WorldWideScience

Sample records for reactor lattice calculations

  1. Methods for thermal reactor lattice calculations

    International Nuclear Information System (INIS)

    Schneider, A.

    1976-12-01

    The American code HAMMER and the British code WIMS, for the analysis of thermal reactor lattices, have been investigated. The primary objective of this investigation was to identify the causes for the discrepancies that exist between the calculated and the experimentally determined reactivity of clean critical experiments. Three phases have been undertaken in the research: (a) Detailed comparison between the group cross-sections used by the codes; (b) Definition of the various approximations incorporated into the codes; (c) Comparison between the values of a variety of reaction rates calculated by the two codes. It was concluded that the main cause of discrepancy between calculations and experiments is due to data inaccuracies, while approximations introduced in solving the transport equation are of smaller importance

  2. WIMSD5, Deterministic Multigroup Reactor Lattice Calculations

    International Nuclear Information System (INIS)

    2004-01-01

    1 - Description of program or function: The Winfrith improved multigroup scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered the choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a successor version of WIMS-D/4. 2 - Method of solution: The treatment of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of temperatures. The collision theory procedure gives accurate spectrum computations in the 69 groups of the library for the principal regions of the lattice using a simplified geometric representation of complicated lattice cells. The computed spectra are then used for the condensation of cross-sections to the number of groups selected for solution of the transport equation in detailed geometry. Solution of the transport equation is provided either by use of the Carlson DSN method or by collision probability methods. Leakage calculations including an allowance for streaming asymmetries may be made using either diffusion theory or the more elaborate B1-method. The output of the code provides Eigenvalues for the cases where a simple buckling mode is applicable or cell-averaged parameters for use in overall reactor calculations. Various reaction rate edits are provided for direct comparison with experimental measurements. 3 - Restrictions on the complexity of

  3. Application of WIMSD-4 for ''MARIA'' reactor lattice calculations

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.

    1993-12-01

    A general description of the WIMSD-4 lattice code is given with the emphasis on available geometrical models. The difficulties encountered while modelling reactor lattices with the tubular type fuel elements are explained. Then the analysis of code options allowing to overcome these difficulties is carried out. Eventually, recommendations of options and input parameters for calculations of MARIA reactor lattice with satisfactory accuracy are given. During the work a set of modifications had to be introduced leading to a new code version called WIMS-S. Another version, under the name WIMS-T has been developed to allow for burnup calculations of the MARIA reactor lattice with improved resonance approach. (author). 14 refs, 6 figs, 10 tabs

  4. On the thermal scattering law data for reactor lattice calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Mattes, M.

    2004-01-01

    Thermal scattering law data for hydrogen bound in water, hydrogen bound in zirconium hydride and deuterium bound in heavy water have been re-evaluated. The influence of the thermal scattering law data on critical lattices has been studied with detailed Monte Carlo calculations and a summary of results is presented for a numerical benchmark and for the TRIGA reactor benchmark. Systematics for a large sequence of benchmarks analysed with the WIMS-D lattice code are also presented. (author)

  5. Calculation methods for advanced concept light water reactor lattices

    International Nuclear Information System (INIS)

    Carmona, S.

    1986-01-01

    In the last few years s several advanced concepts for fuel rod lattices have been studied. Improved fuel utilization is one of the major aims in the development of new fuel rod designs and lattice modifications. By these changes s better performance in fuel economics s fuel burnup and material endurance can be achieved in the frame of the well-known basic Light Water Reactor technology. Among the new concepts involved in these studies that have attracted serious attention are lattices consisting of arrays of annular rods duplex pellet rods or tight multicells. These new designs of fuel rods and lattices present several computational problems. The treatment of resonance shielded cross sections is a crucial point in the analyses of these advanced concepts . The purpose of this study was to assess adequate approximation methods for calculating as accurately as possible, resonance shielding for these new lattices. Although detailed and exact computational methods for the evaluation of the resonance shielding in these lattices are possible, they are quite inefficient when used in lattice codes. The computer time and memory required for this kind of computations are too large to be used in an acceptable routine manner. In order to over- come these limitations and to make the analyses possible with reasonable use of computer resources s approximation methods are necessary. Usual approximation methods, for the resonance energy regions used in routine lattice computer codes, can not adequately handle the evaluation of these new fuel rod lattices. The main contribution of the present work to advanced lattice concepts is the development of an equivalence principle for the calculation of resonance shielding in the annular fuel pellet zone of duplex pellets; the duplex pellet in this treatment consists of two fuel zones with the same absorber isotope in both regions. In the transition from a single duplex rod to an infinite array of this kind of fuel rods, the similarity of the

  6. Introduction to reactor lattice calculations by the WIMSD code

    International Nuclear Information System (INIS)

    Kulikowska, T.

    1998-01-01

    The present report is based on lectures delivered at the Workshop on Nuclear Reaction Data and Nuclear Reactors: Physics, Design and Safety hold in International Centre of Theoretical Physics, Trieste, in March 1998. The main goal of the set of lectures was to give the basis of reactor physics calculations for participants working on nuclear data.The last lectures, devoted to WIMS including the WIMSD code users. Following this general line the material is divided into three parts: The first part includes a short description of neutron transport phenomena limited to those definitions that are necessary to understand the approach to practical solution of the problem given in the second part on reactor lattice transport calculations. The detailed discussion of the neutron cross sections has been skipped as this subject has been treated in detail by other lectures. In the third part those versions of the well-known WIMSD code which are distributed by NEA Data Bank are described. The general structure of the code is given supplied in a more detailed description of aspects being the most common points of misunderstanding for the code users. (author)

  7. EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

    International Nuclear Information System (INIS)

    Jones, D.B.

    1986-01-01

    EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

  8. MOSRA-SRAC. Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA

    International Nuclear Information System (INIS)

    Okumura, Keisuke

    2015-10-01

    MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations. A CD-ROM is attached as an appendix. (J.P.N.)

  9. Neutronic calculations of hexagonal lattice nuclear reactors: Modelling of the CAREM-25 reactor

    International Nuclear Information System (INIS)

    Pacio, Julio Cesar

    2008-01-01

    This work was carried out in the frame of the Cnea CAREM-25 project (Central Argentina de Elementos Modulares).This project involves the development and construction of an argentinian design nuclear reactor for producing electricity. It's a PWR type (light water moderated and enriched U02 fueled) integrated reactor in an hexagonal lattice.The total power of this prototype is 100 MW thermal. In this frame, the main objective of this work is to consolidate and validate a neutronic line of calculus which can be applied to the CAREM-25 core.At a first analysis at cell level, the different fuel elements were modeled with the Dragon code, obtaining homogenised and condensed cross sections.Then a core level analysis with the Puma code was performed at full power condition and room temperature. A comparison of the obtained results is needed.For this reason, a Monte Carlo analysis (at room temperature) was performed.Also a validation of the Dragon code was carried out on the base of experimental data of WWER type lattices (similars to CAREM).The confidence on the results is then granted and their uncertainties were quantified.The Dragon-Puma line of calculus is then established and the main objective of this work is achieved. A full neutronic analysis should be followed by thermohydraulics calculations in an iterative procedure, and it would be the objective of future works.Finally, a burnup analysis was performed, at cell and core level.The design condition for extraction burnup and fuel cycle duration were verified. [es

  10. Accuracy of cell calculation methods used for analysis of high conversion light water reactor lattice

    International Nuclear Information System (INIS)

    Jeong, Chang-Joon; Okumura, Keisuke; Ishiguro, Yukio; Tanaka, Ken-ichi

    1990-01-01

    Validation tests were made for the accuracy of cell calculation methods used in analyses of tight lattices of a mixed-oxide (MOX) fuel core in a high conversion light water reactor (HCLWR). A series of cell calculations was carried out for the lattices referred from an international HCLWR benchmark comparison, with emphasis placed on the resonance calculation methods; the NR, IR approximations, the collision probability method with ultra-fine energy group. Verification was also performed for the geometrical modelling; a hexagonal/cylindrical cell, and the boundary condition; mirror/white reflection. In the calculations, important reactor physics parameters, such as the neutron multiplication factor, the conversion ratio and the void coefficient, were evaluated using the above methods for various HCLWR lattices with different moderator to fuel volume ratios, fuel materials and fissile plutonium enrichments. The calculated results were compared with each other, and the accuracy and applicability of each method were clarified by comparison with continuous energy Monte Carlo calculations. It was verified that the accuracy of the IR approximation became worse when the neutron spectrum became harder. It was also concluded that the cylindrical cell model with the white boundary condition was not so suitable for MOX fuelled lattices, as for UO 2 fuelled lattices. (author)

  11. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    1999-01-01

    The present lecture has a main goal to show how the transport lattice calculations are realised in a standard computer code. This is illustrated on the example of the WIMSD code, belonging to the most popular tools for reactor calculations. Most of the approaches discussed here can be easily modified to any other lattice code. The description of the code assumes the basic knowledge of reactor lattice, on the level given in the lecture on 'Reactor lattice transport calculations'. For more advanced explanation of the WIMSD code the reader is directed to the detailed descriptions of the code cited in References. The discussion of the methods and models included in the code is followed by the generally used homogenisation procedure and several numerical examples of discrepancies in calculated multiplication factors based on different sources of library data. (author)

  12. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    2001-01-01

    The description of reactor lattice codes is carried out on the example of the WIMSD-5B code. The WIMS code in its various version is the most recognised lattice code. It is used in all parts of the world for calculations of research and power reactors. The version WIMSD-5B is distributed free of charge by NEA Data Bank. The description of its main features given in the present lecture follows the aspects defined previously for lattice calculations in the lecture on Reactor Lattice Transport Calculations. The spatial models are described, and the approach to the energy treatment is given. Finally the specific algorithm applied in fuel depletion calculations is outlined. (author)

  13. Fission product model for lattice calculation of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Iijima, S.; Yoshida, T.; Yamamoto, T.

    1988-01-01

    A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

  14. Review of the Lattice Calculations for the CAREM-25 Reactor with Agincd as Absorber Material

    International Nuclear Information System (INIS)

    Zamonsky, Oscar

    2000-01-01

    In this work we compare some models to calculate the fuel elements of the CAREM-25 reactor at lattice level.In particular, we analyze the sensibility of the infinite multiplication factor and the peaking factor to several models and we propose the more accurate one for further calculations.The analysis is made for the cross sections library, the spatial discretization of the fuel element, the length of the burnup steps, the fuel temperature, and the coolant temperature and density.We also analyze several ways to model the AgInCd absorbers

  15. Lattice cell burnup calculation

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1977-01-01

    Accurate burnup prediction is a key item for design and operation of a power reactor. It should supply information on isotopic changes at each point in the reactor core and the consequences of these changes on the reactivity, power distribution, kinetic characters, control rod patterns, fuel cycles and operating strategy. A basic stage in the burnup prediction is the lattice cell burnup calculation. This series of lectures attempts to give a review of the general principles and calculational methods developed and applied in this area of burnup physics

  16. A comparison of neutron resonance absorption in thermal reactor lattices in the AUS neutronics code system with Monte Carlo calculations

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1985-08-01

    The calculation of resonance shielding by the subgroup method, as incorporated in the MIRANDA module of the AUS neutronics code system, is compared with Monte Carlo calculatons for a number of thermal reactor lattices. For the large range of single rod and rod cluster lattices considered, AUS results for resonance absorption were high by up to two per cent

  17. A nodal method of calculating power distributions for LWR-type reactors with square fuel lattices

    International Nuclear Information System (INIS)

    Hoeglund, Randolph.

    1980-06-01

    A nodal model is developed for calculating the power distribution in the core of a light water reactor with a square fuel lattice. The reactor core is divided into a number of more or less cubic nodes and a nodal coupling equation, which gives the thermal power density in one node as a function of the power densities in the neighbour nodes, is derived from the neutron diffusion equations for two energy groups. The three-dimensional power distribution can be computed iteratively using this coupling equation, for example following the point Jacobi, the Gauss-Seidel or the point successive overrelaxation scheme. The method has been included as the neutronic model in a reactor core simulation computer code BOREAS, where it is combined with a thermal-hydraulic model in order to make a simultaneous computation of the interdependent power and void distributions in a boiling water reactor possible. Also described in this report are a method for temporary one-dimensional iteration developed in order to accelerate the iterative solution of the problem and the Haling principle which is widely used in the planning of reloading operations for BWR reactors. (author)

  18. Benchmark calculation of APOLLO-2 and SLAROM-UF in a fast reactor lattice

    International Nuclear Information System (INIS)

    Hazama, T.

    2009-07-01

    A lattice cell benchmark calculation is carried out for APOLLO2 and SLAROM-UF on the infinite lattice of a simple pin cell featuring a fast reactor. The accuracy in k-infinity and reaction rates is investigated in their reference and standard level calculations. In the 1. reference level calculation, APOLLO2 and SLAROM-UF agree with the reference value of k-infinity obtained by a continuous energy Monte Carlo calculation within 50 pcm. However, larger errors are observed in a particular reaction rate and energy range. The major problem common to both codes is in the cross section library of 239 Pu in the unresolved energy range. In the 2. reference level calculation, which is based on the ECCO 1968 group structure, both results of k-infinity agree with the reference value within 100 pcm. The resonance overlap effect is observed by several percents in cross sections of heavy nuclides. In the standard level calculation based on the APOLLO2 library creation methodology, a discrepancy appears by more than 300 pcm. A restriction is revealed in APOLLO2. Its standard cross section library does not have a sufficiently small background cross section to evaluate the self shielding effect on 56 Fe cross sections. The restriction can be removed by introducing the mixture self-shielding treatment recently introduced to APOLLO2. SLAROM-UF original standard level calculation based on the JFS-3 library creation methodology is the best among the standard level calculations. Improvement from the SLAROM-UF standard level calculation is achieved mainly by use of a proper weight function for light or intermediate nuclides. (author)

  19. Determination of space-energy distribution of resonance neutrons in reactor lattice cell and calculation of resonance integrals

    International Nuclear Information System (INIS)

    Zmijarevic, I.

    1980-01-01

    Space-energy distribution of resonance neutrons in reactor lattice cell was determined by solving the Boltzmann equation by spherical harmonics method applying P-3 approximation. Computer code SPLET used for these calculations is described. Resonance absorption and calculation of resonance integrals are described as well. Effective resonance integral values for U-238 resonance at 6.7 Ev are calculated for heavy water reactor cell with metal, oxide and carbide fuel elements

  20. LWR-WIMS, a computer code for light water reactor lattice calculations

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1982-06-01

    LMR-WIMS is a comprehensive scheme of computation for studying the reactor physics aspects and burnup behaviour of typical lattices of light water reactors. This report describes the physics methods that have been incorporated in the code, and the modifications that have been made since the code was issued in 1972. (U.K.)

  1. SPLET - A program for calculating the space-lethargy distribution of epithermal neutrons in a reactor lattice cell

    International Nuclear Information System (INIS)

    Matausek, M.V.; Zmijatevic, I.

    1981-01-01

    A procedure to solve the space-single-lethargy dependent transport equation for epithermal neutrons in a cylindricised multi-region reactor lattice cell has been developed and proposed in the earlier papers. Here, the computational algorithm is comprised and the computing program SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, as well as the related integral quantities as reaction rates and resonance integrals, is described. (author)

  2. Calculational methods for lattice cells

    International Nuclear Information System (INIS)

    Askew, J.R.

    1980-01-01

    At the current stage of development, direct simulation of all the processes involved in the reactor to the degree of accuracy required is not an economic proposition, and this is achieved by progressive synthesis of models for parts of the full space/angle/energy neutron behaviour. The split between reactor and lattice calculations is one such simplification. Most reactors are constructed of repetitions of similar geometric units, the fuel elements, having broadly similar properties. Thus the provision of detailed predictions of their behaviour is an important step towards overall modelling. We shall be dealing with these lattice methods in this series of lectures, but will refer back from time to time to their relationship with overall reactor calculation The lattice cell is itself composed of somewhat similar sub-units, the fuel pins, and will itself often rely upon a further break down of modelling. Construction of a good model depends upon the identification, on physical and mathematical grounds, of the most helpful division of the calculation at this level

  3. Homogenization theory in reactor lattices

    International Nuclear Information System (INIS)

    Benoist, P.

    1986-02-01

    The purpose of the theory of homogenization of reactor lattices is to determine, by the mean of transport theory, the constants of a homogeneous medium equivalent to a given lattice, which allows to treat the reactor as a whole by diffusion theory. In this note, the problem is presented by laying emphasis on simplicity, as far as possible [fr

  4. Validation of CENDL and JEFF evaluated nuclear data files for TRIGA calculations through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors

    International Nuclear Information System (INIS)

    Uddin, M.N.; Sarker, M.M.; Khan, M.J.H.; Islam, S.M.A.

    2009-01-01

    The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69-group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.

  5. Validation study of SRAC2006 code system based on evaluated nuclear data libraries for TRIGA calculations by benchmarking integral parameters of TRX and BAPL lattices of thermal reactors

    International Nuclear Information System (INIS)

    Khan, M.J.H.; Sarker, M.M.; Islam, S.M.A.

    2013-01-01

    Highlights: ► To validate the SRAC2006 code system for TRIGA neutronics calculations. ► TRX and BAPL lattices are treated as standard benchmarks for this purpose. ► To compare the calculated results with experiment as well as MCNP values in this study. ► The study demonstrates a good agreement with the experiment and the MCNP results. ► Thus, this analysis reflects the validation study of the SRAC2006 code system. - Abstract: The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations

  6. Lattice calculations in gauge theory

    International Nuclear Information System (INIS)

    Rebbi, C.

    1985-01-01

    The lattice formulation of quantum gauge theories is discussed as a viable technique for quantitative studies of nonperturbative effects in QCD. Evidence is presented to ascertain that whole classes of lattice actions produce a universal continuum limit. Discrepancies between numerical results from Monto Carlo simulations for the pure gauge system and for the system with gauge and quark fields are discussed. Numerical calculations for QCD require very substantial computational resources. The use of powerful vector processors of special purpose machines, in extending the scope and magnitude or the calculations is considered, and one may reasonably expect that in the near future good quantitative predictions will be obtained for QCD

  7. Computer programs for lattice calculations

    International Nuclear Information System (INIS)

    Keil, E.; Reich, K.H.

    1984-01-01

    The aim of the workshop was to find out whether some standardisation could be achieved for future work in this field. A certain amount of useful information was unearthed, and desirable features of a ''standard'' program emerged. Progress is not expected to be breathtaking, although participants (practically from all interested US, Canadian and European accelerator laboratories) agreed that the mathematics of the existing programs is more or less the same. Apart from the NIH (not invented here) effect, there is a - to quite some extent understandable - tendency to stay with a program one knows and to add to it if unavoidable rather than to start using a new one. Users of the well supported program TRANSPORT (designed for beam line calculations) would prefer to have it fully extended for lattice calculations (to some extent already possible now), while SYNCH users wish to see that program provided with a user-friendly input, rather than spending time and effort for mastering a new program

  8. Reactor dynamics calculations

    International Nuclear Information System (INIS)

    Devooght, J.; Lefvert, T.; Stankiewiez, J.

    1981-01-01

    This chapter deals with the work done in reactor dynamics within the Coordinated Research Program on Transport Theory and Advanced Reactor Calculations by three groups in Belgium, Poland, Sweden and Italy. Discretization methods in diffusion theory, collision probability methods in time-dependent neutron transport and singular perturbation method are represented in this paper

  9. HETERO code, heterogeneous procedure for reactor calculation

    International Nuclear Information System (INIS)

    Jovanovic, S.M.; Raisic, N.M.

    1966-11-01

    This report describes the procedure for calculating the parameters of heterogeneous reactor system taking into account the interaction between fuel elements related to established geometry. First part contains the analysis of single fuel element in a diffusion medium, and criticality condition of the reactor system described by superposition of elements interactions. the possibility of performing such analysis by determination of heterogeneous system lattice is described in the second part. Computer code HETERO with the code KETAP (calculation of criticality factor η n and flux distribution) is part of this report together with the example of RB reactor square lattice

  10. Summary report on the international comparison of NEACRP burnup benchmark calculations for high conversion light water reactor lattices

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Ishiguro, Yukio; Takano, Hideki

    1988-10-01

    The results of the NEACRP HCLWR cell burnup benchmark calculations are summarized in this report. Fifteen organizations from eight countries participated in this benchmark and submitted twenty solutions. Large differences are still observed among the calculated values of void reactivities and conversion ratios. These differences are mainly caused from the discrepancies in the reaction rates of U-238, Pu-239 and fission products. The physics problems related to these results are briefly investigated in the report. In the specialists' meeting on this benchmark calculations held in April 1988, it was recommended to perform continuous energy Monte Carlo calculations in order to obtain reference solutions for design codes. The conclusions resulted from the specialists' meeting are also presented. (author)

  11. Symmetries applied to reactor calculations

    International Nuclear Information System (INIS)

    Makai, M.

    1982-03-01

    Three problems of a reactor-calculational model are discussed with the help of symmetry considerations. 1/ A coarse mesh method applicable to any geometry is derived. It is shown that the coarse mesh solution can be constructed from a few standard boundary value problems. 2/ A second stage homogenization method is given based on the Bloch theorem. This ensures the continuity of the current and the flux at the boundary. 3/ The validity of the micro-macro separation is shown for heterogeneous lattices. A formula for the neutron density is derived for cell homogenization. (author)

  12. Calculation models for a nuclear reactor

    International Nuclear Information System (INIS)

    Tashanii, Ahmed Ali

    2010-01-01

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

  13. Calculating luminosity for a coupled Tevatron lattice

    International Nuclear Information System (INIS)

    Holt, J.A.; Martens, M.A.; Michelotti, L.; Goderre, G.

    1995-05-01

    The traditional formula for calculating luminosity assumes an uncoupled lattice and makes use of one-degree-of-freedom lattice functions, β H and β v , for relating transverse beam widths to emittances. Strong coupling requires changing this approach. It is simplest to employ directly the linear normal form coordinates of the one turn map. An equilibrium distribution in phase space is expressed as a function of the Jacobian's eigenvectors and beam size parameters or emittances. Using the equilibrium distributions an expression for the luminosity was derived and applied to the Tevatron lattice, which was coupled due to a quadrupole roll

  14. Lattice calculation of nonleptonic charm decays

    International Nuclear Information System (INIS)

    Simone, J.N.

    1991-11-01

    The decays of charmed mesons into two body nonleptonic final states are investigated. Weak interaction amplitudes of interest in these decays are extracted from lattice four-point correlation functions using a effective weak Hamiltonian including effects to order G f in the weak interactions yet containing effects to all orders in the strong interactions. The lattice calculation allows a quantitative examination of non-spectator processes in charm decays helping to elucidate the role of effects such as color coherence, final state interactions and the importance of the so called weak annihilation process. For D → Kπ, we find that the non-spectator weak annihilation diagram is not small, and we interpret this as evidence for large final state interactions. Moreover, there is indications of a resonance in the isospin 1/2 channel to which the weak annihilation process contributes exclusively. Findings from the lattice calculation are compared to results from the continuum vacuum saturation approximation and amplitudes are examined within the framework of the 1/N expansion. Factorization and the vacuum saturation approximation are tested for lattice amplitudes by comparing amplitudes extracted from lattice four-point functions with the same amplitude extracted from products of two-point and three-point lattice correlation functions arising out of factorization and vacuum saturation

  15. A Lattice Calculation of Parton Distributions

    International Nuclear Information System (INIS)

    Alexandrou, Constantia; Cichy, Krzysztof; Poznan Univ.; Drach, Vincent; Univ. of Southern Denmark, Odense; Garcia-Ramos, Elena; Humboldt-Universitaet, Berlin; Hadjiyiannakou, Kyriakos; Jansen, Karl; Steffens, Fernanda; Wiese, Christian

    2015-04-01

    We report on our exploratory study for the direct evaluation of the parton distribution functions from lattice QCD, based on a recently proposed new approach. We present encouraging results using N f =2+1+1 twisted mass fermions with a pion mass of about 370 MeV. The focus of this work is a detailed description of the computation, including the lattice calculation, the matching to an infinite momentum and the nucleon mass correction. In addition, we test the effect of gauge link smearing in the operator to estimate the influence of the Wilson line renormalization, which is yet to be done.

  16. APOLLO2 calculations of RBMK lattices

    International Nuclear Information System (INIS)

    Kalashnikov, D.

    1998-01-01

    The purpose of this study is to investigate the use of erbium as burnable poison in RBMK reactors. The neutronic code APOLLO2 has been used and a comparison with the Monte-Carlo code TRIPOLI2 has been made. The first chapter briefly presents the RBMK characteristics, the second chapter deals with the neutronic behaviour of a fuel assembly in an infinite lattice which is an important step in the modelling process. The third chapter presents the analysis of the use of erbium in typical elements of the RBMK lattice. A good agreement is obtained between the 2 codes except in the draining situations. Erbium appears to reduce the positive reactivity effect of the draining configuration. (A.C.)

  17. Theoretical Calculations of the Effect on Lattice Parameters of Emptying the Coolant Channels in a D{sub 2}O- Moderated and Cooled Natural Uranium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Weissglas, P [The Swedish State Power Board, Stockholm (Sweden)

    1960-11-15

    The purpose of the present study was to evaluate theoretically the effect of coolant boiling and subsequent void formation in a pressurized D{sub 2}O moderated and cooled reactor. The fuel rods were arranged in a cluster geometry and clad in Zr-2. The coolant was separated from the moderator by a Zr-2 shroud. In this geometry the following problems have been given special attention: l) calculation of the effective resonance integral, 2) thermal disadvantage factors, 3) fast fission effects, 4) leakage effects, 5) changes in epithermal absorption. No account has up to now been taken of the variation of these effects with position in the reactor and burnup. Some comparisons of the theoretical methods and measurements have been attempted. It is concluded that at the present time it is not possible to calculate the void coefficient with any accuracy but it may be possible to give an upper limit from theoretical consideration.

  18. A proposal for the calculation of the critical buckling of a PWR or undermoderated lattice

    International Nuclear Information System (INIS)

    Benoist, P.

    1989-01-01

    A method improving the calculation of the critical buckling of a PWR or undermorated lattice is proposed. This method takes into account the lattice heterogeneity with more detail than the existing ones; it lies on some approximations. The method requires a relatively small inplementational effort. It could be used in the calculation of fast reactors [fr

  19. Calculating lattice thermal conductivity: a synopsis

    Science.gov (United States)

    Fugallo, Giorgia; Colombo, Luciano

    2018-04-01

    We provide a tutorial introduction to the modern theoretical and computational schemes available to calculate the lattice thermal conductivity in a crystalline dielectric material. While some important topics in thermal transport will not be covered (including thermal boundary resistance, electronic thermal conduction, and thermal rectification), we aim at: (i) framing the calculation of thermal conductivity within the general non-equilibrium thermodynamics theory of transport coefficients, (ii) presenting the microscopic theory of thermal conduction based on the phonon picture and the Boltzmann transport equation, and (iii) outlining the molecular dynamics schemes to calculate heat transport. A comparative and critical addressing of the merits and drawbacks of each approach will be discussed as well.

  20. Argosy 4 - A programme for lattice calculations

    International Nuclear Information System (INIS)

    MacDougall, J.D.

    1965-07-01

    This report contains a detailed description of the methods of calculation used in the Argosy 4 computer programme, and of the input requirements and printed results produced by the programme. An outline of the physics of the Argosy method is given. Section 2 describes the lattice calculation, including the burn up calculation, section 3 describes the control rod calculation and section 4 the reflector calculation. In these sections the detailed equations solved by the programme are given. In section 5 input requirements are given, and in section 6 the printed output obtained from an Argosy calculation is described. In section 7 are noted the principal differences between Argosy 4 and earlier versions of the Argosy programme

  1. Parallel computer calculation of quantum spin lattices

    International Nuclear Information System (INIS)

    Lamarcq, J.

    1998-01-01

    Numerical simulation allows the theorists to convince themselves about the validity of the models they use. Particularly by simulating the spin lattices one can judge about the validity of a conjecture. Simulating a system defined by a large number of degrees of freedom requires highly sophisticated machines. This study deals with modelling the magnetic interactions between the ions of a crystal. Many exact results have been found for spin 1/2 systems but not for systems of other spins for which many simulation have been carried out. The interest for simulations has been renewed by the Haldane's conjecture stipulating the existence of a energy gap between the ground state and the first excited states of a spin 1 lattice. The existence of this gap has been experimentally demonstrated. This report contains the following four chapters: 1. Spin systems; 2. Calculation of eigenvalues; 3. Programming; 4. Parallel calculation

  2. Lattice QCD Calculation of Nucleon Structure

    International Nuclear Information System (INIS)

    Liu, Keh-Fei; Draper, Terrence

    2016-01-01

    It is emphasized in the 2015 NSAC Long Range Plan that 'understanding the structure of hadrons in terms of QCD's quarks and gluons is one of the central goals of modern nuclear physics.' Over the last three decades, lattice QCD has developed into a powerful tool for ab initio calculations of strong-interaction physics. Up until now, it is the only theoretical approach to solving QCD with controlled statistical and systematic errors. Since 1985, we have proposed and carried out first-principles calculations of nucleon structure and hadron spectroscopy using lattice QCD which entails both algorithmic development and large-scale computer simulation. We started out by calculating the nucleon form factors -- electromagnetic, axial-vector, ?NN, and scalar form factors, the quark spin contribution to the proton spin, the strangeness magnetic moment, the quark orbital angular momentum, the quark momentum fraction, and the quark and glue decomposition of the proton momentum and angular momentum. The first round of calculations were done with Wilson fermions in the 'quenched' approximation where the dynamical effects of the quarks in the sea are not taken into account in the Monte Carlo simulation to generate the background gauge configurations. Beginning in 2000, we have started implementing the overlap fermion formulation into the spectroscopy and structure calculations. This is mainly because the overlap fermion honors chiral symmetry as in the continuum. It is going to be more and more important to take the symmetry into account as the simulations move closer to the physical point where the u and d quark masses are as light as a few MeV only. We began with lattices which have quark masses in the sea corresponding to a pion mass at ~ 300 MeV and obtained the strange form factors, charm and strange quark masses, the charmonium spectrum and the D_s meson decay constant f_D__s, the strangeness and charmness, the meson mass decomposition and the strange quark spin from the

  3. Lattice QCD Calculation of Nucleon Structure

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Keh-Fei [University of Kentucky, Lexington, KY (United States). Dept. of Physics and Astronomy; Draper, Terrence [University of Kentucky, Lexington, KY (United States). Dept. of Physics and Astronomy

    2016-08-30

    It is emphasized in the 2015 NSAC Long Range Plan that "understanding the structure of hadrons in terms of QCD's quarks and gluons is one of the central goals of modern nuclear physics." Over the last three decades, lattice QCD has developed into a powerful tool for ab initio calculations of strong-interaction physics. Up until now, it is the only theoretical approach to solving QCD with controlled statistical and systematic errors. Since 1985, we have proposed and carried out first-principles calculations of nucleon structure and hadron spectroscopy using lattice QCD which entails both algorithmic development and large-scale computer simulation. We started out by calculating the nucleon form factors -- electromagnetic, axial-vector, πNN, and scalar form factors, the quark spin contribution to the proton spin, the strangeness magnetic moment, the quark orbital angular momentum, the quark momentum fraction, and the quark and glue decomposition of the proton momentum and angular momentum. The first round of calculations were done with Wilson fermions in the `quenched' approximation where the dynamical effects of the quarks in the sea are not taken into account in the Monte Carlo simulation to generate the background gauge configurations. Beginning in 2000, we have started implementing the overlap fermion formulation into the spectroscopy and structure calculations. This is mainly because the overlap fermion honors chiral symmetry as in the continuum. It is going to be more and more important to take the symmetry into account as the simulations move closer to the physical point where the u and d quark masses are as light as a few MeV only. We began with lattices which have quark masses in the sea corresponding to a pion mass at ~ 300 MeV and obtained the strange form factors, charm and strange quark masses, the charmonium spectrum and the Ds meson decay constant fDs, the strangeness and charmness, the meson mass

  4. Development of a computer code for neutronic calculations of a hexagonal lattice of nuclear reactor using the flux expansion nodal method

    Directory of Open Access Journals (Sweden)

    Mohammadnia Meysam

    2013-01-01

    Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.

  5. HELIOS calculations for UO2 lattice benchmarks

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1998-01-01

    Calculations for the ANS UO 2 lattice benchmark have been performed with the HELIOS lattice-physics code and six of its cross-section libraries. The results obtained from the different libraries permit conclusions to be drawn regarding the adequacy of the energy group structures and of the ENDF/B-VI evaluation for 238 U. Scandpower A/S, the developer of HELIOS, provided Los Alamos National Laboratory with six different cross section libraries. Three of the libraries were derived directly from Release 3 of ENDF/B-VI (ENDF/B-VI.3) and differ only in the number of groups (34, 89 or 190). The other three libraries are identical to the first three except for a modification to the cross sections for 238 U in the resonance range

  6. Effect of cosine current approximation in lattice cell calculations in cylindrical geometry

    International Nuclear Information System (INIS)

    Mohanakrishnan, P.

    1978-01-01

    It is found that one-dimensional cylindrical geometry reactor lattice cell calculations using cosine angular current approximation at spatial mesh interfaces give results surprisingly close to the results of accurate neutron transport calculations as well as experimental measurements. This is especially true for tight light water moderated lattices. Reasons for this close agreement are investigated here. By re-examining the effects of reflective and white cell boundary conditions in these calculations it is concluded that one major reason is the use of white boundary condition necessitated by the approximation of the two-dimensional reactor lattice cell by a one-dimensional one. (orig.) [de

  7. Reactor core performance calculating device

    International Nuclear Information System (INIS)

    Tominaga, Kenji; Bando, Masaru; Sano, Hiroki; Maruyama, Hiromi.

    1995-01-01

    The device of the present invention can calculate a power distribution efficiently at high speed by a plurality of calculation means while taking an amount of the reactor state into consideration. Namely, an input device takes data from a measuring device for the amount of the reactor core state such as a large number of neutron detectors disposed in the reactor core for monitoring the reactor state during operation. An input data distribution device comprises a state recognition section and a data distribution section. The state recognition section recognizes the kind and amount of the inputted data and information of the calculation means. The data distribution section analyzes the characteristic of the inputted data, divides them into a several groups, allocates them to each of the calculation means for the purpose of calculating the reactor core performance efficiently at high speed based on the information from the state recognition section. A plurality of the calculation means calculate power distribution of each of regions based on the allocated inputted data, to determine the power distribution of the entire reactor core. As a result, the reactor core can be evaluated at high accuracy and at high speed irrespective of the whole reactor core or partial region. (I.S.)

  8. Resonance shielding in thermal reactor lattices

    International Nuclear Information System (INIS)

    Rothenstein, W.; Taviv, E.; Aminpour, M.

    1982-01-01

    The theoretical foundations of a new methodology for the accurate treatment of resonance absorption in thermal reactor lattice analysis are presented. This methodology is based on the solution of the point-energy transport equation in its integral or integro-differential form for a heterogeneous lattice using detailed resonance cross-section profiles. The methodology is applied to LWR benchmark analysis, with emphasis on temperature dependence of resonance absorption during fuel depletion, spatial and mutual self-shielding, integral parameter analysis and treatment of cluster geometry. The capabilities of the OZMA code, which implements the new methodology are discussed. These capabilities provide a means against which simpler and more rapid resonance absorption algorithms can be checked. (author)

  9. Effective action calculation in lattice QCD

    International Nuclear Information System (INIS)

    Hoek, J.

    1983-01-01

    A method (called the effective action method) devised to make analytic calculations in Quantum Chromodynamics in the region of strong coupling is presented. First, the author deals with developing the calculation of a strong coupling expansion of the generating functional for gauge systems on a lattice with arbitrary sources. An accompanying manual describes the implementation of this calculation on a computer. The next step consists of substituting the expressions for the one-link free energies for a specific gauge group in the result of the previous calculation. This process of substitution, together with the replacement of the sources by a bilinear combination of fermion fields, is described for the group SU(3). More details on the implementation of the substitution scheme on a computer can be found in the accompanying manual. From the effective action thus obtained in terms of meson fields and baryon fields the Green functions of the theory can be derived. As an illustrative application the effective potential determining the vacuum expectation value of the meson field is calculated. (Auth.)

  10. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1970-04-01

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  11. Benchmarking lattice physics data and methods for boiling water reactor analysis

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Edenius, M.; Harris, D.R.; Hebert, M.J.; Kapitz, D.M.; Pilat, E.E.; VerPlanck, D.M.

    1983-01-01

    The objective of the work reported was to verify the adequacy of lattice physics modeling for the analysis of the Vermont Yankee BWR using a multigroup, two-dimensional transport theory code. The BWR lattice physics methods have been benchmarked against reactor physics experiments, higher order calculations, and actual operating data

  12. On calculation of lattice parameters of refractory metal solid solutions

    International Nuclear Information System (INIS)

    Barsukov, A.D.; Zhuravleva, A.D.; Pedos, A.A.

    1995-01-01

    Technique for calculating lattice periods of solid solutions is suggested. Experimental and calculation values of lattice periods of some solid solutions on the basis of refractory metals (V-Cr, Nb-Zr, Mo-W and other) are presented. Calculation error was correlated with experimental one. 7 refs.; 2 tabs

  13. Propagation calculation for reactor cases

    Energy Technology Data Exchange (ETDEWEB)

    Yang Yanhua [School of Power and Energy Engineering, Shanghai Jiao Tong Univ., Shanghai (China); Moriyama, K.; Maruyama, Y.; Nakamura, H.; Hashimoto, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-11-01

    The propagation of steam explosion for real reactor geometry and conditions are investigated by using the computer code JASMINE-pro. The ex-vessel steam explosion is considered, which is described as follow: during the accident of reactor core meltdown, the molten core melts a hole at the bottom of reactor vessel and causes the higher temperature core fuel being leaked into the water pool below reactor vessel. During the melt-water mixing interaction process, the high temperature melt evaporates the cool water at an extreme high rate and might induce a steam explosion. A steam explosion could experience first the premixing phase and then the propagation explosion phase. For a propagation calculation, we should know the information about the initial fragmentation time, the total melt mass, premixing region size, initial void fraction and distribution of the melt volume fraction, and so on. All the initial conditions used in this calculation are based on analyses from some simple assumptions and the observation from the experiments. The results show that the most important parameter for the initial condition of this phase is the total mass and its initial distribution. This gives the requirement for a premixing calculation. On the other hand, for higher melt volume fraction case, the fragmentation is strong so that the local pressure can exceed over the EOS maximum pressure of the code, which lead to the incorrect calculation or divergence of the calculation. (Suetake, M.)

  14. Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca

    International Nuclear Information System (INIS)

    Bosevski, T.; Altiparmakov, D.; Marinkovic, N.

    1974-01-01

    In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr

  15. The problem of reactivity and reaction-rate calculations for mixed graphite lattices

    International Nuclear Information System (INIS)

    Pitcher, H.H.W.

    1963-05-01

    The dependence of reactor physics quantities, such as η f and Pu239/U235 fission ratio, in a single cell on the environment of the cell, and the relationship of the reactivity of a mixed lattice to the reactivity of its components, in graphite-moderated reactors are investigated. In a particular case, a mixed lattice fuelled with uranium at 0 and 3000 MWD/Te showed at 8 cm. pitch a small but appreciable change (∼ 1%) in cell quantities, and at 25 cm. pitch a smaller change. It is found that the present method of calculating lattice reactivity, ignoring intercell effects, is probably adequate for standard-pitch metal-fuelled graphite-moderated systems. More general mixed-lattice systems, particularly if accurate values of cell quantities are required, may need special calculation techniques; these are discussed, and techniques adequate for most systems are presented. (author)

  16. Neutronic calculation of reactor cells

    International Nuclear Information System (INIS)

    Jaliff, J.O.

    1981-01-01

    Multigroup calculations of cylindrical pin cells were programmed, in Fortran IV, upon the basis of collision probabilities in each energy group. A rational approximation to the fuel-to-fuel collision probability in resonance groups was used. Together with the intermediate resonance theory, cross sections corrected for heterogeneity and absorber interactions were found. For the optimization of the program, the cell of a BWR reactor was taken as reference. Data for such a cell and the reactor's operating conditions are presented. PINCEL is a fast and flexible program, with checked results, around a 69-group library. (M.E.L.) [es

  17. Reactor calculations and nuclear information

    International Nuclear Information System (INIS)

    Lang, D.W.

    1977-12-01

    The relationship of sets of nuclear parameters and the macroscopic reactor quantities that can be calculated from them is examined. The framework of the study is similar to that of Usachev and Bobkov. The analysis is generalised and some properties required by common sense are demonstrated. The form of calculation permits revision of the parameter set. It is argued that any discrepancy between a calculation and measurement of a macroscopic quantity is more useful when applied directly to prediction of other macroscopic quantities than to revision of the parameter set. The mathematical technique outlined is seen to describe common engineering practice. (Author)

  18. Some approximate calculations in SU2 lattice mean field theory

    International Nuclear Information System (INIS)

    Hari Dass, N.D.; Lauwers, P.G.

    1981-12-01

    Approximate calculations are performed for small Wilson loops of SU 2 lattice gauge theory in mean field approximation. Reasonable agreement is found with Monte Carlo data. Ways of improving these calculations are discussed. (Auth.)

  19. Temperature variation of criticality of thermal reactor lattices

    International Nuclear Information System (INIS)

    Velner, S.; Rothenstein, W.

    1975-01-01

    Departures from the asymptotic mode in the experimental setup have been examined in detail for two assemblies, one exponential, the other critical. It was found that the flux shape differed noticeably from the asymptotic mode in the core region especially for the exponential assemblies. On the other hand the departure from the fundamental mode has very little effect on the change of material buckling with temperature. Results of the calculations and their comparison with experiment are presented. The variation of material buckling with temperature is the same for ENDF/B-II and for ENDF/B-IV data, both for asymptotic reactor theory and for the buckling values derived from the flux calculated with the SN code. The results obtained with ENDF/B-IV data for both lattices are shown. In the small exponential assembly the results derived from S-4 calculations are compared with experiment. In the critical assembly the ratio of U-238 to U-235 fissions delta 28 and the relative conversion ratio - the ratio of U-238 captures to U-235 fissions in the lattice compared with the same quantity in a thermal column - are also shown. In both cases the experimental change of buckling with temperature is smaller than the calculated change. (B.G.)

  20. IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments

    International Nuclear Information System (INIS)

    2003-01-01

    Description: B and W has performed and analysed a series of physics experiments basically concerned with the technology of heterogeneous reactors moderated and cooled by a variable mixture of heavy and light water. A reactor so moderated is termed Spectral Shift Control Reactor (S SCR). In the practical application of this concept, the moderator mixture is rich in heavy water at the beginning of core life, so a relatively large fraction of the neutrons are epithermal and are absorbed in the fertile material. As fuel is consumed, the moderator is diluted with light water. In this way the neutron spectrum is shifted, thereby increasing the proportion of thermal neutrons and the reactivity of the system. The general objective of the S SCR Basic Physics Program was to study the nuclear properties of rod lattices moderated by D 2 O-H 2 O mixtures. The volume ratio of moderator to non-moderator in all lattices was approximately 1.0, and the fuel was either 4%-enriched UO 2 clad in stainless steel or 93%-enriched UO 2 -ThO 2 (Nth/N 15) pellets clad in aluminum. The D 2 O concentration in the moderator ranged from zero to about 90 mole %. The experimental program includes critical experiments with both types of fuel, exponential experiments at room temperature with both types of fuel, exponential experiments at elevated temperatures with the 4%-enriched UO 2 fuel, and neutron age measurements in ThO 2 lattices. The theoretical program included the development of calculation methods applicable to these systems, and the analysis and correlation of the experimental data. A first report provides the results of critical experiments performed under the Spectral Shift Control Reactor Basic Physics Program. A second report documents experimental results and theoretical interpretation of a series of twenty uniform lattice critical experiments in which the neutron spectrum is varied over a fairly broad range. A third report addresses issues that bear on the problems associated with

  1. The development of a computer technique for the investigation of reactor lattice parameters

    International Nuclear Information System (INIS)

    Joubert, W.R.

