wall physics over long time scales, and significantly improved the capabilities and performance of legacy components, UEDGE, NUBEAM, GLF23, GYRO, and BOUT++. These improved capabilities leveraged the team’s expertise in applied mathematics (solvers and algorithms) and computer science (performance improvements and language interoperability). The project pioneered new methods for tackling the complexity of simulating the concomitant complexity of tokamak experiments.

  8. 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

  9. 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)

  10. 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)

  11. Nuclear Power Reactor simulator - based training program

    International Nuclear Information System (INIS)

    Abdelwahab, S.A.S.

    2009-01-01

    nuclear power stations will continue playing a major role as an energy source for electric generation and heat production in the world. in this paper, a nuclear power reactor simulator- based training program will be presented . this program is designed to aid in training of the reactor operators about the principles of operation of the plant. also it could help the researchers and the designers to analyze and to estimate the performance of the nuclear reactors and facilitate further studies for selection of the proper controller and its optimization process as it is difficult and time consuming to do all experiments in the real nuclear environment.this program is written in MATLAB code as MATLAB software provides sophisticated tools comparable to those in other software such as visual basic for the creation of graphical user interface (GUI). moreover MATLAB is available for all major operating systems. the used SIMULINK reactor model for the nuclear reactor can be used to model different types by adopting appropriate parameters. the model of each component of the reactor is based on physical laws rather than the use of look up tables or curve fitting.this simulation based training program will improve acquisition and retention knowledge also trainee will learn faster and will have better attitude

  12. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author) [pt

  13. Development of CANDU core monitoring system

    International Nuclear Information System (INIS)

    Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H.

    2003-01-01

    The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1

  14. Development of CANDU core monitoring system

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H. [Institute for Advanced Engineering, Yongin (Korea, Republic of)

    2003-07-01

    The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1.

  15. 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

  16. 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

  17. 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.

  18. Closure simulation of the MSIV of Unit 1 of the Laguna Verde nuclear power plant using the Simulate 3K code; Simulacion del cierre de las MSIV de la Unidad 1 de la central nuclear Laguna Verde empleando el codigo Simulate-3K

    Energy Technology Data Exchange (ETDEWEB)

    Alegria A, A., E-mail: aalegria@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the simulation of closure transient of all main steam isolation valves (MSIV) was performed with the Simulate-3K (S-3K) code for the Unit 1 of the Laguna Verde nuclear power plant (NPP-LV), which operates to thermal power of 2317 MWt, corresponding to the cycle 15 of operation. The set points for the performance of systems correspond to those set out in transient analysis: 3 seconds for the closure of all MSIV; the start of Scram when 121% of the neutron flux is reached, respect from baseline before the transient; the opening by peer of safety relief valves (SRV) in relief mode when the set point of the pressure is reached, the shoot of the feedwater flow seconds after the start of closing of the MSIV and the shoot of the recirculation water pumps when the pressure is reached in the dome of 1048 psig. The simulation time was of 57 seconds, with the top 50 to reach the steady state, from which the closure of all MSIV starts. In this paper the behavior of the pressure in the dome are analyzed, thermal power, neutron flux, the collapsed water level, the flow at the entrance of core, the steam flow coming out of vessel and the flow through of the SRV; the fuel temperature, the minimal critical power ratio, the readings in the instrumentation systems and reactivities. Instrumentation systems were implemented to analyze the neutron flux, these consist of 96 local power range monitors (LPRM) located in different radial and axial positions of the core and 4 channels of average power range monitors, which grouped at 24 LPRM each one. LPRM response to the change of neutron flux in the center of the core, at different axial positions is also shown. Finally, the results show that the safety limit MCPR is not exceeded. (Author)

  19. 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.

  20. 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.

  1. Soft computing simulation tools for nuclear energy systems

    International Nuclear Information System (INIS)

    Kannan Balasubramanian, S.

    2012-01-01

    This chapter deals with simulation, a very powerful tool in designing, constructing and operating nuclear power generating facilities. There are very different types of power plants, and the examples mentioned in this chapter originate from experience with water cooled and water moderated thermal reactors, based on fission of uranium-235. Nevertheless, the methodological achievements in simulation mentioned below can definitely be used not only for this particular type of nuclear power generating reactor. Simulation means: investigation of processes in the time domain. We can calculate the characteristics and properties of different systems, e.g. we can design a bridge over a river, but if we calculate how it would respond to a thunderstorm with high winds, its movement can or can not evolve after a certain time into destructive oscillation - this type of calculations are called simulation

  2. Simulating Neutronic Core Parameters in a Research and Test Reactor

    International Nuclear Information System (INIS)

    Selim, H.K.; Amin, E.A.; Koutb, M.E.

    2011-01-01

    The present study proposes an Artificial Neural Network (ANN) modeling technique that predicts the control rods positions in a nuclear research reactor. The neutron, flux in the core of the reactor is used as the training data for the neural network model. The data used to train and validate the network are obtained by modeling the reactor core with the neutronic calculation code: CITVAP. The type of the network used in this study is the feed forward multilayer neural network with the backpropagation algorithm. The results show that the proposed ANN has good generalization capability to estimate the control rods positions knowing neutron flux for a research and test reactor. This method can be used to predict critical control rods positions to be used for reactor operation after reload

  3. SB LOCA analyses for Krsko Full Scope Simulator verification

    International Nuclear Information System (INIS)

    Prosek, A.; Parzer, I.; Mavko, B.

    2000-01-01

    Nuclear power plant simulators are intended to be used for training and maintaining competence to ensure safe, reliable operation of nuclear power plants throughout the world. The simulator shall be specified to a reference unit and its performance validation testing shall be provided. In this study a small-break loss-of-coolant accident (SB LOCA) response of Krsko nuclear power plant (NPP) was calculated for full scope simulator verification. The investigation included five cases with varying the break size in the cold leg of reactor coolant system. The plant specific and verified RELAP5/MOD2 model of Krsko nuclear power plant (NPP), developed in the past for 1882 MWt power, was adapted for 2000 MWt power (cycle 17) including the model for replacement steam generators. The results showed that the plant system response to breaks with small break area was slower compared to breaks with larger break area. The core heatup occurred in most of the cases analyzed. The acceptance criteria for emergency core cooling system were also met. The predicted results of the SB LOCA analysis for Krsko NPP suggest that they may be used for verification of the Krsko Full Scope Simulator performance. (author)

  4. GENERAL-RELATIVISTIC SIMULATIONS OF THREE-DIMENSIONAL CORE-COLLAPSE SUPERNOVAE

    Energy Technology Data Exchange (ETDEWEB)

