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Sample records for reactor beam calculations

  1. 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

  2. Reactor beam calculations to determine optimum delivery of epithermal neutrons for treatment of brain tumors

    International Nuclear Information System (INIS)

    Wheeler, F.J.; Nigg, D.W.; Capala, J.

    1997-01-01

    Studies were performed to assess theoretical tumor control probability (TCP) for brain-tumor treatment with boron neutron capture therapy (BNCT) using epithermal neutron sources from reactors. The existing epithermal-neutron beams at the Brookhaven Medical Research Reactor Facility (BMRR), the Petten High Flux Reactor Facility (HWR) and the Finnish Research Reactor 1 (FIR1) have been analyzed and characterized using common analytical and measurement methods allowing for this inter-comparison. Each of these three facilities is unique and each offers an advantage in some aspect of BNCT, but none of these existing facilities excel in all neutron-beam attributes as related to BNCT. A comparison is therefore also shown for a near-optimum reactor beam which does not currently exist but which would be feasible with existing technology. This hypothetical beam is designated BNCT-1 and has a spectrum similar to the FIR-1, the mono-directionality of the HFR and the intensity of the BMRR. A beam very similar to the BNCT-1 could perhaps be achieved with modification of the BMRR, HFR, or FIR, and could certainly be realized in a new facility with today's technology

  3. 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

  4. 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.)

  5. BR2 reactor neutron beams

    International Nuclear Information System (INIS)

    Neve de Mevergnies, M.

    1977-01-01

    The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

  6. 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.

  7. Electron beam solenoid reactor concept

    International Nuclear Information System (INIS)

    Bailey, V.; Benford, J.; Cooper, R.; Dakin, D.; Ecker, B.; Lopez, O.; Putman, S.; Young, T.S.T.

    1977-01-01

    The electron Beam Heated Solenoid (EBHS) reactor is a linear magnetically confined fusion device in which the bulk or all of the heating is provided by a relativistic electron beam (REB). The high efficiency and established technology of the REB generator and the ability to vary the coupling length make this heating technique compatible with several radial and axial enery loss reduction options including multiple-mirrors, electrostatic and gas end-plug techniques. This paper addresses several of the fundamental technical issues and provides a current evaluation of the concept. The enhanced confinement of the high energy plasma ions due to nonadiabatic scattering in the multiple mirror geometry indicates the possibility of reactors of the 150 to 300 meter length operating at temperatures > 10 keV. A 275 meter EBHS reactor with a plasma Q of 11.3 requiring 33 MJ of beam eneergy is presented

  8. 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.)

  9. A recycling molecular beam reactor

    International Nuclear Information System (INIS)

    Prada-Silva, G.; Haller, G.L.; Fenn, J.B.

    1974-01-01

    In a Recycling Molecular Beam Reactor, RMBR, a beam of reactant gas molecules is formed from a supersonic free jet. After collision with a target the molecules pass through the vacuum pumps and are returned to the nozzle source. Continuous recycling permits the integration of very small reaction probabilities into measurable conversions which can be analyzed by gas chromatography. Some preliminary experiments have been carried out on the isomerization of cyclopropane

  10. 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

  11. 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)

  12. 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

  13. 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)

  14. 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.)

  15. 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.

  16. 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)

  17. MAPLE research reactor beam-tube performance

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Gillespie, G.E.

    1989-05-01

    Atomic Energy of Canada Limited (AECL) has been developing the MAPLE (Multipurpose Applied Physics Lattice Experimental) reactor concept as a medium-flux neutron source to meet contemporary research reactor applications. This paper gives a brief description of the MAPLE reactor and presents some results of computer simulations used to analyze the neutronic performance. The computer simulations were performed to identify how the MAPLE reactor may be adapted to beam-tube applications such as neutron radiography

  18. 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)

  19. 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

  20. 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

  1. 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)

  2. 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)

  3. 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

  4. 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.).

  5. 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.)