    1982-01-01

    An integrated computer technique was developed whereby all the computer programmes needed to calculate reactor lattice parameters from basic neutron data, could be combined in one system. The theory of the computer programmes is explained in detail. Results are given and compared with experimental values as well as those calculated with a standard system

  2. Status and prospects for lattice calculations in heavy quark physics

    International Nuclear Information System (INIS)

    Wittig, H.; Forschungszentrum Juelich GmbH

    1996-06-01

    The current status of lattice calculation of weak matrix elements for heavy quark systems is reviewed. After an assessment of systematic errors in present simulations, results for the B meson decay constant, the B parameter B B and semi-leptonic heavy-to-light and heavy-to-heavy transitions are discussed. The final topic are lattice results for heavy baryon spectroscopy. (orig.)

  3. Program LATTICE for Calculation of Parameters of Targets with Heterogeneous (Lattice) Structure

    CERN Document Server

    Bznuni, S A; Soloviev, A G; Sosnin, A N

    2002-01-01

    Program LATTICE, with which help it is possible to describe lattice structure for the program complex CASCAD, is created in the C++ language. It is shown that for model-based electronuclear system on a basis of molten salt reactor with graphite moderator at transition from homogeneous structure to heterogeneous at preservation of a chemical compound there is a growth of k_{eff} by approximately 6 %.

  4. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  5. Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

    This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

  6. Interlaboratory computational comparisons of critical fast test reactor pin lattices

    International Nuclear Information System (INIS)

    Mincey, J.F.; Kerr, H.T.; Durst, B.M.

    1979-01-01

    An objective of the Consolidated Fuel Reprocessing Program's (CFRP) nuclear engineering group at Oak Ridge National Laboratory (ORNL) is to ensure that chemical equipment components designed for the reprocessing of spent LMFBR fuel (among other fuel types) are safe from a criticality standpoint. As existing data are inadequate for the general validation of computational models describing mixed plutonium--uranium oxide systems with isotopic compositions typical of LMFBR fuel, a program of critical experiments has been initiated at the Battelle Pacific Northwest Laboratories (PNL). The first series of benchmark experiments consisted of five square-pitched lattices of unirradiated Fast Test Reactor (FTR) fuel moderated and reflected by light water. Calculations of these five experiments have been conducted by both ORNL/CFRP and PNL personnel with the purpose of exploring how accurately various computational models will predict k/sub eff/ values for such neutronic systems and if differences between k/sub eff/ values obtained with these different models are significant

  7. Temperature effects studies in light water reactor lattices

    International Nuclear Information System (INIS)

    Erradi, Lahoussine.

    1982-02-01

    The CREOLE experiments performed in the EOLE critical facility located in the Nuclear Center of CADARACHE - CEA (UO 2 and UO 2 -PuO 2 lattice reactivity temperature coefficient continuous measurements between 20 0 C and 300 0 C; integral measurements by boron equivalent effect in the moderator; water density effects measurements with the use of over cladding aluminium tubes to remove moderator) allow to get an interesting and complete information on the temperature effects in the light water reactor lattices. A very elaborated calcurated scheme using the transport theory and the APOLLO cross sections library, has been developed. The analysed results of the whole lot of experiments show that the discrepancy between theory and experiment strongly depends on the temperature range and on the type of lattices considered. The error is mainly linked with the thermal spectrum effects. A study on the temperature coefficient sensitivity to the different cell neutron parameters has shown that only the shapes of the 235 U and 238 U thermal cross sections have enough weight and uncertainty margins to explain the observed experimental/calculation bias. Instead of arbitrarily fitting the identified wrong data on the calculation of the reactivity temperature coefficient we have defined a procedure of modification of the cross sections based on the consideration of the basic nuclear data: resonance parameters and associated statistic laws. The implementation of this procedure has led to propose new thermal cross sections sets for 235 U and 238 U consistent with the uncertainty margins associated with the previously accepted values and with some experimental data [fr

  8. Three dimensional diffusion calculations of nuclear reactors

    International Nuclear Information System (INIS)

    Caspo, N.

    1981-07-01

    This work deals with the three dimensional calculation of nuclear reactors using the code TRITON. The purposes of the work were to perform three-dimensional computations of the core of the Soreq nuclear reactor and of the power reactor ZION and to validate the TRITON code. Possible applications of the TRITON code in Soreq reactor calculations and in power reactor research are suggested. (H.K.)

  9. Dissecting Reactor Antineutrino Flux Calculations

    Science.gov (United States)

    Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.

    2017-09-01

    Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235U, 239Pu, 241Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In the present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238U contribution as well as the effective charge and the allowed shape assumption used in the conversion method. We observe that including a shape correction of about +6 % MeV-1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.

  10. Recursive integral equations with positive kernel for lattice calculations

    International Nuclear Information System (INIS)

    Illuminati, F.; Isopi, M.

    1990-11-01

    A Kirkwood-Salzburg integral equation, with positive defined kernel, for the states of lattice models of statistical mechanics and quantum field theory is derived. The equation is defined in the thermodynamic limit, and its iterative solution is convergent. Moreover, positivity leads to an exact a priori bound on the iteration. The equation's relevance as a reliable algorithm for lattice calculations is therefore suggested, and it is illustrated with a simple application. It should provide a viable alternative to Monte Carlo methods for models of statistical mechanics and lattice gauge theories. 10 refs

  11. RAHAB calculation of lattice parameters for CANDU-type lattices using Monte Carlo calculations for resolved resonance capture

    International Nuclear Information System (INIS)

    Craig, D.S.; Festarini, G.L.

    1986-07-01

    The Monte Carlo code, REPC, has been used to calculate resonance reaction rates for the thermal test lattices TRX-1 and MIT-4, and for the CRNL lattices ZEEP-1, 19 UO 2 and 37 UO 2 . These reaction rates were used in the RAHAB cell code to calculate k eff , conversion ratios, and fast fission ratios, for comparison with experimental values. The calculations used the cluster geometry for the 19-, 28-, and 37-element clusters. Calculations were also made using annular representations of the cluster for comparison of the rates with those obtained using the discrete ordinate code OZMA

  12. A novel lattice energy calculation technique for simple inorganic crystals

    Energy Technology Data Exchange (ETDEWEB)

    Kaya, Cemal [Department of Chemistry, Faculty of Science, Cumhuriyet University, 58140 Sivas (Turkey); Kaya, Savaş, E-mail: savaskaya@cumhuriyet.edu.tr [Department of Chemistry, Faculty of Science, Cumhuriyet University, 58140 Sivas (Turkey); Banerjee, Priyabrata [Surface Engineering and Tribology Group, CSIR-Central Mechanical Engineering Research Institute, Mahatma Gandhi Avenue, Durgapur 713209 (India)

    2017-01-01

    In this pure theoretical study, a hitherto unexplored equation based on Shannon radii of the ions forming that crystal and chemical hardness of any crystal to calculate the lattice energies of simple inorganic ionic crystals has been presented. To prove the credibility of this equation, the results of the equation have been compared with experimental outcome obtained from Born-Fajans-Haber- cycle which is fundamentally enthalpy-based thermochemical cycle and prevalent theoretical approaches proposed for the calculation of lattice energies of ionic compounds. The results obtained and the comparisons made have demonstrated that the new equation is more useful compared to other theoretical approaches and allows to exceptionally accurate calculation of lattice energies of inorganic ionic crystals without doing any complex calculations.

  13. A critical review of homogenization techniques in reactor lattices

    International Nuclear Information System (INIS)

    Benoist, P.

    1983-01-01

    The determination of the shape of the neutron flux in a whole reactor is, at the time being, a much too complex problem to be treated by transport theory. Since the earlier times of reactor theory, the necessity appeared to solve the problem in two steps. First the reactor is divided into zones, each of them forming a regular lattice. In each of these zones, homogenized parameters are determined by transport theory, in order to define an equivalent smeared medium. In a second step, these parameters are introduced in a diffusion theory scheme in order to treat the reactor as a whole. This is the homogenization procedure. 14 refs

  14. A critical review of homogenization techniques in reactor lattices

    International Nuclear Information System (INIS)

    Benoist, P.

    1983-01-01

    The determination of the shape of the neutron flux in a whole reactor is, at the time being, a much too complex problem to be treated by transport theory. Since the earlier times of reactor theory, the necessity appeared to solve the problem in two steps. First the reactor is divided into zones, each of them forming a regular lattice. In each of these zones, homogenized parameters are determined by transport theory, in order to define an equivalent smeared medium. In a second step, these parameters are introduced in a diffusion theory scheme in order to treat the reactor as a whole. This is the homogenization procedure

  15. theory and calculation of the design of nuclear reactor

    International Nuclear Information System (INIS)

    Refaat, R.A.

    1994-01-01

    For the sake of formation of a complete general code for nuclear power reactor design, this thesis deals with a great part of this code. the code links the solution of the neutron integral transport equation by the multigroup treatment (76 energy groups) for the calculation of the reactor cell parameters by the fuel management program that solves the neutron diffusion equation inside a large number of nuclear fuel assemblies. the lattice cell code is modified to accommodate the calculation of lattice cell parameters for more than one enrichment ( one after the other). it is also modified to calculate the burn up parameters using unequal time steps. these two modifications are complicated but necessary for the link between the cell program and fuel management program. the comparison between the results of the fitted cross sections and that given by the cell calculations shows the necessity of using the cell code cross sections. this is also necessary for the sake of generality for any type of reactors. the comparison for the fuel management calculation depending on fitted data and that depending on cell calculation data insures the necessity for using the cell data i.e. insures the necessity of linking the cell calculation program by the fuel management program

  16. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

    Bouveret, F.

    2001-01-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  17. DRAGON, Reactor Cell Calculation System with Burnup

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: DRAGON is a collection of models to simulate the neutronic behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations which can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. The user must supply cross sections. DRAGON can access directly standard microscopic cross-section libraries in the following formats: DRAGON, MATXS (TRANSX-CTR), WIMSD4, WIMS-AECL, and APOLLO. It has the capability of exchanging macroscopic and microscopic cross-section libraries with a code such as PSR-0206/TRANSX-CTR or PSR-0317/TRANSX-2 by the use of the GOXS and ISOTXS format files. Macroscopic cross sections can also be read in DRAGON via the input data stream. 2 - Method of solution: DRAGON contains a multigroup iterator conceived to control a number of different algorithms for the solution of the neutron transport equation. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are included in a source term. The current version, DRAGON 9 71124 (Release 3.02), which was released in January 1998, contains three such algorithms. The JPM option solves the integral transport equation using the interface current method applied to homogeneous blocks; the SYBIL option solves the integral transport equation using the collision probability method for simple one-dimensional (1-D) or two-dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; and the

  18. Effect of lattice-level adjoint-weighting on the kinetics parameters of CANDU reactors

    International Nuclear Information System (INIS)

    Nichita, Eleodor

    2009-01-01

    Space-time kinetics calculations for CANDU reactors are routinely performed using the Improved Quasistatic (IQS) method. The IQS method calculates kinetics parameters such as the effective delayed-neutron fraction and generation time using adjoint weighting. In the current implementation of IQS, the direct flux, as well as the adjoint, is calculated using a two-group cell-homogenized reactor model which is inadequate for capturing the effect of the softer energy spectrum of the delayed neutrons. Additionally, there may also be fine spatial effects that are lost because the intra-cell adjoint shape is ignored. The purpose of this work is to compare the kinetics parameters calculated using the two-group cell-homogenized model with those calculated using lattice-level fine-group heterogeneous adjoint weighting and to assess whether the differences are large enough to justify further work on incorporating lattice-level adjoint weighting into the IQS method. A second goal is to evaluate whether the use of a fine-group cell-homogenized lattice-level adjoint, such as is the current practice for Light Water Reactors (LWRs), is sufficient to capture the lattice effects in question. It is found that, for CANDU lattices, the generation time is almost unaffected by the type of adjoint used to calculate it, but that the effective delayed-neutron fraction is affected by the type of adjoint used. The effective delayed-neutron fraction calculated using the two-group cell-homogenized adjoint is 5.2% higher than the 'best' effective delayed-neutron fraction value obtained using the detailed lattice-level fine-group heterogeneous adjoint. The effective delayed-neutron fraction calculated using the fine-group cell-homogenized adjoint is only 1.7% higher than the 'best' effective delayed-neutron fraction value but is still not equal to it. This situation is different from that encountered in LWRs where weighting by a fine-group cell-homogenized adjoint is sufficient to calculate the

  19. Investigation of fuel lattice pitch changes influence on reactor performance through evaluate the neutronic parameters

    International Nuclear Information System (INIS)

    Zareian Ronizi, F.; Fadaei, A.H.; Setayeshi, S.; Shahidi, A.R.

    2015-01-01

    Highlights: • One of the most complex issues that Nu-engineers deal with is the design of NR core. • Numerous factors in nuclear core design depend on Fuel-to-Moderator volume ratio. • Aim of this research is to investigate RX performance for different lattice pitches. - Abstract: Nuclear reactor core design is one of the most complex issues that nuclear engineers deal with. The number and complexity of effective parameters and their impact on reactor design, which makes the problem difficult to solve, require precise knowledge of these parameters and their influence on the reactor operation. Numerous factors in a nuclear reactor core design depend on the Fuel-to-Moderator volume ratio, V F /V M , in a fuel cell. This ratio can be modified by changing the lattice pitch which is the thickness of water channels between fuels plates while keeping fuel slab dimensions fixed. Cooling and moderating properties of water are affected by such a change in a reactor core, and hence some parameters related to these properties might be changed. The aim of this research is to provide the suitable knowledge for nuclear core designing. To reach this goal, the first operating core of Tehran Research Reactor (TRR) with different lattice pitches is simulated, and the effect of different lattice pitches on some parameters such as effective multiplication factor (K eff ), reactor life time, distribution of neutron flux and power density in the core, as well as moderator temperature and density coefficient of reactivity are evaluated. The nuclear reactor analysis code, MTR-PC package is employed to carry out the considered calculation. Finally, the results are presented in some tables and graphs that provide useful information for nuclear engineers in the nuclear reactor core design

  20. Status of glueball mass calculations in lattice gauge theory

    International Nuclear Information System (INIS)

    Kronfeld, A.S.

    1989-11-01

    The status of glueball spectrum calculations in lattice gauge theory is briefly reviewed, with focus on the comparison between Monte Carlo simulations and small-volume analytical calculations in SU(3). The agreement gives confidence that the large-volume Monte Carlo results are accurate, at least in the context of the pure gauge theory. An overview of some of the technical questions, which is aimed at non-experts, serves as an introduction. 19 refs., 1 fig

  1. Uncertainty quantification in lattice QCD calculations for nuclear physics

    Energy Technology Data Exchange (ETDEWEB)

    Beane, Silas R. [Univ. of Washington, Seattle, WA (United States); Detmold, William [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Orginos, Kostas [College of William and Mary, Williamsburg, VA (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Savage, Martin J. [Institute for Nuclear Theory, Seattle, WA (United States)

    2015-02-05

    The numerical technique of Lattice QCD holds the promise of connecting the nuclear forces, nuclei, the spectrum and structure of hadrons, and the properties of matter under extreme conditions with the underlying theory of the strong interactions, quantum chromodynamics. A distinguishing, and thus far unique, feature of this formulation is that all of the associated uncertainties, both statistical and systematic can, in principle, be systematically reduced to any desired precision with sufficient computational and human resources. As a result, we review the sources of uncertainty inherent in Lattice QCD calculations for nuclear physics, and discuss how each is quantified in current efforts.

  2. Burnable poison management in light water reactor lattices

    Energy Technology Data Exchange (ETDEWEB)

    Buenemann, D; Mueller, A

    1970-07-01

    For a better reactivity control and power flattening as well as for an increase in dynamic stability the use of burnable poisons in light water reactors has been considered. The main goals for a burnable poison management and its technological realisation are discussed. The poison is assumed to be in the form of separate poison rods or homogeneous or inhomogeneous poisoning in the fuel rods. A new concept with a central poison rod within the fuel rod is discussed. The balance-equation for the needed concentration of burnable poisons for reactivity central as well as the problems of optimization of lumped poisons are treated in connection with the fuel lattice burnup. A first approximation for the design is found. For the calculation of the microburnup of lumped poison and fuel the special code NEUTRA has been developed. The burnup-equation can be chosen either in a simplified burnup-version with 2 pseudo fission products for each fissionable isotope or with an extended system of burnup equations to be used at sophisticated calculations. These burnup equations are coupled to S{sub N}-routines optionally for cylindrical or x-y-geometry for the proper calculation of the microscopic isotope density-, flux-, and power distributions. The theoretical predictions have been checked by means of special experiments so as to determine the accuracy of the computations. Even for a relatively long burnup of the fuel the predicted values are found to be within the experimental error in the case of lumped rods containing a cadmium-alloy or boron carbide. (auth)

  3. Hamiltonian lattice field theory: Computer calculations using variational methods

    International Nuclear Information System (INIS)

    Zako, R.L.

    1991-01-01

    I develop a variational method for systematic numerical computation of physical quantities -- bound state energies and scattering amplitudes -- in quantum field theory. An infinite-volume, continuum theory is approximated by a theory on a finite spatial lattice, which is amenable to numerical computation. I present an algorithm for computing approximate energy eigenvalues and eigenstates in the lattice theory and for bounding the resulting errors. I also show how to select basis states and choose variational parameters in order to minimize errors. The algorithm is based on the Rayleigh-Ritz principle and Kato's generalizations of Temple's formula. The algorithm could be adapted to systems such as atoms and molecules. I show how to compute Green's functions from energy eigenvalues and eigenstates in the lattice theory, and relate these to physical (renormalized) coupling constants, bound state energies and Green's functions. Thus one can compute approximate physical quantities in a lattice theory that approximates a quantum field theory with specified physical coupling constants. I discuss the errors in both approximations. In principle, the errors can be made arbitrarily small by increasing the size of the lattice, decreasing the lattice spacing and computing sufficiently long. Unfortunately, I do not understand the infinite-volume and continuum limits well enough to quantify errors due to the lattice approximation. Thus the method is currently incomplete. I apply the method to real scalar field theories using a Fock basis of free particle states. All needed quantities can be calculated efficiently with this basis. The generalization to more complicated theories is straightforward. I describe a computer implementation of the method and present numerical results for simple quantum mechanical systems

  4. Hamiltonian lattice field theory: Computer calculations using variational methods

    International Nuclear Information System (INIS)

    Zako, R.L.

    1991-01-01

    A variational method is developed for systematic numerical computation of physical quantities-bound state energies and scattering amplitudes-in quantum field theory. An infinite-volume, continuum theory is approximated by a theory on a finite spatial lattice, which is amenable to numerical computation. An algorithm is presented for computing approximate energy eigenvalues and eigenstates in the lattice theory and for bounding the resulting errors. It is shown how to select basis states and choose variational parameters in order to minimize errors. The algorithm is based on the Rayleigh-Ritz principle and Kato's generalizations of Temple's formula. The algorithm could be adapted to systems such as atoms and molecules. It is shown how to compute Green's functions from energy eigenvalues and eigenstates in the lattice theory, and relate these to physical (renormalized) coupling constants, bound state energies and Green's functions. Thus one can compute approximate physical quantities in a lattice theory that approximates a quantum field theory with specified physical coupling constants. The author discusses the errors in both approximations. In principle, the errors can be made arbitrarily small by increasing the size of the lattice, decreasing the lattice spacing and computing sufficiently long. Unfortunately, the author does not understand the infinite-volume and continuum limits well enough to quantify errors due to the lattice approximation. Thus the method is currently incomplete. The method is applied to real scalar field theories using a Fock basis of free particle states. All needed quantities can be calculated efficiently with this basis. The generalization to more complicated theories is straightforward. The author describes a computer implementation of the method and present numerical results for simple quantum mechanical systems

  5. Methods in nuclear reactors calculations

    International Nuclear Information System (INIS)

    Velarde, G.

    1966-01-01

    Studies are made of the neutron transport equation corresponding to the the real and virtual reactors, as well as the starting hypotheses. Methods are developed to solve the transport equation in slab geometry, and P l ; B l ; M l ; S n and discrete ordinates approximations. (Author)

  6. RA-0 reactor. New neutronic calculations

    International Nuclear Information System (INIS)

    Rumis, D.; Leszczynski, F.

    1990-01-01

    An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core's interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author) [es

  7. Neutronic parameters calculations of a CANDU reactor

    International Nuclear Information System (INIS)

    Zamonsky, G.

    1991-01-01

    Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author) [es

  8. Reactor Thermal Hydraulic Numerical Calculation And Modeling

    International Nuclear Information System (INIS)

    Duong Ngoc Hai; Dang The Ba

    2008-01-01

    In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)

  9. Hybrid SN Laplace Transform Method For Slab Lattice Calculations

    International Nuclear Information System (INIS)

    Segatto, Cynthia F.; Vilhena, Marco T.; Zani, Jose H.; Barros, Ricardo C.

    2008-01-01

    In typical lattice cells where a highly absorbing, small fuel element is embedded in the moderator, a large weakly absorbing medium, high-order transport methods become unnecessary. In this paper we describe a hybrid discrete ordinates (S N ) method for slab lattice calculations. This hybrid S N method combines the convenience of a low-order S N method in the moderator with a high-order S N method in the fuel. We use special fuel-moderator interface conditions based on an approximate angular flux interpolation analytical method and the Laplace transform (LTS N ) numerical method to calculate the neutron flux distribution and the thermal disadvantage factor. We present numerical results for a range of typical model problems. (authors)

  10. A parametric thermohydraulic study an advanced pressurized light water reactor with a tight fuel rod lattice

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Hame, W.

    1982-12-01

    A parametric thermohydraulic study for an Advanced Pressurized Light Water Reactor (APWR) with a tight fuel rod lattice has been performed. The APWR improves the uranium utilisation. The APWR core should be placed in a modern German PWR plant. Within this study about 200 different reactors have been calculated. The tightening of the fuel rod lattice implies a decrease of the net electrical output of the plant, which is greater for the heterogeneous reactor than for the homogeneous reactor. APWR cores mean higher core pressure drops and higher water velocities in the core region. The cores tend to be shorter and the number of fuel rods to be higher than for the PWR. At the higher fuel rod pitch to diameter ratios (p/d) the DNB limitation is more stringent than the limitation on the fuel rod linear rating given by the necessity of reflooding after a reactor accident. The contrary is true for the lower p/d ratios. Subcooled boiling in the highest rated coolant channels occurs for the most of the calculated reactors. (orig.) [de

  11. HETERO code, heterogeneous procedure for reactor calculation; Program Hetero, heterogeni postupak proracuna reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, S M; Raisic, N M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1966-11-15

    This report describes the procedure for calculating the parameters of heterogeneous reactor system taking into account the interaction between fuel elements related to established geometry. First part contains the analysis of single fuel element in a diffusion medium, and criticality condition of the reactor system described by superposition of elements interactions. the possibility of performing such analysis by determination of heterogeneous system lattice is described in the second part. Computer code HETERO with the code KETAP (calculation of criticality factor {eta}{sub n} and flux distribution) is part of this report together with the example of RB reactor square lattice.

  12. Resonance integral calculations for high temperature reactors

    International Nuclear Information System (INIS)

    Blake, J.P.H.

    1960-02-01

    Methods of calculation of resonance integrals of finite dilution and temperature are given for both, homogeneous and heterogeneous geometries, together with results obtained from these methods as applied to the design of high temperature reactors. (author)

  13. Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

    International Nuclear Information System (INIS)

    Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.

    2015-01-01

    Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to

  14. ZrH reactor lattice spacing (heat transfer considerations)

    International Nuclear Information System (INIS)

    Felten, L.D.

    1970-01-01

    Temperature calculations for a 295 element ZrH reactor at fuel element spacings from 0.010'' to 0.065'' showed a very small dependence of reactor temperature on element spacing. It was found that one variation in coolant channel area (2 zones) was sufficient to satisfactorily shape the radial flow profile for the core. (U.S.)

  15. CLUB - a multigroup integral transport theory code for lattice calculations of PHWR cells

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    1992-01-01

    The computer code CLUB has been developed to calculate lattice parameters as a function of burnup for a pressurised heavy water reactor (PHWR) lattice cell containing fuel in the form of cluster. It solves the multigroup integral transport equation by the method based on combination of small scale collision probability (CP) method and large scale interface current technique. The calculations are performed by using WIMS 69 group cross section library or its condensed versions of 27 or 28 group libraries. It can also compute Keff from the given geometrical buckling in the input using multigroup diffusion theory in fundamental mode. The first order differential burnup equations can be solved by either Trapezoidal rule or Runge-Kutta method. (author). 17 refs., 2 figs

  16. Pressure drop characteristics in tight-lattice bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Tamai, Hidesada; Kureta, Masatoshi; Yoshida, Hiroyuki; Akimoto, Hajime

    2004-01-01

    The reduced-moderation water reactor (RMWR) consists of several distinctive structures; a triangular tight-lattice configuration and a double-flat core. In order to design the RMWR core from the point of view of thermal-hydraulics, an evaluation method on pressure drop characteristics in the rod bundles at the tight-lattice configuration is required. In this study, calculated results by the Martinelli-Nelson's and Hancox's correlations were compared with experimental results in 4 x 5 rod bundles and seven-rod bundles. Consequently, the friction loss in two-phase flows becomes smaller at the tight-lattice configuration with the hydraulic diameter less than about 3 mm. This reason is due to the difference of the configuration between the multi-rod bundle and the circular tube and due to the effect of the small hydraulic diameter on the two-phase multiplier. (author)

  17. Reactor physics calculations on HTR type configurations

    Energy Technology Data Exchange (ETDEWEB)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.).

  18. Reactor physics calculations on HTR type configurations

    International Nuclear Information System (INIS)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.)

  19. Reactor physics calculations in the Nordic countries

    International Nuclear Information System (INIS)

    Hoeglund, R.

    1995-01-01

    The seventh biennial meeting on reactor physics calculations in the Nordic countries was arranged by VTT Energy on May 8-9, 1995. 26 papers on different subjects in the field of reactor physics were presented by 45 participants representing research establishments, technical universities, utilities, consultants and suppliers. Resent development and verification of the program systems of ABB Atom, Risoe, Scandpower, Studsvik and VTT Energy were the main topic of the meeting. Benchmarking of the two assembly codes CASMO-4 and HELIOS is proceeding. Cross section data calculated with CASMO-HEX have been validated for the Loviisa reactors. On core analysis ABB atom gives a description on its latest core simulator version POLCA7 with the calculation Core Master 2 and the BWR core supervision system Core Watch. Transient calculations with HEXTRAN, HEXTRAN- PLIM, TRAB, RAMONA, SIMULATE-3K and a code based on PRESTO II/POLCA7 were also presented

  20. Nuclear calculation methods for light water moderated reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1961-02-01

    This report is intended as an introductory review. After a brief discussion of problems encountered in the nuclear design of water moderated reactors a comprehensive scheme of calculations is described. This scheme is based largely on theoretical methods and computer codes developed in the U.S.A. but some previously unreported developments made in this country are also described. It is shown that the effective reproduction factor of simple water moderated lattices may be estimated to an accuracy of approximately 1%. Methods for treating water gap flux peaking and control absorbers are presented in some detail, together with a brief discussion of temperature coefficients, void coefficients and burn-up problems. (author)

  1. Development of a reference scheme for MOX lattice physics calculations

    International Nuclear Information System (INIS)

    Finck, P.J.; Stenberg, C.G.; Roy, R.

    1998-01-01

    The US program to dispose of weapons-grade Pu could involve the irradiation of mixed-oxide (MOX) fuel assemblies in commercial light water reactors. This will require licensing acceptance because of the modifications to the core safety characteristics. In particular, core neutronics will be significantly modified, thus making it necessary to validate the standard suites of neutronics codes for that particular application. Validation criteria are still unclear, but it seems reasonable to expect that the same level of accuracy will be expected for MOX as that which has been achieved for UO 2 . Commercial lattice physics codes are invariably claimed to be accurate for MOX analysis but often lack independent confirmation of their performance on a representative experimental database. Argonne National Laboratory (ANL) has started implementing a public domain suite of codes to provide for a capability to perform independent assessments of MOX core analyses. The DRAGON lattice code was chosen, and fine group ENDF/B-VI.04 and JEF-2.2 libraries have been developed. The objective of this work is to validate the DRAGON algorithms with respect to continuous-energy Monte Carlo for a suite of realistic UO 2 -MOX benchmark cases, with the aim of establishing a reference DRAGON scheme with a demonstrated high level of accuracy and no computing resource constraints. Using this scheme as a reference, future work will be devoted to obtaining simpler and less costly schemes that preserve accuracy as much as possible

  2. Neutronic investigations of an equilibrium core for a tight-lattice light water reactor

    International Nuclear Information System (INIS)

    Broeders, C.H.M.

    1992-01-01

    Calculation procedures and first results concerning the neutronic design of an equilibrium core of an advanced pressurized water reactor (APWR) with mixed oxide fuel in a compact light water moderated triangular lattice are presented. Principle and qualification of the cell burnup calculations with the KARBUS program are briefly discussed. The fuel assembly design with single control rod positions filled with control rod material or coolant water requires special transport theory calculations, which are performed with a one-dimensional supercell model. The macroscopic fuel assembly cross section data is collected in a special library to be used in a new calculational procedure, ARCOSI, for multi-cycle reactor core simulations. Its first application for a reference design resulted in an equilibrium configuration with moderator density reactivity coefficients which are satisfactory as regards safety. (orig.) [de

  3. Parton distributions and lattice QCD calculations: A community white paper

    Science.gov (United States)

    Lin, Huey-Wen; Nocera, Emanuele R.; Olness, Fred; Orginos, Kostas; Rojo, Juan; Accardi, Alberto; Alexandrou, Constantia; Bacchetta, Alessandro; Bozzi, Giuseppe; Chen, Jiunn-Wei; Collins, Sara; Cooper-Sarkar, Amanda; Constantinou, Martha; Del Debbio, Luigi; Engelhardt, Michael; Green, Jeremy; Gupta, Rajan; Harland-Lang, Lucian A.; Ishikawa, Tomomi; Kusina, Aleksander; Liu, Keh-Fei; Liuti, Simonetta; Monahan, Christopher; Nadolsky, Pavel; Qiu, Jian-Wei; Schienbein, Ingo; Schierholz, Gerrit; Thorne, Robert S.; Vogelsang, Werner; Wittig, Hartmut; Yuan, C.-P.; Zanotti, James

    2018-05-01

    In the framework of quantum chromodynamics (QCD), parton distribution functions (PDFs) quantify how the momentum and spin of a hadron are divided among its quark and gluon constituents. Two main approaches exist to determine PDFs. The first approach, based on QCD factorization theorems, realizes a QCD analysis of a suitable set of hard-scattering measurements, often using a variety of hadronic observables. The second approach, based on first-principle operator definitions of PDFs, uses lattice QCD to compute directly some PDF-related quantities, such as their moments. Motivated by recent progress in both approaches, in this document we present an overview of lattice-QCD and global-analysis techniques used to determine unpolarized and polarized proton PDFs and their moments. We provide benchmark numbers to validate present and future lattice-QCD calculations and we illustrate how they could be used to reduce the PDF uncertainties in current unpolarized and polarized global analyses. This document represents a first step towards establishing a common language between the two communities, to foster dialogue and to further improve our knowledge of PDFs.

  4. Comparison of measured and calculated reaction rate distributions in an scwr-like test lattice

    Energy Technology Data Exchange (ETDEWEB)

    Raetz, Dominik, E-mail: dominik.raetz@psi.ch [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Jordan, Kelly A., E-mail: kelly.jordan@psi.ch [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Murphy, Michael F., E-mail: mike.murphy@psi.ch [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Perret, Gregory, E-mail: gregory.perret@psi.ch [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Chawla, Rakesh, E-mail: rakesh.chawla@psi.ch [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne, EPFL (Switzerland)

    2011-04-15

    High resolution gamma-ray spectroscopy measurements were performed on 61 rods of an SCWR-like fuel lattice, after irradiation in the central test zone of the PROTEUS zero-power research reactor at the Paul Scherrer Institute in Switzerland. The derived reaction rates are the capture rate in {sup 238}U (C{sub 8}) and the total fission rate (F{sub tot}), and also the reaction rate ratio C{sub 8}/F{sub tot}. Each of these has been mapped rod-wise on the lattice and compared to calculated results from whole-reactor Monte Carlo simulations with MCNPX. Ratios of calculated to experimental values (C/E's) have been assessed for the C{sub 8}, F{sub tot} and C{sub 8}/F{sub tot} distributions across the lattice. These C/E's show excellent agreement between the calculations and the measurements. For the {sup 238}U capture rate distribution, the 1{sigma} level in the comparisons corresponds to an uncertainty of {+-}0.8%, while for the total fission rate the corresponding value is {+-}0.4%. The uncertainty for C{sub 8}/F{sub tot}, assessed as a reaction rate ratio characterizing each individual rod position in the test lattice, is significantly higher at {+-}2.2%. To determine the reproducibility of these results, the measurements were performed twice, once in 2006 and again in 2009. The agreement between these two measurement sets is within the respective statistical uncertainties.

  5. Criticality Analysis Of TCA Critical Lattices With MNCP-4C Monte Carlo Calculation

    International Nuclear Information System (INIS)

    Zuhair

    2002-01-01

    The use of uranium-plutonium mixed oxide (MOX) fuel in electric generation light water reactor (PWR, BWR) is being planned in Japan. Therefore, the accuracy evaluations of neutronic analysis code for MOX cores have been employed by many scientists and reactor physicists. Benchmark evaluations for TCA was done using various calculation methods. The Monte Carlo become the most reliable method to predict criticality of various reactor types. In this analysis, the MCNP-4C code was chosen because various superiorities the code has. All in all, the MCNP-4C calculation for TCA core with 38 MOX critical lattice configurations gave the results with high accuracy. The JENDL-3.2 library showed significantly closer results to the ENDF/B-V. The k eff values calculated with the ENDF/B-VI library gave underestimated results. The ENDF/B-V library gave the best estimation. It can be concluded that MCNP-4C calculation, especially with ENDF/B-V and JENDL-3.2 libraries, for MOX fuel utilized NPP design in reactor core is the best choice

  6. Calculation of hadronic matrix elements using lattice QCD

    International Nuclear Information System (INIS)

    Gupta, R.

    1993-01-01

    The author gives a brief introduction to the scope of lattice QCD calculations in his effort to extract the fundamental parameters of the standard model. This goal is illustrated by two examples. First the author discusses the extraction of CKM matrix elements from measurements of form factors for semileptonic decays of heavy-light pseudoscalar mesons such as D → Keν. Second, he presents the status of results for the kaon B parameter relevant to CP violation. He concludes the talk with a short outline of his experiences with optimizing QCD codes on the CM5

  7. Calculation of hadronic matrix elements using lattice QCD

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, R.

    1993-08-01

    The author gives a brief introduction to the scope of lattice QCD calculations in his effort to extract the fundamental parameters of the standard model. This goal is illustrated by two examples. First the author discusses the extraction of CKM matrix elements from measurements of form factors for semileptonic decays of heavy-light pseudoscalar mesons such as D {yields} Ke{nu}. Second, he presents the status of results for the kaon B parameter relevant to CP violation. He concludes the talk with a short outline of his experiences with optimizing QCD codes on the CM5.

  8. Nuclear Research Center IRT reactor dynamics calculation

    International Nuclear Information System (INIS)

    Aleman Fernandez, J.R.

    1990-01-01

    The main features of the code DIRT, for dynamical calculations are described in the paper. With the results obtained by the program, an analysis of the dynamic behaviour of the Research Reactor IRT of the Nuclear Research Center (CIN) is performed. Different transitories were considered such as variation of the system reactivity, coolant inlet temperature variation and also variations of the coolant velocity through the reactor core. 3 refs

  9. Reactor calculation benchmark PCA blind test results

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.

    1980-01-01

    Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables

  10. Reactor calculation benchmark PCA blind test results

    Energy Technology Data Exchange (ETDEWEB)

    Kam, F.B.K.; Stallmann, F.W.

    1980-01-01

    Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables.

  11. Coarse-mesh method for multidimensional, mixed-lattice diffusion calculations

    International Nuclear Information System (INIS)

    Dodds, H.L. Jr.; Honeck, H.C.; Hostetler, D.E.

    1977-01-01

    A coarse-mesh finite difference method has been developed for multidimensional, mixed-lattice reactor diffusion calculations, both statics and kinetics, in hexagonal geometry. Results obtained with the coarse-mesh (CM) method have been compared with a conventional mesh-centered finite difference method and with experiment. The results of this comparison indicate that the accuracy of the CM method for highly heterogeneous (mixed) lattices using one point per hexagonal mesh element (''hex'') is about the same as the conventional method with six points per hex. Furthermore, the computing costs (i.e., central processor unit time and core storage requirements) of the CM method with one point per hex are about the same as the conventional method with one point per hex

  12. The programme PIP2 for lattice cell thermal calculations

    International Nuclear Information System (INIS)

    Clayton, A.J.

    1964-08-01

    The programme PIP2 solves the multigroup equations obtained by applying the method of collision probabilities to a fuel region (which may contain a cluster of fuel elements), and the SPECTROX flux assumption in a surrounding 'moderator'. The programme does not calculate collision probabilities for the fuel region and any geometry can be treated in the fuel region for which collision probabilities can be calculated. Lattice cell source problems may be treated and it is possible to include part of the physical moderator with the fuel region for treatment by the collision probability method. The programme is primarily intended for thermal fixed source problems, with the sources in the (physical moderator), but by including part of the moderator with the fuel it is possible to include fixed sources in the fuel for the study of fast effects. (author)

  13. Lattice QCD Calculations in Nuclear Physics towards the Exascale

    Science.gov (United States)

    Joo, Balint

    2017-01-01

    The combination of algorithmic advances and new highly parallel computing architectures are enabling lattice QCD calculations to tackle ever more complex problems in nuclear physics. In this talk I will review some computational challenges that are encountered in large scale cold nuclear physics campaigns such as those in hadron spectroscopy calculations. I will discuss progress in addressing these with algorithmic improvements such as multi-grid solvers and software for recent hardware architectures such as GPUs and Intel Xeon Phi, Knights Landing. Finally, I will highlight some current topics for research and development as we head towards the Exascale era This material is funded by the U.S. Department of Energy, Office Of Science, Offices of Nuclear Physics, High Energy Physics and Advanced Scientific Computing Research, as well as the Office of Nuclear Physics under contract DE-AC05-06OR23177.

  14. Hardware matrix multiplier/accumulator for lattice gauge theory calculations

    International Nuclear Information System (INIS)

    Christ, N.H.; Terrano, A.E.

    1984-01-01

    The design and operating characteristics of a special-purpose matrix multiplier/accumulator are described. The device is connected through a standard interface to a host PDP11 computer. It provides a set of high-speed, matrix-oriented instructions which can be called from a program running on the host. The resulting operations accelerate the complex matrix arithmetic required for a class of Monte Carlo calculations currently of interest in high energy particle physics. A working version of the device is presently being used to carry out a pure SU(3) lattice gauge theory calculation using a PDP11/23 with a performance twice that obtainable on a VAX11/780. (orig.)

  15. Nuclear calculation of the thorium reactor

    International Nuclear Information System (INIS)

    Hirakawa, Naohiro

    1998-01-01

    Even if for a reactor using thorium (and 233-U), its nuclear design calculation procedure is similar to the case using conventional 235-U, 238-U and plutonium. As nuclear composition varies with time on operation of nuclear reactor, calculation of its mean cross section should be conducted in details. At that time, one-group cross section obtained by integration over a whole of energy range is used for small member group. And, as the nuclear data for a base of its calculation is already prepared by JENDL3.2 and nuclear data library derived from it, the nuclear calculation of a nuclear reactor using thorium has no problem. From such a veiwpoint, IAEA has organized a coordinated research program of 'Potential of Th-based Fuel Cycles to Constrain Pu and to reduce Long-term Waste Toxicities' since 1996. All nations entering this program were regulated so as to institute by selecting a nuclear fuel cycle thinking better by each nation and to examine what cycle is expected by comparing their results. For a promise to conduct such neutral comparison, a comparison of bench mark calculations aiming at PWR was conducted to protect that the obtained results became different because of different calculation method and cross section adopted by each nation. Therefore, it was promoted by entrance of China, Germany, India, Israel, Japan, Korea, Russia and USA. The SWAT system developed by Tohoku University is used for its calculation code, by using which calculated results on the bench mark calculation at the fist and second stages and the nuclear reactor were reported. (G.K.)

  16. Calculation of tritium release from reactor's stack

    International Nuclear Information System (INIS)

    Akhadi, M.

    1996-01-01

    Method for calculation of tritium release from nuclear to environment has been discussed. Part of gas effluent contain tritium in form of HTO vapor released from reactor's stack was sampled using silica-gel. The silica-gel was put in the water to withdraw HTO vapor absorbed by silica-gel. Tritium concentration in the water was measured by liquid scintillation counter of Aloka LSC-703. Tritium concentration in the gas effluent and total release of tritium from reactor's stack during certain interval time were calculated using simple mathematic formula. This method has examined for calculation of tritium release from JRR-3M's stack of JAERI, Japan. From the calculation it was obtained the value of tritium release as much as 4.63 x 10 11 Bq during one month. (author)

  17. Reference Monte Carlo calculations of Maria reactor core

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.

    2002-01-01

    The reference Monte Carlo calculations of MARIA reactor core have been carried to evaluate accuracy of the calculations at each stage of its neutron-physics analysis using deterministic codes. The elementary cell has been calculated with two main goals; evaluation of effects of simplifications introduced in deterministic lattice spectrum calculations by the WIMS code and evaluation of library data in recently developed WIMS libraries. In particular the beryllium data of those libraries needed evaluation. The whole core calculations mainly the first MARIA critical experiment and the first critical core after the 8-year break in operation. Both cores contained only fresh fuel elements but only in the first critical core the beryllium blocks were not poisoned by Li-6 and He-3. Thus the MCNP k-eff results could be compared with the experiment. The MCNP calculations for the cores with beryllium poisoned suffered the deficiency of uncertainty in the poison concentration, but a comparison of power distribution shows that realistic poison levels have been carried out for the operating reactor MARIA configurations. (author)

  18. A Framework for Lattice QCD Calculations on GPUs

    Energy Technology Data Exchange (ETDEWEB)

    Winter, Frank; Clark, M A; Edwards, Robert G; Joo, Balint

    2014-08-01

    Computing platforms equipped with accelerators like GPUs have proven to provide great computational power. However, exploiting such platforms for existing scientific applications is not a trivial task. Current GPU programming frameworks such as CUDA C/C++ require low-level programming from the developer in order to achieve high performance code. As a result porting of applications to GPUs is typically limited to time-dominant algorithms and routines, leaving the remainder not accelerated which can open a serious Amdahl's law issue. The lattice QCD application Chroma allows to explore a different porting strategy. The layered structure of the software architecture logically separates the data-parallel from the application layer. The QCD Data-Parallel software layer provides data types and expressions with stencil-like operations suitable for lattice field theory and Chroma implements algorithms in terms of this high-level interface. Thus by porting the low-level layer one can effectively move the whole application in one swing to a different platform. The QDP-JIT/PTX library, the reimplementation of the low-level layer, provides a framework for lattice QCD calculations for the CUDA architecture. The complete software interface is supported and thus applications can be run unaltered on GPU-based parallel computers. This reimplementation was possible due to the availability of a JIT compiler (part of the NVIDIA Linux kernel driver) which translates an assembly-like language (PTX) to GPU code. The expression template technique is used to build PTX code generators and a software cache manages the GPU memory. This reimplementation allows us to deploy an efficient implementation of the full gauge-generation program with dynamical fermions on large-scale GPU-based machines such as Titan and Blue Waters which accelerates the algorithm by more than an order of magnitude.