    Ott, Christian D.; Abdikamalov, Ernazar; Moesta, Philipp; Haas, Roland; Drasco, Steve; O' Connor, Evan P.; Reisswig, Christian [TAPIR, Mailcode 350-17, California Institute of Technology, Pasadena, CA 91125 (United States); Meakin, Casey A. [Theoretical Division, Los Alamos National Laboratory, Los Alamos, NM (United States); Schnetter, Erik, E-mail: cott@tapir.caltech.edu [Perimeter Institute for Theoretical Physics, Waterloo, ON (Canada)

    2013-05-10

    We study the three-dimensional (3D) hydrodynamics of the post-core-bounce phase of the collapse of a 27 M{sub Sun} star and pay special attention to the development of the standing accretion shock instability (SASI) and neutrino-driven convection. To this end, we perform 3D general-relativistic simulations with a three-species neutrino leakage scheme. The leakage scheme captures the essential aspects of neutrino cooling, heating, and lepton number exchange as predicted by radiation-hydrodynamics simulations. The 27 M{sub Sun} progenitor was studied in 2D by Mueller et al., who observed strong growth of the SASI while neutrino-driven convection was suppressed. In our 3D simulations, neutrino-driven convection grows from numerical perturbations imposed by our Cartesian grid. It becomes the dominant instability and leads to large-scale non-oscillatory deformations of the shock front. These will result in strongly aspherical explosions without the need for large-scale SASI shock oscillations. Low-l-mode SASI oscillations are present in our models, but saturate at small amplitudes that decrease with increasing neutrino heating and vigor of convection. Our results, in agreement with simpler 3D Newtonian simulations, suggest that once neutrino-driven convection is started, it is likely to become the dominant instability in 3D. Whether it is the primary instability after bounce will ultimately depend on the physical seed perturbations present in the cores of massive stars. The gravitational wave signal, which we extract and analyze for the first time from 3D general-relativistic models, will serve as an observational probe of the postbounce dynamics and, in combination with neutrinos, may allow us to determine the primary hydrodynamic instability.

  5. GENERAL-RELATIVISTIC SIMULATIONS OF THREE-DIMENSIONAL CORE-COLLAPSE SUPERNOVAE

    International Nuclear Information System (INIS)

    Ott, Christian D.; Abdikamalov, Ernazar; Mösta, Philipp; Haas, Roland; Drasco, Steve; O'Connor, Evan P.; Reisswig, Christian; Meakin, Casey A.; Schnetter, Erik

    2013-01-01

    We study the three-dimensional (3D) hydrodynamics of the post-core-bounce phase of the collapse of a 27 M ☉ star and pay special attention to the development of the standing accretion shock instability (SASI) and neutrino-driven convection. To this end, we perform 3D general-relativistic simulations with a three-species neutrino leakage scheme. The leakage scheme captures the essential aspects of neutrino cooling, heating, and lepton number exchange as predicted by radiation-hydrodynamics simulations. The 27 M ☉ progenitor was studied in 2D by Müller et al., who observed strong growth of the SASI while neutrino-driven convection was suppressed. In our 3D simulations, neutrino-driven convection grows from numerical perturbations imposed by our Cartesian grid. It becomes the dominant instability and leads to large-scale non-oscillatory deformations of the shock front. These will result in strongly aspherical explosions without the need for large-scale SASI shock oscillations. Low-l-mode SASI oscillations are present in our models, but saturate at small amplitudes that decrease with increasing neutrino heating and vigor of convection. Our results, in agreement with simpler 3D Newtonian simulations, suggest that once neutrino-driven convection is started, it is likely to become the dominant instability in 3D. Whether it is the primary instability after bounce will ultimately depend on the physical seed perturbations present in the cores of massive stars. The gravitational wave signal, which we extract and analyze for the first time from 3D general-relativistic models, will serve as an observational probe of the postbounce dynamics and, in combination with neutrinos, may allow us to determine the primary hydrodynamic instability.

  6. Simulation of rod drop experiments in the initial cores of Loviisa and Mochovce

    International Nuclear Information System (INIS)

    Kaloinen, E.; Kyrki-Rajamaeki, R.; Wasastjerna, F.

    1999-01-01

    Interpretation of rod drop measurements during startup tests of the Loviisa reactors has earlier been studied with two-dimensional core calculations using a spatial prompt jump approximation. In these calculations the prediction for the reactivity meter reading was lower than the measured values by 25%. Another approach to solve the problem is simulation of the rod drop experiment with dynamic core calculations coupled with out of core calculations to estimate the response of ex-core ionization chambers for the reactivity meter. This report described the calculations performed with the three-dimensional dynamic code HEXTRAN for prediction of the reactivity meter readings in rod drop experiments in initial cores of the WWER-440 reactors. (Authors)

  7. The Cea multi-scale and multi-physics simulation project for nuclear applications

    International Nuclear Information System (INIS)

    Ledermann, P.; Chauliac, C.; Thomas, J.B.

    2005-01-01

    Full text of publication follows. Today numerical modelling is everywhere recognized as an essential tool of capitalization, integration and share of knowledge. For this reason, it becomes the central tool of research. Until now, the Cea developed a set of scientific software allowing to model, in each situation, the operation of whole or part of a nuclear installation and these codes are largely used in nuclear industry. However, for the future, it is essential to aim for a better accuracy, a better control of uncertainties and better performance in computing times. The objective is to obtain validated models allowing accurate predictive calculations for actual complex nuclear problems such as fuel behaviour in accidental situation. This demands to master a large and interactive set of phenomena ranging from nuclear reaction to heat transfer. To this end, Cea, with industrial partners (EDF, Framatome-ANP, ANDRA) has designed an integrated platform of calculation, devoted to the study of nuclear systems, and intended at the same time for industries and scientists. The development of this platform is under way with the start in 2005 of the integrated project NURESIM, with 18 European partners. Improvement is coming not only through a multi-scale description of all phenomena but also through an innovative design approach requiring deep functional analysis which is upstream from the development of the simulation platform itself. In addition, the studies of future nuclear systems are increasingly multidisciplinary (simultaneous modelling of core physics, thermal-hydraulics and fuel behaviour). These multi-physics and multi-scale aspects make mandatory to pay very careful attention to software architecture issues. A global platform is thus developed integrating dedicated specialized platforms: DESCARTES for core physics, NEPTUNE for thermal-hydraulics, PLEIADES for fuel behaviour, SINERGY for materials behaviour under irradiation, ALLIANCES for the performance

  8. Research and development of in-core transducers at the CIAE

    International Nuclear Information System (INIS)

    Huang Yucai; Liu Yupu; Jia Guozhen; Liu Lianping

    1996-01-01

    In this paper, R and D of in-core transducers at the CIAE are briefly summarized. With the construction and commissioning of PWR nuclear power plant in China, fuel rod behaviour need to be studied carefully. As conventional transducers cannot meet the requirements of in-core applications, R and D of in-core transducers are developed. Since 1980's, several kinds of in-core transducers have been successfully fabricated and tested under the conditions simulating PWR. At present, in-pile tests of the transducers combining with the studies of individual behaviour of PWR fuel rod are being planned at the CIAE. (author). 11 refs, 12 figs, 4 tabs

  9. Tinuso: A processor architecture for a multi-core hardware simulation platform

    DEFF Research Database (Denmark)

    Schleuniger, Pascal; Karlsson, Sven

    2010-01-01

    Multi-core systems have the potential to improve performance, energy and cost properties of embedded systems but also require new design methods and tools to take advantage of the new architectures. Due to the limited accuracy and performance of pure software simulators, we are working on a cycle...... accurate hardware simulation platform. We have developed the Tinuso processor architecture for this platform. Tinuso is a processor architecture optimized for FPGA implementation. The instruction set makes use of predicated instructions and supports C/C++ and assembly language programming. It is designed...... to be easy extendable to maintain the exibility required for the research on multi-core systems. Tinuso contains a co-processor interface to connect to a network interface. This interface allow for communication over an on-chip network. A clock frequency estimation study on a deeply pipelined Tinuso...

  10. Development of a research simulator for the study of human factors and experiments

    International Nuclear Information System (INIS)

    Kawano, R.; Shibuya, S.

    1999-01-01

    A research simulator of nuclear power plant for Human Factors was developed. It simulates the behaviors of the 1100MWe BWR nuclear power plant and has almost same functions ant scope of the simulation as a full-scope training simulator. Physical models installed in the system enable us to execute experiments with multi-malfunction scenario. A severe accident simulation package replaces the running simulation code when the maximum core temperature exceeds 1200 deg C and the core approaches meltdown conditions. The central control panel was simulated by soft panels, indicator and operational switches on the panels by computer graphics, displayed on 22 console boxes containing CRT. The introduction of soft panels and EWSs connected with LAN accomplished flexibility and extendibility. Some experiments by using the simulator were executed and the system has been improved based on the experience from the experiments. It is important to evaluate the effectiveness of any new system by using an actual plant size research simulator before its practical application to keep steady and safe operation of nuclear power plants. (author)

  11. Nuclear power plant diagnostics study at the Midland training simulator

    International Nuclear Information System (INIS)

    Reifman, J.; Rank, P.; Lee, J.C.

    1991-01-01

    Training simulators provide a real world environment for testing advanced diagnostic and control systems as an aid to nuclear power plant operators. The simulators not only duplicate the hardware din the actual control room, allowing for analysis of man-machine interface, but also represent the dynamic behavior of the reference plant in real-time, in a realistic manner. Training simulators provide the means to representing the reference plant operations in a wide range of operation conditions including off-normal and emergency conditions. Transient events with very low probability of occurrence can then be represented and used to test the capabilities of advanced diagnostic and control systems. For these reasons, full-scope operator training simulators have been used as a test bed for a number of advanced diagnostic concepts. The University of Michigan and Consumers Power Company have been collaborating in a program devoted to the development and study of advanced concepts for automatic diagnostics and control of nuclear power plants. The program has been focused on the use of the full-scope operator training Midland Nuclear Power Plant Unit 2 (MNP-2) Simulator for development, testing, and verification of advanced diagnostics concepts. In their current efforts, the authors have developed two artificial intelligent (AI) diagnostic concepts that have been applied to the MNP-2 Simulator: the systematic generation and updating of a rule-based knowledge system for nuclear power plant diagnostics and a nonlinear parameter estimation algorithm called the simulation filter. The simulation filter algorithm is used with the MNP-2 Simulator to improve the simulation of the Three Mile Island Unit 2 (TMI-2) accident. 11 refs., 4 figs

  12. Improvement of the nuclear plant analyzer for Korean Standard Nuclear Power Plants

    International Nuclear Information System (INIS)

    Choi, Sung Soo; Han, Byoung Sub; Suh, Jae Seung; Son, Dae Seong

    2005-04-01

    Accurate analysis of the operating characteristics of Nuclear Power Plants provides valuable information for both norman and abnormal operation. The information can be used for the enhancement of plant performance and safety. Usually, such analysis is performed using computer codes used for plant design or simulators. However, their usages are limited because special expertise is required to use the computer codes and simulators are not portable. Therefore, it deemed necessary to develop an NPA which minimizes those limitations and can be used for the analysis and simulation of Nuclear Power Plants. The purpose of this study is to develop a real-time best-estimate NPA for the Korean Nuclear Power Plants(KSNP). The NPA is an interactive, high fidelity engineering simulator. NPA combines the process model simulating the plant behavior with the latest computer technology such as Graphical User Interface(GUI) and simulation executive for enhanced user interface. The process model includes models for a three-dimensional reactor core, the NSSS, secondary system including turbine and feedtrain, safety auxiliary systems, and various control systems. Through the verification and validation of the NPA, it was demonstrated that the NPA can realistically simulate the plant behaviors during transient and accident conditions

  13. Evaluation of nuclear characteristics of minor actinide loaded core. Analyses of BFS-69 and BFS-66-2 critical experiments

    International Nuclear Information System (INIS)

    Hazama, Taira; Sato, Wakaei

    2010-09-01

    Collaboration with Russian Institute of Physics and Power Engineering named 'Investigation of neutronic-physical characteristics and their change when introducing large quantity of neptunium (Np) at different BFS critical assemblies' has been accomplished. This is the second report of the collaboration to describe experimental information and analysis results on BFS-69 and BFS-66-2 critical experiments. In the experiments, various nuclear characteristics were measured in 2 kinds of cores with/without Np loading of about 8 kg. JAEA's standard analysis results were presented with four kinds of nuclear data (JENDL-3.2, JENDL-3.3, JENDL/AC-2008, and ENDF/BVII). Analytical results show: 1) An overestimation trend has been observed in BFS-69 criticality results, especially with JENDL-3.3 and JENDL/AC-2008. The difference from ENDF/B-II having better results mainly lies in the average cosine of the scattering angle around 1 MeV. 2) A small discrepancy exists in BFS-69 Na void reactivity results with the three JENDL nuclear data. The difference from ENDF/B-II mainly lies in scattering cross sections of sodium around 1 MeV and fission cross section of 239 Pu around 1 keV. 3) The analysis results simulate measured Np effects on nuclear characteristics within experimental errors. (author)