  19. Thermal calculations for water cooled research reactors

    International Nuclear Information System (INIS)

    Fabrega, S.

    1979-01-01

    The formulae and the more important numerical data necessary for thermic calculations on the core of a research reactor, cooled with low pressure water, are presented. Most of the problems met by the designer and the operator are dealt with (calculations margins, cooling after shut-down). Particular cases are considered (gas release, rough walls, asymmetric cooling slabs etc.), which are not generally envisaged in works on general thermics

  20. A lattice QCD calculation of the transverse decay constant of the b1(1235) meson

    International Nuclear Information System (INIS)

    Jansen, K.; McNeile, C.; Michael, C.; Urbach, C.

    2009-10-01

    We review various B meson decays that require knowledge of the transverse decay constant of the b 1 (1235) meson. We report on an exploratory lattice QCD calculation of the transverse decay constant of the b 1 meson. The lattice QCD calculations used unquenched gauge configurations, at two lattice spacings, generated with two flavours of sea quarks. The twisted mass formalism is used. (orig.)

  1. Reactor perturbation calculations by Monte Carlo methods

    International Nuclear Information System (INIS)

    Gubbins, M.E.

    1965-09-01

    Whilst Monte Carlo methods are useful for reactor calculations involving complicated geometry, it is difficult to apply them to the calculation of perturbation worths because of the large amount of computing time needed to obtain good accuracy. Various ways of overcoming these difficulties are investigated in this report, with the problem of estimating absorbing control rod worths particularly in mind. As a basis for discussion a method of carrying out multigroup reactor calculations by Monte Carlo methods is described. Two methods of estimating a perturbation worth directly, without differencing two quantities of like magnitude, are examined closely but are passed over in favour of a third method based on a correlation technique. This correlation method is described, and demonstrated by a limited range of calculations for absorbing control rods in a fast reactor. In these calculations control rod worths of between 1% and 7% in reactivity are estimated to an accuracy better than 10% (3 standard errors) in about one hour's computing time on the English Electric KDF.9 digital computer. (author)

  2. Qualification of the WIMS lattice code, for the design, operation and accident analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Lerner, A.M.

    1996-01-01

    A basic problem in nuclear reactor physics in that of the description of the neutron population behaviour in the multiplicative medium of a nuclear fuel. Due to the magnitude of the physical problem involved and the present degree of technological evolution regarding computing resources, of increasing complexity and possibilities, the calculation programs or codes have turned to be a basic auxiliary tool in reactor physics. In order to analyze the global problem, several aspects should be taken into consideration. The first aspect to be considered is that of the availability of the necessary nuclear data. The second one is the existence of a variety of methods and models to perform the calculations. The final phase for this kind of analysis is the qualification of the computing programs to be used, i.e. the verification of the validity domain of its nuclear data and the models involved. The last one is an essential phase, and in order to carry it on great variety of calculations are required, that will check the different aspects contained in the code. We here analyze the most important physical processes that take place in a nuclear reactor cell, and we consider the qualification of the lattice code WIMS, that calculates the neutronic parameters associated with such processes. Particular emphasis has been put in the application to natural uranium fuelled reactor, heavy water cooled and moderated, as the Argentinean power reactors now in operation. A wide set of experiments has been chosen: a.-Fresh fuel in zero-power experimental facilities and power reactors; b.-Irradiated fuel in both types of facilities; c.-Benchmark (prototype) experiments with loss of coolant. From the whole analysis it was concluded that for the research reactors, as well as for the heavy water moderated power reactors presently operating in our country, or those that could operate in a near future, the lattice code WIMS is reliable and produces results within the experimental values and

  3. Methodology of shielding calculation for nuclear reactors

    International Nuclear Information System (INIS)

    Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo

    1982-01-01

    A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt

  4. Reactor calculations for improving utilization of TRIGA reactor

    International Nuclear Information System (INIS)

    Ravnik, M.

    1986-01-01

    A brief review of our work on reactor calculations of 250 kW TRIGA with mixed core (standard + FLIP fuel) will be presented. The following aspects will be treated: - development of computer programs; - optimization of in-core fuel management with respect to fuel costs and irradiation channels utilization. TRIGAP programme package will be presented as an example of computer programs. It is based on 2-group 1-D diffusion approximation and besides calculations offers possibilities for operational data logging and fuel inventory book-keeping as well. It is developed primarily for the research reactor operators as a tool for analysing reactor operation and fuel management. For this reason it is arranged for a small (PC) computer. Second part will be devoted to reactor physics properties of the mixed cores. Results of depletion calculations will be presented together with measured data to confirm some general guidelines for optimal mixed core fuel management. As the results are obtained using TRIGAP program package results can be also considered as an illustration and qualification for its application. (author)

  5. Parallel computer calculation of quantum spin lattices; Calcul de chaines de spins quantiques sur ordinateur parallele

    Energy Technology Data Exchange (ETDEWEB)

    Lamarcq, J. [Service de Physique Theorique, CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France)

    1998-07-10

    Numerical simulation allows the theorists to convince themselves about the validity of the models they use. Particularly by simulating the spin lattices one can judge about the validity of a conjecture. Simulating a system defined by a large number of degrees of freedom requires highly sophisticated machines. This study deals with modelling the magnetic interactions between the ions of a crystal. Many exact results have been found for spin 1/2 systems but not for systems of other spins for which many simulation have been carried out. The interest for simulations has been renewed by the Haldane`s conjecture stipulating the existence of a energy gap between the ground state and the first excited states of a spin 1 lattice. The existence of this gap has been experimentally demonstrated. This report contains the following four chapters: 1. Spin systems; 2. Calculation of eigenvalues; 3. Programming; 4. Parallel calculation 14 refs., 6 figs.

  6. Spherical Harmonics Treatment of Epithermal Neutron Spectra in Reactor lattices

    International Nuclear Information System (INIS)

    Matausek, M.V.

    1972-04-01

    A procedure has been developed to solve the slowing down transport equation for neutrons in a cylindrized reactor lattice cell. Treating the anisotropy of the epithermal neutron flux by the spherical harmonics formalism, which reduces the space-angle-lethargy-dependent transport equation to the matrix integrodifferential equation in space and lethargy, and replacing the lethargy transfer integrals by finite-difference forms, a set of matrix ordinary differential equations, with lethargy and space dependent coefficients, is obtained. In the resonance region this set takes a lower block triangular form and can be directly solved by forward block substitution; in the lethargy range, where the fast fission effects have to be considered, the iterative procedure is introduced. A simple and efficient approximation is then proposed, making possible the analytical solution for the spatial dependence of the spherical harmonics flux moments. The proposed procedure has been numerically examined and approved. Some typical results are presented and discussed. (author)

  7. Automatic fuel lattice design in a boiling water reactor using a particle swarm optimization algorithm and local search

    International Nuclear Information System (INIS)

    Lin Chaung; Lin, Tung-Hsien

    2012-01-01

    Highlights: ► The automatic procedure was developed to design the radial enrichment and gadolinia (Gd) distribution of fuel lattice. ► The method is based on a particle swarm optimization algorithm and local search. ► The design goal were to achieve the minimum local peaking factor. ► The number of fuel pins with Gd and Gd concentration are fixed to reduce search complexity. ► In this study, three axial sections are design and lattice performance is calculated using CASMO-4. - Abstract: The axial section of fuel assembly in a boiling water reactor (BWR) consists of five or six different distributions; this requires a radial lattice design. In this study, an automatic procedure based on a particle swarm optimization (PSO) algorithm and local search was developed to design the radial enrichment and gadolinia (Gd) distribution of the fuel lattice. The design goals were to achieve the minimum local peaking factor (LPF), and to come as close as possible to the specified target average enrichment and target infinite multiplication factor (k ∞ ), in which the number of fuel pins with Gd and Gd concentration are fixed. In this study, three axial sections are designed, and lattice performance is calculated using CASMO-4. Finally, the neutron cross section library of the designed lattice is established by CMSLINK; the core status during depletion, such as thermal limits, cold shutdown margin and cycle length, are then calculated using SIMULATE-3 in order to confirm that the lattice design satisfies the design requirements.

  8. RHEIN, Modular System for Reactor Design Calculation

    International Nuclear Information System (INIS)

    Reiche, Christian; Barz, Hansulrich; Kunzmann, Bernd; Seifert, Eberhard; Wand, Hartmut

    1990-01-01

    1 - Description of program or function: RHEIN is a modular reactor code system for neutron physics calculations. It consists of a small number of system codes for execution control, data management, and handling support, as well as of the physical calculation routines. The execution is controlled by input data containing mathematical and physical parameters and simple commands for routine calls and data manipulations. The calculation routines are in tune with one another and the system takes care of the data transfer between them. Cross-section libraries with self shielding parameters are added to the system. 2 - Method of solution: The calculation routines can be used for solving the following physics problems: - Calculation of cross-section sets for infinite mediums, taking into account chord length. - Zero-dimensional spectrum calculation in diffusion, P1, or B1 approximation. - One-dimensional calculation in diffusion, P1, or collision probability approximation. - Two-dimensional diffusion calculation. - Cell calculation by THERMOS. - Zone-wise homogenized group collapsing within zero, one, or two-dimensional models. - Normalization, summarizing, etc. - Output of cross-section sets to off systems Sn and Monte-Carlo calculations

  9. Thorium-Based Fuels Preliminary Lattice Cell Studies for Candu Reactors

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Rizoiu, A.C.

    2009-01-01

    The choice of nuclear power as a major contributor to the future global energy needs must take into account acceptable risks of nuclear weapon proliferation, in addition to economic competitiveness, acceptable safety standards, and acceptable waste disposal options. Candu reactors offer a proven technology, safe and reliable reactor technology, with an interesting evolutionary potential for proliferation resistance, their versatility for various fuel cycles creating premises for a better utilization of global fuel resources. Candu reactors impressive degree of fuel cycle flexibility is a consequence of its channel design, excellent neutron economy, on-power refueling, and simple fuel bundle. These features facilitate the introduction and exploitation of various fuel cycles in Candu reactors in an evolutionary fashion. The main reasons for our interest in Thorium-based fuel cycles have been, globally, to extend the energy obtainable from natural Uranium and, locally, to provide a greater degree of energy self-reliance. Applying the once through Thorium (OTT) cycle in existing and advanced Candu reactors might be seen as an evaluative concept for the sustainable development both from the economic and waste management points of view. Two Candu fuel bundles project will be used for the proposed analysis, namely the Candu standard fuel bundle with 37 fuel elements and the CANFLEX fuel bundle with 43 fuel elements. Using the Canadian proposed scheme - loading mixed ThO 2 -SEU CANFLEX bundles in Candu 6 reactors - simulated at lattice cell level led to promising conclusions on operation at higher fuel burnups, reduction of the fissile content to the end of the cycle, minor actinide content reduction in the spent fuel, reduction of the spent fuel radiotoxicity, presence of radionuclides emitting strong gamma radiation for proliferation resistance benefit. The calculations were performed using the lattice codes WIMS and Dragon (together with the corresponding nuclear data

  10. Adjustement of Dancoff factor for calculating the cell of fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Borges, V.; Sefidvash, F.

    1988-01-01

    A new nuclear reactor design based on the fluidized bed concept is under reserch and development. It utilized spherical fuel of slightly enriched zircaloy-clad uranium dioxide fluidized by light water under pressure since the Leopard code has been developed for light water reactor analysis, it was necessary to develop a method to determine the dimensions of the hypothetical fuel rod lattice, which are neutronically equivalent to the spherical fuel pellet lattice. This method is shown to calculate the Dancoff factor correctly. (author) [pt

  11. Calculation of the neutron parameters of fast thermal reactor

    International Nuclear Information System (INIS)

    Kukuleanu, V.; Mocioiu, D.; Drutse, E.; Konstantinesku, E.

    1975-01-01

    The system of neutron calculation for fast reactors is given. This system was used for estimation of physical parameters of fast thermal reactors investigated. The results obtained and different specific problems of the reactors of this type are described. (author)

  12. Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)

    2014-07-01

    A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm{sup 2}s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)

  13. Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes

    International Nuclear Information System (INIS)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

    2014-01-01

    A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm 2 s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)

  14. LCEs for Naval Reactor Benchmark Calculations

    International Nuclear Information System (INIS)

    W.J. Anderson

    1999-01-01

    The purpose of this engineering calculation is to document the MCNP4B2LV evaluations of Laboratory Critical Experiments (LCEs) performed as part of the Disposal Criticality Analysis Methodology program. LCE evaluations documented in this report were performed for 22 different cases with varied design parameters. Some of these LCEs (10) are documented in existing references (Ref. 7.1 and 7.2), but were re-run for this calculation file using more neutron histories. The objective of this analysis is to quantify the MCNP4B2LV code system's ability to accurately calculate the effective neutron multiplication factor (k eff ) for various critical configurations. These LCE evaluations support the development and validation of the neutronics methodology used for criticality analyses involving Naval reactor spent nuclear fuel in a geologic repository

  15. NEPTUNE: a modular system for light-water reactor calculation

    International Nuclear Information System (INIS)

    Bouchard, J.; Kanevoky, A.; Reuss, P.

    1975-01-01

    A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

  16. CANDU reactor core simulations using fully coupled DRAGON and DONJON calculations

    International Nuclear Information System (INIS)

    Varin, E.; Marleau, G.

    2006-01-01

    The operating CANDU-6 reactors are refueled on-power to compensate for the reactivity loss due to fuel burnup. In order to predict the core behavior, fuel bundle burnups and local parameter information need to be tracked. The history-based approach has been developed to follow local parameter as well as history effect in CANDU reactors. The finite reactor diffusion code DONJON and the lattice code DRAGON have been coupled to perform reactor follow-up calculations using a history-based approach. A coupled methodology that manages the transfer of information between standard DONJON and DRAGON data structures has been developed. Push-through refueling can be taken into account directly in cell calculations. Using actual on-site information, an isotopic core content database has been generated with coupled DONJON and DRAGON calculations. Moreover calculations have been performed for different local parameters. Results are compared with those obtained using standard cross section generation approaches

  17. Neutron data error estimate of criticality calculations for lattice in shielding containers with metal fissionable materials

    International Nuclear Information System (INIS)

    Vasil'ev, A.P.; Krepkij, A.S.; Lukin, A.V.; Mikhal'kova, A.G.; Orlov, A.I.; Perezhogin, V.D.; Samojlova, L.Yu.; Sokolov, Yu.A.; Terekhin, V.A.; Chernukhin, Yu.I.

    1991-01-01

    Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab

  18. Calculation of the Flux in a Square Lattice Cell and a Comparison with Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Apelqvist, G [State Power Board, Stockholm (Sweden)

    1961-05-15

    A calculation has been made of the thermal neutron flux in a square lattice cell using methods devised by Galanin. The f and L lattice parameters have been expressed in measurable quantities and a comparison made between measured and calculated values.

  19. Neutronics aspects associated to irregular lattices in sodium fast reactors cores

    International Nuclear Information System (INIS)

    Gentili, Michele

    2015-01-01

    The fuel assemblies of SFR cores (sodium fast reactors) are normally arranged in hexagonal regular lattices, whose compactness is ensured in nominal operating conditions by thermal expansion of assemblies pads disposed on the six assembly wrapper faces. During the reactor operations, thermal expansion phenomena and irradiation creep phenomena occur and they cause the fuel assemblies to bow and to deform both radially and axially. The main goal of this PhD is the understanding of the neutronic aspects and phenomena occurring in case of core and lattice deformations, as much as the design and implementation of deterministic neutronic calculation schemes and methods in order to evaluate the consequences for the core design activities and the safety analysis. The first part of this work is focused on the development of an analytical model with the purpose to identify the neutronic phenomena that are the main contributors to the reactivity changes induced by lattice and core deformations. A first scheme based on the spatial mesh projection method has been conceived and implemented for the ERANOS codes (BISTRO, H3D and VARIANT) and to the SNATCH solver. The second calculation scheme propose is based on mesh deformation: the computing mesh is deformed as a function of the assembly displacement field. This methodology has been implemented for the solver SNATCH, which normally allows the Boltzmann equation to be solved for a regular mesh. Finally, an iterative method has been developed in order to fulfill an a-priori estimation of the maximal reactivity insertion as a function of the postulated mechanical energy provided to the core, as much as the deformation causing it. (author) [fr

  20. A technique for analytical calculation of observables in lattice gauge theories

    International Nuclear Information System (INIS)

    Narayanan, R.; Vranas, P.

    1990-01-01

    It is shown that the partition function for a finite lattice factorizes into terms that can be associated with each vertex in the finite lattice. This factorization property forms the basis of well defined and efficient technique developed to calculate partition functions to high accuracy, on finite lattices for gauge theories. This technique along with the expansion in finite lattices, provides a powerful means for calculating observables in lattice gauge theories. This is applied to SU(2) lattice gauge theory in four dimensions. The free energy, expectation value of a plaquette and specific heat are calculated. The results are very good in the strong coupling region, succeed in entering the weak coupling region and describe the crossover region quite well, agreeing all the way with the Monte Carlo data. (orig.)

  1. Determining the asymptotic buckling for the reference RB reactor lattice

    International Nuclear Information System (INIS)

    Martinc, R.; Sotic, O.

    1969-01-01

    Material buckling was measured for reference lattice of the heavy water reflected system with 2% enriched uranium fuel. Experiments were done for cores with lattice pitch values: 8, 8√2, i 16 cm. Each of these cores had heavy water reflector, as well as active reflector - heavy water lattice with natural uranium fuel. The core was reflected by natural uranium lattice in order to approach asymptotic regime in the central zone. Buckling values obtained with the natural uranium lattice as reflector are, as a rule, lower then in case of heavy water reflector [sr

  2. Neutron calculation scheme for coupled reactors

    International Nuclear Information System (INIS)

    Porta, Jacques.

    1980-11-01

    The CABRI and PHEBUS cores are of the low enrichment rod type in which the fuel is made up of uranium oxide pellets encased in tubular cladding but the SCARABEE core has high enrichment plates, the fuel, an aluminium-uranium alloy (UAl) is metal, rolled into plate form. These three cores in well described rectangular geometry, receive in their centres the very heterogeneous cylindrical test loops (numerous containments of different kinds, large void spaces acting as lagging). After a detailed study of these three reactors, it is found that the search for a calculation scheme (common to the three projects) leads to the elimination of the scattering approximation in our calculations. It is therefore necessary to review the various existing models from a theoretical angle and then to select a particular method, according to the available data processing tools, a choice that will be dictated by the optimization of the parameters: cost in calculation time, difficulties (or ease) of use and accuracy achieved. A problem of experiment interpretation by calculation is dealt with in Chapter 3. The determination of the coupling by calculation is closely linked to the geometrical and energy modelization chosen. But from the experimental angle the determination of the coupling also gives rise to problems with respect to the interpretation of the experimental values obtained by thermal balance determinations, counting of the gamma emission of the fission products of fissile detectors and counting of lanthane 140 in the fuel fission products. The method of calculation is discussed as is the use made of detectors and the counting procedures. In chapter 4, it is not a local modelization that is discussed but an overall one in an original three dimensional calculation [fr

  3. Calculation of Kinetic Parameters of TRIGA Reactor

    International Nuclear Information System (INIS)

    Snoj, Luka; Kavcic, Andrej; Zerovnik, Gasper; Ravnik, Matjaz

    2008-01-01

    Modern Monte Carlo transport codes in combination of fast computer clusters enable very accurate calculations of the most important reactor kinetic parameters. Such are the effective delayed neutron fraction, β eff , and mean neutron generation time, Λ. We calculated the β eff and Λ for various realistic and hypothetical annular TRIGA Mark II cores with different types and amount of fuel. It can be observed that the effective delayed neutron fraction strongly depends on the number of fuel elements in the core or on the core size. E.g., for 12 wt. % uranium standard fuel with 20 % enrichment, β eff varies from 0.0080 for a small core (43 fuel rods) to 0.0075 for a full core (90 fuel rods). It is interesting to note that calculated value of β eff strongly depends also on the delayed neutron nuclear data set used in calculations. The prompt neutron life-time mainly depends on the amount (due to either content or enrichment) of 235 U in the fuel as it is approximately inversely proportional to the average absorption cross-section of the fuel. E.g., it varies from 28 μs for 30 wt. % uranium content fuelled core to 48 μs for 8.5 wt. % uranium content LEU fuelled core. The results are especially important for pulse mode operation and analysis of the pulses. (authors)

  4. HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor

    International Nuclear Information System (INIS)

    Finch, D.R.

    1965-01-01

    1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions

  5. Large-scale calculation of ferromagnetic spin systems on the pyrochlore lattice

    Energy Technology Data Exchange (ETDEWEB)

    Soldatov, Konstantin, E-mail: soldatov_ks@students.dvfu.ru [School of Natural Sciences, Far Eastern Federal University, Vladivostok (Russian Federation); Nefedev, Konstantin, E-mail: nefedev.kv@dvfu.ru [School of Natural Sciences, Far Eastern Federal University, Vladivostok (Russian Federation); Institute of Applied Mathematics, Far Eastern Branch, Russian Academy of Science, Vladivostok (Russian Federation); Komura, Yukihiro [CIJ-solutions, Chuo-ku, Tokyo 103-0023 (Japan); Okabe, Yutaka, E-mail: okabe@phys.se.tmu.ac.jp [Department of Physics, Tokyo Metropolitan University, Hachioji, Tokyo 192-0397 (Japan)

    2017-02-19

    We perform the high-performance computation of the ferromagnetic Ising model on the pyrochlore lattice. We determine the critical temperature accurately based on the finite-size scaling of the Binder ratio. Comparing with the data on the simple cubic lattice, we argue the universal finite-size scaling. We also calculate the classical XY model and the classical Heisenberg model on the pyrochlore lattice. - Highlights: • Calculations of the ferromagnetic models on the pyrochlore lattice were performed. • Precise critical temperatures were determined using Binder ratio finite-size scaling. • The universal finite-size scaling was argued.

  6. Monte Carlo sampling strategies for lattice gauge calculations

    International Nuclear Information System (INIS)

    Guralnik, G.; Zemach, C.; Warnock, T.

    1985-01-01

    We have sought to optimize the elements of the Monte Carlo processes for thermalizing and decorrelating sequences of lattice gauge configurations and for this purpose, to develop computational and theoretical diagnostics to compare alternative techniques. These have been applied to speed up generations of random matrices, compare heat bath and Metropolis stepping methods, and to study autocorrelations of sequences in terms of the classical moment problem. The efficient use of statistically correlated lattice data is an optimization problem depending on the relation between computer times to generate lattice sequences of sufficiently small correlation and times to analyze them. We can solve this problem with the aid of a representation of auto-correlation data for various step lags as moments of positive definite distributions, using methods known for the moment problem to put bounds on statistical variances, in place of estimating the variances by too-lengthy computer runs

  7. Nuclear Data Processing for Reactor Physics Calculation

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Pandiangan, Tumpal

    2003-01-01

    Nuclear data processing for reactor physics calculation has been done. Raw nuclear data cross-sections on file ENDF should be prepared and processed before it used in neutronic calculation. The processing code system such as NJOY-PC code has been used from linearization of nuclear cross-sections data and background contribution of resonance parameter (MF2) using RECONR module (0K) with energy range from 10 -5 to 10 7 eV. Afterward, the neutron cross-sections data should be processed and broadened to desire temperature (i.e. 293K) by using BROADR module. The Grouper and Therma modules will be applied for multi-groups calculation which suitable for WIMS/D4 (69 groups) and thermalization of nuclear constants. The final stage of processing nuclear cross-sections is updating WIMS/D4 library. The WIMSR module in NJOY-PC and WILLIE code will be applied in this stage. The evaluated nuclear data file, especially for 1 H 1 isotope, was taken from JENDL-3.2 and ENDF/B-VI for preliminary study. The results of nuclear data processing 1 H 1 shows that the old-WIMS (WIMS-lama) library have much discrepancies comparing with JENDL-3.2 or ENDF/B-VI files, especially in energy around 5 keV

  8. What lattice calculations can tell us about the gluon gas

    International Nuclear Information System (INIS)

    Kaellman, C.G.; Helsinki Univ.; Montonen, C.

    1982-01-01

    Higher order perturbative and nonperturbative corrections to the grand potential of hot QCD are considered qualitatively Comparing with lattice results, it is argued that the nonperturbative parts are small but that the O(g 4 ) term in Ω is large and positive. (orig.)

  9. One-loop calculations in Supersymmetric Lattice QCD

    Directory of Open Access Journals (Sweden)

    Costa M.

    2017-01-01

    We present here results from dimensional regularization, relegating to a forthcoming publication [1] our results along with a more complete list of references. Part of the lattice study regards also the renormalization of quark bilinear operators which, unlike the nonsupersymmetric case, exhibit a rich pattern of operator mixing at the quantum level.

  10. Lattice QCD calculations on commodity clusters at DESY

    International Nuclear Information System (INIS)

    Gellrich, A.; Pop, D.; Wegner, P.; Wittig, H.; Hasenbusch, M.; Jansen, K.

    2003-06-01

    Lattice Gauge Theory is an integral part of particle physics that requires high performance computing in the multi-Tflops regime. These requirements are motivated by the rich research program and the physics milestones to be reached by the lattice community. Over the last years the enormous gains in processor performance, memory bandwidth, and external I/O bandwidth for parallel applications have made commodity clusters exploiting PCs or workstations also suitable for large Lattice Gauge Theory applications. For more than one year two clusters have been operated at the two DESY sites in Hamburg and Zeuthen, consisting of 32 resp. 16 dual-CPU PCs, equipped with Intel Pentium 4 Xeon processors. Interconnection of the nodes is done by way of Myrinet. Linux was chosen as the operating system. In the course of the projects benchmark programs for architectural studies were developed. The performance of the Wilson-Dirac Operator (also in an even-odd preconditioned version) as the inner loop of the Lattice QCD (LQCD) algorithms plays the most important role in classifying the hardware basis to be used. Using the SIMD streaming extensions (SSE/SSE2) on Intel's Pentium 4 Xeon CPUs give promising results for both the single CPU and the parallel version. The parallel performance, in addition to the CPU power and the memory throughput, is nevertheless strongly influenced by the behavior of hardware components like the PC chip-set and the communication interfaces. The paper starts by giving a short explanation about the physics background and the motivation for using PC clusters for Lattice QCD. Subsequently, the concept, implementation, and operating experiences of the two clusters are discussed. Finally, the paper presents benchmark results and discusses comparisons to systems with different hardware components including Myrinet-, GigaBit-Ethernet-, and Infiniband-based interconnects. (orig.)

  11. Status and prospects for the calculation of hadron structure from lattice QCD

    International Nuclear Information System (INIS)

    Renner, Dru B.

    2010-02-01

    Lattice QCD calculations of hadron structure are a valuable complement to many experimental programs as well as an indispensable tool to understand the dynamics of QCD. I present a focused review of a few representative topics chosen to illustrate both the challenges and advances of our community: the momentum fraction, axial charge and charge radius of the nucleon. I will discuss the current status of these calculations and speculate on the prospects for accurate calculations of hadron structure from lattice QCD. (orig.)

  12. Neutron multipilication factors as a function of temperature: a comparison of calculated and measured values for lattices using 233UO2-ThO2 fuel in graphite

    International Nuclear Information System (INIS)

    Newman, D.F.; Gore, B.F.

    1978-01-01

    Neutron multiplication factors calculated as a function of temperature for three graphite-moderated 233 UO 2 -ThO 2 -fueled lattices are correlated with the values measured for these lattices in the high-temperature lattice test reactor (HTLTR). The correlation analysis is accomplished by fitting calculated values of k/sub infinity/(T) to the measured values using two least-squares-fitted correlation coefficients: (a) a normalization factor and (b) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross-section data. Use of an alternate cross-section data set for thorium, which has a smaller resonance integral than ENDF/B-IV data, improved the agreement between calculated and measured temperature coefficients of reactivity for the three experimental lattices. The results of the correlations are used to estimate the bias in the temperature coefficient of reactivity calculated for a lattice typical of fresh 233 U recycle fuel for a high-temperature gas-cooled reactor (HTGR). This extrapolation to a lattice having a heavier fissile loading than the experimental lattices is accomplished using a sensitivity analysis of the estimated bias to alternate thorium cross-section data used in calculations of k/sub infinity/(T). The envelope of uncertainty expected to contain the actual values for the temperature coefficient of the reactivity for the 233 U-fueled HTGR lattice studied remains negative at 1600 K (1327 0 C). Although a broader base of experimental data with improved accuracy is always desirable, the existing data base provided by the HTLTR experiments is judged to be adequate for the verification of neutronic calculations for the HTGR containing 233 U fuel at its current state of development

  13. Early fusion reactor neutronic calculations: A reevaluation

    International Nuclear Information System (INIS)

    Perry, R.T.

    1996-01-01

    Several fusion power plant design studies were made at a number of universities and laboratories in the late 1960s and early 1970s. These studies included such designs as the Princeton Plasma Physics Laboratory Fusion Power Plan and the University of Wisconsin UWMAK-I Reactor Neutronic analyses of the blankets and shields were part of the studies. During this time there were dissertations written on neutronic analysis systems and the results of neutronic analysis on several blanket and shield designs. The results were presented in the literature. Now in the fifth decade of fusion research, investigators often return to the earlier analyses for the neutronic results that are applicable to current blanket and shield designs, with the idea of using the older work as a basis for the new. However, the analyses of the past were made with cross-section data sets that have long been replaced with more modern versions. In addition, approximations were often made to the cross sections used because more exact data were not available. Because these results are used as guides, it is important to know if they are reproducible using more modern data. In this paper, several of the neutronic calculations made in the early studies are repeated using the MATXS-11 data library. This library is the ENDF/B-VI version of the MATXS-5 library. The library has 80 neutron groups. Tritium breeding ratios, heating rates, and fluxes are calculated and compared. This transport code used here is the one- dimensional S n code, ONEDANT. It is important to note that the calculations here are not to be considered as benchmarks because parameter and sensitivity studies were not made. They are used only to see if the results of older calculations are in reasonable agreement with a more modern library

  14. Reactor calculations in aid of isotope production at SAFARI-1

    International Nuclear Information System (INIS)

    Ball, G.

    2003-01-01

    Varying levels of reactor physics support is given to the isotope production industry. As the pressures on both the safety limits and economical production of reactor produced isotopes mount, reactor physics calculational support is playing an ever increasing role. Detailed modelling of the reactor, irradiation rigs and target material enables isotope production in reactors to be maximised with respect to yields and quality. NECSA's methodology in this field is described and some examples are given. (author)

  15. Uncertainty Analysis of Light Water Reactor Fuel Lattices

    Directory of Open Access Journals (Sweden)

    C. Arenas

    2013-01-01

    Full Text Available The study explored the calculation of uncertainty based on available cross-section covariance data and computational tool on fuel lattice levels, which included pin cell and the fuel assembly models. Uncertainty variations due to temperatures changes and different fuel compositions are the main focus of this analysis. Selected assemblies and unit pin cells were analyzed according to the OECD LWR UAM benchmark specifications. Criticality and uncertainty analysis were performed using TSUNAMI-2D sequence in SCALE 6.1. It was found that uncertainties increase with increasing temperature, while kinf decreases. This increase in the uncertainty is due to the increase in sensitivity of the largest contributing reaction of uncertainty, namely, the neutron capture reaction 238U(n, γ due to the Doppler broadening. In addition, three types (UOX, MOX, and UOX-Gd2O3 of fuel material compositions were analyzed. A remarkable increase in uncertainty in kinf was observed for the case of MOX fuel. The increase in uncertainty of kinf in MOX fuel was nearly twice the corresponding value in UOX fuel. The neutron-nuclide reaction of 238U, mainly inelastic scattering (n, n′, contributed the most to the uncertainties in the MOX fuel, shifting the neutron spectrum to higher energy compared to the UOX fuel.

  16. The validity of METHUSELAH II in water moderated lattice calculations

    International Nuclear Information System (INIS)

    Hicks, D.; Hopkins, D.R.

    1964-09-01

    An improved version of the METHUSELAH code has been developed, which embodies some refinements in the treatment of the thermal spectrum, improved cross-section data, and a neutron balance output. The changes in nuclear data and physical models are summarised in this report; a detailed description of the programme modifications will be published separately. In this report METHUSELAH II predictions are compared with published lattice reactivity and reaction rate data. The systems examined include British S.G.H.W. type lattices (with H 2 O and D 2 O moderation), Canadian natural uranium/D 2 0 experiments, U.S. low enrichment H 2 O systems, and the Hanford Pu A1/H 2 O experiments. In general the agreement is sufficiently good to demonstrate the value of METHUSELAH II as an assessment tool and to indicate clear improvements over METHUSELAH I. A number of discrepancies are, however, observed and are the subject of comment. (author)

  17. Efficiency of free-energy calculations of spin lattices by spectral quantum algorithms

    International Nuclear Information System (INIS)

    Master, Cyrus P.; Yamaguchi, Fumiko; Yamamoto, Yoshihisa

    2003-01-01

    Ensemble quantum algorithms are well suited to calculate estimates of the energy spectra for spin-lattice systems. Based on the phase estimation algorithm, these algorithms efficiently estimate discrete Fourier coefficients of the density of states. Their efficiency in calculating the free energy per spin of general spin lattices to bounded error is examined. We find that the number of Fourier components required to bound the error in the free energy due to the broadening of the density of states scales polynomially with the number of spins in the lattice. However, the precision with which the Fourier components must be calculated is found to be an exponential function of the system size

  18. WIMSD4 calculations of the Westinghouse 'EDASA' lattices with plutonium dioxide fuel

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1977-03-01

    A series of Westinghouse critical PuO 2 /UO 2 pin-cell assemblies is analysed using the lattice code WIMSD4. The results are presented in terms of computed k-effective values, with comment on the choice of method for calculating high leakage systems and on the adequacy of WIMSD4 for evaluating plutonium enriched lattices. (author)

  19. Statistic method of research reactors maximum permissible power calculation

    International Nuclear Information System (INIS)

    Grosheva, N.A.; Kirsanov, G.A.; Konoplev, K.A.; Chmshkyan, D.V.

    1998-01-01

    The technique for calculating maximum permissible power of a research reactor at which the probability of the thermal-process accident does not exceed the specified value, is presented. The statistical method is used for the calculations. It is regarded that the determining function related to the reactor safety is the known function of the reactor power and many statistically independent values which list includes the reactor process parameters, geometrical characteristics of the reactor core and fuel elements, as well as random factors connected with the reactor specific features. Heat flux density or temperature is taken as a limiting factor. The program realization of the method discussed is briefly described. The results of calculating the PIK reactor margin coefficients for different probabilities of the thermal-process accident are considered as an example. It is shown that the probability of an accident with fuel element melting in hot zone is lower than 10 -8 1 per year for the reactor rated power [ru

  20. Radionuclide inventory calculation in VVER and BWR reactor

    International Nuclear Information System (INIS)

    Bouhaddane, A.; Farkas, F.; Slugen, V.; Ackermann, L.; Schienbein, M.

    2014-01-01

    The paper shows different aspects in the radionuclide inventory determination. Precise determination of the neutron flux distribution, presented for a BRW reactor, is vital for the activation calculations. The precision can be improved utilizing variance reduction methods as importance treatment, weight windows etc. Direct calculation of the radionuclide inventory via Monte Carlo code is presented for a VVER reactor. Burn-up option utilized in this calculation appears to be proper for reactor internal components. However, it will not be probably effective outside the reactor core. Further calculations in this area are required to support the forth-set findings. (authors)

  1. Method and program for complex calculation of heterogeneous reactor

    International Nuclear Information System (INIS)

    Kalashnikov, A.G.; Glebov, A.P.; Elovskaya, L.F.; Kuznetsova, L.I.

    1988-01-01

    An algorithm and the GITA program for complex one-dimensional calculation of a heterogeneous reactor which permits to conduct calculations for the reactor and its cell simultaneously using the same algorithm are described. Multigroup macrocross sections for reactor zones in the thermal energy range are determined according to the technique for calculating a cell with complicate structure and then the continuous multi group calculation of the reactor in the thermal energy range and in the range of neutron thermalization is made. The kinetic equation is solved using the Pi- and DSn- approximations [fr

  2. Seismic calculations for underground reactor buildings

    International Nuclear Information System (INIS)

    Altes, J.; Koschmieder, D.

    1977-08-01

    Embedding the buildings in soil changes their seismic response behaviour as compared to surface buildings, i.e. higher stiffness and increased radiation damping is attained. Finite element models are best suited for determinig the effects of embedment and of layered subsoil. The code used was the LUSH2-programme, which is applicable to 2-dimensional problems and provides an approximate treatment for non-linear dynamic soil behaviour. For embedded buildings there is a good agreement between 2- and 3-dimensional models of the response for points below the soil surface. It is therefore permissible to use the less costly 2-dimensional programmes. To simulate earthquake, three different acceleration-time histories, derived from actual measurements and from artificial synthesis, with differing response spectra were fed in. The soil characteristics assumed are applicable to a representative site in Germany. Three different types of models were examined, using analytical models with only a few elements for parametric studies and with up to 716 elements for more precise calculations. A comparison was made between the semi-embedment, the total embedment, and installation of the reactor building above-ground. (orig.) [de

  3. Study on the output from programs in calculating lattice with transverse coupling

    International Nuclear Information System (INIS)

    Xu Jianming

    1994-01-01

    SYNCH and MAD outputs in calculating lattice with coordinate rotation have been studied. The result shows that the four dispersion functions given by SYNCH output in this case are wrong. There are large discrepancies between the Twiss Parameters given by these two programs. One has to be careful in using these programs to calculate or match lattices with coordinate rotations (coupling between two transverse motions) so that to avoid wrong results

  4. IRT-type research reactor physical calculation methodology

    International Nuclear Information System (INIS)

    Carrera, W.; Castaneda, S.; Garcia, F.; Garcia, L.; Reyes, O.

    1990-01-01

    In the present paper an established physical calculation procedure for the research reactor of the Nuclear Research Center (CIN) is described. The results obtained by the method are compared with the ones reported during the physical start up of a reactor with similar characteristics to the CIN reactor. 11 refs

  5. Transmutation of waste actinides in thermal reactors: survey calculations of candidate irradiation schemes

    International Nuclear Information System (INIS)

    Gorrell, T.C.

    1978-11-01

    Actinide recycle and transmutation calculations were made for twelve specific thermal reactor environments. The calculations included H 2 O-moderated reactor lattices with enriched U, recycled Pu, and 233 ' 235 U-Th. In addition two D 2 O reactor cases were calculated. When all actinides were recycled into 235 U-enriched fuel, about 10 percent of the transuranic actinides were fissioned per 3-year fuel cycle. About 9 percent of the actinides were fissioned per 3-year fuel cycle when waste actinides (no U or Pu) were irradiated in separate target rods in a U-fuel assembly. When actinides were recycled in separate target assemblies, the fission rate was strongly dependent on the specific loading of the target. Fission rates of 5 to 10 percent per 3-year fuel cycle were observed

  6. Development of M3C code for Monte Carlo reactor physics criticality calculations

    International Nuclear Information System (INIS)

    Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.

    2015-06-01

    The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)

  7. Strongly coupled gauge theories: What can lattice calculations teach us?

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    Electroweak symmetry breaking and the dynamical origin of the Higgs boson are central questions today. Strongly coupled systems predicting the Higgs boson as a bound state of a new gauge-fermion interaction are candidates to describe beyond Standard Model physics. The phenomenologically viable models are strongly coupled, near the conformal boundary, requiring non-perturbative studies to reveal their properties. Lattice studies show that many of the beyond-Standard Model candidates have a relatively light isosinglet scalar state that is well separated from the rest of the spectrum. When the scale is set via the vev of electroweak symmetry breaking, a 2 TeV vector resonance appears to be a general feature of many of these models with several other resonances that are not much heavier.

  8. Calculation of research reactor RA power at uncontrolled reactivity changes

    International Nuclear Information System (INIS)

    Cupac, S.

    1978-01-01

    The safety analysis of research reactor RA involves also the calculation of reactor power at uncontrolled reactivity changes. The corresponding computer code, based on Point Kinetics Model has been made. The short review of method applied for solving kinetic equations is given and several examples illustrating the reactor behaviour at various reactivity changes are presented. The results already obtained are giving rather rough picture of reactor behaviour in considered situations. This is the consequence of using simplified feed back and reactor cooling models, as well as temperature reactivity coefficients, which do not correspond to the actual reactor RA structure (which is now only partly fulfilled with 80% enriched uranium fuel). (author) [sr

  9. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  10. Calculation of the void reactivity of CANDU lattices using the SCALE code system

    Energy Technology Data Exchange (ETDEWEB)

    Valko, J. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Feher, S. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Slobben, J. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1995-11-01

    The reactivity effect of coolant voiding in CANDU-type fuel lattices has been calculated with different methods using the SCALE code system. The known positive void reactivity coefficient of the original lattice was correctly obtained. A modified fuel bundle containing dysprosium and slightly enriched uranium to eliminate the positive reactivity effect was also calculated. Owing to the increased heterogeneity of this modified fuel the one-dimensional cylindrical calculation with XSDRN proved to be inadequate. Code options allowing bundle geometry were successfully used for the calculation of the strongly space dependent flux and spectrum changes which determine the void reactivity. (orig.).

  11. Burnup calculation for a tokamak commercial hybrid reactor

    International Nuclear Information System (INIS)

    Feng Kaiming; Xie Zhongyou

    1990-08-01

    A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

  12. Local structure theory: calculation on hexagonal arrays, and interaction of rule and lattice

    International Nuclear Information System (INIS)

    Gutowitz, H.A.; Victor, J.D.

    1989-01-01

    Local structure theory calculations are applied to the study of cellular automata on the two-dimensional hexagonal lattice. A particular hexagonal lattice rule denoted (3422) is considered in detail. This rule has many features in common with Conway's Life. The local structure theory captures many of the statistical properties of this rule; this supports hypotheses raised by a study of Life itself. As in Life, the state of a cell under (3422) depends only on the state of the cell itself and the sum of states in its neighborhood at the previous time step. This property implies that evolution rules which operate in the same way can be studied on different lattices. The differences between the behavior of these rules on different lattices are dramatic. The mean field theory cannot reflect these differences. However, a generalization of the mean field theory, the local structure theory, does account for the rule-lattice interaction

  13. Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

    2016-06-15

    Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

  14. Impact of neutron resonance treatments on reactor calculation

    International Nuclear Information System (INIS)

    Leszczynski, F.

    1988-01-01

    The neutron resonance treatment on reactor calculation is one of the not completely resolved problems of reactor theory. The calculation required on design, fuel management and accident analysis of nuclear reactors contains adjust coefficients and semi-empirical values introduced on the computer codes; these values are obtained comparing calculation results with experimental values and more exact calculation results. This is made when the characteristics of the analyzed system are such that this type of comparisons are possible. The impact that one fixed resonance treatment method have on the final evaluation of physics reactor parameters, reactivity, power distribution, etc., is useful to know. In this work, the differences between calculated parameters with two different methods of resonance treatment in cell calculations are shown. It is concluded that improvements on resonance treatment are necessary for growing the reliability on core calculations results. Finally, possible improvements, easy to implement in current computer codes, are presented. (Author) [es

  15. Calculated investigation of actinide transmutation in the BOR-60 reactor

    International Nuclear Information System (INIS)

    Zhemkov, I.Yu.; Ishunina, O.V.; Yakovleva, I.V.