  14. Experimental study on air ingress during a primary pipe rupture accident with a graphite reactor core simulator

    International Nuclear Information System (INIS)

    Takeda, Tetsuaki; Hishida, Makoto; Baba, Shinichi

    1991-11-01

    When a primary coolant pipe of a High Temperature Gas Cooled Reactor (HTGR) ruptures, helium gas in the reactor core blows out into the container, and the primary cooling system reduces the pressure. After the pressures are balanced between the reactor and the container, air is expected to enter into the reactor core from the breach. It seems to be probable that the graphite structures is oxidized by air. Hence, it is necessary to investigate the air ingress process and the behavior of the generating gases by the oxidation reactions. The previous experimental study is performed on the molecular diffusion and natural convection of the two component gas mixtures using a test model simulating simply the reactor. Objective of the study was to investigate the air ingress process during the early stage of the primary pipe rupture accident. However, since the model did not have any kind of graphite components, the reaction between graphite and oxygen was not simulated. The present model includes the reactor core and the high temperature plenum simulators made of graphite. The major results obtained in the present study are summarized in the followings: (1) The air ingress process with graphite oxidation reaction is similar to that without the reaction qualitatively. (2) When the reactor core simulator is maintained at low temperatures (lower than 450degC), the initiation time of the natural circulation of air is almost equal to that of the natural circulation of nitrogen. On the other hand, when the temperature of the reactor core simulator is high (more than 500degC), the initiation time of the natural circulation of air is earlier than that of nitrogen. (3) When the temperature of the reactor core simulator is higher than 600degC, oxygen is almost dissipated by the graphite structures. When the temperature of the reactor core simulator is below 700degC, carbon dioxide mainly is generated by the oxidation reactions. (author)

  15. High-Fidelity Space-Time Adaptive Multiphysics Simulations in Nuclear Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Solin, Pavel [Univ. of Reno, NV (United States); Ragusa, Jean [Texas A & M Univ., College Station, TX (United States)

    2014-03-09

    We delivered a series of fundamentally new computational technologies that have the potential to significantly advance the state-of-the-art of computer simulations of transient multiphysics nuclear reactor processes. These methods were implemented in the form of a C++ library, and applied to a number of multiphysics coupled problems relevant to nuclear reactor simulations.

  16. High-Fidelity Space-Time Adaptive Multiphysics Simulations in Nuclear Engineering

    International Nuclear Information System (INIS)

    Solin, Pavel; Ragusa, Jean

    2014-01-01

    We delivered a series of fundamentally new computational technologies that have the potential to significantly advance the state-of-the-art of computer simulations of transient multiphysics nuclear reactor processes. These methods were implemented in the form of a C++ library, and applied to a number of multiphysics coupled problems relevant to nuclear reactor simulations.

  17. Entanglement measures in embedding quantum simulators with nuclear spins

    Science.gov (United States)

    Xin, Tao; Pedernales, Julen S.; Solano, Enrique; Long, Gui-Lu

    2018-02-01

    We implement an embedding quantum simulator (EQS) in nuclear spin systems. The experiment consists of a simulator of up to three qubits, plus a single ancillary qubit, where we are able to efficiently measure the concurrence and the three-tangle of two-qubit and three-qubit systems as they undergo entangling dynamics. The EQS framework allows us to drastically reduce the number of measurements needed for this task, which otherwise would require full-state reconstruction of the qubit system. Our simulator is built of the nuclear spins of four 13C atoms in a molecule of trans-crotonic acid manipulated with NMR techniques.

  18. Parity simulation for nuclear plant analysis

    International Nuclear Information System (INIS)

    Hansen, K.F.; Depiente, E.

    1986-01-01

    The analysis of the transient performance of nuclear plants is sufficiently complex that simulation tools are needed for design and safety studies. The simulation tools are needed for design and safety studies. The simulation tools are normally digital because of the speed, flexibility, generality, and repeatability of digital computers. However, communication with digital computers is an awkward matter, requiring special skill or training. The designer wishing to gain insight into system behavior must expend considerable effort in learning to use computer codes, or else have an intermediary communicate with the machine. There has been a recent development in analog simulation that simplifies the user interface with the simulator, while at the same time improving the performance of analog computers. This development is termed parity simulation and is now in routine use in analyzing power electronic network transients. The authors describe the concept of parity simulation and present some results of using the approach to simulate neutron kinetics problems

  19. Equipment and performance upgrade of compact nuclear simulator

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Chang; Kwon, Kee Choon; Lee, Dong Young; Cha, Kyung Ho; Park, Won Man; Song, Soon Ja; Lee, Jung Woon; Oh, In Seok; Moon, Byung Sou; Ham, Chang Shik; Seh, Kyung Won; Joo, Young Chang; Lee, Eui Jin

    1999-12-01

    We developed techniques for upgrading nuclear power plant simulators that lost some function due to aging and obsolescence of parts. This project will provide a guidance for resolving the problems with aged simulators. We transplanted the simulation code operated in Macro-VAX II into the HP workstation with operating system version 10.20. If further research is proceeded, it will be possible to build simulator system on PCs. (author)

  20. Simulation of the PHEBUS FPT-1 experiment using MELCOR and exploration of the primary core degradation mechanism

    International Nuclear Information System (INIS)

    Wang, Jun; Corradini, Michael L.; Fu, Wen; Haskin, Troy; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-01-01

    Highlights: • Core degradation evaluation is an important process in risk analysis. • PHEBUS experiment was simulated using MELCOR. • The results confirm the validity of MELCOR’s simulation of the PHEBUS experiment. • These results are used to analyze the mode and behavior of core degradation. - Abstract: Core degradation evaluation of probability, progression and consequences of a core degradation accident is critical for evaluation of risk as well as its mitigation. However, research and modeling of severe accidents to date are limited, and their accuracy in predicting severe accident consequences is still insufficient. It is therefore important to explore the mechanisms of core degradation and to develop mitigation measures for severe accidents. PHEBUS FPT1 is a typical and classic core degradation experiment. MELCOR is a world famous severe accident analysis code developed by Sandia National Lab that has seen wide application, a broad user base, and a number of supporting experiments. The PHEBUS experiment was simulated using MELCOR in this paper. Experimental data on, thermal power and steam mass flow rates are used to determine average pressure, energy distribution, molten mass, temperature of the fuel, and hydrogen generation. Data from the PHEBUS experiment and Cho’s calculations are used to compare the average pressure, several fuel temperatures and the hydrogen generation rate. The results confirm the validity of MELCOR’s simulation of the PHEBUS experiment. The temperature distribution of the core is provided. These results are used to determine the mode and behavior of core degradation with the intent of building a foundation for further research

  1. Core power capability verification for PWR NPP

    International Nuclear Information System (INIS)

    Xian Chunyu; Liu Changwen; Zhang Hong; Liang Wei

    2002-01-01

    The Principle and methodology of pressurized water reactor nuclear power plant core power capability verification for reload are introduced. The radial and axial power distributions of normal operation (category I or condition I) and abnormal operation (category II or condition II) are simulated by using neutronics calculation code. The linear power density margin and DNBR margin for both categories, which reflect core safety, are analyzed from the point view of reactor physics and T/H, and thus category I operating domain and category II protection set point are verified. Besides, the verification results of reference NPP are also given

  2. 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.)

  3. 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.

  4. The effect of core configuration on temperature coefficient of reactivity in IRR-1

    Energy Technology Data Exchange (ETDEWEB)

    Bettan, M.; Silverman, I.; Shapira, M.; Nagler, A. [Soreq Nuclear Research Center, Yavne (Israel)

    1997-08-01

    Experiments designed to measure the effect of coolant moderator temperature on core reactivity in an HEU swimming pool type reactor were performed. The moderator temperature coefficient of reactivity ({alpha}{sub {omega}}) was obtained and found to be different in two core loadings. The measured {alpha}{sub {omega}} of one core loading was {minus}13 pcm/{degrees}C at the temperature range of 23-30{degrees}C. This value of {alpha}{sub {omega}} is comparable to the data published by the IAEA. The {alpha}{sub {omega}} measured in the second core loading was found to be {minus}8 pcm/{degrees}C at the same temperature range. Another phenomenon considered in this study is core behavior during reactivity insertion transient. The results were compared to a core simulation using the Dynamic Simulator for Nuclear Power Plants. It was found that in the second core loading factors other than the moderator temperature influence the core reactivity more than expected. These effects proved to be extremely dependent on core configuration and may in certain core loadings render the reactor`s reactivity coefficient undesirable.

  5. Hydrodynamical simulations of the stream-core interaction in the slow merger of massive stars

    Science.gov (United States)

    Ivanova, N.; Podsiadlowski, Ph.; Spruit, H.

    2002-08-01

    We present detailed simulations of the interaction of a stream emanating from a mass-losing secondary with the core of a massive supergiant in the slow merger of two stars inside a common envelope. The dynamics of the stream can be divided into a ballistic phase, starting at the L1 point, and a hydrodynamical phase, where the stream interacts strongly with the core. Considering the merger of a 1- and 5-Msolar star with a 20-Msolar evolved supergiant, we present two-dimensional hydrodynamical simulations using the PROMETHEUS code to demonstrate how the penetration depth and post-impact conditions depend on the initial properties of the stream material (e.g. entropy, angular momentum, stream width) and the properties of the core (e.g. density structure and rotation rate). Using these results, we present a fitting formula for the entropy generated in the stream-core interaction and a recipe for the determination of the penetration depth based on a modified Bernoulli integral.

  6. Coupled full core neutron transport/CFD simulations of pressurized water reactors

    International Nuclear Information System (INIS)

    Kochunas, B.; Stimpson, S.; Collins, B.; Downar, T.; Brewster, R.; Baglietto, E.; Yan, J.

    2012-01-01

    Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)

  7. 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

  8. Verification of results of core physics on-line simulation by NGFM code

    International Nuclear Information System (INIS)

    Zhao Yu; Cao Xinrong; Zhao Qiang

    2008-01-01

    Nodal Green's Function Method program NGFM/TNGFM has been trans- planted to windows system. The 2-D and 3-D benchmarks have been checked by this program. And the program has been used to check the results of QINSHAN-II reactor simulation. It is proved that the NGFM/TNGFM program is applicable for reactor core physics on-line simulation system. (authors)

  9. Building of Nuclear Ship Engineering Simulation System development of the simulator for the integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Teruo; Shimazaki, Junya; Yabuuchi, Noriaki; Fukuhara, Yosifumi; Kusunoki, Takeshi; Ochiai, Masaaki [Department of Nuclear Energy Systems, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Nakazawa, Toshio [Department of HTTR Project, Oarai Research Establishment, Japan Atomic Energy Research Institute, Oarai, Ibaraki (Japan)

    2000-03-01

    JAERI had carried out the design study of a light-weight and compact integral type reactor of power 100 MW{sub th} with passive safety as a power source for the future nuclear ships, and completed an engineering design. To confirm the design and operation performance and to utilize the study of automation of the operations of reactor, we developed a real-time simulator for the integral type reactor. This simulator is a part of Nuclear Ship Engineering Simulation System (NESSY) and on the same hardware as 'Mutsu' simulator which was developed to simulate the first Japanese nuclear ship Mutsu'. Simulation accuracy of 'Mutsu' simulator was verified by comparing the simulation results With data got in the experimental voyage of 'Mutsu'. The simulator for the integral type reactor uses the same programs which were used in 'Mutsu' simulator for the separate type PWR, and the simulated results are approximately consistent with the calculated values using RELAP5/MOD2 (The later points are reported separately). Therefore simulation accuracy of the simulator for the integral type reactor is also expected to be reasonable, though it is necessary to verify by comparing with the real plant data or experimental data in future. We can get the perspectives to use as a real-time engineering simulator and to achieve the above-mentioned aims. This is a report on development of the simulator for the integral type reactor mainly focused on the contents of the analytical programs expressed the structural features of reactor. (author)

  10. Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Energy Technology Data Exchange (ETDEWEB)

    Turinsky, Paul J., E-mail: turinsky@ncsu.edu [North Carolina State University, PO Box 7926, Raleigh, NC 27695-7926 (United States); Kothe, Douglas B., E-mail: kothe@ornl.gov [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6164 (United States)

    2016-05-15

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics “core simulator” based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL

  11. NUFACTS-nuclear fuel cycle activity simulator: reference manual. Final report

    International Nuclear Information System (INIS)

    Triplett, M.B.; Waddell, J.D.; Breese, T.A.