    2000-01-01

    One of the prospective actinide burner reactor type is the fast reactor with a 'hard' spectrum and small breeding factor, which is the BOR-60. The calculated investigations demonstrate that Loading up to 40% of minor-actinides to the BOR-60 reactor did not lead to the considerable change of neutron-physical characteristics. The performed calculations show that the BOR- 60 reactor possesses a high efficiency of the minor-actinide and plutonium bum-up (up to 37 kg/(TW · h)) hat is comparable with properties of the actinide burner-reactors under design. The BOR-60 reactor can provide a homogeneous minor-actinide Loading (minor-actinide addition to the standard fuel) to the core and heterogeneous Loading (as separate assemblies-targets with a high minor-actinide fraction) to the first rows of a radial blanket that allows the optimum usage of the reactor and its characteristics. (authors)

  16. Lattice dynamics and thermal conductivity of lithium fluoride via first-principles calculations

    Science.gov (United States)

    Liang, Ting; Chen, Wen-Qi; Hu, Cui-E.; Chen, Xiang-Rong; Chen, Qi-Feng

    2018-04-01

    The lattice thermal conductivity of lithium fluoride (LiF) is accurately computed from a first-principles approach based on an iterative solution of the Boltzmann transport equation. Real-space finite-difference supercell approach is employed to generate the second- and third-order interatomic force constants. The related physical quantities of LiF are calculated by the second- and third- order potential interactions at 30 K-1000 K. The calculated lattice thermal conductivity 13.89 W/(m K) for LiF at room temperature agrees well with the experimental value, demonstrating that the parameter-free approach can furnish precise descriptions of the lattice thermal conductivity for this material. Besides, the Born effective charges, dielectric constants and phonon spectrum of LiF accord well with the existing data. The lattice thermal conductivities for the iterative solution of BTE are also presented.

  17. Calculation of power density with MCNP in TRIGA reactor

    International Nuclear Information System (INIS)

    Snoj, L.; Ravnik, M.

    2006-01-01

    Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (author)

  18. Fast reactor calculational route for Pu burning core design

    Energy Technology Data Exchange (ETDEWEB)

    Hunter, S. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted. The document includes comments on and explanations of the modelling assumptions in the various calculations. Practical information on the use of the calculational route and the computer systems is also given. (J.P.N.)

  19. Lattice calculation of hadronic weak matrix elements: the ΔI = 1/2 rule

    International Nuclear Information System (INIS)

    Bernard, C.

    1984-01-01

    A lattice Monte Carlo technique for calculating the matrix elements of weak operators is described. Emphasis is placed on the ΔI = 1/2 rule, which is such a large effect that the significant errors associated with current lattice methods (statistics, finite size, finite lattice spacing, extrapolations in quark mass, etc.) should not disguise the important qualitative features. A detailed exposition of the analytic bases for the calculation is given, and an attempt is made to avoid the questionable phenomenological assumptions (such as some of those inherent in the Penguin approach) which were necessary when matrix elements could not be calculated. The current state of the calculation-in-progress is described. This work is being done in collaboration with A. Soni, T. Draper, G. Hockney, and M. Rushton

  20. Calculation Of A Lattice Physics Parameter For SBWR Fuel Bundle Design

    International Nuclear Information System (INIS)

    Sardjono, Y.

    1996-01-01

    The maximum power peaking factor for Nuclear Power Plant SBWR type is 1.5. The precision for that calculation is related with the result of unit cell analysis each rod in the fuel bundles. This analysis consist of lattice eigenvalue, lattice average diffusion cross section as well as relative power peaking factor in the fuel rod for each fuel bundles. The calculation by using TGBLA computer code which is based on the transport and 168 group diffusion theory. From this calculation can be concluded that the maximum relative power peaking factor is 1.304 and lower than design limit

  1. Calculations on neutron irradiation damage in reactor materials

    International Nuclear Information System (INIS)

    Sone, Kazuho; Shiraishi, Kensuke

    1976-01-01

    Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

  2. MULTI - A multigroup or multipoint P{sub 3} programme for calculating thermal neutron spectra in a reactor cell

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1968-06-15

    Programme MULTI calculates the space energy distribution of thermal neutrons in a multizone, cylindrical, infinitely long reactor lattice by using the multigroup or multipoint P{sub 3} approximation. This report presents a short description of the algorithm and the programme and gives the instructions for its exploitation. (author)

  3. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  4. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  5. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-01-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  6. Numerical calculation of hadron masses in lattice quantum chromodynamics

    International Nuclear Information System (INIS)

    Montvay, I.

    1985-07-01

    Recent numerical Monte Carlo simulations of the hadron spectrum are reviewed. After a general introduction, different ways of calculating the hadron masses in the ''quenched approximation'' (i.e. neglecting virtual quark loops) are described and the latest results are summarized. The pseudofermion method and the iterative hopping expansion method for the introduction of dynamical quarks is discussed, and the first results about the hadron spectrum including the effect of virtual quark loops are reviewed. A separate section is devoted to the discussion of the questions related to scaling with dynamical quarks. (orig./HSI)

  7. Thermodynamic cycle calculations for a pumped gaseous core fission reactor

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Van Dam, H.

    1991-01-01

    Finite and 'infinitesimal' thermodynamic cycle calculations have been performed for a 'solid piston' model of a pumped Gaseous Core Fission Reactor with dissociating reactor gas, consisting of Uranium, Carbon and Fluorine ('UCF'). In the finite cycle calculations the influence has been investigated of several parameters on the thermodynamics of the system, especially on the attainable direct (nuclear to electrical) energy conversion efficiency. In order to facilitate the investigation of the influence of dissociation, a model gas, 'Modelium', was developed, which approximates, in a simplified, analytical way, the dissociation behaviour of the 'real' reactor gas. Comparison of the finite cycle calculation results with those of a so-called infinitesimal Otto cycle calculation leads to the conclusion that the conversion efficiency of a finite cycle can be predicted, without actually performing the finite cycle calculation, with reasonable accuracy, from the so-called 'infinitesimal efficiency factor', which is determined only by the thermodynamic properties of the reactor gas used. (author)

  8. Reconstruction calculation of pin power for ship reactor core

    International Nuclear Information System (INIS)

    Li Haofeng; Shang Xueli; Chen Wenzhen; Wang Qiao

    2010-01-01

    Aiming at the limitation of the software that pin power distribution for ship reactor core was unavailable, the calculation model and method of the axial and radial pin power distribution were proposed. Reconstruction calculations of pin power along axis and radius was carried out by bicubic and bilinear interpolation and cubic spline interpolation, respectively. The results were compared with those obtained by professional reactor physical soft with fine mesh difference. It is shown that our reconstruction calculation of pin power is simple and reliable as well as accurate, which provides an important theoretic base for the safety analysis and operating administration of the ship nuclear reactor. (authors)

  9. Application of fuel management calculation codes for CANDU reactor

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun

    2003-01-01

    Qinshan Phase III Nuclear Power Plant adopts CANDU-6 reactors. It is the first time for China to introduce this heavy water pressure tube reactor. In order to meet the demands of the fuel management calculation, DRAGON/DONJON code is developed in this paper. Some initial fuel management calculations about CANDU-6 reactor of Qinshan Phase III are carried out using DRAGON/DONJON code. The results indicate that DRAGON/DONJON can be used for the fuel management calculation for Qinshan Phase III

  10. Experimental determination of lattice parameters for 2% enriched uranium heavy water reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Takac, S; Markovic, H; Bosevski, T [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1963-04-15

    Systematic measurements of the buckling, infinite multiplication factor and the thermal utilization factor were made on a series of lattices for 2% enriched uranium tubular fuel elements in heavy water. This work represents the first phase of experimental verification of standard theoretical methods used for the determination of reactor parameters.

  11. Iterative resonance self-shielding methods using resonance integral table in heterogeneous transport lattice calculations

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Kim, Kang-Seog

    2011-01-01

    This paper describes the iteration methods using resonance integral tables to estimate the effective resonance cross sections in heterogeneous transport lattice calculations. Basically, these methods have been devised to reduce an effort to convert resonance integral table into subgroup data to be used in the physical subgroup method. Since these methods do not use subgroup data but only use resonance integral tables directly, these methods do not include an error in converting resonance integral into subgroup data. The effective resonance cross sections are estimated iteratively for each resonance nuclide through the heterogeneous fixed source calculations for the whole problem domain to obtain the background cross sections. These methods have been implemented in the transport lattice code KARMA which uses the method of characteristics (MOC) to solve the transport equation. The computational results show that these iteration methods are quite promising in the practical transport lattice calculations.

  12. MURLI, 1-D Flux, Reaction Rate in Cylindrical Geometry Thermal Reactor Lattice by Transport

    International Nuclear Information System (INIS)

    Huria, H.C.

    1985-01-01

    1 - Description of problem or function: MURLI is an integral transport theory code to calculate fluxes and reaction rates in one- dimensional cylindrical geometry lattice cells of a thermal reactor. For a specified buckling, it computes k-effective using few-group diffusion theory and a few-group collapsed set of Cross sections. The code can optionally be used to solve a first order differential equation for the number density of fissile, fertile and fission product nuclei as a function of time, and to recalculate fluxes, reaction rates and k-effective at different stages of burnup. A 27-group cross section data library is included. There are four pseudo-fission products each associated with the decay chains of plutonium and uranium isotopes in addition to Rh-105, Xe-135, Np-239, U-236, Am-241, Am-242 and Am-243. There is also data for one lumped pseudo-fission product. 2 - Method of solution: Multiple collision probabilities and escape probabilities are calculated for each cylindrical shell region assuming protons are born uniformly and isotropically over the entire region volume. The equations of integral transport theory can then be solved for neutron flux. The first order differential burnup equation is solved by a fourth order Runge-Kutta method. 3 - Restrictions on the complexity of the problem: There are maxima of 8 fissionable elements, 8 resonant elements, and 20 spatial regions

  13. Calculations of radiation levels during reactor operations for safeguard inspections

    International Nuclear Information System (INIS)

    Sobhy, M.

    1996-01-01

    When an internal core spent fuel storage is used in the shield tank to accommodate a large number of spent fuel baskets, physical calculations are performed to determine the number of these spent fuel elements which can be accommodated and still maintain the gamma activity outside under the permissible limit. The corresponding reactor power level is determined. The radioactivity calculations are performed for this internal storage at different axial levels to avoid the criticality of the reactor core. Transport theory is used in calculations based on collision probability for multi group cell calculations. Diffusion theory is used in three dimensions in the core calculations. The nuclear fuel history is traced and radioactive decay is calculated, since reactor fission products are very sensitive to power level. The radioactivity is calculated with a developed formula based on fuel basket loading integrity. (author)

  14. Errors due to the cylindrical cell approximation in lattice calculations

    Energy Technology Data Exchange (ETDEWEB)

    Newmarch, D A [Reactor Development Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1960-06-15

    It is shown that serious errors in fine structure calculations may arise through the use of the cylindrical cell approximation together with transport theory methods. The effect of this approximation is to overestimate the ratio of the flux in the moderator to the flux in the fuel. It is demonstrated that the use of the cylindrical cell approximation gives a flux in the moderator which is considerably higher than in the fuel, even when the cell dimensions in units of mean free path tend to zero; whereas, for the case of real cells (e.g. square or hexagonal), the flux ratio must tend to unity. It is also shown that, for cylindrical cells of any size, the ratio of the flux in the moderator to flux in the fuel tends to infinity as the total neutron cross section in the moderator tends to zero; whereas the ratio remains finite for real cells. (author)

  15. Calculations for accidents in water reactors during operation at power

    International Nuclear Information System (INIS)

    Blanc, H.; Dutraive, P.; Fabrega, S.; Millot, J.P.

    1976-07-01

    The behaviour of a water reactor on an accident occurring as the reactor is normally operated at power may be calculated through the computer code detailed in this article. Reactivity accidents, loss of coolant ones and power over-running ones are reviewed. (author)

  16. TORT application in reactor pressure vessel neutron flux calculations

    International Nuclear Information System (INIS)

    Belousov, S.I.; Ilieva, K.D.; Antonov, S.Y.

    1994-01-01

    The neutron flux values onto reactor pressure vessel for WWER-1000 and WWER-440 reactors, at the places important for metal embrittlement surveillance have been calculated by 3 dimensional code TORT and synthesis method. The comparison of the results received by both methods confirms their good consistency. (authors). 13 refs., 4 tabs

  17. Elaboration and qualification of a reference calculation routes for the absorbers in the PWR reactors

    International Nuclear Information System (INIS)

    Blanc-Tranchant, P.

    1999-11-01

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B 4 C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  18. Development and qualification of reference calculation schemes for absorbers in pressured water reactor

    International Nuclear Information System (INIS)

    Blanc-Tranchant, P.

    2001-01-01

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code APOLLO2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B4C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI4. They were then checked against experimental data measured during French experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  19. Reactivity and reaction rate ratio changes with moderator voidage in a light water high converter reactor lattice

    International Nuclear Information System (INIS)

    Chawla, R.; Gmuer, K.; Hager, H.; Seiler, R.

    1986-01-01

    Integral reaction rate ratios and other k ∞ related parameters have been measured in the first three cores of the experimental program on light water high converter reactor (LWHCR) test lattices in the PROTEUS reactor. The reference tight-pitch lattice consisted of two rod types, with an average fissile-plutonium enrichment of 6% and a fuel/moderator ratio of 2.0. The moderators were H 2 O, Dowtherm (simulating an H 2 O voidage of 42.5%), and air (100% void). Comparisons of the measured parameters have been made with calculational results based mainly on the use of two separate codes and their associated data libraries, namely, WIMS-D and EPRI-CPM. A reconstruction of individual components of the k-infinity void coefficient has been carried out on the basis of the measured changes with voidage of the various reaction rate ratios, as well as of k-infinity itself. The subsequent more detailed comparisons between experiment and calculation should provide a useful basis for resolving the conflicting calculational results that have been reported in the past for the void coefficient characteristics of LWHCRs. (author)

  20. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)

    2008-07-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  1. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    International Nuclear Information System (INIS)

    Shwageraus, E.; Fridman, E.

    2008-01-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO 2 fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO 2 LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  2. Measurement and Calculation of Gamma Radiation from HWZPR Reactor

    International Nuclear Information System (INIS)

    Jalali, Majid

    2006-01-01

    HWZPR is a research reactor with natural uranium fuel, D 2 O moderator and graphite reflector with maximum power of 100 W. It is a suitable means for theoretical research and heavy water reactor experiments. Neutrons from the core participate in different nuclear reactions by interactions with fuel, moderator, graphite and the concrete around the reactor. The results of these interactions are the production of prompt gammas in the environment. Useful information is gained by the reactor gamma spectrum measurement from point of view of relative quantity and energy distribution of direct and scattered radiations. Reactor gamma ray spectrum has been gathered in different places around the reactor by HPGe detector. In analysis of these spectra, 1 H(n,γ) 2 H, 16 O(n,n'γ) 16 O, 2 H(n,γ) 3 H and 238 U(n,γ) 239 U reactions occurring in reactor moderator and fuel, are important. The measured spectrum has been primarily estimated by the MCNP code. There is agreement between the code and the experiments in some points. The scattered gamma rays from 27 Al (n,γ) 28 Al reaction in the reactor tank, are the most among the gammas scattered in the reactor environment. Also the dose calculations by MCNP code show that 72% of gamma dose belongs to the energy range 3-11 MeV from reactor gamma spectrum and the danger of exposure from the reactor high-energy photons is serious. (author)

  3. Adaptation of GRS calculation codes for Soviet reactors

    International Nuclear Information System (INIS)

    Langenbuch, S.; Petri, A.; Steinborn, J.; Stenbok, I.A.; Suslow, A.I.

    1994-01-01

    The use of ATHLET for incident calculation of WWER has been tested and verified in numerous calculations. Further adaptation may be needed for the WWER 1000 plants. Coupling ATHLET with the 3D nuclear model BIPR-8 for WWER cores clearly improves studies of the influence of neutron kinetics. In the case of FBMK reactors ATHLET calculations show that typical incidents in the complex RMBK reactors can be calculated even though verification still has to be worked on. Results of the 3D-core model QUABOX/CUBBOX-HYCA show good correlation of calculated and measured values in reactor plants. Calculations carried out to date were used to check essential parameters influencing RBMK core behaviour especially dependence of effective voidre activity on the number of control rods. (orig./HP) [de

  4. Calculation of induced activity in the V-230 reactor

    International Nuclear Information System (INIS)

    Bouhahhane, A.; Farkas, G.

    2013-01-01

    In this paper, we focused on the calculation of the neutron induced activity of nuclear reactor components for decommissioning purposes. The results confirm, that the most important radionuclides in the reactor components dismantling process are 55 Fe (1 st decade), 60 Co (10 - 50 y) and 63 Ni (during the whole process). Another aim of this paper was to refer to the possibility to improve the accuracy of the calculations using continuous energy Monte Carlo methods. (authors)

  5. Time-independent lattice Boltzmann method calculation of hydrodynamic interactions between two particles

    Science.gov (United States)

    Ding, E. J.

    2015-06-01

    The time-independent lattice Boltzmann algorithm (TILBA) is developed to calculate the hydrodynamic interactions between two particles in a Stokes flow. The TILBA is distinguished from the traditional lattice Boltzmann method in that a background matrix (BGM) is generated prior to the calculation. The BGM, once prepared, can be reused for calculations for different scenarios, and the computational cost for each such calculation will be significantly reduced. The advantage of the TILBA is that it is easy to code and can be applied to any particle shape without complicated implementation, and the computational cost is independent of the shape of the particle. The TILBA is validated and shown to be accurate by comparing calculation results obtained from the TILBA to analytical or numerical solutions for certain problems.

  6. Study of heterogeneous nuclear reactor lattice properties in multizone systems

    International Nuclear Information System (INIS)

    Raisic, N.M.

    1964-12-01

    Described analysis of substitution experiments and its comparison with the classic procedures showed that it could be successfully applied for processing the experimental results. This method shows some advantages and some deficiencies compared to classic methods. Precision of the method is considered sufficient for design of nuclear facilities. Routine standards used in design of nuclear facilities demand following precision of nuclear parameters: about 5% during feasibility study and design of heavy water facility with natural uranium; 2 - 3% in the phase of parameters optimisation and preparing the main project; 1% during optimization of operation. Accuracy of core parameters obtained by analysis of substitution experiments show that they could be successfully used in the phase of reactor parameters optimisation. It is possible to increase the precision of critical parameters and thus apply the proposed method for analysis of reactor parameters needed in the phase of operation optimization

  7. Application of perturbation theory to lattice calculations based on method of cyclic characteristics

    Science.gov (United States)

    Assawaroongruengchot, Monchai

    computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and keff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and keff-EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these techniques to the CVR

  8. Application of perturbation theory to lattice calculations based on method of cyclic characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Assawaroongruengchot, M

    2007-07-01

    computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k{sub eff} at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and k{sub eff}-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and k{sub eff}-EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these

  9. Application of perturbation theory to lattice calculations based on method of cyclic characteristics

    International Nuclear Information System (INIS)

    Assawaroongruengchot, M.

    2007-01-01

    computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and k eff -EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and k eff -EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these techniques to the CVR

  10. Calculation of the evolution of molten salt breeder reactor

    International Nuclear Information System (INIS)

    Esteves, Fernando de Avelar

    1999-01-01

    A forecast for the future electrical consumption in Brazil and forecast of the nuclear electrical generation demand are discussed in this paper, which includes also an analysis on advanced nuclear reactors concept to supply that demand. This paper presents a concise description of the Molten Salt Breeder Reactor, considered the most appropriated to meet that demand. This paper also presents the burnup calculation modeling, including the operation modeling of this type of reactor from an initial load o 233 U up to the equilibrium cycle, the results of these calculations and its analysis. (author)

  11. Subchannel analysis of 37-rod tight-lattice bundle experiments for reduced-moderation water reactor

    International Nuclear Information System (INIS)

    Nakatsuka, Toru; Tamai, Hidesada; Akimoto, Hajime

    2005-01-01

    R and D project to investigate thermal-hydraulic performance of tight-lattice fuel bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in collaboration with utilities, reactor vendors and universities from 2002. The RMWR realizes a high conversion ratio larger than 0.1 for sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The reactor core comprises tight-lattice fuel assemblies with gap clearance of around 1.0 mm to reduce the water volume ratio to achieve the high conversion ratio. A problem of utmost importance from a thermal-hydraulic point of view is the coolability of the tight-lattice assembly with such a small gap width. JAERI has been carrying out experimental study to investigate the system parameter effects on the thermal-hydraulic performance and to confirm the feasibility of the core. In the present study, the subchannel analysis code NASCA was applied to 37-rod tight-lattice bundle experiments. The NASCA can give good predictions of critical power for the gap width of 1.3 mm while the prediction accuracy decreases for the gap width of 1.0 mm. To improve the prediction accuracy, the code will be modified to take the effect of film thickness distribution around fuel rods on boiling transition. (author)

  12. Enrichment reduction calculations for the DIDO reactor. App. B

    International Nuclear Information System (INIS)

    Constantine, G.; Javadi, M.; Thick, E.

    1985-01-01

    The possibility has been raised that DIDO/PLUTO type heavy water moderated reactors can be operated with fuel of lower than the 75% enrichment material currently in use with the object of increasing the proliferation resistance of the fuel cycle. This paper sets out to examine the reactor physics aspects of enrichment reductions to 45% and 20% for Harwell's MTR's as part of an IAEA collaborative exercise currently being conducted to examine the topic in a more general way for the whole class of heavy water moderated reactors. The reactor physics tool used at Harwell is WIMSE, the Winfrith Improved Multigroup Scheme, a suite of linked reactor physics codes which has been used extensively for light water, heavy water and graphite moderated thermal reactors. The course of the calculations and the WIMSE modules involved in this study are described briefly

  13. Progress in the improved lattice calculation of direct CP-violation in the Standard Model

    Science.gov (United States)

    Kelly, Christopher

    2018-03-01

    We discuss the ongoing effort by the RBC & UKQCD collaborations to improve our lattice calculation of the measure of Standard Model direct CP violation, ɛ', with physical kinematics. We present our progress in decreasing the (dominant) statistical error and discuss other related activities aimed at reducing the systematic errors.

  14. Neutron thermalization in reactor lattice cells: An NPY-project report

    International Nuclear Information System (INIS)

    Stamm'ler, R.J.J.; Takac, S.M.; Weiss, Z.J.

    1966-01-01

    The NPY-Project is a joint research programme in reactor physics between Norway, Poland, Yugoslavia and the International Atomic Energy Agency. One of the tasks of the project was to make a theoretical and experimental investigation of the phenomena of neutron thermalization in lattice cells, and this work is covered by the present monograph. The different lattices of the zero-power assemblies in the NPY countries offered ample opportunity for the theoreticians and experimentalists to test and compare their methods, and the exchange of experiences was stimulating and valuable. 85 refs, 26 figs, 19 tabs

  15. Variable stiffness lattice support system for a condenser type nuclear reactor containment

    International Nuclear Information System (INIS)

    George, J.A.; Sutherland, J.D.

    1979-01-01

    A support structure for the lattice supporting a fusible material in the annular condenser region of a nuclear reactor containment, the flexibility of which structure can be selectively adjusted in accordance with seismic or other loading requirements. The lattice is affixed to a flexible member in a manner which allows relative movement between the two components. The flexible member is affixed to a rigid support member in a manner which selectively adjusts the resiliency of the flexible member. The support member is rigidly affixed to a wall of the containment annulus, and can also be utilized to support cooling ducts. 6 claims

  16. Description of the CAREM Reactor Neutronic Calculation Codes

    International Nuclear Information System (INIS)

    Villarino, Eduardo; Hergenreder, Daniel

    2000-01-01

    In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included

  17. Calculations in the weak and crossover regions of SU(2) lattice gauge theory

    International Nuclear Information System (INIS)

    Greensite, J.; Hansson, T.H.; Hari Dass, N.D.; Lauwers, P.G.

    1981-07-01

    A calculational scheme for lattice gauge theory is proposed which interpolates between lowest order mean-field and full Monte-Carlo calculations. The method is to integrate over a restricted set of link variables in the functional integral, with the remainder fixed at their mean-field value. As an application the authors compute small SU(2) Wilson loops near and above the weak-to-strong coupling transition point. (Auth.)

  18. Calculation of photon dose for Dalat research reactor in case of loss of reactor tank water

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2007-01-01

    Photon sources of actinides and fission products were estimated by ORIGEN2 code with the modified cross-section library for Dalat research reactor (DRR) using new cross-section generated by WIMS-ANL code. Photon sources of reactor tank water calculated from the experimental data. MCNP4C2 with available non-analog Monte Carlo model and ANSI/ANL-6.1.1-1977 flux-to-dose factors were used for dose estimation. The agreement between calculation results and those of measurements showed that the methods and models used to get photon sources and dose were acceptable. In case the reactor water totally leaks out from the reactor tank, the calculated dose is very high at the top of reactor tank while still low in control room. In the reactor hall, the operation staffs can access for emergency works but with time limits. (author)

  19. Parameter analysis calculation on characteristics of portable FAST reactor

    International Nuclear Information System (INIS)

    Otsubo, Akira; Kowata, Yasuki

    1998-06-01

    In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)

  20. An improved geometric algorithm for calculating the topology of lattice gauge fields

    International Nuclear Information System (INIS)

    Pugh, D.J.R.; Teper, M.; Oxford Univ.

    1989-01-01

    We implement the algorithm of Phillips and Stone on a hypercubic, periodic lattice and show that at currently accessible couplings the SU(2) topological charge so calculated is dominated by short-distance fluctuations. We propose and test an improvement to rid the measure of such lattice artifacts. We find that the improved algorithm produces a topological susceptibility that is consistent with that obtained by the alternative cooling method, thus resolving the controversial discrepancy between geometric and cooling methods. We briefly discuss the reasons for this and point out that our improvement is likely to be particularly effective when applied to the case of SU(3). (orig.)

  1. Extracting scattering phase shifts in higher partial waves from lattice QCD calculations

    Energy Technology Data Exchange (ETDEWEB)

    Luu, Thomas; Savage, Martin J.

    2011-06-01

    Lüscher’s method is routinely used to determine meson-meson, meson-baryon, and baryon-baryon s-wave scattering amplitudes below inelastic thresholds from lattice QCD calculations—presently at unphysical light-quark masses. In this work we review the formalism and develop the requisite expressions to extract phase shifts describing meson-meson scattering in partial waves with angular momentum l≤6 and l=9. The implications of the underlying cubic symmetry, and strategies for extracting the phase shifts from lattice QCD calculations, are presented, along with a discussion of the signal-to-noise problem that afflicts the higher partial waves.

  2. Dispersion parameters: impact on calculated reactor accident consequences

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.

    1979-01-01

    Much attention has been given in recent years to the modeling of the atmospheric dispersion of pollutants released from a point source. Numerous recommendations have been made concerning the choice of appropriate dispersion parameters. A series of calculations has been performed to determine the impact of these recommendations on the calculated consequences of large reactor accidents. Results are presented and compared in this paper.

  3. Modelling of Control Bars in Calculations of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Khlaifi, A.; Buiron, L.

    2004-01-01

    The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

  4. Burnable poisons in the light water reactor design, microburnup experiments and calculations. Part of a coordinated programme on burnup calculations and experiments for thermal reactors

    International Nuclear Information System (INIS)

    Penndorf, K.

    1976-04-01

    Investigations on Research Agreement N 1519/CF (1.8.1974 - 31.7.1975) entitled ''Burnable poisons in light water reactor design, microburnup experiments and calculations'' were carried out in the frame of the IAEA's coordinated research programme on ''Burn-up calculation and experiments for thermal reactors''. The theoretical and experimental work on application of solid burnable poison used for reduction of the amount of boric acid necessary to control of PWR or to lower the number of control rods needed in a BWR. Solid burnable poisons are needed in present PWR designs for the reduction of the boron acid concentration in order to prevent positive coefficients of reactivity. The special operational conditions of a ship reactor lead to the application of this kind of poison for compensation of almost all burnup reactivity. This strengthens the necessity of a very accurate and many dimensional calculations because an appropriate binding of reactivity has to be kept over the whole cycle time. Several burnup experiments had been run in the 15 MW material test reactor FRG-II. The following devices have been irradiated: poison pins within and without PWR fuel pin lattice segments and fuel pins containing pellets with a poison core. Measurements of reactivity, fluence, fission product concentration have been performed. Methods applied were γ-scanning and neutron pulse, radiography and transmission measurement techniques. Evaluation of the experiments was done by one and two dimensional Ssub(N) transport burnup calculations. In parallel a collision probability transport burnup code for current PWR design work is being developed, the main feature of which is economy in manpower and computer time

  5. A procedure validation for high conversion reactors fuel elements calculation

    International Nuclear Information System (INIS)

    Ishida, V.N.; Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.

    1990-01-01

    The present work includes procedure validation of cross sections generation starting from nuclear data and the calculation system actually used at the Bariloche Atomic Center Reactor and Neutrons Division for its application to fuel elements calculation of a high conversion reactor (HCR). To this purpose, the fuel element calculation belonging to a High Conversion Boiling water Reactor (HCBWR) was chosen as reference problem, employing the Monte Carlo method. Various cases were considered: with and without control bars, cold of hot, at different vacuum fractions. Multiplication factors, reaction rates, power maps and peak factors were compared. A sensitivity analysis of typical cells used, the approximations employed to solve the transport equation (Sn or Diffusion), the 1-D or 2-D representation and densification of the spatial network used, with the aim of evaluating their influence on the parameters studied and to come to an optimum combination to be used in future design calculations. (Author) [es

  6. Assessment of uncertainty in full core reactor physics calculations using statistical methods

    International Nuclear Information System (INIS)

    McEwan, C.

    2012-01-01

    The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

  7. Assessment of uncertainty in full core reactor physics calculations using statistical methods

    Energy Technology Data Exchange (ETDEWEB)

    McEwan, C., E-mail: mcewac2@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

  8. A lattice QCD calculation of the transverse decay constant of the b{sub 1}(1235) meson

    Energy Technology Data Exchange (ETDEWEB)

    Jansen, K. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany). John von Neumann-Inst. fuer Computing NIC; McNeile, C. [Wuppertal Univ. (Germany). Theoretische Teilchenphysik; Michael, C. [Liverpool Univ. (United Kingdom). Theoretical Physics Division, Dept. of Mathematical Sciences; Urbach, C. [Humboldt-Univ., Berlin (Germany). Theorie der Elementarteilchen

    2009-10-15

    We review various B meson decays that require knowledge of the transverse decay constant of the b{sub 1}(1235) meson. We report on an exploratory lattice QCD calculation of the transverse decay constant of the b{sub 1} meson. The lattice QCD calculations used unquenched gauge configurations, at two lattice spacings, generated with two flavours of sea quarks. The twisted mass formalism is used. (orig.)

  9. Effect of Pu-rich agglomerate in MOX fuel on a lattice calculation

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Yamamoto, Toru; Namekawa, Masakazu

    2007-01-01

    The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) fuel on a lattice calculation has been demonstrated. The Pu-rich agglomerate parameters are defined based on the measurement data of MIMAS-MOX and the focus is on the highly enriched MOX fuel in accordance with increased burnup resulting in a higher volume fraction of the Pu-rich agglomerates. The lattice calculations with a heterogeneous fuel model and a homogeneous fuel model are performed simulating the PWR 17x17 fuel assembly. The heterogeneous model individually treats the Pu-rich agglomerate and U-Pu matrix, whereas the homogeneous model homogenizes the compositions within the fuel pellet. A continuous-energy Monte Carlo burnup code, MVP-BURN, is used for burnup calculations up to 70 GWd/t. A statistical geometry model is applied in modeling a large number of Pu-rich agglomerates assuming that they are distributed randomly within the MOX fuel pellet. The calculated nuclear characteristics include k-inf, Pu isotopic compositions, power density and burnup of the Pu-rich agglomerates, as well as the pellet-averaged Pu compositions as a function of burnup. It is shown that the effect of Pu-rich agglomerates on the lattice calculation is negligibly small. (author)

  10. Thorium Fuel Performance in a Tight-Pitch Light Water Reactor Lattice

    International Nuclear Information System (INIS)

    Kim, Taek Kyum; Downar, Thomas J.

    2002-01-01

    Research on the utilization of thorium-based fuels in the intermediate neutron spectrum of a tight-pitch light water reactor (LWR) lattice is reported. The analysis was performed using the Studsvik/Scandpower lattice physics code HELIOS. The results show that thorium-based fuels in the intermediate spectrum of tight-pitch LWRs have considerable advantages in terms of conversion ratio, reactivity control, nonproliferation characteristics, and a reduced production of long-lived radiotoxic wastes. Because of the high conversion ratio of thorium-based fuels in intermediate spectrum reactors, the total fissile inventory required to achieve a given fuel burnup is only 11 to 17% higher than that of 238 U fertile fuels. However, unlike 238 U fertile fuels, the void reactivity coefficient with thorium-based fuels is negative in an intermediate spectrum reactor. This provides motivation for replacing 238 U with 232 Th in advanced high-conversion intermediate spectrum LWRs, such as the reduced-moderator reactor or the supercritical reactor

  11. Calculation of the geometric buckling for reactors of various shapes

    Energy Technology Data Exchange (ETDEWEB)

    Sjoestrand, N E

    1958-05-15

    A systematic investigation is made of the eleven coordinate systems in which the reactor equation {nabla}{sup 2}{phi} + B{sup 2}{phi} = 0 is separable. The fundamental solution and geometric buckling are given for those cases where the separated equations lead to known functions. It is especially shown that reactors of prolate and oblate spheroidal shape can be calculated in detail, and the results are given in extensive tables.

  12. Nuclear data sets for reactor design calculations - approved 1975

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  13. American National Standard: nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    1983-01-01

    This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include criteria for acceptance of evaluated nuclear data sets, criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  14. American National Standard nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    A standard is presented which identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  15. Application of MCNP in the criticality calculation for reactors

    International Nuclear Information System (INIS)

    Zhong Zhaopeng; Shi Gong; Hu Yongming

    2003-01-01

    The criticality calculation is carried out with 3-D Monte Carlo code (MCNP). The author focuses on the introduction of modelling of the core and reflector. The core description is simplified by using repetition structure function of MCNP. k eff in different control rods positions are calculated for the case of JRR3, and the results is consistent with that of the reference. This work shows that MCNP is applicable for reactor criticality calculation

  16. Comparison of calculational methods for EBT reactor nucleonics

    International Nuclear Information System (INIS)

    Henninger, R.J.; Seed, T.J.; Soran, P.D.; Dudziak, D.J.

    1980-01-01

    Nucleonic calculations for a preliminary conceptual design of the first wall/blanket/shield/coil assembly for an EBT reactor are described. Two-dimensional Monte Carlo, and one- and two-dimensional discrete-ordinates calculations are compared. Good agreement for the calculated values of tritium breeding and nuclear heating is seen. We find that the three methods are all useful and complementary as a design of this type evolves

  17. A gamma heating calculation methodology for research reactor application

    International Nuclear Information System (INIS)

    Lee, Y.K.; David, J.C.; Carcreff, H.

    2001-01-01

    Gamma heating is an important issue in research reactor operation and fuel safety. Heat deposition in irradiation targets and temperature distribution in irradiation facility should be determined so as to obtain the optimal irradiation conditions. This paper presents a recently developed gamma heating calculation methodology and its application on the research reactors. Based on the TRIPOLI-4 Monte Carlo code under the continuous-energy option, this new calculation methodology was validated against calorimetric measurements realized within a large ex-core irradiation facility of the 70 MWth OSIRIS materials testing reactor (MTR). The contributions from prompt fission neutrons, prompt fission γ-rays, capture γ-rays and inelastic γ-rays to heat deposition were evaluated by a coupled (n, γ) transport calculation. The fission product decay γ-rays were also considered but the activation γ-rays were neglected in this study. (author)

  18. Calculation of the Nucleon Axial Form Factor Using Staggered Lattice QCD

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Aaron S. [Fermilab; Hill, Richard J. [Perimeter Inst. Theor. Phys.; Kronfeld, Andreas S. [Fermilab; Li, Ruizi [Indiana U.; Simone, James N. [Fermilab

    2016-10-14

    The nucleon axial form factor is a dominant contribution to errors in neutrino oscillation studies. Lattice QCD calculations can help control theory errors by providing first-principles information on nucleon form factors. In these proceedings, we present preliminary results on a blinded calculation of $g_A$ and the axial form factor using HISQ staggered baryons with 2+1+1 flavors of sea quarks. Calculations are done using physical light quark masses and are absolutely normalized. We discuss fitting form factor data with the model-independent $z$ expansion parametrization.

  19. Nuclear data and multigroup methods in fast reactor calculations

    International Nuclear Information System (INIS)

    Gur, Y.

    1975-03-01

    The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)

  20. Static Q anti Q force from instanton gas and numerical lattice calculations

    International Nuclear Information System (INIS)

    Ilgenfrits, E.M.; Mueller-Preussker, M.

    1982-01-01

    Lattice Monte Carlo calculation predictions for the static strength between quarks are compared with the results obtained in the framework of instanton gas model and a typical instanton size is determined. Yang-Mills theory data for different ratios of Wilson loops in case of SU(3) for the string tension are presented. The instanton corrections to perturbation strength turn to be essential to reach an agreement with obtained by lattice calculations data inside the small-distance region up to approximately 0.3 fm. Arguments in favour of the statement that data difference in this region from the phenomenologically known value is connected with the notion of infinitely heavy quarks but not with neglect of virtual quark loops are presented

  1. Efficient implementation of the Monte Carlo method for lattice gauge theory calculations on the floating point systems FPS-164

    International Nuclear Information System (INIS)

    Moriarty, K.J.M.; Blackshaw, J.E.

    1983-01-01

    The computer program calculates the average action per plaquette for SU(6)/Z 6 lattice gauge theory. By considering quantum field theory on a space-time lattice, the ultraviolet divergences of the theory are regulated through the finite lattice spacing. The continuum theory results can be obtained by a renormalization group procedure. Making use of the FPS Mathematics Library (MATHLIB), we are able to generate an efficient code for the Monte Carlo algorithm for lattice gauge theory calculations which compares favourably with the performance of the CDC 7600. (orig.)

  2. Exact Calculation of the Thermodynamics of Biomacromolecules on Cubic Recursive Lattice.

    Science.gov (United States)

    Huang, Ran

    The thermodynamics of biomacromolecules featured as foldable polymer with inner-linkage of hydrogen bonds, e. g. protein, RNA and DNA, play an impressive role in either physical, biological, and polymer sciences. By treating the foldable chains to be the two-tolerate self-avoiding trails (2T polymer), abstract lattice modeling of these complex polymer systems to approach their thermodynamics and subsequent bio-functional properties have been developed for decades. Among these works, the calculations modeled on Bethe and Husimi lattice have shown the excellence of being exactly solvable. Our project extended this effort into the 3D situation, i.e. the cubic recursive lattice. The preliminary exploration basically confirmed others' previous findings on the planar structure, that we have three phases in the grand-canonical phase diagram, with a 1st order transition between non-polymerized and polymer phases, and a 2nd order transition between two distinguishable polymer phases. However the hydrogen bond energy J, stacking energy ɛ, and chain rigidity energy H play more vigorous effects on the thermal behaviors, and this is hypothesized to be due to the larger number of possible configurations provided by the complicated 3D model. By the so far progress, the calculation of biomacromolecules may be applied onto more complex recursive lattices, such as the inhomogeneous lattice to describe the cross-dimensional situations, and beside the thermal properties of the 2T polymers, we may infer some interesting insights of the mysterious folding problem itself. National Natural Science Foundation of China.

  3. Radiological shielding of low power compact reactor: calculation and design

    International Nuclear Information System (INIS)

    Marino, Raul

    2004-01-01

    The development of compact reactors becoming a technology that offers great projection and innumerable use possibilities, both in electricity generation and in propulsion.One of the requirements for the operation of this type of reactor is that it must include a radiological shield that will allow for different types of configurations and that, may be moved with the reactor if it needs to be transported.The nucleus of a reactor emits radiation, mainly neutrons and gamma rays in the heat of power, and gamma radiation during the radioactive decay of fission products.This radiation must be restrained in both conditions of operation to avoid it affecting workers or the public.The combination of different materials and properties in layers results in better performance in the form of a decrease in radiation, hence causing the dosage outside the reactor, whether in operation or shut down, to fall within the allowed limits.The calculations and design of radiological shields is therefore of paramount importance in reactor design.The choice of material and the design of the shield have a strong impact on the cost and the load capacity, the latter being one of the characteristics to optimize.The imposed condition of design is that the reactor can be transported together with the decay shield in a standard container of 40 foot [es

  4. The validation of neutron kinetic calculations of CEGB reactors

    International Nuclear Information System (INIS)

    Emmett, J.C.A.; Hutt, P.K.; Nunn, D.L.; Waterson, R.H.

    1982-01-01

    Reactor kinetic calculations are required by the CEGB to predict space and time varying neutron fluxes through the course of various hypothesized core transients. These transients arise through flow or reactivity perturbations occurring in a part of the core. A description is given of the results of dual programmes of work undertaken at BNL to validate such calculations. Firstly, analyses have been carried out to establish how data for these calculations should best be derived. Secondly, experimental measurements have been compared against the predictions of such calculations with data derived in the recommended way. (author)

  5. Neutron energy spectra calculations in the low power research reactor

    International Nuclear Information System (INIS)

    Omar, H.; Khattab, K.; Ghazi, N.

    2011-01-01

    The neutron energy spectra have been calculated in the fuel region, inner and outer irradiation sites of the zero power research reactor using the MCNP-4C code and the combination of the WIMS-D/4 transport code for generation of group constants and the three-dimensional CITATION diffusion code for core analysis calculations. The neutron energy spectrum has been divided into three regions and compared with the proposed empirical correlations. The calculated thermal and fast neutron fluxes in the low power research reactor MNSR inner and outer irradiation sites have been compared with the measured results. Better agreements have been noticed between the calculated and measured results using the MCNP code than those obtained by the CITATION code. (author)

  6. Systematic assembly homogenization and local flux reconstruction for nodal method calculations of fast reactor power distributions

    International Nuclear Information System (INIS)

    Dorning, J.J.