    1978-01-01

    The Nuclear Fuel Cycle Activity Simulator (NUFACTS) is a package of FORTRAN subroutines which facilitate the simulation of a diversity of nuclear power growth scenarios. An approach to modeling the nuclear fuel cycle has been developed that is highly adaptive and capable of addressing a variety of problems. Being a simulation model rather than an optimization model, NUFACTS mimics the events and processes that are characteristic of the nuclear fuel cycle. This approach enables the model user to grasp the modeling approach rather quickly. Within this report descriptions of the model and its components are provided with several emphases. First, a discussion of modeling approach and basic assumptions is provided. Next, instructions are provided for generating data, inputting the data properly, and running the code. Finally, detailed descriptions of individual program element are given as an aid to modifying and extending the present capabilities

  12. A real-time computer simulation of nuclear simulator software using standard PC hardware and linux environments

    International Nuclear Information System (INIS)

    Cha, K. H.; Kweon, K. C.

    2001-01-01

    A feasibility study, which standard PC hardware and Real-Time Linux are applied to real-time computer simulation of software for a nuclear simulator, is presented in this paper. The feasibility prototype was established with the existing software in the Compact Nuclear Simulator (CNS). Throughout the real-time implementation in the feasibility prototype, we has identified that the approach can enable the computer-based predictive simulation to be approached, due to both the remarkable improvement in real-time performance and the less efforts for real-time implementation under standard PC hardware and Real-Time Linux envrionments

  13. 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.

  14. 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.

  15. 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

  16. 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)

  17. Module-based Simulation System for efficient development of nuclear simulation programs

    International Nuclear Information System (INIS)

    Yoshikawa, Hidekazu; Wakabayashi, Jiro

    1990-01-01

    Module-based Simulation System (MSS) has been developed to realize a new software environment enabling versatile dynamic simulation of a complex nuclear power plant system flexibly. Described in the paper are (i) fundamental methods utilized in MMS and its software systemization, (ii) development of human interface system to help users in generating integrated simulation programs automatically, and (iii) development of an intelligent user support system for helping users in the two phases of automatical semantic diagnosis and consultation to automatic input data setup for the MSS-generated programs. (author)

  18. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    Energy Technology Data Exchange (ETDEWEB)

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  19. Nuclear Engine System Simulation (NESS). Version 2.0: Program user's guide. Final Report

    International Nuclear Information System (INIS)

    Pelaccio, D.G.; Scheil, C.M.; Petrosky, L.

    1993-03-01

    This Program User's Guide discusses the Nuclear Thermal Propulsion (NTP) engine system design features and capabilities modeled in the Nuclear Engine System Simulation (NESS): Version 2.0 program (referred to as NESS throughout the remainder of this document), as well as its operation. NESS was upgraded to include many new modeling capabilities not available in the original version delivered to NASA LeRC in Dec. 1991, NESS's new features include the following: (1) an improved input format; (2) an advanced solid-core NERVA-type reactor system model (ENABLER 2); (3) a bleed-cycle engine system option; (4) an axial-turbopump design option; (5) an automated pump-out turbopump assembly sizing option; (6) an off-design gas generator engine cycle design option; (7) updated hydrogen properties; (8) an improved output formnd (9) personal computer operation capability. Sample design cases are presented in the user's guide that demonstrate many of the new features associated with this upgraded version of NESS, as well as design modeling features associated with the original version of NESS

  20. 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

  1. Proceedings of the 1994 nuclear simulation symposium

    Energy Technology Data Exchange (ETDEWEB)

    Laughton, P J [ed.; Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.

    1994-12-31

    This symposium on nuclear simulation included sessions on reactor physics, on thermalhydraulics, and on fuel and fuel channels of CANDU type reactors. The individual papers have all been abstracted separately.

  2. Fast multipurpose Monte Carlo simulation for proton therapy using multi- and many-core CPU architectures

    Energy Technology Data Exchange (ETDEWEB)

    Souris, Kevin, E-mail: kevin.souris@uclouvain.be; Lee, John Aldo [Center for Molecular Imaging and Experimental Radiotherapy, Institut de Recherche Expérimentale et Clinique, Université catholique de Louvain, Avenue Hippocrate 54, 1200 Brussels, Belgium and ICTEAM Institute, Université catholique de Louvain, Louvain-la-Neuve 1348 (Belgium); Sterpin, Edmond [Center for Molecular Imaging and Experimental Radiotherapy, Institut de Recherche Expérimentale et Clinique, Université catholique de Louvain, Avenue Hippocrate 54, 1200 Brussels, Belgium and Department of Oncology, Katholieke Universiteit Leuven, O& N I Herestraat 49, 3000 Leuven (Belgium)

    2016-04-15

    Purpose: Accuracy in proton therapy treatment planning can be improved using Monte Carlo (MC) simulations. However the long computation time of such methods hinders their use in clinical routine. This work aims to develop a fast multipurpose Monte Carlo simulation tool for proton therapy using massively parallel central processing unit (CPU) architectures. Methods: A new Monte Carlo, called MCsquare (many-core Monte Carlo), has been designed and optimized for the last generation of Intel Xeon processors and Intel Xeon Phi coprocessors. These massively parallel architectures offer the flexibility and the computational power suitable to MC methods. The class-II condensed history algorithm of MCsquare provides a fast and yet accurate method of simulating heavy charged particles such as protons, deuterons, and alphas inside voxelized geometries. Hard ionizations, with energy losses above a user-specified threshold, are simulated individually while soft events are regrouped in a multiple scattering theory. Elastic and inelastic nuclear interactions are sampled from ICRU 63 differential cross sections, thereby allowing for the computation of prompt gamma emission profiles. MCsquare has been benchmarked with the GATE/GEANT4 Monte Carlo application for homogeneous and heterogeneous geometries. Results: Comparisons with GATE/GEANT4 for various geometries show deviations within 2%–1 mm. In spite of the limited memory bandwidth of the coprocessor simulation time is below 25 s for 10{sup 7} primary 200 MeV protons in average soft tissues using all Xeon Phi and CPU resources embedded in a single desktop unit. Conclusions: MCsquare exploits the flexibility of CPU architectures to provide a multipurpose MC simulation tool. Optimized code enables the use of accurate MC calculation within a reasonable computation time, adequate for clinical practice. MCsquare also simulates prompt gamma emission and can thus be used also for in vivo range verification.

  3. Fast multipurpose Monte Carlo simulation for proton therapy using multi- and many-core CPU architectures

    International Nuclear Information System (INIS)

    Souris, Kevin; Lee, John Aldo; Sterpin, Edmond

    2016-01-01

    Purpose: Accuracy in proton therapy treatment planning can be improved using Monte Carlo (MC) simulations. However the long computation time of such methods hinders their use in clinical routine. This work aims to develop a fast multipurpose Monte Carlo simulation tool for proton therapy using massively parallel central processing unit (CPU) architectures. Methods: A new Monte Carlo, called MCsquare (many-core Monte Carlo), has been designed and optimized for the last generation of Intel Xeon processors and Intel Xeon Phi coprocessors. These massively parallel architectures offer the flexibility and the computational power suitable to MC methods. The class-II condensed history algorithm of MCsquare provides a fast and yet accurate method of simulating heavy charged particles such as protons, deuterons, and alphas inside voxelized geometries. Hard ionizations, with energy losses above a user-specified threshold, are simulated individually while soft events are regrouped in a multiple scattering theory. Elastic and inelastic nuclear interactions are sampled from ICRU 63 differential cross sections, thereby allowing for the computation of prompt gamma emission profiles. MCsquare has been benchmarked with the GATE/GEANT4 Monte Carlo application for homogeneous and heterogeneous geometries. Results: Comparisons with GATE/GEANT4 for various geometries show deviations within 2%–1 mm. In spite of the limited memory bandwidth of the coprocessor simulation time is below 25 s for 10"7 primary 200 MeV protons in average soft tissues using all Xeon Phi and CPU resources embedded in a single desktop unit. Conclusions: MCsquare exploits the flexibility of CPU architectures to provide a multipurpose MC simulation tool. Optimized code enables the use of accurate MC calculation within a reasonable computation time, adequate for clinical practice. MCsquare also simulates prompt gamma emission and can thus be used also for in vivo range verification.

  4. Fast multipurpose Monte Carlo simulation for proton therapy using multi- and many-core CPU architectures.

    Science.gov (United States)

    Souris, Kevin; Lee, John Aldo; Sterpin, Edmond

    2016-04-01

    Accuracy in proton therapy treatment planning can be improved using Monte Carlo (MC) simulations. However the long computation time of such methods hinders their use in clinical routine. This work aims to develop a fast multipurpose Monte Carlo simulation tool for proton therapy using massively parallel central processing unit (CPU) architectures. A new Monte Carlo, called MCsquare (many-core Monte Carlo), has been designed and optimized for the last generation of Intel Xeon processors and Intel Xeon Phi coprocessors. These massively parallel architectures offer the flexibility and the computational power suitable to MC methods. The class-II condensed history algorithm of MCsquare provides a fast and yet accurate method of simulating heavy charged particles such as protons, deuterons, and alphas inside voxelized geometries. Hard ionizations, with energy losses above a user-specified threshold, are simulated individually while soft events are regrouped in a multiple scattering theory. Elastic and inelastic nuclear interactions are sampled from ICRU 63 differential cross sections, thereby allowing for the computation of prompt gamma emission profiles. MCsquare has been benchmarked with the gate/geant4 Monte Carlo application for homogeneous and heterogeneous geometries. Comparisons with gate/geant4 for various geometries show deviations within 2%-1 mm. In spite of the limited memory bandwidth of the coprocessor simulation time is below 25 s for 10(7) primary 200 MeV protons in average soft tissues using all Xeon Phi and CPU resources embedded in a single desktop unit. MCsquare exploits the flexibility of CPU architectures to provide a multipurpose MC simulation tool. Optimized code enables the use of accurate MC calculation within a reasonable computation time, adequate for clinical practice. MCsquare also simulates prompt gamma emission and can thus be used also for in vivo range verification.

  5. Development of adaptive core emulator for PMS-XRBP of CE type plant

    International Nuclear Information System (INIS)

    Song, Jae Seung; Zee, Sung Quun; Lee, Chung Chan; Lee, Ki Bog; Rhy, Hyo Sang; Chang, Jong Hwa; Lee, Young Ouk; Baek, Seung Min; Seo, Ho Joon.

    1996-12-01

    The purpose of this report is to develop ONED-based adaptive core emulator (ACE) for Korean Standard Nuclear Power Plant. This report is first year report and includes 1) augmentation of ONED94 I/O system 2) non-equilibrium xenon initialization for core transient simulation 3) ONED94 verification via plant measurements 4) automatic data link system from PMS and personal computer. (author). 4 tabs., 4 figs., 8 refs

  6. Development of adaptive core emulator for PMS-XRBP of CE type plant

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jae Seung; Zee, Sung Quun; Lee, Chung Chan; Lee, Ki Bog; Rhy, Hyo Sang; Chang, Jong Hwa; Lee, Young Ouk; Baek, Seung Min; Seo, Ho Joon

    1996-12-01

    The purpose of this report is to develop ONED-based adaptive core emulator (ACE) for Korean Standard Nuclear Power Plant. This report is first year report and includes (1) augmentation of ONED94 I/O system (2) non-equilibrium xenon initialization for core transient simulation (3) ONED94 verification via plant measurements (4) automatic data link system from PMS and personal computer. (author). 4 tabs., 4 figs., 8 refs.

  7. A Multi-Physics simulation of the Reactor Core using CUPID/MASTER

    International Nuclear Information System (INIS)

    Lee, Jae Ryong; Cho, Hyoung Kyu; Yoon, Han Young; Cho, Jin Young; Jeong, Jae Jun

    2011-01-01

    KAERI has been developing a component-scale thermal hydraulics code, CUPID. The aim of the code is for multi-dimensional, multi-physics and multi-scale thermal hydraulics analysis. In our previous papers, the CUPID code has proved to be able to reproduce multidimensional thermal hydraulic analysis by validated with various conceptual problems and experimental data. For the numerical closure, it adopts a three dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer. For the multi-scale analysis, the CUPID is on progress to merge into system-scale thermal hydraulic code, MARS. In the present paper, a multi-physics simulation was performed by coupling the CUPID with three dimensional neutron kinetics code, MASTER. The MASTER is merged into the CUPID as a dynamic link library (DLL). The APR1400 reactor core during control rod drop/ejection accident was simulated as an example by adopting a porous media approach to employ fuel assembly. The following sections present the numerical modeling for the reactor core, coupling of the kinetics code, and the simulation results

  8. Simulations of cold nuclear matter at sub-saturation densities

    Energy Technology Data Exchange (ETDEWEB)