    1991-01-01

    A simultaneous pin lattice cell and fuel bundle homogenization theory has been developed for use with nodal diffusion calculations of practical reactors. The theoretical development of the homogenization theory, which is based on multiple-scales asymptotic expansion methods carried out through fourth order in a small parameter, starts from the transport equation and systematically yields: a cell-homogenized bundled diffusion equation with self-consistent expressions for the cell-homogenized cross sections and diffusion tensor elements; and a bundle-homogenized global reactor diffusion equation with self-consistent expressions for the bundle-homogenized cross sections and diffusion tensor elements. The continuity of the angular flux at cell and bundle interfaces also systematically yields jump conditions for the scaler flux or so-called flux discontinuity factors on the cell and bundle interfaces in terms of the two adjacent cell or bundle eigenfunctions. The expressions required for the reconstruction of the angular flux or the 'de-homogenization' theory were obtained as an integral part of the development; hence the leading order transport theory angular flux is easily reconstructed throughout the reactor including the regions in the interior of the fuel bundles or computational nodes and in the interiors of the pin lattice cells. The theoretical development shows that the exact transport theory angular flux is obtained to first order from the whole-reactor nodal diffusion calculations, done using the homogenized nuclear data and discontinuity factors, is a product of three computed quantities: a ''cell shape function''; a ''bundle shape function''; and a ''global shape function''. 10 refs

  7. Towards the reanalysis of void coefficients measurements at proteus for high conversion light water reactor lattices

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, M.; Koeberl, O.; Perret, G. [Paul Scherrer Institut PSI, 5232 Villigen (Switzerland)

    2012-07-01

    High Conversion Light Water Reactors (HCLWR) allows a better usage of fuel resources thanks to a higher breeding ratio than standard LWR. Their uses together with the current fleet of LWR constitute a fuel cycle thoroughly studied in Japan and the US today. However, one of the issues related to HCLWR is their void reactivity coefficient (VRC), which can be positive. Accurate predictions of void reactivity coefficient in HCLWR conditions and their comparisons with representative experiments are therefore required. In this paper an inter comparison of modern codes and cross-section libraries is performed for a former Benchmark on Void Reactivity Effect in PWRs conducted by the OECD/NEA. It shows an overview of the k-inf values and their associated VRC obtained for infinite lattice calculations with UO{sub 2} and highly enriched MOX fuel cells. The codes MCNPX2.5, TRIPOLI4.4 and CASMO-5 in conjunction with the libraries ENDF/B-VI.8, -VII.0, JEF-2.2 and JEFF-3.1 are used. A non-negligible spread of results for voided conditions is found for the high content MOX fuel. The spread of eigenvalues for the moderated and voided UO{sub 2} fuel are about 200 pcm and 700 pcm, respectively. The standard deviation for the VRCs for the UO{sub 2} fuel is about 0.7% while the one for the MOX fuel is about 13%. This work shows that an appropriate treatment of the unresolved resonance energy range is an important issue for the accurate determination of the void reactivity effect for HCLWR. A comparison to experimental results is needed to resolve the presented discrepancies. (authors)

  8. Burnup calculation in microcells of high conversion reactors

    International Nuclear Information System (INIS)

    Gomez, S.E.; Salvatore, M.; Patino, N.E.; Abbate, M.J.

    1991-01-01

    The development of high converter reactors (HCR) requires careful burnup calculations because their main goals are reach high discharge burnup levels (Up to 50 GWd/T) and a close to one conversion ratio. Then, it is necessary a revision of design elements used for this type of calculation. In this work, a burnup module (BUM) developed in order to use nuclear data directly from evaluated data files is presented; these was included in the AMPX system. (author)

  9. Calculation device for fuel power history in BWR type reactors

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

  10. Fast reactors. Thermal calculations of annulus application to Phenix

    International Nuclear Information System (INIS)

    Kung, J.P.; Gama, J.M.

    1975-01-01

    The gas convection phenomena involved in the annuli of the penetrations of the heat exchanger of the Phenix reactor are analyzed and the calculations performed using the BINIX program developed by GAAA to study the same phenomena are presented. The theory/experience comparison led to a better understanding of thermo-siphon phenomena [fr

  11. Calculation of the quadrupole magnet strengths in the PEP lattice for SCORE

    International Nuclear Information System (INIS)

    King, A.S.; Lee, M.J.

    1978-03-01

    The code, QUADS, which determines the step size in making configuration changes and calculates the field strengths of the 11 main ring quadrupole magnet families at each configuration has been completed. This code has been designed to have minimum computation time while keeping the necessary features for making future modifications of the beam lattice. It is being incorporated into SCORE, the program for the strength computation of the ring elements. The purpose of this note is to describe the method used in this calculation. 4 figs

  12. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    International Nuclear Information System (INIS)

    Faghihi, F.; Mirvakili, S.M.

    2009-01-01

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ ex ), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10 3 Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  13. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of)

    2009-06-15

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ({rho}{sub ex}), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10{sup 3}Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  14. Calculation of Added Mass for Submerged Reactor with Complex Shape

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Jong-Oh; Kim, Gyeongho; Choo, Yeon-Seok; Yoo, Yeon-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Kijang Research Reactor (KJRR) is currently under construction. Its reactor is located on the bottom of a reactor pool which is filled with water to a depth of 12m. Some components are installed on or inside the reactor and their structural integrity and safety performance need to be verified under seismic situations. For the verification, time history data or Floor Response Spectrum (FRS) on their support location, which is the reactor, should be obtained. A Finite Element (FE) model with fluid elements can give very accurate results for the matter; however, it costs too many resources and takes too much time for the transient analyses. In order to make the model more efficient and simple, added masses are often used to simulate the effect of water instead of the fluid elements. Many literatures introduce methods to calculate the added mass according to the exterior shape of structures. In this paper, how to calculate added masses for complex shaped structure was suggested. The proposed method was applied to RSA for KJRR and its accuracy was verified through comparison of the natural frequencies of RSA with fluid elements and the added masses. They showed the differences less than 1.5% between two models. Finally, it is concluded that the proposed method is quite useful to obtain added masses for complex shaped structure.

  15. The Calculation Of Total Radioactivity Of Kartini Reactor Fuel Element

    International Nuclear Information System (INIS)

    Budisantoso, Edi Trijono; Sardjono, Y.

    1996-01-01

    The total radioactivity of Kartini reactor fuel element has been calculated by using ORIGEN2. In this case, the total radioactivity is the sum of alpha, beta, and gamma radioactivity from activation products nuclides, actinide nuclides and fission products nuclides in the fuel element. The calculation was based on irradiation history of fuel in the reactor core. The fuel element no 3203 has location history at D, E, and F core zone. The result is expressed in graphics form of total radioactivity and photon radiations as function of irradiation time and decay time. It can be concluded that the Kartini reactor fuel element in zone D, E, and F has total radioactivity range from 10 Curie to 3000 Curie. This range is for radioactivity after decaying for 84 days and that after reactor shut down. This radioactivity is happened in the fuel element for every reactor operation and decayed until the fuel burn up reach 39.31 MWh. The total radioactivity emitted photon at the power of 0.02 Watt until 10 Watt

  16. On the calculation of lattice sums arising in Bose-Einstein statistics of quasiparticle excitations

    International Nuclear Information System (INIS)

    Millev, Y.; Faehnle, M.

    1994-05-01

    A new method for the calculations of the average occupation number of bosonic quasi-particle excitations valid for any type of lattice is proposed. The method is based on the recognition of the connection with lattice Green's functions and generalized Watson integrals, on one hand, and on a very simple differentiation technique which renders unnecessary and artificial to this problem more sophisticated Laplace transform summation procedures. The mean-field approximation to Green's function theories of ferromagnetism arises naturally as the zeroth term in the obtained summation formulae. The results have been specified completely for the three cubic lattices. They are new for the simple cubic and face-centred cases, whereas certain redundancy is removed form the known body-centred cubic results. Applications of the method to more complex sums as, for instance, the thermodynamic sum for the total energy of the quasiparticles, are straightforward. There has also been found a new three-position recursion relation for the calculation of frequently occurring triple geometric integrals in the face-centred cubic case. It originates form a corresponding relation for a relevant Heun function. (author). 29 refs, 1 tab

  17. Lattice calculation of electric dipole moments and form factors of the nucleon

    Science.gov (United States)

    Abramczyk, M.; Aoki, S.; Blum, T.; Izubuchi, T.; Ohki, H.; Syritsyn, S.

    2017-07-01

    We analyze commonly used expressions for computing the nucleon electric dipole form factors (EDFF) F3 and moments (EDM) on a lattice and find that they lead to spurious contributions from the Pauli form factor F2 due to inadequate definition of these form factors when parity mixing of lattice nucleon fields is involved. Using chirally symmetric domain wall fermions, we calculate the proton and the neutron EDFF induced by the C P -violating quark chromo-EDM interaction using the corrected expression. In addition, we calculate the electric dipole moment of the neutron using a background electric field that respects time translation invariance and boundary conditions, and we find that it decidedly agrees with the new formula but not the old formula for F3. Finally, we analyze some selected lattice results for the nucleon EDM and observe that after the correction is applied, they either agree with zero or are substantially reduced in magnitude, thus reconciling their difference from phenomenological estimates of the nucleon EDM.

  18. Calculation of prefabricated part of WWR-K reactor building

    International Nuclear Information System (INIS)

    Belyashova, N.N.; Aptikaev, F.F.; Kopnichev, Yu.F.

    1998-01-01

    According of factual characteristics a strength and deformation of over-land part of carrier constructions under construction movement is defined. Direct dynamical calculation of design elements under action of inertial loads from supports shifts shows, that seismic stability of enclosing construction is not ensured. Possibly practically total collapse of coating construction is possibly, under which following levels of damages of internal design constructions of reactor central room have been forecasted: 1. Fall of destroyed design construction on reactor vessel in time moment (1.56-1.59 s) after coming to building of earthquake seismic waves of 10 balls. 2. It is possibly cracks formation in radial direction in lower part of reactor cap, but destroying of cap does not incident; 3. It is possibly cracks formation within stretched concrete zone of reactor construction at the mark from - 0.859 up to 0.100. Destroy of concrete's compressive zone of reactor construction have not being expected. 4. Collapse of reactor first contour coating constructions have not being expected

  19. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  20. Physics calculations for the Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Kalimullah; Kier, P.H.; Hummel, H.H.

    1977-06-01

    Calculations of distributions of power and sodium void reactivity, unvoided and voided Doppler coefficients and steel and fuel worths have been performed using diffusion theory and first-order perturbation theory for the LWR discharge Pu-fueled CRBR at BOL, the FFTF-grade Pu-fueled CRBR at BOL and for the beginning and end of equilibrium cycle of the LWR-Pu-fueled CRBR. The results of the burnup and breeding ratio calculations performed for obtaining the reactor compositions during the equilibrium cycle are also reported. Effects of sodium and steel contents on the distributions of sodium void reactivity and steel worth have also been studied. Errors and uncertainties in the reactivity coefficients due to cross-sections and the two-dimensional geometric representations of the reactor used in the calculations have also been estimated. Comparisons of the results with those in the CRBR PSAR are also discussed

  1. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Klippel, H.T.; Hogenbirk, A.; Oppe, J.; Sciolla, C.M.; Stad, R.C.L. van der; Zhang, B.C.

    1997-06-01

    As part of the activities within the framework of the development of INCOGEN, a 'Dutch' conceptual design of a smaller HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRs, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (orig.)

  2. Exposure calculation code module for reactor core analysis: BURNER

    International Nuclear Information System (INIS)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules

  3. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Hass, J.B.M. De; Klippel, H.Th.; Hogenbirk, A.; Oppe, J.; Sciolla, C.; Stad, R.C.L. Van Der; Zhang, B.C.

    1997-01-01

    As part of the activities within the framework of the development of INCOGEN, a ''Dutch'' conceptual design of a small HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRS, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (author)

  4. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

    1999-01-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  5. A lattice calculation of the decay constants of heavy-light pseudoscalars

    International Nuclear Information System (INIS)

    Labrenz, J.N.

    1992-08-01

    A lattice calculation of the decay constants for D and B mesons is described. Results are obtained (in the quenched approximation) from wall-source lattices in Coulomb gauge at β = 6.3, through a procedure that interpolates smoothly between the static approximation of Eichten and the conventional (''heavy'' Wilson fermion) method. The previously observed discrepancy between these two approaches has been understood, and we discuss the resolution and its limitations. We find f D = 206(9) ± 37 MeV, f D s = 231(7) ± 39 MeV, f B = 179(10) ± 39 MeV, and f B s = 203(8) ± 42 MeV. The first error in each result is statistical, resulting from the jackknife procedure applied to the full analysis. The second is our estimate of systematic errors due to scale-breaking, axial current renormalization, and fitting or extrapolation uncertainties

  6. A framework for the calculation of the ΔNγ* transition form factors on the lattice

    International Nuclear Information System (INIS)

    Agadjanov, Andria; Bernard, Véronique; Meißner, Ulf-G.; Rusetsky, Akaki

    2014-01-01

    Using the non-relativistic effective field theory framework in a finite volume, we discuss the extraction of the ΔNγ * transition form factors from lattice data. A counterpart of the Lüscher approach for the matrix elements of unstable states is formulated. In particular, we thoroughly discuss various kinematic settings, which are used in the calculation of the above matrix element on the lattice. The emerging Lüscher–Lellouch factor and the analytic continuation of the matrix elements into the complex plane are also considered in detail. A full group-theoretical analysis of the problem is made, including the partial-wave mixing and projecting out the invariant form factors from data

  7. OPAL reactor calculations using the Monte Carlo code serpent

    Energy Technology Data Exchange (ETDEWEB)

    Ferraro, Diego; Villarino, Eduardo [Nuclear Engineering Dept., INVAP S.E., Rio Negro (Argentina)

    2012-03-15

    In the present work the Monte Carlo cell code developed by VTT Serpent v1.1.14 is used to model the MTR fuel assemblies (FA) and control rods (CR) from OPAL (Open Pool Australian Light-water) reactor in order to obtain few-group constants with burnup dependence to be used in the already developed reactor core models. These core calculations are performed using CITVAP 3-D diffusion code, which is well-known reactor code based on CITATION. Subsequently the results are compared with those obtained by the deterministic calculation line used by INVAP, which uses the Collision Probability Condor cell-code to obtain few-group constants. Finally the results are compared with the experimental data obtained from the reactor information for several operation cycles. As a result several evaluations are performed, including a code to code cell comparison at cell and core level and calculation-experiment comparison at core level in order to evaluate the Serpent code actual capabilities. (author)

  8. Contribution to the qualification of Gd calculation in PWR reactors

    International Nuclear Information System (INIS)

    Chaucheprat, Patrick.

    1982-06-01

    This thesis presents the state of knowledge on gadolinium and the advantages of its use as burnable poison. A study on the behaviour of gadolinium makes it possible to bring out the essential parameters to which it is sensitive. The most important part of this work is devoted to the measurements by oscillations carried out in Minerve in 1981. The conceiving and implementation of this campaign are reported. The experimental results and the amending factors linked to the interpretation are presented. To complete this study at zero time, it seemed useful to process configurations with fuel clusters of UO 2 - Gd 2 O 3 in order to see the effect of UO 2 - Gd 2 O 3 rods in interaction. To this end, efficiency determinations of UO 2 - Gd 2 O 3 rod clusters were carried out in the Melodie lattice. The second part of this work involves the change in the gadolinium. Two main points are tackled here. The first concerns the determinations by oscillations of ''reconstituted'' samples that are composed of two concentric rings with various 235 U enrichments and gadolinium levels so as to simulate irradiated UO 2 - Gd 2 O 3 fuel. The second point is devoted to the description of the GEDEON experiment. UO 2 - Gd 2 O 3 rods will be irradiated in a 13 x 13 lattice of which the spectrum is representative of that of a PWR. This experiment will take place in the centre of the Melusine reactor at Grenoble [fr

  9. Contribution to a neutronic calculation scheme for pressurized water reactors

    International Nuclear Information System (INIS)

    Martin Del Campo, C.

    1987-01-01

    This research thesis aims at developing and validating the set of data and codes which build up the neutron computation scheme of pressurized water reactors. More precisely, it focuses on the improvement of the precision of calculation of command clusters (absorbing components which can be inserted into the core to control the reactivity), and on the modelling of reflector representation (material placed around the core and reflecting back the escaping neutrons). For the first case, a precise calculation is performed, based on the transport theory. For the second case, diffusion constants obtained in the previous case and simplified equations are used to reduce the calculation cost

  10. Improvements in the model of neutron calculations for research reactors

    International Nuclear Information System (INIS)

    Calzetta, Osvaldo; Leszczynski, Francisco

    1987-01-01

    Within the research program in the field of neutron physics calculations being carried out in the Nuclear Engineering Division at the Centro Atomico Bariloche, the errors which due to some typical approximations appear in the final results are researched. For research MTR type reactors, two approximations, for high and low enrichment are investigated: the treatment of the geometry and the method of few-group cell cross-sections calculation, particularly in the resonance energy region. Commonly, the cell constants used for the entire reactor calculation are obtained making an homogenization of the full fuel elements, by one-dimensional calculations. An improvement is made that explicitly includes the fuel element frames in the core calculation geometry. Besides, a detailed treatment-in energy and space- is used to find the resonance few-group cross sections, and a comparison of the results with detailed and approximated calculations is made. The least number and the best mesh of energy groups needed for cell calculations is fixed too. (Author) [es

  11. Measurement and calculation of fast neutron flux in a zero-energy reactor

    International Nuclear Information System (INIS)

    Day, D.H.; Fox, W.N.; Hyder, H.R.

    1963-05-01

    An activation technique for measuring relative fast neutron fluxes is described which has some advantages over the normal method using U238 fission. The technique is based on the formation of Rh 103 after inelastic scattering of neutrons above 100 keV in energy. This isomer decays with a 57.4 minute half-life giving an easily measurable γ-activity. The energy dependence of the inelastic scattering cross-section of Rh 103 is similar to that of the fission cross-section of U 238 thus making the results of direct relevance to reactor calculations. Using the Rh 103 activation technique, measurements have been made of the fast neutron flux distribution in a typical pressure tube heavy water lattice and are compared in this report with theoretical calculations using the MONTE CARLO method. (author)

  12. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

    2015-01-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  13. Coupled map lattice (CML) approach to power reactor dynamics (I) - preservation of normality

    International Nuclear Information System (INIS)

    Konno, H.

    1996-01-01

    An application of coupled map lattice (CML) model for simulating power fluctuations in nuclear power reactors is presented. (1) Preservation of Gaussianity in the point model is studied in a chaotic force driven Langevin equation in conjunction with the Gaussian-white noise driven Langevin equation. (2) Preservation of Guassianity is also studied in the space-dependent model with the use of a CML model near the onset of the Hopf bifurcation point. It is shown that the spatial dimensionality decreases as the maximum eigenvalue of the system increases. The result is consistent with the observation of neutron fluctuation in a BWR. (author)

  14. Minimizing the power peaking factor of fuel lattices using factors of group for boiling water reactors

    International Nuclear Information System (INIS)

    Guzman, J. R.; Longoria, L. C.; De la Cruz, E.; Arredondo, C.

    2010-10-01

    A method to design the distribution and composition of nuclear fuel for the array of fuel rods in a lattice for BWRs is presented in this work. The aim of the method is to minimize the power peaking factor until an adequate value is reached. Also, this method uses a few calculations of lattice. The method is based on the classification of the fuel rods in two groups: the group of fuel rods with the higher power level (group pow ), and the other group of fuel rods (no-group pow ). The enrichment of 235 U of each fuel rod of the group pow is multiplied by a factor called group fissile factor (f group ), and the enrichment of 235 U of each fuel rod of the no-group pow is multiplied by a factor called no-group fissile factor (f no-group ). These factors are fitted so that the power peaking factor is minimized. The importance of the method with the use of these two factors is applied to the design of a fuel lattice for BWRs as the Laguna Verde nuclear power plant. The calculations of lattice are made by means of the Helios code. (Author)

  15. Comparison of calculational methods for liquid metal reactor shields

    International Nuclear Information System (INIS)

    Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.

    1985-09-01

    A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs

  16. Extended hadron and two-hadron operators of definite momentum for spectrum calculations in lattice QCD

    CERN Document Server

    Morningstar, C; Fahy, B; Foley, J; Jhang, Y C; Juge, K J; Lenkner, D; Wong, C C H

    2013-01-01

    Multi-hadron operators are crucial for reliably extracting the masses of excited states lying above multi-hadron thresholds in lattice QCD Monte Carlo calculations. The construction of multi-hadron operators with significant coupling to the lowest-lying states of interest involves combining single hadron operators of various momenta. The design and implementation of large sets of spatially-extended single-hadron operators of definite momentum and their combinations into two-hadron operators are described. The single hadron operators are all assemblages of gauge-covariantly-displaced, smeared quark fields. Group-theoretical projections onto the irreducible representations of the symmetry group of a cubic spatial lattice are used in all isospin channels. Tests of these operators on 24^3 x 128 and 32^3 x 256 anisotropic lattices using a stochastic method of treating the low-lying modes of quark propagation which exploits Laplacian Heaviside quark-field smearing are presented. The method provides reliable estimat...

  17. RA-0 reactor. New neutronic calculations; Reactor RA-0. Nuevos calculos neutronicos

    Energy Technology Data Exchange (ETDEWEB)

    Rumis, D; Leszczynski, F

    1991-12-31

    An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core`s interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author). [Espanol] En este trabajo se actualizan los calculos neutronicos realizados para el reactor RA-0, instalado en la Facultad de Ciencias Exactas, Fisicas y Naturales de la Universidad Nacional de Cordoba. Se describen los calculos realizados hasta la fecha y los resultados obtenidos. Las tecnicas incorporadas al calculo de un reactor como el RA-0 permiten predecir en detalle el comportamiento del flujo en el interior del nucleo y en el reflector, lo que sera una importante ayuda en el diseno de experimentos. En particular, el empleo del codigo WIMSD4 para calculos del reactor completo constituye una novedad en las posibles aplicaciones de ese codigo para resolver problemas que se presentan en la practica. (Autor).

  18. ChPT calculations for the analysis of lattice QCD data

    International Nuclear Information System (INIS)

    Greil, Ludwig

    2014-01-01

    We present calculations within the framework of three-flavor chiral perturbation theory (ChPT) for several observables (first moments of parton distributions, baryon octet masses and vector meson masses including phi-omega-mixing). We use lattice QCD data to determine the local couplings appearing in this chosen effective theory and we use these extrapolations to study the convergence of the chiral expansion around the symmetric point where all light quark masses have the same value. We also comment on the various benefits that stem from an expansion around the symmetric point.

  19. Identity of the conjugate gradient and Lanczos algorithms for matrix inversion in lattice fermion calculations

    International Nuclear Information System (INIS)

    Burkitt, A.N.; Irving, A.C.

    1988-01-01

    Two of the methods that are widely used in lattice gauge theory calculations requiring inversion of the fermion matrix are the Lanczos and the conjugate gradient algorithms. Those algorithms are already known to be closely related. In fact for matrix inversion, in exact arithmetic, they give identical results at each iteration and are just alternative formulations of a single algorithm. This equivalence survives rounding errors. We give the identities between the coefficients of the two formulations, enabling many of the best features of them to be combined. (orig.)

  20. Calculating the Jet Transport Coefficient q-hat in Lattice Gauge Theory

    International Nuclear Information System (INIS)

    Majumder, Abhijit

    2013-01-01

    The formalism of jet modification in the higher twist approach is modified to describe a hard parton propagating through a hot thermalized medium. The leading order contribution to the transverse momentum broadening of a high energy (near on-shell) quark in a thermal medium is calculated. This involves a factorization of the perturbative process of scattering of the quark from the non-perturbative transport coefficient. An operator product expansion of the non-perturbative operator product which represents q -hat is carried out and related via dispersion relations to the expectation of local operators. These local operators are then evaluated in quenched SU(2) lattice gauge theory

  1. An analytical method for neutron thermalization calculations in heterogenous reactors

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

    1965-07-01

    It is well known that the use of the diffusion approximation for stureactors may result in considerable errors. On the other hand, more exact numerical methods are rather laborious and require the use of large digital computers. In this paper, the use of the diffusion approximation in absorbing media has been avoided, but the treatment remained analytical, thus simplifying practical calculations.

  2. Calculation qualification of gadolinium burnable poisons in water reactors

    International Nuclear Information System (INIS)

    Chaucheprat, P.

    1988-01-01

    The work presented in this thesis constitutes the qualification on the one end of Appolo-Neptune scheme for the gadolinium burnable poison in a pressurized water reactor, and on the other end of basis nuclear data on natural gadolinium. This study has permitted to reduce by a factor 3 the actual incertitude on the gadolinium poison comparatively at precisions cited in international benchmarks calculations [fr

  3. An analytical method for neutron thermalization calculations in heterogenous reactors

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1965-01-01

    It is well known that the use of the diffusion approximation for studying neutron thermalization in heterogeneous reactors may result in considerable errors. On the other hand, more exact numerical methods are rather laborious and require the use of large digital computers. In this paper, the use of the diffusion approximation in absorbing media has been avoided, but the treatment remained analytical, thus simplifying practical calculations

  4. Application of the perturbation theory for sensitivity calculations in thermalhydraulics reactor calculations

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de

    1986-01-01

    The sensitivity of non linear responses associated with physical quantities governed by non linear differential systems can be studied using perturbation theory. The equivalence and formal differences between the differential and GPT formalisms are shown and both are used for sensitivity calculations of transient problems in a typical PWR coolant channel. The results obtained are encouraging with respect to the potential of the method for thermalhydraulics calculations normally performed for reactor design and safety analysis. (Author) [pt

  5. Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

    International Nuclear Information System (INIS)

    Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

    2004-01-01

    As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

  6. Direct calculation of the spin stiffness on square, triangular and cubic lattices using the coupled cluster method

    OpenAIRE

    Krüger, S. E.; Darradi, R.; Richter, J.; Farnell, D. J. J

    2006-01-01

    We present a method for the direct calculation of the spin stiffness by means of the coupled cluster method. For the spin-half Heisenberg antiferromagnet on the square, the triangular and the cubic lattices we calculate the stiffness in high orders of approximation. For the square and the cubic lattices our results are in very good agreement with the best results available in the literature. For the triangular lattice our result is more precise than any other result obtained so far by other a...

  7. Preliminary physics calculations for the Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Kalimullah.

    1975-01-01

    Calculations of sodium void, fuel, and clad worths, power distribution, and control rod worths have been carried out for an R-Z model of the CRBR, using diffusion theory and first-order perturbation theory for material worths. The power distribution and control rod worths have also been calculated in two-dimensional triangular mesh geometry. The present results are preliminary because of inaccuracy of the reactor model and the cross sections used, but the final results are not expected to be greatly different. (U.S.)

  8. COPDIRC - calculation of particle deposition in reactor coolants

    International Nuclear Information System (INIS)

    Reeks, M.W.

    1982-06-01

    A description is given of a computer code COPDIRC intended for the calculation of the deposition of particulate onto smooth perfectly sticky surfaces in a gas cooled reactor coolant. The deposition is assumed to be limited by transport in the boundary layer adjacent to the depositing surface. This implies that the deposition velocity normalised with respect to the local friction velocity, is an almost universal function of the normalised particle relaxation time. Deposition is assumed similar to deposition in an equivalent smooth perfectly absorbing pipe. The deposition is calculated using 2 models. (author)

  9. Alternative methodology for irradiation reactor experimental shielding calculation

    International Nuclear Information System (INIS)

    Vellozo, Sergio de Oliveira; Vital, Helio de Carvalho

    1996-01-01

    Due to a change in the project of the Experimental Irradiation Reactor, its shielding design had to be recalculated according to an alternative simplified analytical approach, since the standard transport calculations were temporarily unavailable. In the calculation of the new width for the shielding made up of steel and high-density concrete layers, the following radiation components were considered: fast neutrons and primary gammas (produced by fission and beta decay), from the core; and secondary gammas, produced by thermal neutron capture in the shielding. (author)

  10. A method to calculate spatial xenon oscillations in PWR reactors

    International Nuclear Information System (INIS)

    Ronig, H.

    1976-01-01

    The new digital computer programme SEXI for the calculation of spatial Xe oscillations is described. A series expansion of the flux density and the particle densities following the geometrical eigenfunctions of a homogeneous block reactor is chosen as an approach to the solution of the system of differential equations describing this feedback process between neutron flux density and Xe particle density. To calculate the neutron flux density, the time-dependent form of the diffusion equation is used instead of the more common stationary form. Integration is carried out using formal time differential quotients of the Fourier coefficients. (orig./RW) [de

  11. Safety analysis calculations for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S Y; MacDonald, R; MacFarlane, D [Argonne National Laboratory, Argonne, IL (United States)

    1983-08-01

    calculations performed with existing computer codes, most suited for each type of reactor, are presented.

  12. Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

    International Nuclear Information System (INIS)

    Viais J, J.

    1994-01-01

    Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

  13. Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, M.; Perret, G. [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland)

    2012-07-01

    The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

  14. Qualification of γ-heating calculation in nuclear reactors

    International Nuclear Information System (INIS)

    Ravaux, Simon

    2013-01-01

    During the last few years, the γ-heating issue has gained in stature, mainly for the safety of the 3. generation reactors in which a stainless steel reflector is inserted. The purpose of this work is the qualification of the needed tools for calculation of the γ-heating in the nuclear reactors. In a nuclear reactor, all the photons are directly or indirectly produced by the neutron-matter interactions. Thus, the first phase of this work is a critical analysis of the photon production data in the standard nuclear data library. New evaluations have been proposed to the next version of the JEFF library after that some omissions have been found. They have partly been accepted for JEFF-3.2. Two particle-transport codes are currently developed in the CEA: the deterministic code APOLLO2 and the Monte Carlo code TRIPOLI4. The second part of this work is the qualification of both these codes by interpreting an integral experiment called PERLE. The experimental set-up is made by a LWR pin assembly surrounded by a stainless steel reflector in which the γ-heating is measured by Thermo-luminescent Detector (TLD). A calculation scheme has been proposed for both APOLLO2 and TRIPOLI4 in order to calculate the TLD's responses. Comparisons between calculations and measurements have shown that TRIPOLI4 gives a satisfactory estimation of the γ-heating in the reflector. These discrepancies are within the experimental 1 σ uncertainty. Before the qualification, APOLLO2 has been previously validated against TRIPOLI4 reference calculation. This validation gives an estimation of the bias due to the deterministic approximations of the transport equation resolution. The qualification has shown that the discrepancies between APOLLO2 predictions and TLD's measurements are in the same range as experimental uncertainties. (author) [fr

  15. Advances in neutronics calculation of fast neutron reactors - Demonstration on Super-Phenix reactor

    International Nuclear Information System (INIS)

    Czernecki, Sebastien

    1998-01-01

    The fast reactor european neutronics calculations system, ERANOS, has integrated recent improvements both in nuclear data, with the use of the adjusted nuclear library ERALIB 1 from the JEF2.2 library, and calculation methods, with the use of the new european cell code, ECCO, and the deterministic code, TGV/VARIANT. This code performs full 3-D reactor calculation in the transport theory with variational method. The aim of this work is to create and validate a new calculational scheme for fast spectrum systems offering good compromise between accuracy and running time. The new scheme is based on these improvements plus a special procedure accounting for control rod heterogeneity, which uses a reactivity equivalence homogenization. The new scheme has been validated by means of experiment/calculation comparisons, using the extensive start-up program measurements performed in Super-Phenix reactor. The validation uses also recent measurements performed in the Phenix reactor. The results are very satisfactory and show a significant improvement for almost all core parameters, especially for critical mass, control rod worth and radial subassembly power distribution. A detailed analysis of the discrepancies between the old scheme and the new one for this parameter allows to understand the separate effects of methods and nuclear data on the radial power distribution shape. (author) [fr

  16. Calculation of the anti-trap factor in heavy water lattices

    International Nuclear Information System (INIS)

    Naudet, R.; Mougey, J.

    1965-01-01

    The calculation of the anti-trap factor of a lattice is complex when a large fraction of captures occurs in a range of energies where the spectrum in the fuel is considerably different from the simple dE/E law. This is particularly true for heavy water lattices in which the distances. between the bars are generally fairly large with respect to the slowing-down length. In order to take into account this effect it is necessary both to know the constitution of the effective resonance integral as a function of the energy, and to be able to calculate the distribution in the fuel. This report is devoted to these two problems. An improved method of treating the statistical domain makes it possible to plot the curves of the cross-sections per unit lethargy for various shapes of the fuel. Furthermore, the slowing-down of the neutrons is studied using a Monte-Carlo method which makes it possible in particular to take into account the perturbations caused by the non-moderating rods. A study is also made of the problem of shielding effects due to the captures themselves. (authors) [fr

  17. Safety analysis calculations for research and test reactors

    International Nuclear Information System (INIS)

    Chen, S.Y.; MacDonald, R.; MacFarlane, D.

    1983-01-01

    Safety issues for the two general types of reactors, i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl/sub x/) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with HEU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods (approx. 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. Results of transient calculations performed with existing computer codes, most suited for each type of reactor, are presented

  18. Parallelizing the QUDA Library for Multi-GPU Calculations in Lattice Quantum Chromodynamics

    International Nuclear Information System (INIS)

    Babich, Ronald; Clark, Michael; Joo, Balint

    2010-01-01

    Graphics Processing Units (GPUs) are having a transformational effect on numerical lattice quantum chromodynamics (LQCD) calculations of importance in nuclear and particle physics. The QUDA library provides a package of mixed precision sparse matrix linear solvers for LQCD applications, supporting single GPUs based on NVIDIA's Compute Unified Device Architecture (CUDA). This library, interfaced to the QDP++/Chroma framework for LQCD calculations, is currently in production use on the '9g' cluster at the Jefferson Laboratory, enabling unprecedented price/performance for a range of problems in LQCD. Nevertheless, memory constraints on current GPU devices limit the problem sizes that can be tackled. In this contribution we describe the parallelization of the QUDA library onto multiple GPUs using MPI, including strategies for the overlapping of communication and computation. We report on both weak and strong scaling for up to 32 GPUs interconnected by InfiniBand, on which we sustain in excess of 4 Tflops.

  19. Parallelizing the QUDA Library for Multi-GPU Calculations in Lattice Quantum Chromodynamics

    Energy Technology Data Exchange (ETDEWEB)

    Ronald Babich, Michael Clark, Balint Joo

    2010-11-01

    Graphics Processing Units (GPUs) are having a transformational effect on numerical lattice quantum chromodynamics (LQCD) calculations of importance in nuclear and particle physics. The QUDA library provides a package of mixed precision sparse matrix linear solvers for LQCD applications, supporting single GPUs based on NVIDIA's Compute Unified Device Architecture (CUDA). This library, interfaced to the QDP++/Chroma framework for LQCD calculations, is currently in production use on the "9g" cluster at the Jefferson Laboratory, enabling unprecedented price/performance for a range of problems in LQCD. Nevertheless, memory constraints on current GPU devices limit the problem sizes that can be tackled. In this contribution we describe the parallelization of the QUDA library onto multiple GPUs using MPI, including strategies for the overlapping of communication and computation. We report on both weak and strong scaling for up to 32 GPUs interconnected by InfiniBand, on which we sustain in excess of 4 Tflops.

  20. A lattice calculation of the nucleon's spin-dependent structure function g2 revisited

    International Nuclear Information System (INIS)

    Goeckeler, M.; Rakow, P.E.L.; Schaefer, A.; Schierholz, G.

    2000-11-01

    Our previous calculation of the spin-dependent structure function g 2 is revisited. The interest in this structure function is to a great extent motivated by the fact that it receives contributions from twist-two as well as from twist-three operators already in leading order of 1/Q 2 thus offering the unique possibility of directly assessing higher-twist effects. In our former calculation the lattice operators were renormalized perturbatively and mixing with lower-dimensional operators was ignored. However, the twist-three operator which gives rise to the matrix element d 2 mixes non-perturbatively with an operator of lower dimension. Taking this effect into account leads to a considerably smaller value of d 2 , which is consistent with the experimental data. (orig.)

  1. Handbook for the calculation of reactor protections; Formulaire sur le calcul de la protection des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-07-01

    This note constitutes the first edition of a Handbook for the calculation of reactor protections. This handbook makes it possible to calculate simply the different neutron and gamma fluxes and consequently, to fix the minimum quantities of materials necessary under general safety conditions both for the personnel and for the installations. It contains a certain amount of nuclear data, calculation methods, and constants corresponding to the present state of our knowledge. (authors) [French] Cette note constitue la premiere edition du 'Formulaire sur le calcul de la protection des reacteurs'. Ce formulaire permet de calculer de facon simple les difterents flux de neutrons et de gamma et, par suite, de fixer les quantites minima de materiaux a utiliser pour que les conditions generales de securite soient respectees, tant pour le personnel que pour les installations. Il contient un certain nombre de donnees nucleaires, de methodes de calcul et de constantes correspondant a l'etat actuel de nos connaissances. (auteurs)

  2. Stresses in reactor pressure vessel nozzles -- Calculations and experiments

    International Nuclear Information System (INIS)

    Brumovsky, M.; Polachova, H.

    1995-01-01

    Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber's, Hardrath-Ohman's as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared

  3. Comparison of standard fast reactor calculations (Baker model)

    Energy Technology Data Exchange (ETDEWEB)

    Voropaev, A I; Van' kov, A A; Tsybulya, A M

    1978-12-01

    Compared are standard fast reactor calculations performed at different laboratories using several nuclear data files: BNAB-70 and OSKAR-75 (the USSR), CARNAVAL-4 (France), FD-5 (Great Britain), KFK-INR (West Germany), ENDF/B4 (the USA). Three fuel compositions were chosen: (1) /sup 239/Pu and /sup 238/U; (2) /sup 239/Pu, /sup 238/U and fission products; (3) /sup 239/Pu, /sup 240/Pu, /sup 238/U and fission products. Medium temperature was 300K. The calculations have been conducted in the diffusion approximation. Data on critical masses and breeding ratios are tabulated. Discrepancies in the calculations of all the characteristics are small since all the countries possess practically the same nuclear data files.

  4. Monte Carlo reactor calculation with substantially reduced number of cycles

    International Nuclear Information System (INIS)

    Lee, M. J.; Joo, H. G.; Lee, D.; Smith, K.

    2012-01-01

    A new Monte Carlo (MC) eigenvalue calculation scheme that substantially reduces the number of cycles is introduced with the aid of coarse mesh finite difference (CMFD) formulation. First, it is confirmed in terms of pin power errors that using extremely many particles resulting in short active cycles is beneficial even in the conventional MC scheme although wasted operations in inactive cycles cannot be reduced with more particles. A CMFD-assisted MC scheme is introduced as an effort to reduce the number of inactive cycles and the fast convergence behavior and reduced inter-cycle effect of the CMFD assisted MC calculation is investigated in detail. As a practical means of providing a good initial fission source distribution, an assembly based few-group condensation and homogenization scheme is introduced and it is shown that efficient MC eigenvalue calculations with fewer than 20 total cycles (including inactive cycles) are possible for large power reactor problems. (authors)

  5. Calculation of β-effective of a molten salt reactor

    International Nuclear Information System (INIS)

    Hirakawa, N.; Sakaba, H.

    1987-01-01

    A method to calculate the β eff of a molten salt reactor was developed taking the effect of the flow of the molten salt into account. The method was applied to the 1000MW MSR design made by ORNL. The change in β eff due to the change in the residence time outside of the core of the fuel salt and to the change in the flow velocity when the total amount of the fuel salt is kept constant were investigated. It was found that β eff was reduced to 47.9% of the value when the fuel salt is at rest for the present design. (author)

  6. Neutron flux shape effects in large fast reactor safety calculations

    International Nuclear Information System (INIS)

    Galati, A.; Loizzo, P.; Musco, A.

    1978-01-01

    Three classes of accidents in a large fast reactor were studied by the two-dimensional core dynamics code NADYP-2. A Modified version of the code, including a point kinetics module, allowed comparison between 2D and 0D power, reactivity and temperature histories. A strong shape effect was evidenced by these calculations in the boiling phase of LOF accidents as well as in the accident generated by control rod removal. Some future possibilities of by passing the consequences of this effect are indicated

  7. Concentration creation of system and applied software of reactor calculations

    International Nuclear Information System (INIS)

    Zizin, M.N.

    1995-01-01

    Basic concept provisions, including modularity, openness and machine-independent programs; accumulation of procedure knowledge in form of computer module library; creation of medium facilitating module development and accompanying; possibility of head programs automated generation and user-friendly interface are presented. Intellectual program shell ShIPR for mathematical reactor modeling, the final goal whereof consists in automated generation of machine-independent programs in the Fortran 77 language on the basis of calculation moduli and accumulated knowledge base, is developed on the basis of the above concept

  8. ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

    1976-07-01

    The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

  9. Calculations of thermodynamic properties of PuO{sub 2} by the first-principles and lattice vibration

    Energy Technology Data Exchange (ETDEWEB)

    Minamoto, Satoshi [Energy and Industrial Systems Department, ITOCHU Techno-Solutions Corporation, Kasumigaseki 3-chome, Chiyoda-ku, Tokyo 100-6080 (Japan)], E-mail: satoshi.minamoto@ctc-g.co.jp; Kato, Masato [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1194 (Japan); Konashi, Kenji [Institute for Materials Research, Tohoku University, 2145-2 Narita-chou, Oarai-chou, Ibaraki 311-1313 (Japan); Kawazoe, Yoshiyuki [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai 980-8577 (Japan)

    2009-03-15

    Plutonium dioxide (PuO{sub 2}) is a key compound of mixed oxide fuel (MOX fuel). To predict the thermal properties of PuO{sub 2} at high temperature, it is important to understand the properties of MOX fuel. In this study, thermodynamic properties of PuO{sub 2} were evaluated by coupling of first-principles and lattice dynamics calculation. Cohesive energy was estimated from first-principles calculations, and the contribution of lattice vibration to total energy was evaluated by phonon calculations. Thermodynamic properties such as volume thermal expansion, bulk modulus and specific heat of PuO{sub 2} were investigated up to 1500 K.

  10. Calculations of thermodynamic properties of PuO2 by the first-principles and lattice vibration

    International Nuclear Information System (INIS)

    Minamoto, Satoshi; Kato, Masato; Konashi, Kenji; Kawazoe, Yoshiyuki

    2009-01-01

    Plutonium dioxide (PuO 2 ) is a key compound of mixed oxide fuel (MOX fuel). To predict the thermal properties of PuO 2 at high temperature, it is important to understand the properties of MOX fuel. In this study, thermodynamic properties of PuO 2 were evaluated by coupling of first-principles and lattice dynamics calculation. Cohesive energy was estimated from first-principles calculations, and the contribution of lattice vibration to total energy was evaluated by phonon calculations. Thermodynamic properties such as volume thermal expansion, bulk modulus and specific heat of PuO 2 were investigated up to 1500 K

  11. 3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

    Energy Technology Data Exchange (ETDEWEB)

    Palau, J M [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)

    2005-07-01

    This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

  12. 3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

    International Nuclear Information System (INIS)

    Palau, J.M.

    2005-01-01

    This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U 235 , U 238 , Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

  13. Debye–Einstein approximation approach to calculate the lattice specific heat and related parameters for a Si nanowire

    Directory of Open Access Journals (Sweden)

    A. KH. Alassafee

    2017-11-01

    Full Text Available The modified Debye–Einstein approximation model is used to calculate nanoscale size-dependent values of Gruneisen parameters and lattice specific heat capacity for Si nanowires. All parameters forming the model, including Debye temperatures, bulk moduli, the lattice thermal expansion and the lattice volume, are calculated according to their nanoscale size dependence. Values for lattice volume Gruneisen parameters increase with the decrease of the nanowires’ diameter, while all other parameters decrease. The nanosize dependence of lattice thermal parameters agree with other reported theoretical results. Keywords: Lattice specific heat capacity, Gruneisen parameter, Debye–Einstein model, Si nanowires

  14. Computing Moment-Based Probability Tables for Self-Shielding Calculations in Lattice Codes

    International Nuclear Information System (INIS)

    Hebert, Alain; Coste, Mireille

    2002-01-01

    As part of the self-shielding model used in the APOLLO2 lattice code, probability tables are required to compute self-shielded cross sections for coarse energy groups (typically with 99 or 172 groups). This paper describes the replacement of the multiband tables (typically with 51 subgroups) with moment-based tables in release 2.5 of APOLLO2. An improved Ribon method is proposed to compute moment-based probability tables, allowing important savings in CPU resources while maintaining the accuracy of the self-shielding algorithm. Finally, a validation is presented where the absorption rates obtained with each of these techniques are compared with exact values obtained using a fine-group elastic slowing-down calculation in the resolved energy domain. Other results, relative to the Rowland's benchmark and to three assembly production cases, are also presented

  15. A direct hybrid SN method for slab-geometry lattice calculations

    International Nuclear Information System (INIS)

    Silva, Davi J.M.; Barros, Ricardo C.; Zani, Jose H.

    2011-01-01

    In this work we describe a hybrid direct method for calculating the thermal disadvantage factor and the neutron flux distribution in fuel-moderator lattices. For the mathematical model, we use the one-speed slab-geometry discrete ordinates (S N ) transport equation with linearly anisotropic scattering. The basic idea is to use higher order angular quadrature set in the highly absorbing fuel region (S NF ) and lower order angular quadrature set in the diffusive moderator region (S NM ) , i.e., N F > N M . We apply special continuity conditions based on the equivalence of the S N and P N-1 equations, which characterize the hybrid model. Numerical results to a typical model problem are given to illustrate the accuracy and the efficiency of the offered hybrid method. (author)

  16. A direct hybrid S{sub N} method for slab-geometry lattice calculations

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Davi J.M.; Barros, Ricardo C., E-mail: rcbarros@pq.cnpq.b [Universidade do Estado do Rio de Janeiro (IPRJ/UERJ), Nova Friburgo, RJ (Brazil). Programa de Pos-graduacao em Modelagem Computacional; Zani, Jose H. [Fundacao Educacional Serra dos Orgaos, Teresopolis, RJ (Brazil). Ciencia da Computacao

    2011-07-01

    In this work we describe a hybrid direct method for calculating the thermal disadvantage factor and the neutron flux distribution in fuel-moderator lattices. For the mathematical model, we use the one-speed slab-geometry discrete ordinates (S{sub N}) transport equation with linearly anisotropic scattering. The basic idea is to use higher order angular quadrature set in the highly absorbing fuel region (S{sub NF}) and lower order angular quadrature set in the diffusive moderator region (S{sub NM}) , i.e., N{sub F} > N{sub M}. We apply special continuity conditions based on the equivalence of the S{sub N} and P{sub N-1} equations, which characterize the hybrid model. Numerical results to a typical model problem are given to illustrate the accuracy and the efficiency of the offered hybrid method. (author)

  17. Response matrix method for neutron transport in reactor lattices using group symmetry properties

    International Nuclear Information System (INIS)

    Mund, E.H.

    1991-01-01

    This paper describes a response matrix method for the approximate solution of one-velocity, multi-dimensional transport problems in reactor lattices, with isotropic neutron scattering. The transport equation is solved on a homogeneous cell by using a Petrov-Galerkin technique based on a set of trial and test functions (including polynomials and exponential functions) closely related to transport problems in infinite media. The number of non-zero elements of the response matrices reduces to a minimum when the symmetry properties of the cell are included ab initio in the span of the basis functions. To include these properties, use is made of projection operations which are performed very efficiently on symbolic manipulation programs. Numerical results of model problems in square geometry show a good agreement with reference solutions

  18. SPECIFICATIONS FOR HIGH TEMPERATURE LATTICE TEST REACTOR BUILDING 318 PROJECT CAH-100

    Energy Technology Data Exchange (ETDEWEB)

    Vitro Engineering Company

    1964-07-15

    This is the specifications for the High Temperature Lattice Test Reactor Building 318 and it is divided into the following 21 divisions or chapters: (1) Excavating, Backfill & Grading; (2) Reinforced Concrete; (3) Masonry; (4) Structural Steel & Miscellaneous Metal Items, Contents - Division V; (5) Plumbing, Process & Service Piping; (6) Welding; (7) Insulated Metal Siding; (8) Roof Decks & Roofing; (9) Plaster Partitions & Ceiling; (10) Standard Doors, Windows & Hardware; (11) Shielding Doors; (12) Sprinkler System & Fire Extinguishers, Contents - Division XIII; (13) Heating, Ventilating & Air Conditioning; (14) Painting, Protective Coating & Floor Covering, Contents - Division XV; (15) Electrical; (16) Communications & Alarm Systems; (17) Special Equipment & Furnishings; (18) Overhead Bridge Crane; (19) Prefabricated Steel Building; (20) Paved Drive; and (21) Landscaping & Irrigation Sprinklers.

  19. Subcriticality calculation in nuclear reactors with external neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mails: asilva@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br

    2007-07-01

    The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

  20. Subcriticality calculation in nuclear reactors with external neutron sources

    International Nuclear Information System (INIS)

    Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    2007-01-01

    The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

  1. Analysis of core calculation schemes for advanced water reactors

    International Nuclear Information System (INIS)

    Nicolas, Anne

    1989-01-01

    This research thesis addresses the analysis of the core control of sub-moderated water reactors with plutonium fuel and varying spectrum. Firstly, a calculation scheme is defined, based on transport theory for the three existing assembly configurations. It is based on the efficiency analysis of the control cluster and of the flow sheet shape in the assembly. Secondly, studies of the assembly with control cluster and within a theory of diffusion with homogenization or detailed assembly representation are performed by taking the environment into account in order to assess errors. Thirdly, due to the presence of a very efficient absorbent in control clusters, a deeper physical analysis requires the study of the flow gradient existing at the interface between assemblies. A parameter is defined to assess this gradient, and theoretically calculated by using finite elements. Developed software is validated [fr

  2. Methods for reactor physics calculations for control rods in fast reactors

    International Nuclear Information System (INIS)

    Grimstone, M.J.; Rowlands, J.L.

    1988-12-01

    The IAEA Specialists' Meeting on ''Methods for Reactor Physics Calculations for Control Rods in Fast Reactors'' was held in Winfrith, United Kingdom, on 6-8 December, 1988. The meeting was attended by 23 participants from nine countries. The purpose of the meeting was to review the current calculational methods and their accuracy as assessed by theoretical studies and comparisons with measurements, and then to identify the requirements for improved methods or additional studies and comparisons. The control rod properties or effects to be considered were their reactivity worths, their effect on the power distribution through the core, and the reaction rates and energy deposition both within and adjacent to the rods. The meeting was divided into five sessions, in the first of which each national delegation presented a brief overview of their programme of work on calculational methods for fast reactor control rods. In the next three sessions a total of seventeen papers were presented describing calculational methods and assessments of their accuracy. The final session was a discussion to draw conclusions regarding the current status of methods and the further developments and validation work required. A separate abstract was prepared for each of the 23 papers presented at the meeting. Refs, figs and tabs

  3. Novel and Efficient Methods for Calculating Pressure in Polymer Lattice Models

    Science.gov (United States)

    Zhang, Pengfei; Wang, Qiang

    2014-03-01

    Pressure calculation in polymer lattice models is an important but nontrivial subject. The three existing methods - thermodynamic integration, repulsive wall, and sedimentation equilibrium methods - all have their limitations and cannot be used to accurately calculate the pressure at all polymer volume fractions φ. Here we propose two novel methods. In the first method, we combine Monte Carlo simulation in an expanded grand-canonical ensemble with the Wang-Landau - Optimized Ensemble (WL-OE) simulation to calculate the pressure as a function of polymer volume fraction, which is very efficient at low to intermediate φ and exhibits negligible finite-size effects. In the second method, we introduce a repulsive plane with bridging bonds, which is similar to the repulsive wall method but eliminates its confinement effects, and estimate the two-dimensional density of states (in terms of the number of bridging bonds and the contact number) using the 1/ t version of Wang-Landau algorithm. This works well at all φ, especially at high φ where all the methods involving chain insertion trial moves fail.

  4. The spectral code Apollo2: from lattice to 2D core calculations

    International Nuclear Information System (INIS)

    Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I.; Santamarina, A.

    2005-01-01

    Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations

  5. The spectral code Apollo2: from lattice to 2D core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I. [CEA Saclay (DEN/DANS/SERMA), 91 - Gif-sur-Yvette (France); Santamarina, A. [CEA Cadarache (CEA/DEN/DER/SPRC), 13 - Saint Paul lez Durance (France)

    2005-07-01

    Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations.

  6. Calculation of forces on reactor containment fan cooler piping

    International Nuclear Information System (INIS)

    Miller, J.S.; Ramsden, K.

    2004-01-01

    The purpose of this paper is to present the results of the Reactor Containment Fan Cooler (RCFC) system piping load calculations. These calculations are based on piping loads calculated using the EPRI methodology and RELAP5 to simulate the hydraulic behavior of the system. The RELAP5 generated loads were compared to loads calculated using the EPRI GL-96-06 methodology. This evaluation was based on a pressurized water reactor's RCFC coils thermal hydraulic behavior during a Loss of Offsite Power (LOOP) and a loss of coolant accident (LOCA). The RCFC consist of two banks of service water and chill water coils. There are 5 SX and 5 chill water coils per bank. Therefore, there are 4 RCFC units in the containment with 2 banks of coils per RCFC. Two Service water pumps provide coolant for the 4 RCFC units (8 banks total, 2 banks per RCFC unit and 2 RCFC units per pump). Following a LOOP/LOCA condition, the RCFC fans would coast down and upon being re-energized, would shift to low-speed operation. The fan coast down is anticipated to occur very rapidly due to the closure of the exhaust damper as a result of LOCA pressurization effects. The service water flow would also coast down and be restarted in approximately 43 seconds after the initiation of the event. The service water would drain from the RCFC coils during the pump shutdown and once the pumps restart, water is quickly forced into the RCFC coils causing hydraulic loading on the piping. Because of this scenario and the potential for over stressing the piping, an evaluation was performed by the utility using RELAP5 to assess the piping loads. Subsequent to the hydraulic loads being analyzed using RELAP5, EPRI through GL-96-06 provided another methodology to assess loads on the RCFC piping system. This paper presents the results of using the EPRI methodology and RELAP5 to perform thermal hydraulic load calculations. It is shown that both EPRI methodology and RELAP5 calculations can be used to generate hydraulic loads

  7. Hybrid Reduced Order Modeling Algorithms for Reactor Physics Calculations

    Science.gov (United States)

    Bang, Youngsuk

    Reduced order modeling (ROM) has been recognized as an indispensable approach when the engineering analysis requires many executions of high fidelity simulation codes. Examples of such engineering analyses in nuclear reactor core calculations, representing the focus of this dissertation, include the functionalization of the homogenized few-group cross-sections in terms of the various core conditions, e.g. burn-up, fuel enrichment, temperature, etc. This is done via assembly calculations which are executed many times to generate the required functionalization for use in the downstream core calculations. Other examples are sensitivity analysis used to determine important core attribute variations due to input parameter variations, and uncertainty quantification employed to estimate core attribute uncertainties originating from input parameter uncertainties. ROM constructs a surrogate model with quantifiable accuracy which can replace the original code for subsequent engineering analysis calculations. This is achieved by reducing the effective dimensionality of the input parameter, the state variable, or the output response spaces, by projection onto the so-called active subspaces. Confining the variations to the active subspace allows one to construct an ROM model of reduced complexity which can be solved more efficiently. This dissertation introduces a new algorithm to render reduction with the reduction errors bounded based on a user-defined error tolerance which represents the main challenge of existing ROM techniques. Bounding the error is the key to ensuring that the constructed ROM models are robust for all possible applications. Providing such error bounds represents one of the algorithmic contributions of this dissertation to the ROM state-of-the-art. Recognizing that ROM techniques have been developed to render reduction at different levels, e.g. the input parameter space, the state space, and the response space, this dissertation offers a set of novel

  8. Calculation device for amount of heavy element nuclide in reactor fuels and calculation method therefor

    International Nuclear Information System (INIS)

    Naka, Takafumi; Yamamoto, Munenari.

    1995-01-01

    When there are two or more origins of deuterium nuclides in reactor fuels, there are disposed a memory device for an amount of deuterium nuclides for every origin in a noted fuel segment at a certain time point, a device for calculating the amount of nuclides for every origin and current neutron fluxes in the noted fuel segment, and a device for separating and then displaying the amount of deuterium nuclides for every origin. Equations for combustion are dissolved for every origin of the deuterium nuclides based on the amount of the deuterium nuclides for every origin and neutron fluxes, to calculate the current amount of deuterium nuclides for every origin. The amount of deuterium nuclides originated from uranium is calculated ignoring α-decay of curium, while the amount of deuterium nuclides originated from plutonium is calculated ignoring the generation of plutonium formed from neptunium. Deuterium nuclides can be measured and controlled accurately for every origin of the reactor fuels. Even when nuclear fuel materials have two or more nationalities, the measurement and control thereof can be conducted for every country. (N.H.)

  9. Comparison study on cell calculation method of fast reactor

    International Nuclear Information System (INIS)

    Chiba, Gou

    2002-10-01

    Effective cross sections obtained by cell calculations are used in core calculations in current deterministic methods. Therefore, it is important to calculate the effective cross sections accurately and several methods have been proposed. In this study, some of the methods are compared to each other using a continuous energy Monte Carlo method as a reference. The result shows that the table look-up method used in Japan Nuclear Cycle Development Institute (JNC) sometimes has a difference over 10% in effective microscopic cross sections and be inferior to the sub-group method. The problem was overcome by introducing a new nuclear constant system developed in JNC, in which the ultra free energy group library is used. The system can also deal with resonance interaction effects between nuclides which are not able to be considered by other methods. In addition, a new method was proposed to calculate effective cross section accurately for power reactor fuel subassembly where the new nuclear constant system cannot be applied. This method uses the sub-group method and the ultra fine energy group collision probability method. The microscopic effective cross sections obtained by this method agree with the reference values within 5% difference. (author)

  10. Investigating heavy water zero power reactors with a new core configuration based on experiment and calculation results

    Energy Technology Data Exchange (ETDEWEB)

    Nasrazadani, Zahra; Salimi, Raana; Askari, Afrooz; Khorsandi, Jamshid; Mirvakili, Mohammad; Mashayekh, Mohammad [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Esfahan (Iran, Islamic Republic of)

    2017-02-15

    The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor (Keff) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of D2O, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

  11. On the influence of spatial discretization in LWR cell- and lattice calculations with HELIOS 1.9

    International Nuclear Information System (INIS)

    Merk, B.; Koch, R.

    2008-01-01

    Cell- and lattice calculations are the fundamental for all deterministic static and transient 3D full core calculations. The spatial discretization used for the cell- and lattice calculations influences the results for these transport solutions significantly. The arising differences in the neutron flux distribution due to different spatial discretization are demonstrated. These differences in the flux distribution cause significant changes in the k inf value. An evaluation of the k inf value for the case of infinitely fine discretization is made. The influence of the discretization on the calculation of homogenized few group cross-sections which are forwarded to the 3D full core calculations is investigated. Strategies for improving the discretization are developed and their influence on the calculation time is evaluated

  12. Qualification of the WIMS lattice code, for the design, operation and accident analysis of nuclear reactors; Calificacion del programa WIMS de calculo neutronico para diseno, seguimiento de operacion y analisis de accidentes de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Lerner, A M [Ente Nacional Regulador Nuclear, Buenos Aires (Argentina)

    1997-12-31

    A basic problem in nuclear reactor physics in that of the description of the neutron population behaviour in the multiplicative medium of a nuclear fuel. Due to the magnitude of the physical problem involved and the present degree of technological evolution regarding computing resources, of increasing complexity and possibilities, the calculation programs or codes have turned to be a basic auxiliary tool in reactor physics. In order to analyze the global problem, several aspects should be taken into consideration. The first aspect to be considered is that of the availability of the necessary nuclear data. The second one is the existence of a variety of methods and models to perform the calculations. The final phase for this kind of analysis is the qualification of the computing programs to be used, i.e. the verification of the validity domain of its nuclear data and the models involved. The last one is an essential phase, and in order to carry it on great variety of calculations are required, that will check the different aspects contained in the code. We here analyze the most important physical processes that take place in a nuclear reactor cell, and we consider the qualification of the lattice code WIMS, that calculates the neutronic parameters associated with such processes. Particular emphasis has been put in the application to natural uranium fuelled reactor, heavy water cooled and moderated, as the Argentinean power reactors now in operation. A wide set of experiments has been chosen: a.-Fresh fuel in zero-power experimental facilities and power reactors; b.-Irradiated fuel in both types of facilities; c.-Benchmark (prototype) experiments with loss of coolant. From the whole analysis it was concluded that for the research reactors, as well as for the heavy water moderated power reactors presently operating in our country, or those that could operate in a near future, the lattice code WIMS is reliable and produces results within the experimental values and

  13. Lattice calculation of heavy-light decay constants with two flavors of dynamical quarks

    International Nuclear Information System (INIS)

    Bernard, C.; Datta, S.; DeGrand, T.; DeTar, C.; Gottlieb, Steven; Heller, Urs M.; McNeile, C.; Orginos, K.; Sugar, R.; Toussaint, D.

    2002-01-01

    We present results for f B , f B s , f D , f D s and their ratios in the presence of two flavors of light sea quarks (N f =2). We use Wilson light valence quarks and Wilson and static heavy valence quarks; the sea quarks are simulated with staggered fermions. Additional quenched simulations with nonperturbatively improved clover fermions allow us to improve our control of the continuum extrapolation. For our central values the masses of the sea quarks are not extrapolated to the physical u, d masses; that is, the central values are ''partially quenched.'' A calculation using 'fat-link clover' valence fermions is also discussed but is not included in our final results. We find, for example, f B =190(7)( -17 +24 )( -2 +11 )( -0 +8 ) MeV, f B s /f B =1.16(1)(2)(2)( -0 +4 ), f D s =241(5)( -26 +27 )( -4 +9 )( -0 +5 ) MeV, and f B /f D s =0.79(2)( -4 +5 )(3)( -0 +5 ), where in each case the first error is statistical and the remaining three are systematic: the error within the partially quenched N f =2 approximation, the error due to the missing strange sea quark and to partial quenching, and an estimate of the effects of chiral logarithms at small quark mass. The last error, though quite significant in decay constant ratios, appears to be smaller than has been recently suggested by Kronfeld and Ryan, and Yamada. We emphasize, however, that as in other lattice computations to date, the lattice u,d quark masses are not very light and chiral log effects may not be fully under control

  14. First lattice calculation of the B-meson binding and kinetic energies

    CERN Document Server

    Crisafulli, M; Martinelli, G; Sachrajda, Christopher T C

    1995-01-01

    We present the first lattice calculation of the B-meson binding energy \\labar and of the kinetic energy -\\lambda_1/2 m_Q of the heavy-quark inside the pseudoscalar B-meson. This calculation has required the non-perturbative subtraction of the power divergences present in matrix elements of the Lagrangian operator \\bar h D_4 h and of the kinetic energy operator \\bar h \\vec D^2 h. The non-perturbative renormalisation of the relevant operators has been implemented by imposing suitable renormalisation conditions on quark matrix elements, in the Landau gauge. Our numerical results have been obtained from several independent numerical simulations at \\beta=6.0 and 6.2, and using, for the meson correlators, the results obtained by the APE group at the same values of \\beta. Our best estimate, obtained by combining results at different values of \\beta, is \\labar =190 \\err{50}{30} MeV. For the \\overline{MS} running mass, we obtain \\overline {m}_b(\\overline {m}_b) =4.17 \\pm 0.06 GeV, in reasonable agreement with previous...

  15. Calculation method for control rod dropping time in reactor

    International Nuclear Information System (INIS)

    Nogami, Takeki; Kato, Yoshifumi; Ishino, Jun-ichi; Doi, Isamu.

    1996-01-01

    If a control rod starts dropping, the dropping speed is rapidly increased, then settled substantially constant, rapidly decreased when it reaches a dash pot. A second detection signal generated by removing an AC component from a first detection signal is differentiated twice. The time when the maximum value among the twice differentiated values is generated is determined as a time when the control rods starts dropping. The time when minimum value among the twice differentiated values is generated is determined as a time when the control rod reaches the dash pot of the reactor. The measuring time within a range from the time when the control rod starts dropping to the time when the control rod reaches the dash pot of the reactor is determined. As a result, processing for the calculation of the dropping start time and dash pot reaching time of the control rod can be automatized. Further, it is suffice to conduct differentiation twice till the reaching time, which can facilitate the processing thereby enabling to determine a reliable time range. (N.H.)

  16. Nuclear data for the calculation of thermal reactor reactivity coefficients

    International Nuclear Information System (INIS)

    1989-01-01

    On its 15th meeting in Vienna, 16-20 June 1986, the International Nuclear Data Committee (INDC) considered it important to review the accuracy with which changes in thermal reactor reactivity resulting from changes in temperature and coolant density can be predicted. It was noted that reactor physicists in several countries had to adjust the thermal neutron cross-section data base in order to reproduce measured reactivity coefficients. Consequently, it appeared to be essential to examine the consistency of the integral and differential cross-section data and to make all the information available which has a bearing on reactivity coefficient prediction. Following the recommendation of the INDC, the Nuclear Data Section of the International Atomic Energy Agency, therefore, convened the Advisory Group Meeting on Nuclear Data for the Calculation of Thermal Reaction Reactivity Coefficients, in Vienna, Austria, 7-10 Dec. 1987. The Conclusions and Recommendations of the meeting together with the papers presented, are submitted in the present document. A separate abstract was prepared for each of these 12 papers. Refs, figs and tabs

  17. Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

    International Nuclear Information System (INIS)

    Pinedo V, J.L.

    1979-01-01

    One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

  18. Static analysis of material testing reactor cores:critical core calculations

    International Nuclear Information System (INIS)

    Nawaz, A. A.; Khan, R. F. H.; Ahmad, N.

    1999-01-01

    A methodology has been described to study the effect of number of fuel plates per fuel element on critical cores of Material Testing Reactors (MTR). When the number of fuel plates are varied in a fuel element by keeping the fuel loading per fuel element constant, the fuel density in the fuel plates varies. Due to this variation, the water channel width needs to be recalculated. For a given number of fuel plates, water channel width was determined by optimizing k i nfinity using a transport theory lattice code WIMS-D/4. The dimensions of fuel element and control fuel element were determined using this optimized water channel width. For the calculated dimensions, the critical cores were determined for the given number of fuel plates per fuel element by using three dimensional diffusion theory code CITATION. The optimization of water channel width gives rise to a channel width of 2.1 mm when the number of fuel plates is 23 with 290 g ''2''3''5U fuel loading which is the same as in the case of Pakistan Reactor-1 (PARR-1). Although the decrease in number of fuel element results in an increase in optimal water channel width but the thickness of standard fuel element (SFE) and control fuel element (CFE) decreases and it gives rise to compact critical and equilibrium cores. The criticality studies of PARR-1 are in good agreement with the predictions

  19. Achievement and qualification of multigroup cross-section library for light water reactor calculation

    International Nuclear Information System (INIS)

    Gastaldi, B.

    1986-07-01

    This study intends to improve then to check on integral experiments, the calculation of the main neutronic parameters in light water moderated lattices: Uranium 238 capture and consequently Plutonium 239 build-up, multiplication factor, temperature coefficient. The first part of this work concerns the resonant reaction rate calculation method implemented in the APOLLO code, the so-called LIVOLANT and JEANPIERRE formalism. The errors introduced by the corresponding assumptions are quantified and we propose substitution methods which avoid large biases and supply satisfactory results. The second part is dedicated to the cross-section evaluation of uranium major isotopes and to the achievement of APOLLO multigroup cross-sections. This cross-section set takes into considerations on the one hand the recent differential information and the other hand the various integral information obtained in the French Atomic Energy Commission facilities. The nuclear data file (JEF abd ENDF/B5) processing, for multigroup and self-shielded cross-sections achieving enable us to check the new THEMIS computer code. In the last part, the experimental validation of the proposed procedure (accurate formalism mutuel shielding and new multigroup library) is presented. This qualification is based on the reinterpretation of critical experiments performed in the EOLE reactor at Cadarache and spent fuel analysis. The corresponding results demonstrate that our propositions provide improvements on the computation of the PWR neutronic parameters; calculation-experiment discrepancies are now consistent with experimental uncertainty margins. 46 refs; 31 figs; 23 tabl [fr

  20. Measurement and CFD calculation of spacer loss coefficient for a tight-lattice fuel bundle

    International Nuclear Information System (INIS)

    In, Wang Kee; Shin, Chang Hwan; Kwack, Young Kyun; Lee, Chi Young

    2015-01-01

    Highlights: • Experiment and CFD analysis evaluated the pressure drop in a spacer grid. • The measurement and CFD errors for the spacer loss coefficient were estimated. • The spacer loss coefficient for the dual-cooled annular fuel bundle was determined. • The CFD prediction agrees with the measured spacer loss coefficient within 8%. - Abstract: An experiment and computational fluid dynamics (CFD) analysis were performed to evaluate the pressure drop in a spacer grid for a dual-cooled annular fuel (DCAF) bundle. The DCAF bundle for the Korean optimum power reactor (OPR1000) is a 12 × 12 tight-lattice rod array with a pitch-to-diameter ratio of 1.08 owing to a larger outer diameter of the annular fuel rod. An experiment was conducted to measure the pressure drop in spacer grid for the DCAF bundle. The test bundle is a full-size 12 × 12 rod bundle with 11 spacer grid. The test condition covers a Reynolds number range of 2 × 10 4 –2 × 10 5 by changing the temperature and flow rate of water. A CFD analysis was also performed to predict the pressure drop through a spacer grid using the full-size and partial bundle models. The pressure drop and loss coefficient of a spacer grid were predicted and compared with the experimental results. The CFD predictions of spacer pressure drop and loss coefficient agree with the measured values within 8%. The spacer loss coefficient for the DCAF bundle is estimated to be approximately 1.50 at a nominal operating condition of OPR1000, i.e., Re = 4 × 10 5

  1. Calculation of the permeability in porous media using the lattice Boltzmann method

    International Nuclear Information System (INIS)

    Eshghinejadfard, Amir; Daróczy, László; Janiga, Gábor; Thévenin, Dominique

    2016-01-01

    Highlights: • Lattice Boltzmann simulation of fluid flow in porous media delivers a high accuracy. • Domain size, relaxation time and force scheme affect the calculated permeability. • Multiple relaxation time model shows very low viscosity dependence as compared to single relaxation time. • The choice of relaxation time and force scheme is a trade-off between the required accuracy and computational cost. - Abstract: In this paper, the lattice Boltzmann method (LBM) is used to simulate three-dimensional laminar flows in porous media and to calculate the associated permeability. An in-house, parallelized code using the message passing interface technique is employed for the study. Three different flow configurations are studied: first, by manually specifying solid cells in a face-centered cube (FCC); then, doing the same in a body-centered cube (BCC); and finally by reading the solid cells for a real 3D geometry from a set of experimental 2D computed tomography images. In all simulations, the Reynolds number is kept well below 1. It was found that the current LBM simulations yield good estimates for the permeability value. The impact of the employed force scheme and single- or multiple-relaxation time (SRT, MRT) was also studied. Although each force scheme (Guo-SRT, Guo-MRT and Shan-Chen-SRT) may show better results in some regions, the strong dependency of SRT models on relaxation time suggests that the proper choice of the force scheme, relaxation time and domain resolution is a compromise between the required accuracy and computational cost. First, higher resolutions lead as expected to increasingly accurate results but requires more computational cost and time. Second, the MRT model shows a lower viscosity dependence in comparison with SRT models but is somewhat slower. Also, the results are more sensitive to the relaxation time value for coarser domains. Furthermore, lower relaxation times necessitate a higher number of iterations to reach the steady

  2. DRAGON 3.05D, Reactor Cell Calculation System with Burnup

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: The computer code DRAGON contains a collection of models that can simulate the neutron behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: the interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations that can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. 2 - Methods: The code DRAGON contains a multigroup flux solver conceived that can use a various algorithms to solve the neutron transport equation for the spatial and angular distribution of the flux. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are considered as sources. The current release of DRAGON contains five such algorithms. The JPM option that solves the integral transport equation using the J+- method, (interface current method applied to homogeneous blocks); the SYBIL option that solves the integral transport equation using the collision probability method for simple one dimensional (1-D) or two dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; the EXCELL/NXT option to solve the integral transport equation using the collision probability method for more general 2-D geometries and for three dimensional (3-D) assemblies; the MOCC option to solve the transport equation using the method of cyclic characteristics in 2-D Cartesian, and finally the MCU option to solve the transport equation using the method of characteristics (non cyclic) for 3-D Cartesian geometries. The execution of DRAGON is

  3. Lattice vibrations and thermal properties of carbon nitride with defect ZnS structure from first-principles calculations

    NARCIS (Netherlands)

    Fang, C.M.; Wijs, G.A. de

    2004-01-01

    The phonon spectrum Of C3N4 with defect zincblende-type structure (deltaC(3)N(4)) was calculated by density functional theory (DFT) techniques. The results permit an assessment of important mechanical and thermodynamical properties such as the bulk modulus, lattice specific heat, vibration energy,

  4. Physics methods for calculating light water reactor increased performances

    International Nuclear Information System (INIS)

    Vandenberg, C.; Charlier, A.

    1988-01-01

    The intensive use of light water reactors (LWRs) has induced modification of their characteristics and performances in order to improve fissile material utilization and to increase their availability and flexibility under operation. From the conceptual point of view, adequate methods must be used to calculate core characteristics, taking into account present design requirements, e.g., use of burnable poison, plutonium recycling, etc. From the operational point of view, nuclear plants that have been producing a large percentage of electricity in some countries must adapt their planning to the need of the electrical network and operate on a load-follow basis. Consequently, plant behavior must be predicted and accurately followed in order to improve the plant's capability within safety limits. The Belgonucleaire code system has been developed and extensively validated. It is an accurate, flexible, easily usable, fast-running tool for solving the problems related to LWR technology development. The methods and validation of the two computer codes LWR-WIMS and MICROLUX, which are the main components of the physics calculation system, are explained

  5. Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

    International Nuclear Information System (INIS)

    Taryo, Taswanda

    2002-01-01

    Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

  6. Aspects of cell calculations in deterministic reactor core analysis

    International Nuclear Information System (INIS)

    Varvayanni, M.; Savva, P.; Catsaros, N.

    2011-01-01

    Τhe capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available for

  7. Comparative analysis of calculations and experiment for uranium-graphite lattices with natural and slightly-enriched uranium

    International Nuclear Information System (INIS)

    Khrennikov, N.N.; Shchukin, A.V.

    1988-01-01

    Three sets of experiments carried out at different times and in different laboratories on measuring the material parameter for uranium-graphite lattices using natural and slightly enriched uranium are analyzed. Comparison with the calculations by the TRIFOGR and MCU (the Monte Carlo method) codes reveals resonable agreement between the calculation and experiment (of the order of 0.4% in K eff ). 17 refs.; 3 tabs

  8. Development of a fine and ultra-fine group cell calculation code SLAROM-UF for fast reactor analyses

    International Nuclear Information System (INIS)

    Hazama, Taira; Chiba, Go; Sugino, Kazuteru

    2006-01-01

    A cell calculation code SLAROM-UF has been developed for fast reactor analyses to produce effective cross sections with high accuracy in practical computing time, taking full advantage of fine and ultra-fine group calculation schemes. The fine group calculation covers the whole energy range in a maximum of 900-group structure. The structure is finer above 52.5 keV with a minimum lethargy width of 0.008. The ultra-fine group calculation solves the slowing down equation below 52.5 keV to treat resonance structures directly and precisely including resonance interference effects. Effective cross sections obtained in the two calculations are combined to produce effective cross sections over the entire energy range. Calculation accuracy and improvements from conventional 70-group cell calculation results were investigated through comparisons with reference values obtained with continuous energy Monte Carlo calculations. It was confirmed that SLAROM-UF reduces the difference in k-infinity from 0.15 to 0.01% for a JOYO MK-I fuel subassembly lattice cell calculation, and from - 0.21% to less than a statistical uncertainty of the reference calculation of 0.03% for a ZPPR-10A core criticality calculation. (author)

  9. Shield design and streaming calculations for the sodium cooled PEC reactor

    International Nuclear Information System (INIS)

    Prosperi, M.; Tavoni, R.; Travaglini, N.

    1977-01-01

    This paper summarises the shielding calculations carried out for the PEC reactor. A brief description of calculation methods and of the work carried out to set them up is given; the most representative calculations with the relative isoflux curves are also referred. A general outline is then given for the main shielding problems of the PEC reactor

  10. Calculation of the CAREM reactor with the HUEMUL-PUMA-THERMIT chain of codes

    International Nuclear Information System (INIS)

    Notari, Carla; Grant, Carlos R.

    2000-01-01

    The purpose of the work was the evaluation of the the CAREM 25 reactor core, using a chain of codes (HUEMUL-PUMA-THERMIT) different to the one used in the original design (CONDOR-CITVAP-THERMIT). First, we performed a partial validation of the our codes in lattices similar to CAREM and reproduced a benchmark for simulation of gadolinium burnup. The results were considered satisfactory for this stage of the project. Then, we calculated the core along the normal operating equilibrium cycle and in hot and cold shut-down conditions. The main outcome of our evaluation confirms the general behaviour of the reference calculations except in one important point referring to the cold shut down. In this condition, the failure of one single rod of bank number 13 of the shut down system, leaves the core in a supercritical state at the beginning of the cycle and this anomaly persists during almost a third of the overall cycle. A new design of the core is proposed with minor modifications of the reference one, without introducing new types of fuel elements and keeping the same fuel management scheme. This new core fulfills all the design requirements. (author)

  11. Integral transport multiregion geometrical shadowing factor for the approximate collision probability matrix calculation of infinite closely packed lattices

    International Nuclear Information System (INIS)

    Jowzani-Moghaddam, A.

    1981-01-01

    An integral transport method of calculating the geometrical shadowing factor in multiregion annular cells for infinite closely packed lattices in cylindrical geometry is developed. This analytical method has been programmed in the TPGS code. This method is based upon a consideration of the properties of the integral transport method for a nonuniform body, which together with Bonalumi's approximations allows the determination of the approximate multiregion collision probability matrix for infinite closely packed lattices with sufficient accuracy. The multiregion geometrical shadowing factors have been calculated for variations in fuel pin annular segment rings in a geometry of annular cells. These shadowing factors can then be used in the calculation of neutron transport from one annulus to another in an infinite lattice. The result of this new geometrical shadowing and collision probability matrix are compared with the Dancoff-Ginsburg correction and the probability matrix using constant shadowing on Yankee fuel elements in an infinite lattice. In these cases the Dancoff-Ginsburg correction factor and collision probability matrix using constant shadowing are in difference by at most 6.2% and 6%, respectively

  12. Evaluation of the use of color-set geometry during lattice physics constants generation for boiling water reactor simulation

    International Nuclear Information System (INIS)

    Evans, S.; Ivanov, K.

    2013-01-01

    Current methods for BWR nuclear design and analysis consist of using lattice physics neutron transport methods to generate the two-group homogenized cross-sections that are then used in a nodal diffusion theory code. The lattice transport solutions are performed for a single assembly with reflective boundary conditions, which is a practical approximation. A method is developed to account for assembly exposure distributions (environment) in the core within the lattice transport calculations with the use of color-sets (2x2) geometry. The loading pattern is examined and an appropriate number of characteristic color-set cells are selected for analysis. Treatment of the co-resident exposed fuel within this method is also presented. The calculation process was followed for a recent BWR cycle design with comparisons being performed on both a lattice and core-wide basis to evaluate the proposed method. The lattice based comparisons show noticeable differences in the pin power distribution predictions, which require further investigation to see how this translates into core performance calculations. The core-wide comparisons show minor differences and are generally in a good agreement, which is expected with this small perturbation. A slight improvement was noticed in the reduction of the power distribution uncertainty. However, given the additional amount of work and computer run time increase, further evaluation, especially of core pin power predictions, is needed to consider this method for production level design and safety analysis calculations. (authors)

  13. Criticality calculations in reactor accelerator coupling experiment (Race)

    International Nuclear Information System (INIS)

    Reda, M.A.; Spaulding, R.; Hunt, A.; Harmon, J.F.; Beller, D.E.

    2005-01-01

    A Reactor Accelerator Coupling Experiment (RACE) is to be performed at the Idaho State University Idaho Accelerator Center (IAC). The electron accelerator is used to generate neutrons by inducing Bremsstrahlung photon-neutron reactions in a Tungsten- Copper target. This accelerator/target system produces a source of ∼1012 n/s, which can initiate fission reactions in the subcritical system. This coupling experiment between a 40-MeV electron accelerator and a subcritical system will allow us to predict and measure coupling efficiency, reactivity, and multiplication. In this paper, the results of the criticality and multiplication calculations, which were carried out using the Monte Carlo radiation transport code MCNPX, for different coupling design options are presented. The fuel plate arrangements and the surrounding tank dimensions have been optimized. Criticality using graphite instead of water for reflector/moderator outside of the core region has been studied. The RACE configuration at the IAC will have a criticality (k-effective) of about 0,92 and a multiplication of about 10. (authors)

  14. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  15. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  16. Calculation of fuel and moderator temperature coefficients in APR1400 nuclear reactor by MVP code

    International Nuclear Information System (INIS)

    Pham Tuan Nam; Le Thi Thu; Nguyen Huu Tiep; Tran Viet Phu

    2014-01-01

    In this project, these fuel and moderator temperature coefficients were calculated in APR1400 nuclear reactor by MVP code. APR1400 is an advanced water pressurized reactor, that was researched and developed by Korea Experts, its electric power is 1400 MW. The neutronics calculations of full core is very important to analysis and assess a reactor. Results of these calculation is input data for thermal-hydraulics calculations, such as fuel and moderator temperature coefficients. These factors describe the self-safety characteristics of nuclear reactor. After obtaining these reactivity parameters, they were used to re-run the thermal hydraulics calculations in LOCA and RIA accidents. These thermal-hydraulics results were used to analysis effects of reactor physics parameters to thermal hydraulics situation in nuclear reactors. (author)

  17. Neutronic study of nuclear reactors. Complete calculation of TRIGA MARKII reactor and calculations of fuel temperature coefficients. (Qualification of WIMS code)

    International Nuclear Information System (INIS)

    Benmansour, L.

    1992-01-01

    The present work shows a group of results, obtained by a neutronic study, concerning the TRIGA MARK II reactor and LIGHT WATER reactors. These studies aim to make cell and diffusion calculations. WIMS D-4 with extended library and DIXY programs are used and tested for those purposes. We also have proceeded to a qualification of WIMS code based on the fuel temperature coefficient calculations. 33 refs.; 23 figs.; 30 tabs. (author)

  18. Parameter definition for reactor physics calculation of Obrigheim KWO PWR type reactor using the Gels and Erebus codes

    International Nuclear Information System (INIS)

    Faya, A.G.; Nakata, H.; Rodrigues, V.G.; Oosterkamp, W.J.

    1974-01-01

    The main variables for Obrigheim Reactor - KWO diffusion theory calculations, using the EREBUS code were defined. The variables under consideration were: mesh spacing for reactor description, time-step in burn-up calculation, and the temperature in both the moderator and the fuel. The best mesh spacing and time-step were defined considering the relative deviations and the computer time expended in each case. It has been verified that the error involved in the mean fuel temperature calculation (1317 0 K as given by SIEMENS and 1028 0 K as calculated by Dr. Penndorf) does not change substancially the calculation results

  19. Higgs compositeness in Sp(2N) gauge theories - Determining the low-energy constants with lattice calculations

    Science.gov (United States)

    Bennett, Ed; Ki Hong, Deog; Lee, Jong-Wan; David Lin, C.-J.; Lucini, Biagio; Piai, Maurizio; Vadacchino, Davide

    2018-03-01

    As a first step towards a quantitative understanding of the SU(4)/Sp(4) composite Higgs model through lattice calculations, we discuss the low energy effective field theory resulting from the SU(4) → Sp(4) global symmetry breaking pattern. We then consider an Sp(4) gauge theory with two Dirac fermion flavours in the fundamental representation on a lattice, which provides a concrete example of the microscopic realisation of the SU(4)/Sp(4) composite Higgs model. For this system, we outline a programme of numerical simulations aiming at the determination of the low-energy constants of the effective field theory and we test the method on the quenched theory. We also report early results from dynamical simulations, focussing on the phase structure of the lattice theory and a calculation of the lowest-lying meson spectrum at coarse lattice spacing. Combined contributions of B. Lucini (e-mail: b.lucini@swansea.ac.uk) and J.-W. Lee (e-mail: wlee823@pusan.ac.kr).

  20. High-pressure lattice dynamics and thermodynamic properties of zinc-blende BN from first-principles calculation

    International Nuclear Information System (INIS)

    Wang Huanyou; Xu Hui; Wang Xianchun; Jiang Chunzhi

    2009-01-01

    The density function perturbation theory (DFPT) is employed to study the lattice dynamics and thermodynamic properties (with quasiharmonic approximation) of zinc-blende BN. First we discuss the structural properties and compare the phonon spectrum with available Raman scattering experiments. Thereafter using the calculated phonon dispersions we obtain the PTV equation of state from the free energy. Our results for the above properties are generally speaking in good agreement with experiments and with similar theoretical calculations. Owing to the anharmonic effect at high temperature, the calculated linear thermal expansion coefficients (CTE) are low to experimental data.

  1. Measurements and calculations of integral capture cross-sections of structural materials in fast reactor spectra

    International Nuclear Information System (INIS)

    Seth, S.; Brunson, G.; Gmuer, K.; Jermann, M.; McCombie, C.; Richmond, R.; Schmocker, U.

    1979-01-01

    This paper relates the checking of integral data of steel and iron in fast reactor lattices. The fully-rodded GCFR benchmark lattice of the zero-energy reactor PROTEUS was successively modified by replacing the PuO 2 -UO 2 fuel rods by steel-18/8 or steel-37 (iron) rods. The neutron spectra of the modified lattices in fact have median energies close to that of a typical LMFBR. The replacement of fuel by the structural material of interest was such that in each case the value of k(infinity) was reduced to near-unity. This allowed the measurement of the lattice-k(infinity) by the null-reactivity technique. In addition, the principal reaction rates (namely U238 capture and fission, relative to Pu239 fission) and the neutron spectrum were measured. These directly measured integral data which are particularly sensitive to the steel cross-sections can be used for the checking and systematic adjustment of data sets. The results may also be analysed so as to derive specific values for the integral capture cross-sections of steel and iron. Neutron balance equations were set-up for each of the lattices using the measured k(infinity) and reaction rates

  2. On the theories, techniques, and computer codes used in numerical reactor criticality and burnup calculations

    International Nuclear Information System (INIS)

    El-Osery, I.A.

    1981-01-01

    The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented

  3. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor

    International Nuclear Information System (INIS)

    Alcala Ruiz, F.

    1973-01-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  4. A first look at maximally twisted mass lattice QCD calculations at the physical point

    Energy Technology Data Exchange (ETDEWEB)

    Abdel-Rehim, A. [The Cyprus Institute, Nicosia (Cyprus). CaSToRC; Boucaud, P. [Paris XI Univ., Orsay (France). Laboratoire de Physique Theorique; Carrasco, N. [Valencia-CSIC Univ. (Spain). Dept. de Fisica Teorica; IFIC, Valencia (Spain); and others

    2013-11-15

    In this contribution, a first look at simulations using maximally twisted mass Wilson fermions at the physical point is presented. A lattice action including clover and twisted mass terms is presented and the Monte Carlo histories of one run with two mass-degenerate flavours at a single lattice spacing are shown. Measurements from the light and heavy-light pseudoscalar sectors are compared to previous N{sub f}=2 results and their phenomenological values. Finally, the strategy for extending simulations to N{sub f}=2+1+1 is outlined.

  5. A first look at maximally twisted mass lattice QCD calculations at the physical point

    International Nuclear Information System (INIS)

    Abdel-Rehim, A.

    2013-11-01

    In this contribution, a first look at simulations using maximally twisted mass Wilson fermions at the physical point is presented. A lattice action including clover and twisted mass terms is presented and the Monte Carlo histories of one run with two mass-degenerate flavours at a single lattice spacing are shown. Measurements from the light and heavy-light pseudoscalar sectors are compared to previous N f =2 results and their phenomenological values. Finally, the strategy for extending simulations to N f =2+1+1 is outlined.