    Giménez Molinelli, P.A., E-mail: pagm@df.uba.ar [Departamento de Física, Facultad de Ciencias Exactas y Naturales, Universidad de Buenos Aires and IFIBA, CONICET, Ciudad Universitaria, Buenos Aires 1428 (Argentina); Nichols, J.I. [Departamento de Física, Facultad de Ciencias Exactas y Naturales, Universidad de Buenos Aires and IFIBA, CONICET, Ciudad Universitaria, Buenos Aires 1428 (Argentina); López, J.A. [Department of Physics, University of Texas at El Paso, El Paso, TX 79968 (United States); Dorso, C.O. [Departamento de Física, Facultad de Ciencias Exactas y Naturales, Universidad de Buenos Aires and IFIBA, CONICET, Ciudad Universitaria, Buenos Aires 1428 (Argentina)

    2014-03-01

    Ideal nuclear matter is expected to undergo a first order phase transition at the thermodynamic limit. At such phase transitions the size of density fluctuations (bubbles or droplets) scale with the size of the system. This means that simulations of nuclear matter at sub-saturation densities will inexorably suffer from what is vaguely referred to as “finite size effects”. It is usually thought that these finite size effects can be diminished by imposing periodic boundary conditions and making the system large enough, but as we show in this work, that is actually not the case at sub-saturation densities. In this paper we analyze the equilibrium configurations of molecular dynamics simulations of a classical model for symmetric ideal (uncharged) nuclear matter at sub-saturation densities and low temperatures, where phase coexistence is expected at the thermodynamic limit. We show that the most stable configurations in this density range are almost completely determined by artificial aspects of the simulations (i.e. boundary conditions) and can be predicted analytically by surface minimization. This result is very general and is shown to hold true for several well known semi-classical models of nuclear interaction and even for a simple Lennard-Jones potential. Also, in the limit of very large systems, when “small size” effects can be neglected, those equilibrium configurations seem to be restricted to a few structures reminiscent to the “Pasta Phases” expected in Neutron Star matter, but arising from a completely different origin: In Neutron Star matter, the non-homogeneous structures arise from a competition between nuclear and Coulomb interactions while for ideal nuclear matter they emerge from finite (yet not “small”) size effects. The role of periodic boundary conditions and finite size effects in Neutron Star matter simulations are reexamined.

  9. Application of a Russian nuclear reactor simulator VVER-1000

    International Nuclear Information System (INIS)

    Lopez-Peniche S, A.; Salazar S, E.

    2012-10-01

    The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

  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. 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)

  12. 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''

  13. Innovative real time simulation training and nuclear probabilistic risk assessment

    International Nuclear Information System (INIS)

    Reisinger, M.F.

    1991-01-01

    Operator errors have been an area of public concern for the safe operation of nuclear power plants since the TMI2 incident. Simply stated, nuclear plants are very complex systems and the public is skeptical of the operators' ability to comprehend and deal with the vast indications and complexities of potential nuclear power plant events. Prior to the TMI2 incident, operator errors and human factors were not included as contributing factors in the Probabilistic Risk Assessment (PRA) studies of nuclear power plant accidents. More recent efforts in nuclear risk assessment have addressed some of the human factors affecting safe nuclear plant operations. One study found four major factors having significant impact on operator effectiveness. This paper discusses human factor PRAs, new applications in simulation training and the specific potential benefits from simulation in promoting safer operation of future power plants as well as current operating power plants

  14. Radiation dose assessment in nuclear plants through virtual simulations using a game engine

    International Nuclear Information System (INIS)

    Jorge, Carlos A.F.; Mol, Antonio C. A.; Aghina, Mauricio Alves C.

    2008-01-01

    Full text: This paper reports an R and D which has the purpose of performing dose assessment of workers in nuclear plants, through virtual simulations using a game engine. The main objective of this R and D is to support the planning of operational and maintenance routines in nuclear plants, aiming to reduce the dose received by workers. Game engine is the core of a computer game, that is usually made independent of both the scenarios and the original applications, and thus can be adapted for any other purposes, including scientific or technological ones. Computer games have experienced a great development in the last years, regarding computer graphics, 3D image rendering and the representation of the physics needed for the virtual simulations, such as gravity effect and collision among virtual components within the games. Thus, researchers do not need to develop an entire platform for virtual simulations, what would be a hard work itself, but they can rather take advantage of such well developed platforms, adapting them for their own applications. The game engine used in this R and D is part of a computer game widely used, Unreal, that has its source code partially open, and can be pursued for low cost. A nuclear plant in our Institution, Argonauta research reactor, has been virtually modeled in 3D, and trainees can navigate virtually through it, with realistic walking velocity, and experiencing collision. The modified game engine computes and displays in real-time the dose received by a virtual person, the avatar, as it walks through the plant, from the radiation dose rate distribution assigned to the virtual environment. In the beginning of this R and D, radiation dose rate measurements were previously collected by the radiological protection service, and input off-line to the game engine. Currently, on-line measurements can be also input to it, by taking advantage of the game's networking capabilities. A real radiation monitor has been used to collect real

  15. Equipment and performance upgrade of compact nuclear simulator

    International Nuclear Information System (INIS)

    Park, Jae Chang; Kwon, Kee Choon; Cha, Hyung Ho; Park, Won Man; Song, Soon Ja; Seh, Kyung Won; Joo, Young Chang

    1998-12-01

    In this project, the techniques to upgrade obsolescent nuclear power plant simulation were developed: the simulation code operated in Macro-VAX II has been transplanted to the HP workstation: commercial equipment PLC was adopted for the interface system: the protection logic was developed in the prototype level: and the display system was strengthened. These techniques can be directly used to upgrade other obsolescent simulators

  16. Equipment and performance upgrade of compact nuclear simulator

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Chang; Kwon, Kee Choon; Cha, Hyung Ho; Park, Won Man; Song, Soon Ja; Seh, Kyung Won; Joo, Young Chang

    1998-12-01

    In this project, the techniques to upgrade obsolescent nuclear power plant simulation were developed: the simulation code operated in Macro-VAX II has been transplanted to the HP workstation: commercial equipment PLC was adopted for the interface system: the protection logic was developed in the prototype level: and the display system was strengthened. These techniques can be directly used to upgrade other obsolescent simulators.

  17. 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

  18. A method of simulating and visualizing nuclear reactions

    International Nuclear Information System (INIS)

    Atwood, C.H.; Paul, K.M.

    1994-01-01

    Teaching nuclear reactions to students is difficult because the mechanisms are complex and directly visualizing them is impossible. As a teaching tool, the authors have developed a method of simulating nuclear reactions using colliding water droplets. Videotaping of the collisions, taken with a high shutter speed camera and run frame-by-frame, shows details of the collisions that are analogous to nuclear reactions. The method for colliding the water drops and videotaping the collisions are shown