  6. Lattice Boltzmann equation calculation of internal, pressure-driven turbulent flow

    International Nuclear Information System (INIS)

    Hammond, L A; Halliday, I; Care, C M; Stevens, A

    2002-01-01

    We describe a mixing-length extension of the lattice Boltzmann approach to the simulation of an incompressible liquid in turbulent flow. The method uses a simple, adaptable, closure algorithm to bound the lattice Boltzmann fluid incorporating a law-of-the-wall. The test application, of an internal, pressure-driven and smooth duct flow, recovers correct velocity profiles for Reynolds number to 1.25 x 10 5 . In addition, the Reynolds number dependence of the friction factor in the smooth-wall branch of the Moody chart is correctly recovered. The method promises a straightforward extension to other curves of the Moody chart and to cylindrical pipe flow

  7. Development of a core follow calculational system for research reactors

    International Nuclear Information System (INIS)

    Muller, E.Z.; Ball, G.; Joubert, W.R.; Schutte, H.C.; Stoker, C.C.; Reitsma, F.

    1994-01-01

    Over the last few years a comprehensive Pressurized Water Reactor and Materials Testing Reactor core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments. 10 refs., 1 tab., 1 fig

  8. Numerical effects in the neutron flux calculations into WWER-type reactor vessels by Monte Carlo method

    International Nuclear Information System (INIS)

    Alvarez Cardona, C.M.; Rodriguez Gual, M.; Hernandez Valle, S.

    2001-01-01

    The calculation of neutron fluxes and fluence into reactor pressure vessel is a regulatory requirement in the stages of the design, operation and plan lifetime extension. The reactor vessel is considered a unique and non-substitutable part of the NPP that undergoes degradation. The main source of the aging comes from the fast neutron damage induced in the steel crystalline lattice. Due to the proximity of the core edge to the vessel inner surface; the vessel steel is exposed to high fast neutron fluence. The effect of this irradiation on the mechanical properties becomes more acute because of the impurities measured in the Russian steel alloys. In the present paper, a PC version of the Monte Carlo 3-D HEXANN-EVALU system is used for the estimation of the WWER reactor pressure vessel irradiation. It was selected on the basis of its flexible options that on the other hand need to be quantified in connection with the desired magnitudes. The parameters that control the random walk of neutrons as well as the efficiency increasing options included in the code are studied in order to identify their impact in the final results for fluxes and fluence in the reactor pressure vessel. As a result an optimal set of parameters is suggested. (authors)

  9. Fuel lattice design in a boiling water reactor using a knowledge-based automation system

    Energy Technology Data Exchange (ETDEWEB)

    Tung, Wu-Hsiung, E-mail: wstong@iner.gov.tw; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

    2015-11-15

    Highlights: • An automation system was developed for the fuel lattice radial design of BWRs. • An enrichment group peaking equalizing method is applied to optimize the design. • Several heuristic rules and restrictions are incorporated to facilitate the design. • The CPU time for the system to design a 10x10 lattice was less than 1.2 h. • The beginning-of-life LPF was improved from 1.319 to 1.272 for one of the cases. - Abstract: A knowledge-based fuel lattice design automation system for BWRs is developed and applied to the design of 10 × 10 fuel lattices. The knowledge implemented in this fuel lattice design automation system includes the determination of gadolinium fuel pin location, the determination of fuel pin enrichment and enrichment distribution. The optimization process starts by determining the gadolinium distribution based on the pin power distribution of a flat enrichment lattice and some heuristic rules. Next, a pin power distribution flattening and an enrichment grouping process are introduced to determine the enrichment of each fuel pin enrichment type and the initial enrichment distribution of a fuel lattice design. Finally, enrichment group peaking equalizing processes are performed to achieve lower lattice peaking. Several fuel lattice design constraints are also incorporated in the automation system such that the system can accomplish a design which meets the requirements of practical use. Depending on the axial position of the lattice, a different method is applied in the design of the fuel lattice. Two typical fuel lattices with U{sup 235} enrichment of 4.471% and 4.386% were taken as references. Application of the method demonstrates that improved lattice designs can be achieved through the enrichment grouping and the enrichment group peaking equalizing method. It takes about 11 min and 1 h 11 min of CPU time for the automation system to accomplish two design cases on an HP-8000 workstation, including the execution of CASMO-4

  10. Fuel lattice design in a boiling water reactor using a knowledge-based automation system

    International Nuclear Information System (INIS)

    Tung, Wu-Hsiung; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

    2015-01-01

    Highlights: • An automation system was developed for the fuel lattice radial design of BWRs. • An enrichment group peaking equalizing method is applied to optimize the design. • Several heuristic rules and restrictions are incorporated to facilitate the design. • The CPU time for the system to design a 10x10 lattice was less than 1.2 h. • The beginning-of-life LPF was improved from 1.319 to 1.272 for one of the cases. - Abstract: A knowledge-based fuel lattice design automation system for BWRs is developed and applied to the design of 10 × 10 fuel lattices. The knowledge implemented in this fuel lattice design automation system includes the determination of gadolinium fuel pin location, the determination of fuel pin enrichment and enrichment distribution. The optimization process starts by determining the gadolinium distribution based on the pin power distribution of a flat enrichment lattice and some heuristic rules. Next, a pin power distribution flattening and an enrichment grouping process are introduced to determine the enrichment of each fuel pin enrichment type and the initial enrichment distribution of a fuel lattice design. Finally, enrichment group peaking equalizing processes are performed to achieve lower lattice peaking. Several fuel lattice design constraints are also incorporated in the automation system such that the system can accomplish a design which meets the requirements of practical use. Depending on the axial position of the lattice, a different method is applied in the design of the fuel lattice. Two typical fuel lattices with U"2"3"5 enrichment of 4.471% and 4.386% were taken as references. Application of the method demonstrates that improved lattice designs can be achieved through the enrichment grouping and the enrichment group peaking equalizing method. It takes about 11 min and 1 h 11 min of CPU time for the automation system to accomplish two design cases on an HP-8000 workstation, including the execution of CASMO-4 lattice

  11. TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

    1983-01-01

    TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

  12. Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Ohnuki, A.; Kureta, M.; Liu, W.; Tamai, H.; Akimoto, H.

    2004-01-01

    Research and development project for investigating thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured light-water reactor technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important issues for the RMWR because of the tight-lattice configuration. The project has mainly consisted of a large-scale thermal-hydraulic test and development of analytical methods named modeling engineering. In the large-scale test, 37-rod bundle experiments can be performed. Steady-state critical power experiments have been achieved in the test facility and the experimental data reveal the feasibility of RMWR

  13. Advances of study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Akira Ohnuki; Kazuyuki Takase; Masatoshi Kureta; Hiroyuki Yoshida; Hidesada Tamai; Wei Liu; Toru Nakatsuka; Hajime Akimoto

    2005-01-01

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R and D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods, grid spacer configuration etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.0 mm) and the experimental data reveal the feasibility of RMWR. (authors)

  14. Lattice location of dopant atoms: An N-body model calculation

    Indian Academy of Sciences (India)

    Here we applied the superior -body model to study the yield from bismuth in silicon. The finding that bismuth atom occupies a position close to the silicon substitutional site is new. The transverse displacement of the suggested lattice site from the channelling direction is consistent with the experimental results. The above ...

  15. Calculation of anti-seismic design for Xi'an pulsed reactor

    International Nuclear Information System (INIS)

    Li Shuian

    2002-01-01

    The author describes the reactor safety rule, safety regulation and design code that must be observed to anti-seismic design in Xi'an pulsed reactor. It includes the classification of reactor installation, determination of seismic loads, calculate contents, program, method, results and synthetically evaluation. According to the different anti-seismic structure character of reactor installation, an appropriate method was selected to calculate the seismic response. The results were evaluated synthetically using the design code and design requirement. The evaluate results showed that the anti-seismic design function of reactor installation of Xi'an pules reactor is well, and the structure integrality and normal property of reactor installation can be protect under the designed classification of the earthquake

  16. Three-dimensional discrete ordinates reactor assembly calculations on GPUs

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Thomas M [ORNL; Joubert, Wayne [ORNL; Hamilton, Steven P [ORNL; Johnson, Seth R [ORNL; Turner, John A [ORNL; Davidson, Gregory G [ORNL; Pandya, Tara M [ORNL

    2015-01-01

    In this paper we describe and demonstrate a discrete ordinates sweep algorithm on GPUs. This sweep algorithm is nested within a multilevel comunication-based decomposition based on energy. We demonstrated the effectiveness of this algorithm on detailed three-dimensional critical experiments and PWR lattice problems. For these problems we show improvement factors of 4 6 over conventional communication-based, CPU-only sweeps. These sweep kernel speedups resulted in a factor of 2 total time-to-solution improvement.

  17. Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

    2009-01-01

    In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

  18. Calculation of the main neutron parameters of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Ojima, Mario Katsuhiko

    1977-01-01

    The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined

  19. Calculating the Unit Cost Factors for Decommissioning Cost Estimation of the Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Jeong, Kwan Seong; Lee, Dong Gyu; Jung, Chong Hun; Lee, Kune Woo

    2006-01-01

    The estimated decommissioning cost of nuclear research reactor is calculated by applying a unit cost factor-based engineering cost calculation method on which classification of decommissioning works fitted with the features and specifications of decommissioning objects and establishment of composition factors are based. Decommissioning cost of nuclear research reactor is composed of labor cost, equipment and materials cost. Labor cost of decommissioning costs in decommissioning works are calculated on the basis of working time consumed in decommissioning objects. In this paper, the unit cost factors and work difficulty factors which are needed to calculate the labor cost in estimating decommissioning cost of nuclear research reactor are derived and figured out.

  20. Calculation of lattice sums and electrical field gradients for the rhombic and tetragonal phases of YBa2Cu3Ox

    International Nuclear Information System (INIS)

    Lyubutin, I.S.; Terziev, V.G.; Gor'kov, V.P.

    1989-01-01

    The point charge model is used to calculate the lattice sums and determine the electrical field gradients (EFG) as well as the asymmetry parameters η for all cation sites of the rhombic and tetragonal phases of the superconductor YBa 2 Cu 3 O x . The cases of copper of different valency at the Cu 1 sites are considered separately and EFG and η values are calculated in the vicinity of local defects caused by differences in the number and ordering of the oxygen vacancies at the Cu1 sites

  1. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  2. Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2005-07-01

    At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

  3. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    International Nuclear Information System (INIS)

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  4. Topics in quantum chromodynamics: two loop Feynman gauge calculation of the meson nonsinglet evolution potential and fourier acceleration of the calculation of the fermion propagator in lattice QCD

    International Nuclear Information System (INIS)

    Katz, G.R.

    1986-01-01

    Part I of this thesis is a perturbative QCD calculation to two loops of the meson nonsinglet evolution potential in the Feynman gauge. The evolution potential describes the momentum dependence of the distribution amplitude. This amplitude is needed for the calculation to beyond leading order of exclusive amplitudes and form factors. Techniques are presented that greatly simplify the calculation. The results agree with an independent light-cone gauge calculation and disagree with predictions made using conformal symmetry. In Part II the author presents a Fourier acceleration method that is effective in accelerating the computation of the fermion propagator in lattice QCD. The conventional computation suffers from critical slowing down: the long distance structure converges much slower than the short distance structure. by evaluating the fermion propagator in momentum space using fast Fourier transforms, it is possible to make different length scales converge at a more equal rate. From numerical experiments made on a 8 4 lattice, the author obtained savings of a factor of 3 to 4 by using Fourier acceleration. He also discusses the important of gauge fixing when using Fourier acceleration

  5. Calculation analysis of the neutronic experimental data coming from the NUR reactor start-up

    International Nuclear Information System (INIS)

    Madariaga, M.; Villarino, E.; Relloso, J.; Rubio, R

    1991-01-01

    NUR is a new MTR reactor located in Argelia which became critical in march 1989. It is loaded with a 19 plates LEU Fe. This paper contains: a) Reactivity measurements in the first cores technical information about the Fe and some other data necessary for performing cell and reactor calculations b) calculation comparisons with the measured values (2-D and 3-D calculations) with an statistical analysis of the data set from the control rod calibration. (orig.)

  6. Prediction of Low-Thermal-Conductivity Compounds with First-Principles Anharmonic Lattice-Dynamics Calculations and Bayesian Optimization

    Science.gov (United States)

    Seko, Atsuto; Togo, Atsushi; Hayashi, Hiroyuki; Tsuda, Koji; Chaput, Laurent; Tanaka, Isao

    2015-11-01

    Compounds of low lattice thermal conductivity (LTC) are essential for seeking thermoelectric materials with high conversion efficiency. Some strategies have been used to decrease LTC. However, such trials have yielded successes only within a limited exploration space. Here, we report the virtual screening of a library containing 54 779 compounds. Our strategy is to search the library through Bayesian optimization using for the initial data the LTC obtained from first-principles anharmonic lattice-dynamics calculations for a set of 101 compounds. We discovered 221 materials with very low LTC. Two of them even have an electronic band gap <1 eV , which makes them exceptional candidates for thermoelectric applications. In addition to those newly discovered thermoelectric materials, the present strategy is believed to be powerful for many other applications in which the chemistry of materials is required to be optimized.

  7. Coupled hydro-neutronic calculations for fast burst reactor accidents

    International Nuclear Information System (INIS)

    Paternoster, R.; Kimpland, R.; Jaegers, P.; McGhee, J.

    1994-01-01

    Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by either single point kinetics or in a decoupled manner. This methodology is enabled by the use of modem supercomputers (CM-200). Two examples of this capability are presented: an unreflected metal fast burst assembly, and a reflected fast burst assembly typical of the Skua or SPR-III class of fast burst reactor

  8. First Lattice Calculation of the QED Corrections to Leptonic Decay Rates

    Science.gov (United States)

    Giusti, D.; Lubicz, V.; Tarantino, C.; Martinelli, G.; Sachrajda, C. T.; Sanfilippo, F.; Simula, S.; Tantalo, N.

    2018-02-01

    The leading-order electromagnetic and strong isospin-breaking corrections to the ratio of Kμ 2 and πμ 2 decay rates are evaluated for the first time on the lattice, following a method recently proposed. The lattice results are obtained using the gauge ensembles produced by the European Twisted Mass Collaboration with Nf=2 +1 +1 dynamical quarks. Systematic effects are evaluated and the impact of the quenched QED approximation is estimated. Our result for the correction to the tree-level Kμ 2/πμ 2 decay ratio is -1.22 (16 )%, to be compared to the estimate of -1.12 (21 )% based on chiral perturbation theory and adopted by the Particle Data Group.

  9. Results of Koo measurements of HTGR lattice by oscillated zero reactivity technique using the AGIP-NUCLEARE RB-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, F; Brighenti, G.; Chiodi, P. L.; Ghilardotti, G.; Giuliani, C.

    1974-10-15

    This paper describes k-infinity measurements conducted using an assembly of loose HTGR coated particles in the BR-2 reactor by means of null reactivity oscillating method comparing the effect of poisoned and unpoisoned lattices like tests performed in the Physical Constants Test Reactor (PCTR) at Hanford. The RB-2 reactor was the property of the Italian firm AGIP NUCLEARE and operated at the Montecuccolino Center in Bologna.

  10. Accurate calculation of Green functions on the d-dimensional hypercubic lattice

    International Nuclear Information System (INIS)

    Loh, Yen Lee

    2011-01-01

    We write the Green function of the d-dimensional hypercubic lattice in a piecewise form covering the entire real frequency axis. Each piece is a single integral involving modified Bessel functions of the first and second kinds. The smoothness of the integrand allows both real and imaginary parts of the Green function to be computed quickly and accurately for any dimension d and any real frequency, and the computational time scales only linearly with d.

  11. Dependence of calculated void reactivity on film boiling representation in a CANDU lattice

    Energy Technology Data Exchange (ETDEWEB)

    Whitlock, J [McMaster Univ., Hamilton, ON (Canada). Dept. of Engineering Physics

    1994-12-31

    The distribution dependence of void reactivity in a CANDU (CANada Deuterium Uranium) lattice is studied, specifically in the regime of film boiling. A heterogeneous model of this phenomenon predicts a 4% increase in void reactivity over a homogeneous model for fresh fuel, and 11% at discharge. An explanation for this difference is offered, with regard to differing changes in neutron mean free path upon voiding. (author). 9 refs., 4 tabs., 6 figs.

  12. Accident analysis of RB reactor dependent on the lattice pitch; Akcidentalna analiza reaktora ''RB'' pri promeni koraka resetke

    Energy Technology Data Exchange (ETDEWEB)

    Lolic, B; Lazarevic, B [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1963-07-01

    This analysis was concerned with reactor core with 52-56 fuel rods, lattice pitch being, 8, 10, 12, 16, 18, and 20 cm. Measured values of excess reactivity above critical level of 3.85 cm, total anti reactivity of regulating rod, reactivity changes caused by pumping heavy water and reactivity variations due to movement of control rod were used. Three types of accidents were analyzed: movement of regulating rod to the position of zero reactivity worths, increase of heavy water level at rate of 2.5 cm/min, combination of two previous accidents.

  13. Simplified polynomial representation of cross sections for reactor calculation

    International Nuclear Information System (INIS)

    Dias, A.M.; Sakai, M.

    1985-01-01

    It is shown a simplified representation of a cross section library generated by transport theory using the cell model of Wigner-Seitz for typical PWR fuel elements. The effect of burnup evolution through tables of reference cross sections and the effect of the variation of the reactor operation parameters considered by adjusted polynomials are presented. (M.C.K.) [pt

  14. Reactor accident calculation models in use in the Nordic countries

    International Nuclear Information System (INIS)

    Tveten, U.

    1984-01-01

    The report relates to a subproject under a Nordic project called ''Large reactor accidents - consequences and mitigating actions''. In the first part of the report short descriptions of the various models are given. A systematic list by subject is then given. In the main body of the report chapter and subchapter headings are by subject. (Auth.)

  15. Evaluating variability with atomistic simulations: the effect of potential and calculation methodology on the modeling of lattice and elastic constants

    Science.gov (United States)

    Hale, Lucas M.; Trautt, Zachary T.; Becker, Chandler A.

    2018-07-01

    Atomistic simulations using classical interatomic potentials are powerful investigative tools linking atomic structures to dynamic properties and behaviors. It is well known that different interatomic potentials produce different results, thus making it necessary to characterize potentials based on how they predict basic properties. Doing so makes it possible to compare existing interatomic models in order to select those best suited for specific use cases, and to identify any limitations of the models that may lead to unrealistic responses. While the methods for obtaining many of these properties are often thought of as simple calculations, there are many underlying aspects that can lead to variability in the reported property values. For instance, multiple methods may exist for computing the same property and values may be sensitive to certain simulation parameters. Here, we introduce a new high-throughput computational framework that encodes various simulation methodologies as Python calculation scripts. Three distinct methods for evaluating the lattice and elastic constants of bulk crystal structures are implemented and used to evaluate the properties across 120 interatomic potentials, 18 crystal prototypes, and all possible combinations of unique lattice site and elemental model pairings. Analysis of the results reveals which potentials and crystal prototypes are sensitive to the calculation methods and parameters, and it assists with the verification of potentials, methods, and molecular dynamics software. The results, calculation scripts, and computational infrastructure are self-contained and openly available to support researchers in performing meaningful simulations.

  16. Development and qualification of reference calculation schemes for absorbers in pressured water reactor; Elaboration et qualification de schemas de calcul de reference pour les absorbants dans les reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Blanc-Tranchant, P

    2001-07-01

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code APOLLO2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B4C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI4. They were then checked against experimental data measured during French experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  17. Elaboration and qualification of a reference calculation routes for the absorbers in the PWR reactors; Elaboration et qualification des schemas de calcul de reference pour les absorbants dans les reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Blanc-Tranchant, P

    1999-11-01

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B{sub 4}C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  18. The general formulation and practical calculation of the diffusion coefficient in a lattice containing cavities; Formulation generale et calcul pratique du coefficient de diffusion dans un reseau comportant des cavites

    Energy Technology Data Exchange (ETDEWEB)

    Benoist, P [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    The calculation of diffusion coefficients in a lattice necessitates the knowledge of a correct method of weighting the free paths of the different constituents. An unambiguous definition of this weighting method is given here, based on the calculation of leakages from a zone of a reactor. The formulation obtained, which is both simple and general, reduces the calculation of diffusion coefficients to that of collision probabilities in the different media; it reveals in the expression for the radial coefficient the series of the terms of angular correlation (cross terms) recently shown by several authors. This formulation is then used to calculate the practical case of a classical type of lattice composed of a moderator and a fuel element surrounded by an empty space. Analytical and numerical comparison of the expressions obtained with those inferred from the theory of BEHRENS shows up the importance of several new terms some of which are linked with the transparency of the fuel element. Cross terms up to the second order are evaluated. A practical formulary is given at the end of the paper. (author) [French] Le calcul des coefficients de diffusion dans un reseau suppose la connaissance d'un mode de ponderation correct des libres parcours des differents constituants. On definit ici sans ambiguite ce mode de ponderation a partir du calcul des fuites hors d'une zone de reacteur. La formulation obtenue, simple et generale, ramene le calcul des coefficients de diffusion a celui des probabilites de collision dans les differents milieux; elle fait apparaitre dans l'expression du coefficient radial la serie des termes de correlation angulaire (termes rectangles), mis en evidence recemment par plusieurs auteurs. Cette formulation est ensuite appliquee au calcul pratique d'un reseau classique, compose d'un moderateur et d'un element combustible entoure d'une cavite; la comparaison analytique et numerique des expressions obtenues avec celles deduites de la theorie de BEHRENS

  19. Calculation of toroidal fusion reactor blankets by Monte Carlo

    International Nuclear Information System (INIS)

    Macdonald, J.L.; Cashwell, E.D.; Everett, C.J.

    1977-01-01

    A brief description of the calculational method is given. The code calculates energy deposition in toroidal geometry, but is a continuous energy Monte Carlo code, treating the reaction cross sections as well as the angular scattering distributions in great detail

  20. Installation and testing of the ERANOS computer code for fast reactor calculations

    International Nuclear Information System (INIS)

    Gren, Milan

    2010-12-01

    The French ERANOS computer code was acquired and tested by solving benchmark problems. Five problems were calculated: 1D XZ Model, 1D RZ Model, 3D HEX SNR 300 reactor, 2S HEX and 3D HEX VVER 440 reactor. The multi-group diffuse approximation was used. The multiplication coefficients were compared within the first problem, neutron flux density in the calculation points was obtained within the second problem, and powers in the various reactor areas and in the assemblies were calculated within the remaining problems. (P.A.)

  1. Monte carlo calculation of the neutron effective dose rate at the outer surface of the biological shield of HTR-10 reactor

    International Nuclear Information System (INIS)

    Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina

    2012-01-01

    Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.

  2. A study of the literature on nodal methods in reactor physics calculations

    International Nuclear Information System (INIS)

    Van de Wetering, T.F.H.

    1993-01-01

    During the last few decades several calculation methods have been developed for the three-dimensional analysis of a reactor core. A literature survey was carried out to gain insights in the starting points and method of operation of the advanced nodal methods. These methods are applied in reactor core analyses of large nuclear power reactors, because of their high computing speed. The so-called Nodal-Expansion method is described in detail

  3. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    International Nuclear Information System (INIS)

    Estes, B.F.; Berry, D.T.

    1980-02-01

    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters

  4. Development of a model for the primary system CAREM reactor's stationary thermohydraulic calculation

    International Nuclear Information System (INIS)

    Gaspar, C.; Abbate, P.

    1990-01-01

    The ESCAREM program oriented to CAREM reactors' stationary thermohydraulic calculation is presented. As CAREM gives variations in relation to models for BWR (Boiling Water Reactors)/PWR (Pressurized Water Reactors) reactors, it was decided to develop a suitable model which allows to calculate: a) if the Steam Generator design is adequate to transfer the power required; b) the circulation flow that occurs in the Primary System; c) the temperature at the entrance (cool branch) and d) the contribution of each component to the pressure drop in the circulation connection. Results were verified against manual calculations and alternative numerical models. An experimental validation at the Thermohydraulic Essays Laboratory is suggested. A parametric analysis series is presented on CAREM 25 reactor, demonstrating operating conditions, at different power levels, as well as the influence of different design aspects. (Author) [es

  5. Comparison of radiation measurements and calculations of reactor surroundings for skyshine analysis

    Energy Technology Data Exchange (ETDEWEB)

    Tsubosaka, A.; Nomura, Y. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kawabe, T. [Japan Research Institute, Limited, Osaka (Japan); Zharkov, V.P.; Kartashev, I.A.; Netecha, M.E.; Orlov, Y.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    2000-03-01

    ISTC Project 'Experimental Studies of Radiation Scattering in the Atmosphere' were conducted using the IVG-1M and RA reactors by RDIPE in collaboration with IAE NNC RK and JAERI during 1996-1998. The radial distributions of fast neutron flux, thermal neutron flux and gamma radiation dose rate were measured above these two reactors at three heights. Neutron spectra above these two reactors and thermal and fast neutron fluxes over the hollow pipe height in the IVG-1M reactor were also measured in order to determine the radiation characteristics for skyshine analysis. For verifying the computer codes the calculations of reactor surroundings were performed using MCNP and DORT/DOT-3.5. The comparisons between the measurements and the calculations show that MCNP and DORT/DOT-3.5 codes can be widely applied to the shielding problems by selecting properly the calculation conditions. (author)

  6. Two dimensional neutron transport calculation system for plate-reactors: experimental design and qualification with SILOE

    International Nuclear Information System (INIS)

    Roussos, N.

    1982-01-01

    The main objective of this work is to create a neutronic calculations system for the SILOE-SILOETTE reactors, adaptable to other types of plate reactors. The author presents the methodology and the development of the APOLLO 1D (99 gr.) calculations for the creation of cross sections libraries. After a recall of the Discrete Ordinate Method (DOT), the method accuracy is studied in order to optimize the spatial discretization of the calculations; calculations of DOT 3.5 and of SILOETTE core are conducted and their convergence and costs are examined. DOT calculations of SILOETTE and experimental tests results are then compared [fr

  7. Qualification of JEFF3.1.1 library for high conversion reactor calculations using the ERASME/R experiment

    Energy Technology Data Exchange (ETDEWEB)

    Vidal, J. F.; Noguere, G.; Peneliau, Y.; Santamarina, A. [CEA, DEN, DER/SPRC/LEPh, Cadarache, F-13108 Saint-Paul-lez-Durance (France)

    2012-07-01

    With its low CO{sub 2} production, Nuclear Energy appears to be an efficient solution to the global warming due to green-house effect. However, current LWR reactors are poor uranium users and, pending the development of Fast Neutron Reactors, alternative concepts of PWR with higher conversion ratio (HCPWR) are being studied again at CEA, first studies dating from the middle 80's. In these French designs, low moderation ratio has been performed by tightening the lattice pitch, achieving a conversion ratio of 0.8-0.9 with a MOX fuel coming from PWR UOX recycling. Theses HCPWRs are characterized by a harder neutron spectrum and the calculation uncertainties on the fundamental neutronics parameters are increased by a factor 3 regarding a standard PWR lattice, due to the major contribution of the Plutonium isotopes and of the epithermal energy range to the reaction rates. In order to reduce these uncertainties, a 3-year experimental validation program called ERASME has been performed by CEA from 1984 to 1986 in the EOLE reactor. Monte Carlo analysis of the ERASME/R experiments with the Monte Carlo code TRIPOLI4 allowed the qualification of the recommended JEFF.3.1.1 library for major neutronics parameters. K{sub eff} of the MOX under-moderated lattice is over-predicted by 440 {+-} 830 pcm (2{sigma}); the conversion ratio, indicator of the good use of uranium, is also slightly over-predicted: 2 % {+-} 4 % (2{sigma}) and the same for B4C absorber rods worth and soluble boron worth, over-predicted by 2 %, both in the 2 standard deviations range. The radial fission maps of heterogeneities (water-holes, B4C and fertile rods) are well reproduced: maximal (C-E)/E dispersion is 1.3 %, maximal power peak error is 2.7 %. The void reactivity worth is the only parameter poorly calculated with an overprediction of +12.4% {+-} 1.5%. ERASME/R analysis of MOX reactivity, void effect and spectral indexes will contribute to the reevaluation of {sup 241}Am and Plutonium isotopes

  8. Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

    International Nuclear Information System (INIS)

    Pan, Dongqing; Chien Jen, Tien; Li, Tao; Yuan, Chris

    2014-01-01

    This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domain with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired

  9. Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Dongqing; Chien Jen, Tien [Department of Mechanical Engineering, University of Wisconsin-Milwaukee, Milwaukee, Wisconsin 53201 (United States); Li, Tao [School of Mechanical Engineering, Dalian University of Technology, Dalian 116024 (China); Yuan, Chris, E-mail: cyuan@uwm.edu [Department of Mechanical Engineering, University of Wisconsin-Milwaukee, 3200 North Cramer Street, Milwaukee, Wisconsin 53211 (United States)

    2014-01-15

    This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domain with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired.

  10. Diffusion calculation's for the SLOWPOKE-2 reactor using DONJON

    International Nuclear Information System (INIS)

    Noceir, S.; El Hajjaji, O.; Varin, E.

    1997-01-01

    The SLOWPOKE reactor at Ecole Polytechnique will be refueled with a Low Enriched Uranium (LEU) fuel in place of a High Enriched Uranium (HEU) fuel used until now. The purpose of this study is to provide various models, using the reactor physics chain of codes DRAGON/DONJON, in order to predict the behavior of the new LEU Slowpoke. In particle, we will present some numerical results concerning the separate temperature effects of the main components of the core, the effect of a partial void appearing near the fuel pins and the axial and radial flux distributions. Finally the difference between the present HEU and the future LEU fuel power will be given. (author)

  11. Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR [High Flux Isotope Reactor] Reactor

    International Nuclear Information System (INIS)

    Childs, R.L.; Rhoades, W.A.; Williams, L.R.

    1988-01-01

    The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs

  12. Calculation of deuteron interactions within micro-cracks of a D2 loaded lattice at room temperature

    International Nuclear Information System (INIS)

    Fulvio, F.

    2007-01-01

    We have analyzed the possibility that the coefficient of lattice deformation, linked to the formation of micro-cracks at room temperature and low energies, could influence the process of fusion. The calculated probability of fusion within a micro-crack, in the presence of D 2 loading at room temperature and for impure metals, shows moderately elevated values compared with the probability of fusion on the surface. For all the temperatures in the 150-350 K range and for all the energies between 150 and 250 eV, the formation of micro-cracks increases the probability of fusion compared to non-deformed lattices, and also reduces the thickness of the Coulomb barrier. Using the trend of the curve of potential to evaluate the influence of the concentration of impurities, a very high barrier is found within the pure lattice (J ∼ 0.25%). However, under the same thermodynamic conditions, the probability of fusion in the impure metal (J ∼ 0.75%) could be higher, with a total energy less than the potential so that the tunneling effect is amplified. Finally, we have analysed the influence of forced D 2 loading on the process. (author)

  13. Lattice calculation of the leading strange quark-connected contribution to the muon $g-2$

    CERN Document Server

    Blum, T.; Del Debbio, L.; Hudspith, R.J.; Izubuchi, T.; Jüttner, A.; Lehner, C.; Lewis, R.; Maltman, K.; Krstić Marinković, M.; Portelli, A.; Spraggs, M.

    2016-04-11

    We present results for the leading hadronic contribution to the muon anomalous magnetic moment due to strange quark-connected vacuum polarisation effects. Simulations were performed using RBC--UKQCD's $N_f=2+1$ domain wall fermion ensembles with physical light sea quark masses at two lattice spacings. We consider a large number of analysis scenarios in order to obtain solid estimates for residual systematic effects. Our final result in the continuum limit is $a_\\mu^{(2)\\,{\\rm had},\\,s}=53.1(9)\\left(^{+1}_{-3}\\right)\\times10^{-10}$.

  14. Lattice calculation of the leading strange quark-connected contribution to the muon g−2

    Energy Technology Data Exchange (ETDEWEB)

    Blum, T. [Physics Department, University of Connecticut,Storrs, CT 06269-3046 (United States); Boyle, P.A.; Debbio, L. Del [School of Physics and Astronomy, University of Edinburgh,Peter Guthrie Tait Road, Edinburgh EH9 3JZ (United Kingdom); Hudspith, R.J. [Department of Physics and Astronomy, York University,4700 Keele Street, Toronto, Ontario, M3J 1P3 (Canada); Izubuchi, T. [Physics Department, Brookhaven National Laboratory,Upton, NY 11973 (United States); RIKEN-BNL Research Center, Brookhaven National Laboratory,Upton, NY 11973 (United States); Jüttner, A. [School of Physics and Astronomy, University of Southampton,Southampton SO17 1BJ (United Kingdom); Lehner, C. [Physics Department, Brookhaven National Laboratory,Upton, NY 11973 (United States); Lewis, R. [Department of Physics and Astronomy, York University,4700 Keele Street, Toronto, Ontario, M3J 1P3 (Canada); Maltman, K. [Department of Mathematics and Statistics, York University,4700 Keele Street, Toronto, Ontario, M3J 1P3 (Canada); CSSM, University of Adelaide,Adelaide, SA 5005 (Australia); Marinković, M. Krstić [School of Physics and Astronomy, University of Southampton,Southampton SO17 1BJ (United Kingdom); CERN, Theoretical Physics Department, CERN,Geneva (Switzerland); Portelli, A. [School of Physics and Astronomy, University of Edinburgh,Peter Guthrie Tait Road, Edinburgh EH9 3JZ (United Kingdom); School of Physics and Astronomy, University of Southampton,Southampton SO17 1BJ (United Kingdom); Spraggs, M. [School of Physics and Astronomy, University of Southampton,Southampton SO17 1BJ (United Kingdom); Collaboration: The RBC/UKQCD collaboration

    2016-04-11

    We present results for the leading hadronic contribution to the muon anomalous magnetic moment due to strange quark-connected vacuum polarisation effects. Simulations were performed using RBC-UKQCD’s N{sub f}=2+1 domain wall fermion ensembles with physical light sea quark masses at two lattice spacings. We consider a large number of analysis scenarios in order to obtain solid estimates for residual systematic effects. Our final result in the continuum limit is a{sub μ}{sup (2)} {sup had,} {sup s}=53.1(9)({sub −3}{sup +1})×10{sup −10}.

  15. Lattice calculation of the leading strange quark-connected contribution to the muon g−2

    International Nuclear Information System (INIS)

    Blum, T.; Boyle, P.A.; Debbio, L. Del; Hudspith, R.J.; Izubuchi, T.; Jüttner, A.; Lehner, C.; Lewis, R.; Maltman, K.; Marinković, M. Krstić; Portelli, A.; Spraggs, M.

    2016-01-01

    We present results for the leading hadronic contribution to the muon anomalous magnetic moment due to strange quark-connected vacuum polarisation effects. Simulations were performed using RBC-UKQCD’s N f =2+1 domain wall fermion ensembles with physical light sea quark masses at two lattice spacings. We consider a large number of analysis scenarios in order to obtain solid estimates for residual systematic effects. Our final result in the continuum limit is a μ (2) had, s =53.1(9)( −3 +1 )×10 −10 .

  16. Desk top calculation strategies for reactor analysis using Mathematica

    International Nuclear Information System (INIS)

    Kullberg, C.

    1991-01-01

    Mathematics is one of several recently developed equations analysis programs that is particularly well suited for solving a broad range of intermediate engineering problems. The objective of this paper is to demonstrate, using a couple of reactor-related examples, how Mathematica can be exploited as a user-friendly analysis tool to symbolically and numerically handle systems of algebraic and differential equations. 7 refs., 3 figs

  17. Calculation of limit cycle amplitudes in commercial boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Perez, R.B.; Cacuci, D.G.

    1984-01-01

    This paper describes an investigation of the dynamic behavior of a boiling water reactor (BWR) in the nonlinear region corresponding to linearly unstable conditions. A nonlinear model of a typical BWR was developed. The equations underlying this model represent a one-dimensional void reactivity feedback, point kinetics with a single delayed neutron group, fuel behavior, and recirculation loop dynamics (described by a single-node integral momentum equation)

  18. Transport calculation of neutron flux distribution in reflector of PW reactor

    International Nuclear Information System (INIS)

    Remec, I.

    1982-01-01

    Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

  19. Calculation of radiation heat generation on a graphite reflector side of IAN-R1 Reactor

    International Nuclear Information System (INIS)

    Duque O, J.; Velez A, L.H.

    1987-01-01

    Calculation methods for radiation heat generation in nuclear reactor, based on the point kernel approach are revisited and applied to the graphite reflector of IAN-R1 reactor. A Fortran computer program was written for the determination of total heat generation in the reflector, taking 1155 point in it

  20. The new electricity of France PWR: calculation scheme of neutron leakages from the reactor cavity

    International Nuclear Information System (INIS)

    Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.; Champion, G.

    1987-04-01

    A new calculation scheme is adapted to evaluate neutron fluxes in the reactor cavity and the containment of next french PWR. In this scheme a large part is given to Monte Carlo method, coupled with SN-method, in order to take into account multiple neutron diffusions and the complexity of the reactor geometry

  1. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    International Nuclear Information System (INIS)

    Moiseyev, A.V.

    2008-01-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k eff , control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  2. Decay heat measurement on fusion reactor materials and validation of calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)

  3. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  4. Calculation of reactivity of control rods in graphite moderated reactors

    International Nuclear Information System (INIS)

    Nakata, H.

    1978-01-01

    A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies [pt

  5. Multi-parameter variational calculations for the (2+1)-dimensional U(1) lattice gauge theory and the XY model

    International Nuclear Information System (INIS)

    Heys, D.W.; Stump, D.R.

    1987-01-01

    Variational calculations are described that use multi-parameter trial wave functions for the U(1) lattice gauge theory in two space dimensions, and for the XY model. The trial functions are constructed as the exponential of a linear combination of states from the strong-coupling basis of the model, with the coefficients treated as variational parameters. The expectation of the hamiltonian is computed by the Monte Carlo method, using a reweighting technique to evaluate expectation values in finite patches of the parameter space. The trial function for the U(1) gauge theory involves six variational parameters, and its weak-coupling behaviour is in reasonable agreement with theoretical expectations. (orig.)

  6. Guidelines for calculation of atmospheric dispersion and radiological consequences of design basis reactor accidents - Severe accident calculation guidelines, EPR

    International Nuclear Information System (INIS)

    Martens, R.; Schmitz, B.M.; Horn, M.

    1999-01-01

    The activities carried out within the (reduced) project period (1. Sept. until 31. Dec. 1998) for coordinated harmonization between France and Germany, of guidelines for calculation of the radiological consequences of a severe reactor accident, are summarized. (orig./CB) [de

  7. Validation of The Deterministic Diffusion Method For The Neutronic Calculations of Thermal Research Reactors of TRIGA-Type Using The Wisdom-IAEA-69 Nuclear Data Library

    International Nuclear Information System (INIS)

    Hussein, H.M.; Sakr, A.M.; Amin, E.H.

    2011-01-01

    The objective of this paper is to assess the suitability and the accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA type research reactors in proposed condensed energy spectra of five and seven groups with one and three thermal groups respectively, using the calculational line: WIMSD-IAEA-69 nuclear data library/ WIMSD-5B lattice and cell calculations code/ CITVAP v3.1 core calculations code. Firstly, The assessment goes through analyzing the integral parameters - k e ff, ρ 238 , σ 235 , σ 238 , and C * - of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra, which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark- III Thai research reactor, using the CITVAP v3.1 code and macroscopic cross-section libraries generated using the WIMSD-5B code at the proposed energy spectra separately. The results include the excess reactivities and the worth of control rods, which were compared with previous Monte Carlo results and experimental values, that show good agreement with the references at both energy spectra, albeit better accuracies are shown with the five groups spectrum. The results also includes neutron flux distributions which are settled for future comparisons with other calculational techniques, even, they are comparable to reactors and fuels of the same type. The study reflects the adequacy of using the pre-stated calculational line at the condensed energy spectra for evaluation of the neutronic parameters of the TRIGA type reactors, and future comparisons of the un-benchmarked results could assure this result for wider range of neutronics or safety-related parameters

  8. Chapter 10: Calculation of the temperature coefficient of reactivity of a graphite-moderated reactor

    International Nuclear Information System (INIS)

    Brown, G.; Richmond, R.; Stace, R.H.W.

    1963-01-01

    The temperature coefficients of reactivity of the BEPO, Windscale and Calder reactors are calculated, using the revised methods given by Lockey et al. (1956) and by Campbell and Symonds (1962). The results are compared with experimental values. (author)

  9. Determination of D2O - 2% enriched uranium lattice parameters by means of a critical system

    International Nuclear Information System (INIS)

    Raisic, N.; Takac, S.; Markovic, H.; Bosevski, T.

    1963-01-01

    In order to specify experimental procedures for few standard measurements sufficient to provide consistent set of lattice parameters, a series of experiments were performed at the RB reactor using 2% enriched tubular fuel elements. Obtained results were compared to standard two-group diffusion calculation indicating high degree of accuracy for a broad variety of reactor lattice configurations

  10. Nuclear data requirements for fission reactor neutronics calculations

    International Nuclear Information System (INIS)

    Finck, P.

    1998-01-01

    The paper discusses current European nuclear data measurement and evaluation requirements for fission reactor technology applications and problems involved in meeting the requirements. Reference is made to the NEA High Priority Nuclear Data Request List and to the production of the new JEFF-3 library of evaluated nuclear data. There are requirements for both differential (or basic) nuclear data measurements and for different types of integral measurement critical facility measurements and isotopic sample irradiation measurements. Cross-section adjustment procedures are being used to take into account the simpler types of integral measurement, and to define accuracy needs for evaluated nuclear data

  11. Calculated investigation of actinide transmutation in the BOR-60 reactor

    International Nuclear Information System (INIS)

    Zhemkov, I.Yu.; Ishunina, O.V.; Yakovleva, I.V.

    2001-01-01

    In the course of reactor operation the formation of fission products and accumulation of minor-actinides and plutonium take place in the nuclear fuel. These materials define the radiation hazard to a great extent. Of one possible ways lowering the activity of irradiated nuclear fuel is transmutation of long-lived radioactive isotopes in the stable or short-lived ones, that allows to facilitate the problem of the high-level waste and to improve the efficiency of nuclear fuel use at the expense of its recycling and burnup increasing. (authors)

  12. Investigation of the possibility of a calculative reactor safety estimation in the licence procedure for nuclear reactors

    International Nuclear Information System (INIS)

    Adler, B.; Kampf, T.

    1975-12-01

    Up to now it is impossible to calculate completely the safety of nuclear reactors. Therefore the authors have collected and employed a number of at a high degree independent safety parameters for mathematical evaluation of the reactor safety. By means of computer programs such parameters from about 400 research reactors have been analysed and the fluctuation ranges of their greatest density were determined. The limits of these fluctuation ranges are quickly available and can be used as recommended values for the layout and for the safety estimation of research reactors. A comparison of the existing layout recommendations and the determined fluctuation ranges in most cases shows a good agreement. In some cases corrections and new layout recommendations have been proposed. The determined fluctuation ranges found their first practical application in the estimation of the Rossendorf Equipment for Critical Experiments (RAKE). (author)

  13. Calculation to experiment comparison of SPND signals in various nuclear reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance, (France); Snoj, Luka [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana, (Slovenia); Tarchalski, Mikolaj [National Centre for Nuclear Research, ulica Andrzeja Soltana 7, 05-400 Otwock (Swierk), (Poland); Dewynter-Marty, Veronique [CEA, DEN, DANS, DRSN, SIREN, LESCI, Saclay, F-91191 Gif sur Yvette, (France); Malouch, Fadhel [CEA, DEN, DANS, DM2S, SERMA, Saclay, F-91191 Gif sur Yvette, (France)

    2015-07-01

    In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first part of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)

  14. Problems in calculating reactor model (primary circuit) for nuclear power plant diagnostics

    International Nuclear Information System (INIS)

    Markov, P.

    1986-01-01

    Some results are presented of the calculation of eigen-vibrations of the system of WWER-440 nuclear reactor vessels in a vacuum and in a liquid. Computer code BOSOR 4 has been written for calculating forced vibrations of shells with axial symmetry and of a simplified system of reactor vessels. A description is given of this code, which is based on the so-called energy method of finite differences. Briefly discussed is the feasibility of applying the results of the latest computation techniques in the diagnostics of the major components of a nuclear reactor. (Z.M.)

  15. Calculation of static harmonics of a nuclear reactor using CITATION code

    International Nuclear Information System (INIS)

    Belchior Junior, A.; Moreira, J.M.L.

    1989-01-01

    The CITATION code, which solves the multigroup diffusion equation by the finite difference method, calculates the fundamental λ-mode (harmonic) for nuclear reactors. In this work, two fission source correction methods are attempted to obtain higher λ-modes through the CITATION code. The two methods are compared, their advantages and disadvantages analysed and verified against analytical solutions. Two dimensional harmonic modes are calculated for the IEA-R1 research reactor and for the ANGRA-I power reactor. The results are shown in graphics and tables. (author) [pt

  16. Calculation of integral parameters sensitivity in fast reactors

    International Nuclear Information System (INIS)

    Renke, C.A.C.

    1981-01-01

    The variational formulation, incorporated to VARI-1D computer code is used the sensitivity calculations. At a first stage the direct method was also used with the objective of establishing a parallel between the two methods.(E.G.) [pt

  17. Whole core calculations of power reactors by Monte Carlo method

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Mori, Takamasa

    1993-01-01

    Whole core calculations have been performed for a commercial size PWR and a prototype LMFBR by using vectorized Monte Carlo codes. Geometries of cores were precisely represented in a pin by pin model. The calculated parameters were k eff , control rod worth, power distribution and so on. Both multigroup and continuous energy models were used and the accuracy of multigroup approximation was evaluated through the comparison of both results. One million neutron histories were tracked to considerably reduce variances. It was demonstrated that the high speed vectorized codes could calculate k eff , assembly power and some reactivity worths within practical computation time. For pin power and small reactivity worth calculations, the order of 10 million histories would be necessary. Required number of histories to achieve target design accuracy were estimated for those neutronic parameters. (orig.)

  18. Frequency Calculation For Loss Coolant Accident In The Nuclear Reactor

    International Nuclear Information System (INIS)

    Sony, DT

    1996-01-01

    LOCA as initiating event is engineering judgement, because it is rare condition. So, to determine LOCA frequency used be probability and statistic method. By probability and statistic method was estimated from size, weld, age, learning curve and quality, etc. it has been calculated for LOCA frequency in the simplified piping system model, especially estimates from size and weld factors. From calculation, LOCA frequency is 9,82.10 - 6/year

  19. First principle calculation of structure and lattice dynamics of Lu2Si2O7

    Directory of Open Access Journals (Sweden)

    Nazipov D.V.

    2017-01-01

    Full Text Available Ab initio calculations of crystal structure and Raman spectra has been performed for single crystal of lutetium pyrosilicate Lu2Si2O7. The types of fundamental vibrations, their frequencies and intensities in the Raman spectrum has been obtained for two polarizations. Calculations were made in the framework of density functional theory (DFT with hybrid functionals. The isotopic substitution was calculated for all inequivalent ions in cell. The results in a good agreement with experimental data.

  20. Optimization of the neutron calculation model for the RA-6 reactor

    International Nuclear Information System (INIS)

    Coscia, G.A.

    1981-01-01

    A model for the neutronic calculation of the RA-6 reactor which includes the codes ANISN and EQUIPOSE is analyzed. Starting with a brief description of the reactor, the core and its parts, the general scheme of calculation applied is presented. The fuel elements used were those which are utilized in the RA-3 reactor; this is of the MTR type with 90% enriched uranium. With the approximations used, an analysis of such model of calculation was made, trying to optimize it by reducing, if possible, the calculation time without loosing accuracy. In order to improve the calculation model, it is recomended a cross section data library specific for the enrichment of the fuel considered 90% and the incorporation of a more advanced code than EQUIPOISE which would be DIXYBAR. (M.E.L.) [es

  1. Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Sakr, A.M.; Amin, E.H.

    2011-01-01

    Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

  2. Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

    International Nuclear Information System (INIS)

    Moukhamadeev, R.; Suvorov, A.

    2000-01-01

    This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

  3. Calculation of fuel burn-up and fuel reloading for the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lan, Nguyen Phuoc; Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Binh, Do Quang [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Calculation of fuel burnup and fuel reloading for the Dalat Nuclear Research Reactor was carried out by using a new programme named HEXA-BURNUP, realized in a PC. The programme is used to calculate the following parameters of the Dalat reactor: a/Critical configurations of the core loaded with 69, 72, 74, 86, 88, 89 and 92 fuel elements. The effective multiplication coefficients equal 1 within the error ranges of less than 0.38%. b/ The thermal neutron flux distribution in the reactor. The calculated results agree with the experimental data measured at 11 typical positions. c/The average fuel burn-up for the period from Feb. 1984 to Sep. 1992. The difference between calculation and experiment is only about 1.9%. 10 fuel reloading versions are calculated, from which an optimal version is proposed. (author). 9 refs., 4 figs., 5 tabs.

  4. An assessment of methods of calculating sodium voiding reactivity in plutonium fuelled fast reactors

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Simmons, W.N.; Stevenson, J.M.

    1979-01-01

    After a survey of the requirements an assessment of the accuracy of calculations of the sodium void effect using UK methods and data is made on the basis of the following work. First, the analysis of small and large sodium voids in the MOZART and Zebra 13 small (300 MW(E)) fast reactor mock-ups and the BIZET large fast reactor mock-ups, all of conventional design. The analysis was carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code, whole reactor diffusion theory calculations of the neutron flux and perturbation theory methods. Exact perturbation theory was used in many cases, otherwise first order perturbation theory calculations were adjusted to give results equivalent to exact perturbation theory. Second, theoretical studies of some effects, including, the effects of extrapolating to fuel operating temperatures, fuel cycle and burn-up effects, and the heterogeneity effects of large fuelled subassemblies in pin geometry. Third, theoretical studies of approximations in the calculational methods including, the importance in the whole reactor calculation of the energy group structure and the spatial mesh, the importance of reactor material boundaries in the calculation of resonance shielding effects, and the use of neutron fluxes calculated using neutron diffusion theory rather than transport theory. (U.K.)

  5. Gamma-point lattice free energy estimates from O(1) force calculations

    DEFF Research Database (Denmark)

    Voss, Johannes; Vegge, Tejs

    2008-01-01

    We present a new method for estimating the vibrational free energy of crystal (and molecular) structures employing only a single force calculation, for a particularly displaced configuration, in addition to the calculation of the ground state configuration. This displacement vector is the sum...

  6. SRAC-95, Cell Calculation with Burnup, Fuel Management for Thermal Reactors

    International Nuclear Information System (INIS)

    Tsuchihashi, K.; Ishiguro, Y.; Kaneko, K.; Ido, M.

    2004-01-01

    1 - Description of program or function: General neutronics calculation including cell calculation with burn-up, core calculation for any type of thermal reactor. Core burn-up calculation and fuel management by an auxiliary code. 2 - Method of solution: Collision probability method, 1D and 2D Sn for cell calculation; 1D, 2D and 3D diffusion for core calculation. 3 - Restrictions on the complexity of the problem: 20 regions for a continuous energy resonance absorption calculation and 16 steps for cell burn-up

  7. Parallel diffusion length on thermal neutrons in rod type lattices

    International Nuclear Information System (INIS)

    Ahmed, T.; Siddiqui, S.A.M.M.; Khan, A.M.

    1981-11-01

    Calculation of diffusion lengths of thermal neutrons in lead-water and aluminum water lattices in direction parallel to the rods are performed using one group diffusion equation together with Shevelev transport correction. The formalism is then applied to two practical cases, the Kawasaki (Hitachi) and the Douglas point (Candu) reactor lattices. Our results are in good agreement with the observed values. (author)

  8. Model for calculating the boron concentration in PWR type reactors

    International Nuclear Information System (INIS)

    Reis Martins Junior, L.L. dos; Vanni, E.A.

    1986-01-01

    A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

  9. A model for steady-state and transient determination of subcooled boiling for calculations coupling a thermohydraulic and a neutron physics calculation program for reactor core calculation

    International Nuclear Information System (INIS)

    Mueller, R.G.

    1987-06-01

    Due to the strong influence of vapour bubbles on the nuclear chain reaction, an exact calculation of neutron physics and thermal hydraulics in light water reactors requires consideration of subcooled boiling. To this purpose, in the present study a dynamic model is derived from the time-dependent conservation equations. It contains new methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. Furthermore, it enables the complete two-phase flow region to be treated in a consistent manner. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement was reached. The results from the coupling of the new calculation model with a neutron kinetics program proved its suitability for the steady-state and transient calculation of reactor cores. (orig.) [de

  10. Discrete lattice plane broken bond interfacial energy calculations and the use of the dividing surface concept

    International Nuclear Information System (INIS)

    Ramanujan, R.V.

    2003-01-01

    The concept of the dividing surface has been extensively used to define the relationships between thermodynamic quantities at the interface between two phases; it is also useful in calculations of interfacial energy (γ). However, in the original formulation, the two phases are continuum phases, the atomistic nature of the interface was not considered. It is, therefore, useful to examine the use of the dividing surface in the context of atomistic interfacial energy calculations. The case of a planar fcc:hcp interface is considered and the dividing surface positions which are useful in atomistic interfacial energy calculations are stated, one position equates γ to the excess internal energy, the other position allows us to use the Gibbs adsorption equation. An example of a calculation using the convenient dividing surface positions is presented

  11. Calculations of lattice vibrational mode lifetimes using Jazz: a Python wrapper for LAMMPS

    International Nuclear Information System (INIS)

    Gao, Y; Wang, H; Daw, M S

    2015-01-01

    Jazz is a new python wrapper for LAMMPS [1], implemented to calculate the lifetimes of vibrational normal modes based on forces as calculated for any interatomic potential available in that package. The anharmonic character of the normal modes is analyzed via the Monte Carlo-based moments approximation as is described in Gao and Daw [2]. It is distributed as open-source software and can be downloaded from the website http://jazz.sourceforge.net/. (paper)

  12. Calculations of lattice vibrational mode lifetimes using Jazz: a Python wrapper for LAMMPS

    Science.gov (United States)

    Gao, Y.; Wang, H.; Daw, M. S.

    2015-06-01

    Jazz is a new python wrapper for LAMMPS [1], implemented to calculate the lifetimes of vibrational normal modes based on forces as calculated for any interatomic potential available in that package. The anharmonic character of the normal modes is analyzed via the Monte Carlo-based moments approximation as is described in Gao and Daw [2]. It is distributed as open-source software and can be downloaded from the website http://jazz.sourceforge.net/.

  13. Delayed Neutron Fraction (beta-effective) Calculation for VVER 440 Reactor

    International Nuclear Information System (INIS)

    Hascik, J.; Michalek, S.; Farkas, G.; Slugen, V.

    2008-01-01

    Effective delayed neutron fraction (β eff ) is the main parameter in reactor dynamics. In the paper, its possible determination methods are summarized and a β eff calculation for a VVER 440 power reactor as well as for training reactor VR1 using stochastic transport Monte Carlo method based code MCNP5 is made. The uncertainties in determination of basic delayed neutron parameters lead to the unwished conservatism in the reactor control system design and operation. Therefore, the exact determination of the β eff value is the main requirement in the field of reactor dynamics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of experiments and measurements what do you mean differences between different calculation approaches and experimental results. In consequence of difficulties in β eff experimental measurement, this value in exact state is determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. An accurate estimate of β eff is essential for converting reactivity, as measured in dollars, to an absolute reactivity and/or to an absolute k eff . In the past, k eff has been traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum-weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. The summary of the possible β eff determination methods can be found in this work and also a calculation of β eff first for the training reactor VR1 in one operation state and then for VVER 440 power reactor in two different operation states are made using the prompt method, by MCNP5 code.(author)

  14. The concerted calculation of the BN-600 reactor for the deterministic and stochastic codes

    Science.gov (United States)

    Bogdanova, E. V.; Kuznetsov, A. N.

    2017-01-01

    The solution of the problem of increasing the safety of nuclear power plants implies the existence of complete and reliable information about the processes occurring in the core of a working reactor. Nowadays the Monte-Carlo method is the most general-purpose method used to calculate the neutron-physical characteristic of the reactor. But it is characterized by large time of calculation. Therefore, it may be useful to carry out coupled calculations with stochastic and deterministic codes. This article presents the results of research for possibility of combining stochastic and deterministic algorithms in calculation the reactor BN-600. This is only one part of the work, which was carried out in the framework of the graduation project at the NRC “Kurchatov Institute” in cooperation with S. S. Gorodkov and M. A. Kalugin. It is considering the 2-D layer of the BN-600 reactor core from the international benchmark test, published in the report IAEA-TECDOC-1623. Calculations of the reactor were performed with MCU code and then with a standard operative diffusion algorithm with constants taken from the Monte - Carlo computation. Macro cross-section, diffusion coefficients, the effective multiplication factor and the distribution of neutron flux and power were obtained in 15 energy groups. The reasonable agreement between stochastic and deterministic calculations of the BN-600 is observed.

  15. PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method

    International Nuclear Information System (INIS)

    Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua

    1990-01-01

    1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant

  16. Program MCU for Monte-Carlo calculations of neutron-physical characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    Abagyan, L.P.; Alekseev, N.I.; Bryzgalov, V.I.; Glushkov, A.E.; Gomin, E.A.; Gurevich, M.I.; Kalugin, M.A.; Majorov, L.V.; Marin, S.V.; Yhdkevich, M.S.

    1994-01-01

    A description of the MCU data modification is presented. The calculation results by the MCU-2 and MCU-3 codes are compared for the critical assemblies of a different reactor types. The full list of the critical assemblies calculation results obtained by all MCU code versions is given. 32 refs.; 32 tabs

  17. Neutronic calculations for the conceptual design of an in-reactor solid breeder experiment, TRIO-01

    International Nuclear Information System (INIS)

    Childs, R.L.; Gabriel, T.A.; Lillie, R.A.

    1981-03-01

    Neutronics calculations have been performed to obtain tritium production and heat generation rates for the irradiation of solid tritium breeding materials in the Oak Ridge Research Reactor (ORR). Two breeder materials, Li 2 O and LiAlO 2 , were considered. Burnup calculations were performed to estimate the amount of 6 Li present as a function of time

  18. Program system for calculating streaming neutron radiation field in reactor cavity

    International Nuclear Information System (INIS)

    He Zhongliang; Zhao Shu.

    1986-01-01

    The A23 neutron albedo data base based on Monte Carlo method well agrees with SAIL albedo data base. RSCAM program system, using Monte Carlo method with albedo approach, is used to calculate streaming neutron radiation field in reactor cavity and containment operating hall. The dose rate distributions calculated with RSCAM in square concrete duct well agree with experiments

  19. Monte Carlo calculation of the nuclear temperature coefficient in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matthes, W.

    1974-04-15

    A Monte Carlo program for the calculation of the nuclear temperature coefficient for fast reactors is described. The special difficulties for this problem are the energy and space dependence of the cross sections and the calculation of differential eifects. These difficulties are discussed in detail and the way for their solution chosen in this program is described. (auth)

  20. The impact of ENDF/B-VI Rev. 3 data on thermal reactor lattices

    International Nuclear Information System (INIS)

    Trkov, A.

    1995-10-01

    The ENDF/B-VI Revision 3 files have been released through the International Atomic Energy Agency. The data for hydrogen, aluminium and uranium-235 were processed to prepare an updated WIMS-D library. Thermal benchmark lattices TRX, BAPL and DIMPLE were analyzed. The new data for the thermal scattering laws of hydrogen bound in water had no significant influence on the integral parameters. The effect of the new uranium-235 data was to reduce the lattice multiplication factor by up to 0.3% Δ k/k. The effect of the new aluminium data was also non-negligible. It was traced to the change in the interpolation law for the total and the capture cross sections, which seems incorrect. (author). 8 refs, 1 fig., 2 tabs

  1. Calculation of control rods in rectangular reactor, and applications (1960)

    International Nuclear Information System (INIS)

    Goshen, S.; Pazy, A.

    1960-01-01

    The aim of this report is to find a method for estimating the anti-reactivity of control rods perpendicular to the axis in a cylindrical pile. The paper is divided into two parts. In the first is given a method of calculating control rods in a rectangular pile, similar to the Nordheim-Scalettar method for cylindrical piles. As an example the formulas are given for the theories of one and two neutron groups, the generalisation for several groups being evident. In the second part we find by a variation method a formula for estimating the Laplacian of a pile, which may be divided into parallelepipeds for which the Laplacian are given. Finally, this formula is used to calculate the anti-reactivity of rods perpendicular to the axis in a cylindrical pile. (author) [fr

  2. Preliminary topical report on comparison reactor disassembly calculations

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1975-11-01

    Preliminary results of comparison disassembly calculations for a representative LMFBR model (2100-l voided core) and arbitrary accident conditions are described. The analytical methods employed were the computer programs: FX2-POOL, PAD, and VENUS-II. The calculated fission energy depositions are in good agreement, as are measures of the destructive potential of the excursions, kinetic energy, and work. However, in some cases the resulting fuel temperatures are substantially divergent. Differences in the fission energy deposition appear to be attributable to residual inconsistencies in specifying the comparison cases. In contrast, temperature discrepancies probably stem from basic differences in the energy partition models inherent in the codes. Although explanations of the discrepancies are being pursued, the preliminary results indicate that all three computational methods provide a consistent, global characterization of the contrived disassembly accident

  3. Evaluation of WIMS-D/4 nuclear data library used on TRIGA reactor calculation

    International Nuclear Information System (INIS)

    Chen Wei; Xie Zhongsheng; Jiang Xinbiao; Chen Da

    1997-01-01

    The 69 groups constants of H in ZrH, 166 Er and 167 Er generated by NJOY and GASKET codes are inserted into WIMS nuclear data library WIMS-CNDC and WIMS-NINT libraries used on RTIGA reactor calculation are obtained. In order to check WIMS-CNDC and WIMS-NINT libraries, the scattering cross-section is compared with that in WIMS-IJS library. The group constant, K ∞ and temperature coefficient are calculated by using WIMS-CNDC, WIMS-NINT and WIMS-IJS. The results show the both libraries are suitable for calculation of TRIGA reactor

  4. Calculation of pressure drop and flow redistribution in the core of LMFBR type reactors

    International Nuclear Information System (INIS)

    Botelho, D.A.; Morgado, O.J.

    1985-01-01

    It is studied the flow redistribution through of fuel elements to the pressure drop calculation in the core of sodium cooled reactors. Using the quasi-static formulation of equations of the conservation of mass, energy and momentum, it was developed a computer program to flow redistribution calculations and pressure drop for different power levels and total flow simulating an arbitrary number of channels for sodium flowing . An optimization of the number of sufficient channels for calculations of this nature is done. The method is applied in studies of transients in the same reactor. (M.C.K.) [pt

  5. A calculation methodology applied for fuel management in PWR type reactors using first order perturbation theory

    International Nuclear Information System (INIS)

    Rossini, M.R.

    1992-01-01

    An attempt has been made to obtain a strategy coherent with the available instruments and that could be implemented with future developments. A calculation methodology was developed for fuel reload in PWR reactors, which evolves cell calculation with the HAMMER-TECHNION code and neutronics calculation with the CITATION code.The management strategy adopted consists of fuel element position changing at the beginning of each reactor cycle in order to decrease the radial peak factor. The bi-dimensional, two group First Order perturbation theory was used for the mathematical modeling. (L.C.J.A.)

  6. To the problem of reinforced concrete reactor vessel design and calculation

    International Nuclear Information System (INIS)

    Kirillov, A.P.; Artem'ev, V.P.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Paushkin, A.G.

    1980-01-01

    Modern methods for calculating reactor vessels of prestressed reinforced concrete are analyzed. It is shown that during the stage of technical and economical substantiation of reactor vessel structure for determining its stressed-deformed state engineering methods of calculation must be used, in particular, fragmentation method, method of rings and plates, and during the stages of contract and detail designs - method of finite elements and dynamic relaxation method. It is concluded that when solving cyclic symmetrical problems as well as asymmetrical problems, calculational algorithms for axis-symmetrical distributions of stresses in the vessel with provision for elastic properties of structural material may be used

  7. Calculation of fuel element temperature TRIGA 2000 reactor in sipping test tubes using CFD

    International Nuclear Information System (INIS)

    Sudjatmi KA

    2013-01-01

    It has been calculated the fuel element temperature in the sipping test of Bandung TRIGA 2000 reactor. The calculation needs to be done to ascertain that the fuel element temperatures are below or at the limit of the allowable temperature fuel elements during reactor operation. ensuring that the implementation of the test by using this device, the temperature is still within safety limits. The calculation is done by making a model sipping test tubes containing a fuel element surrounded by 9 fuel elements. according to the position sipping test tubes in the reactor core. by using Gambit. Dimensional model adapted to the dimensions of the tube and the fuel element in the reactor core of Bandung TRIGA 2000 reactor. Sipping test Operation for each fuel element performed for 30 minutes at 300 kW power. Calculations were performed using CFD software and as input adjusted parameters of TRIGA 2000 reactor. Simulations carried out on the operation of the 30, 60, 90, 120, 150, 180 and 210 minutes. The calculation result shows that the temperature of the fuel in tubes sipping test of 236.06 °C, while the temperature of the wall is 87.58 °C. The maximum temperature in the fuel center of TRIGA 2000 reactor in normal operation is 650 °C. and the boiling is not allowed in the reactor. So it can be concluded that the operation of the sipping test device are is very safe because the fuel center temperature is below the temperature limits the allowable fuel under normal operating conditions as well as the fuel element wall temperature is below the boiling temperature of water. (author)

  8. NEPTUNE: a modular scheme for the calculation of light water reactors

    International Nuclear Information System (INIS)

    Kavenoky, A.

    1975-01-01

    The NEPTUNE modular scheme has been developed to provide the physicist and the design engineer with a single system of codes for the calculation of light water reactors. The APOLLO code is included in NEPTUNE for the multigroup transport treatment of cells, groups of cells and complete fuel assemblies; few groups cross section libraries are automatically transmitted to the reactor multidimensional diffusion modules. In the reactor phase, 1D and 2D diffusion calculations can be performed by use of the finite difference method; 2D and 3D calculations are done respectively by the BILAN and TRIDENT modules using the finite element method. For the depletion calculation coarse and refined computations are offered. NEPTUNE is characterized by two special features for the data processing: the OTOMAT system which provides a virtual memory simulation and the intervention Monitor which allow to disconnect the computation modules and the control modules [fr

  9. MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

    International Nuclear Information System (INIS)

    Relloso, J.M.

    1990-01-01

    This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author) [es

  10. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities.

  11. Calculation of the mechanical equilibrium in a lattice of deformed hexagonal subassemblies

    International Nuclear Information System (INIS)

    Bernard, A.

    1979-01-01

    Stainless steel swelling and irradiation creep in the hexagonal wrappers of fast breeder cores induce deformations (mostly bowing), hence mutual interaction (displacements, forces and stresses, which must be calculated). The HARMONIE code was developed to meet these requirements. In this three dimensional code, one minimizes the elastic potential bending energy (quadratic form), with given linear conditions (no overlapping between adjacent subassemblies). The convergence of this function is obtained through a numerical method (parallel gradient). The free bowing of the subassemblies are given as input datas; the output gives the equilibrium displacements and forces while stresses are calculated in a classical manner

  12. Application of mesh free lattice Boltzmann method to the analysis of very high temperature reactor lower plenum

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Woon [Dongguk Univ., Gyeongju (Korea, Republic of). Dept. of Energy and Environment

    2011-11-15

    Inside a helium-cooled very high temperature reactor (VHTR) lower plenum, hot gas jets from upper fuel channels with very high velocities and temperatures and is mixed before flowing out. One of the major concerns is local hot spots in the plenum due to inefficient mixing of the helium exiting from differentially heated fuel channels and it involves complex fluid flow physics. For this situation, mesh-free technique, especially Lattice Boltzmann Method (LBM), is thus of particular interest owing to its merit of no mesh generation. As an attempt to find efficiency of the method in such a problem, 3 dimensional flow field inside a scaled test model of the VHTR lower plenum is computed with commercial XFLOW code. Large eddy simulation (LES) and classical Smagorinsky eddy viscosity (EV) turbulence models are employed to investigate the capability of the LBM in capturing large scale vortex shedding. (orig.)

  13. Calculation of the CAREM reactor with the HUEMUL-PUMA-THERMIT chain of codes; Calculo del reactor CAREM con la cadena de codigos HUEMUL-PUMA-THERMIT

    Energy Technology Data Exchange (ETDEWEB)

    Notari, Carla; Grant, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina)

    2000-07-01

    The purpose of the work was the evaluation of the the CAREM 25 reactor core, using a chain of codes (HUEMUL-PUMA-THERMIT) different to the one used in the original design (CONDOR-CITVAP-THERMIT). First, we performed a partial validation of the our codes in lattices similar to CAREM and reproduced a benchmark for simulation of gadolinium burnup. The results were considered satisfactory for this stage of the project. Then, we calculated the core along the normal operating equilibrium cycle and in hot and cold shut-down conditions. The main outcome of our evaluation confirms the general behaviour of the reference calculations except in one important point referring to the cold shut down. In this condition, the failure of one single rod of bank number 13 of the shut down system, leaves the core in a supercritical state at the beginning of the cycle and this anomaly persists during almost a third of the overall cycle. A new design of the core is proposed with minor modifications of the reference one, without introducing new types of fuel elements and keeping the same fuel management scheme. This new core fulfills all the design requirements. (author)

  14. CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

    International Nuclear Information System (INIS)

    Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu

    2002-10-01

    In order to improve plutonium utilization, design studies of reduced moderation water reactors which have hard neutron energy spectrum have been carried out at Division of Energy System Research of Japan Atomic Energy Research Institute (JAERI). At present, triangle, tight pitch lattice cores with about 1 mm gap width between fuel rods have been focused in the neutronic core design. Since a degradation of the heat removal from the fuel rods is worried, an evaluation of heat removal capability i.e. critical heat flux becomes one of important evaluation items in the feasibility study. However, any of published data base, which can be applicable to the evaluation on such narrow gap width cores, does not exist. Therefore, in the present study, in order to accumulate applicable data and to confirm applicability of an evaluation methodology of critical heat flux, basic experiments on the critical heat flux were performed using the test sections consisted of 7 heater rods bundles with the gap widths of 1.5, 1.0 and 0.6 mm under the PWR pressure conditions. The present report describes the experimental apparatus, experimental conditions and accumulated data. Analysis results of the data and the applicability of the evaluation methodology used for the design work are also discussed in this report. As the results of the experiment, it was found that the critical heat flux increased as the mass flux and the inlet subcooling increased. In the region of the mass flux less than about 2,000 kg/m 2 /s, the critical heat flux decreased as the gap width decreased. In the larger mass flux region, obvious trend of effects of the gap width on critical heat flux were not observed due to data scatterings. The flow-area-averaged thermal-equilibrium quality at the CHF position was in the higher ranges from 0.3 to 0.8 in the cases of gap widths of 1.0 and 0.6 mm, and 0.1 to 0.3 in the 1.5 mm case. Based on the experimental results such that the CHFs occurred in the higher quality range and

  15. Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

    International Nuclear Information System (INIS)

    Khedr, H.

    2008-01-01

    The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

  16. International comparison calculations for a BWR lattice with adjacent gadolinium pins

    International Nuclear Information System (INIS)

    Maeder, C.; Wydler, P.

    1984-09-01

    The results of burnup calculations for a simplified BWR fuel element with two adjacent gadolinium rods are presented and discussed. Ten complete solutions were contributed by Denmark, France, Italy (3), Japan (3), Switzerland and the UK. Partial results obtained from Poland and the USA are included in an Appendix. (Auth.)

  17. A lattice calculation of the hadronic vacuum polarization contribution to (g-2)μ

    DEFF Research Database (Denmark)

    Della Morte, Michele; Francis, Anthony; Gerardin, A.

    2018-01-01

    We present results of calculations of the hadronic vacuum polarisation contribution to the muon anomalous magnetic moment. Specifically, we focus on controlling the infrared regime of the vacuum polarisation function. Our results are corrected for finite-size effects by combining the Gounaris-Sak...

  18. CCSD(T)/CBS fragment-based calculations of lattice energy of molecular crystals

    Czech Academy of Sciences Publication Activity Database

    Červinka, C.; Fulem, Michal; Růžička, K.

    2016-01-01

    Roč. 144, č. 6 (2016), 1-15, č. článku 064505. ISSN 0021-9606 Institutional support: RVO:68378271 Keywords : density-functional theory * organic oxygen compounds * quantum -mechanical calculations Subject RIV: BJ - Thermodynamics Impact factor: 2.965, year: 2016

  19. Influence of FRAPCON-1 evaluation models on fuel behavior calculations for commercial power reactors

    International Nuclear Information System (INIS)

    Chambers, R.; Laats, E.T.

    1981-01-01

    A preliminary set of nine evaluation models (EMs) was added to the FRAPCON-1 computer code, which is used to calculate fuel rod behavior in a nuclear reactor during steady-state operation. The intent was to provide an audit code to be used in the United States Nuclear Regulatory Commission (NRC) licensing activities when calculations of conservative fuel rod temperatures are required. The EMs place conservatisms on the calculation of rod temperature by modifying the calculation of rod power history, fuel and cladding behavior models, and materials properties correlations. Three of the nine EMs provide either input or model specifications, or set the reference temperature for stored energy calculations. The remaining six EMs were intended to add thermal conservatism through model changes. To determine the relative influence of these six EMs upon fuel behavior calculations for commercial power reactors, a sensitivity study was conducted. That study is the subject of this paper

  20. Core calculations for the upgrading of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

    1998-01-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  1. NERON-Computing system for PHWR reactor cells and heterogeneous parameter calculations

    International Nuclear Information System (INIS)

    Cristian, I.; Cirstoiu, B.; Slavnicu, S.D.

    1976-04-01

    A system of codes for PHWR type reactors is presented. The system includes the cell code NERO and a code PARETE for monopolar and dipolar heterogeneous calculations. A general theory of dipolar flux is necessary for a more accurate evaluation of void coefficient and diffusion moderator coefficient is given. The determination of monopolar and dipolar heterogeneous parameters is very useful for heterogeneous methods developped especially for HWR reactors during the last years. (author)

  2. Calculational prediction of fuel burn-up for the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Phuoc Lan; Do Quang Binh

    2016-01-01

    In this paper, the method of expanding operators and functions in the neutron diffusion equations as chains of time variable is used for calculation of fuel burn-up of the Dalat nuclear reactors. A computer code, named BURREF, programmed in language Fortran-77 running on IBM PC-AT, has been developed based on this method to predict the fuel burn-up of the Dalat reactor. Some results will be presented here. (author)

  3. Non-iterative method to calculate the periodical distribution of temperature in reactors with thermal regeneration

    International Nuclear Information System (INIS)

    Sanchez de Alsina, O.L.; Scaricabarozzi, R.A.

    1982-01-01

    A matrix non-iterative method to calculate the periodical distribution in reactors with thermal regeneration is presented. In case of exothermic reaction, a source term will be included. A computer code was developed to calculate the final temperature distribution in solids and in the outlet temperatures of the gases. The results obtained from ethane oxidation calculation in air, using the Dietrich kinetic data are presented. This method is more advantageous than iterative methods. (E.G.) [pt

  4. Temperature and void reactivity coefficient calculations for the high flux isotope reactor safety analysis report

    International Nuclear Information System (INIS)

    Engle, W.W. Jr.; Williams, L.R.

    1994-07-01

    This report provides documentation of a series of calculations performed in 1991 in order to provide input for the High Flux Isotope Reactor Safety Analysis Report. In particular, temperature and void reactivity coefficients were calculated for beginning-of-life, end-of-life, and xenon equilibrium (29 h) conditions. Much of the data used to prepare the computer models for these calculations was derived from the original HFIR nuclear design study

  5. A contribution to the method of fast reactor thermal output calculation

    International Nuclear Information System (INIS)

    Harant, M.

    1978-01-01

    The method of stating the heat sources is discussed as being one of the factors influencing the accuracy of the thermal output calculation of fast reactors. The distribution of heat sources in the core and in other inner parts of the fast reactor is described using the least square fit method. Relations are derived of outputs of both individual components of fuel elements and of whole inner parts of the reactor. A comparison is made of various methods used for obtaining source integrals. The optimum integration method was found. (author)

  6. Calculation and experimental measurements in the Argonauta reactor subcritical and exponential facility

    International Nuclear Information System (INIS)

    Voi, Dante L.; Furieri, Rosane C.A.A.; Renke, Carlos A.C.; Bastos, Wilma S.; Ferreira, Francisco J.O.

    1997-01-01

    Initial measurements were performed on the exponential and subcritical facility installed on the internal thermal column of the Argonauta reactor at IEN-CNEN-Rio de Janeiro, Brazil. The measurements are include in the reactor physics experimental program for integral parameters determination, for both valid and confirmed theoretical models for reactor calculation. Gamma doses and neutron fluxes were measured with telescopic, proportional counters, wire and foil detectors. Experimental data were compared with results obtained by application of CITATION code. (author). 4 refs., 8 figs

  7. Application of the REMIX thermal mixing calculation program for the Loviisa reactor

    International Nuclear Information System (INIS)

    Kokkonen, I.; Tuomisto, H.

    1987-08-01

    The REMIX computer program has been validated to be used in the pressurized thermal shock study of the Loviisa reactor pressure vessel. The program has been verified against the data from the thermal and fluid mixing experiments. These experiments have been carried out in Imatran voima Oy to study thermal mixing of the high-pressure safety injection water in the Loviisa VVER-440 type pressurized water reactor. The verified REMIX-versions were applied to reactor calculations in the probabilistic pressurized thermal shock study of the Loviisa Plant

  8. Methods in nuclear reactors calculations; Metodos de calculo en reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, G

    1966-07-01

    Studies are made of the neutron transport equation corresponding to the the real and virtual reactors, as well as the starting hypotheses. Methods are developed to solve the transport equation in slab geometry, and P{sub l}; B{sub l}; M{sub l}; S{sub n} and discrete ordinates approximations. (Author)

  9. Phase diagram of the Blume-Emery-Griffiths model on the simple cubic lattice calculated by the linear chain approximation

    International Nuclear Information System (INIS)

    Albayrak, Erhan; Keskin, Mustafa

    2000-01-01

    The linear chain approximation is used to study the temperature dependence of the order parameters and the phase diagrams of the Blume-Emery-Griffiths model on the simple cubic lattice with dipole-dipole, quadrupole-quadrupole coupling strengths and a crystal-field interaction. The problem is approached introducing first a trial one-dimensional Hamiltonian whose free energy can be calculated exactly by the transfer matrix method. Then using the Bogoliubov variational principle, the free energy of the model is determined. It is assumed that the dipolar and quadrupolar intrachain coupling constants are much stronger than the corresponding interchain constants and confined the attention to the case of nearest-neighbor interactions. The phase transitions are examined and the phase diagrams are obtained for several values of the coupling strengths in the three different planes. A comparison with other approximate techniques is also made

  10. Phase diagram of the Blume-Emery-Griffiths model on the simple cubic lattice calculated by the linear chain approximation

    CERN Document Server

    Albayrak, E

    2000-01-01

    The linear chain approximation is used to study the temperature dependence of the order parameters and the phase diagrams of the Blume-Emery-Griffiths model on the simple cubic lattice with dipole-dipole, quadrupole-quadrupole coupling strengths and a crystal-field interaction. The problem is approached introducing first a trial one-dimensional Hamiltonian whose free energy can be calculated exactly by the transfer matrix method. Then using the Bogoliubov variational principle, the free energy of the model is determined. It is assumed that the dipolar and quadrupolar intrachain coupling constants are much stronger than the corresponding interchain constants and confined the attention to the case of nearest-neighbor interactions. The phase transitions are examined and the phase diagrams are obtained for several values of the coupling strengths in the three different planes. A comparison with other approximate techniques is also made.

  11. Quantification of TRISO fuel heterogeneity effects in HTGR lattice physics calculations

    International Nuclear Information System (INIS)

    Perfetti, C. M.; Anghaie, S.; Dugan, E.; Marcille, T.

    2010-01-01

    A large number of LEU-MHR fuel compact models were generated with randomly distributed TRISO particle fuel and were simulated using MCNP5, and it was determined how several neutronic parameters, including k-infinite, the thermal and fast diffusion coefficients, and the four factors, varied across the randomly-generated cases. A sensitivity study was also performed to determine how the four factors depend on the definition of the thermal energy group. Values of k-infinite for the cases had a sample standard deviation of 248 pcm and were found to follow an approximately normal distribution about the mean value of k-infinite. Although all of the four factors were found to have similar sample standard deviations, the resonance escape probability was found to be the most variable parameter with a sample relative standard deviation between 0.07% and 0.08%. HTGR fuel compact homogenization methods typically examine only one reference fuel compact that contains a uniform distribution of TRISO particles, but in reality the TRISO particles are randomly distributed throughout the fuel compact. Thus, the neutronic parameters for actual fuel compacts differ randomly from those in the reference model. To license next-generation High-Temperature Gas Reactors engineers must quantify all uncertainties of the design and this random variation in neutron parameters is a previously unmeasured quantity; this study measures this uncertainty by examining the variation in k-infinite for HTGR fuel compact models with randomly distributed TRISO fuel. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-05-15

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  14. Ab initio calculations of ideal strength and lattice instability in W-Ta and W-Re alloys

    Science.gov (United States)

    Yang, Chaoming; Qi, Liang

    2018-01-01

    An important theoretical criterion to evaluate the ductility of metals with a body-centered cubic (bcc) lattice is the mechanical failure mode of their perfect crystals under tension along ; directions. When the tensile stress reaches the ideal tensile strength, the pure W crystal fails by a cleavage fracture along the {100 } plane so that it is intrinsically brittle. To discover the strategy to improve its ductility, we performed density functional theory and density functional perturbation theory calculations to study the ideal tensile strength and the lattice instability under tension for both W-Ta and W-Re alloys. Anisotropic linear elastic fracture mechanics (LEFM) theory and Rice's criterion were also applied to analyze the mechanical instability at the crack tip under tension based on the competition between cleavage propagation and dislocation emission. The results show that the intrinsic ductility can be achieved in both W-Ta and W-Re, however, by different mechanisms. Even though W-Ta alloys with low Ta concentrations are still intrinsically brittle, the intrinsic ductility of W-Ta alloys with high Ta concentrations is promoted by elastic shear instability before the cleavage failure. The intrinsic ductility of W-Re alloys is produced by unstable transverse phonon waves before the cleavage failure, and the corresponding phonon mode is related to the generation of 1/2 {2 ¯11 } dislocation in bcc crystals. The ideal tensile calculations, phonon analyses, and anisotropic LEFM examinations are mutually consistent in the evaluation of intrinsic ductility. These results bring us physical insights on the ductility-brittle mechanisms of W alloys under extreme stress conditions.

  15. Calculation of the anti-trap factor in heavy water lattices; Calcul du facteur antitrappe dans les reseaux a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Naudet, R; Mougey, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    The calculation of the anti-trap factor of a lattice is complex when a large fraction of captures occurs in a range of energies where the spectrum in the fuel is considerably different from the simple dE/E law. This is particularly true for heavy water lattices in which the distances. between the bars are generally fairly large with respect to the slowing-down length. In order to take into account this effect it is necessary both to know the constitution of the effective resonance integral as a function of the energy, and to be able to calculate the distribution in the fuel. This report is devoted to these two problems. An improved method of treating the statistical domain makes it possible to plot the curves of the cross-sections per unit lethargy for various shapes of the fuel. Furthermore, the slowing-down of the neutrons is studied using a Monte-Carlo method which makes it possible in particular to take into account the perturbations caused by the non-moderating rods. A study is also made of the problem of shielding effects due to the captures themselves. (authors) [French] Le calcul du facteur antitrappe dans un reseau est complique lorsqu'une fraction importante des captures a lieu dans un domaine d'energie ou le spectre dans le combustible s'ecarte sensiblement de la loi simple en dE/E. Ceci est particulierement vrai pour les reseaux a eau lourde dans lesquels les distances entre barres sont en general assez grandes vis-a-vis de la longueur de ralentissement. Pour tenir compte de cet effet il faut connaitre d'une part la decomposition de l'integrale de resonance effective en fonction de l'energie, d'autre part savoir calculer le spectre dans le combustible. Le rapport est consacre a ces deux problemes. Un traitement ameliore du domaine statistique permet de tracer des courbes de sections de capture par unite de lethargie pour differentes geometries de barreaux. D'autre part le ralentissement des neutrons est etudie par une methode de Monte Carlo, qui permet

  16. Calculation Of Fuel Burnup And Radionuclide Inventory In The Syrian Miniature Neutron Source Reactor Using The GETERA Code

    International Nuclear Information System (INIS)

    Khattab, K.; Dawahra, S.

    2011-01-01

    Calculations of the fuel burnup and radionuclide inventory in the Syrian Miniature Neutron Source Reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer and has a bigger library of isotopes and more accurate. (author)

  17. Absorbing device for stationary arrangement in the lattice of a boiling water reactor

    International Nuclear Information System (INIS)

    Fredin, B.; Nylund, O.

    1980-01-01

    The invention refers to an absorbing device for stationary arrangement in the lattice of a BWR in a gap between two bundles of vertical fuel rods. It consists of at least one absorbing plate containing burnable absorbing material. Both lateral surfaces of this plate are directed to one surface each of the bundles mentioned above. According to the invention the absorbing material is contained in channels formed by welding together two adjacent sheet elements, at least one of which being corrugated. The welds will be made at the points where to tops of the waves touch the other sheet element. (orig.) [de

  18. Tables of formulae for calculating the mechanics of stacks in gas-graphite reactors

    International Nuclear Information System (INIS)

    1968-01-01

    This collection of formulae only gives, for nuclear graphite stacks. The mechanical effects due to the strains, thermal or not, of steel structures supporting or surrounding graphite blocks. Equations have been established by mean of experiments made at Chinon with large pile models. Thus, it is possible to calculate displacement, strain and stress in the EDF type stacks of horizontal triangular block lattice. (authors) [fr

  19. Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4

    International Nuclear Information System (INIS)

    Tombakoglu, M.; Cecen, Y.

    2001-01-01

    In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)

  20. An assessment of fission product data for decay power calculation in fast reactors

    International Nuclear Information System (INIS)

    Sridharan, M.S.; Murthy, K.P.N.

    1987-01-01

    A review of our present capability at IGC, Kalpakkam to predict fission product decay power in fast reactors is presented. This is accomplished by comparing our summation calculations with the calculations of others and the reported experimental measurements. Our calculations are based on Chandy code developed at our Centre. The fission product data base of Chandy is essentially drawn from the yield data compiled by Crouch (1977) and the data on halflives etc. compiled by Tobias (1973). In general, we find good agreement amongst the different calculations (within ±5%) and our calculations also compare well with experimental measurements of AKIAMA et al and MURPHY et al