  6. E-beam heated linear solenoid reactors

    International Nuclear Information System (INIS)

    Benford, J.; Bailey, V.; Oliver, D.

    1976-01-01

    A conceptual design and system analysis shows that electron beam heated linear solenoidal reactors are attractive for near term applications which can use low gain fusion sources. Complete plant designs have been generated for fusion based breeders of fissile fuel over a wide range of component parameters (e.g., magnetic fields, reactor lengths, plasma densities) and design options (e.g., various radial and axial loss mechanisms). It appears possible that a reactor of 100 to 300 meters length operating at power levels of 1000 MWt can economically produce 2000 to 8000 kg/yr of 233 U to supply light water reactor fuel needs beyond 2000 A.D. Pure fusion reactors of 300 to 500 meter lengths are possible. Physics and operational features of reactors are described. Beam heating by classical and anomalous energy deposition is reviewed. The technology of the required beams has been developed to MJ/pulse levels, within a factor of 20 of that needed for full scale production reactors. The required repetitive pulsing appears practical

  7. 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

  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. Beam dynamics calculations for the linac booster beam line

    International Nuclear Information System (INIS)

    Lu, J.Q.; Cramer, J.G.; Storm, D.W.

    1987-01-01

    Beam optics focusing characteristics both in the transverse and longitudinal directions of the superconducting linac booster beam line are calculated for different particles. Three computer programs, which are TRANSPORT, LYRA and ENTIME, are used to simulate particle motions. The first one is used to simulate the particle radial motions. The effects of energy increase on to the transverse phase space area are considered by putting in accelerating matrices of each resonators. The second program is used to simulate particle longitudinal motions. Beam longitudinal motions are calculated with program ENTIME also, with which visual pictures in the Energy-Time phase space can be displayed on the terminal screen. Besides, the stability of the particle periodic motions in the radial directions are considered and calculated

  12. 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.)

  13. Focal spot size predictions for beam transport through a gas-filled reactor

    International Nuclear Information System (INIS)

    Yu, S.S.; Lee, E.P.; Buchanan, H.L.

    1980-01-01

    Results from calculations of focal spot size for beam transport through a gas-filled reactor are summarized. In the converging beam mode, we find an enlargement of the focal spot due to multiple scattering and zeroth order self-field effects. This enlargement can be minimized by maintaining small reactors together with a careful choice of the gaseous medium. The self-focused mode, on the other hand, is relatively insensitive to the reactor environment, but is critically dependent upon initial beam quality. This requirement on beam quality can be significantly eased by the injection of an electron beam of modest current from the opposite wall

  14. 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)

  15. 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

  16. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    International Nuclear Information System (INIS)

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

  17. 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)

  18. 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

  19. 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)

  20. Shielding calculations for the TFTR neutral beam injectors

    International Nuclear Information System (INIS)

    Santoro, R.T.; Lillie, R.A.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1979-07-01

    Two-dimensional discrete ordinates calculations have been performed to determine the location and thickness of concrete shielding around the Tokamak Fusion Test Reactor (TFTR) neutral beam injectors. Two sets of calculations were performed: one to determine the dose equivalent rate on the roof and walls of the test cell building when no injectors are present, and one to determine the contribution to the dose equivalent rate at these locations from radiation streaming through the injection duct. Shielding the side and rear of the neutral beam injector with 0.305 and 0.61 m of concrete, respectively, and lining the inside of the test cell wall with an additional layer of concrete having a thickness of 0.305 m and a height above the axis of deuteron injection of 3.10 m are sufficient to maintain the biological dose equivalent rate outside the test cell to approx. 1 mrem/DT pulse

  1. CALCULATION ALGORITHM TRUSS UNDER CRANE BEAMS

    Directory of Open Access Journals (Sweden)

    N. K. Akaev1

    2016-01-01

    Full Text Available Aim.The task of reducing the deflection and increase the rigidity of single-span beams are made. In the article the calculation algorithm for truss crane girders is determined.Methods. To identify the internal effort required for the selection of cross section elements the design uses the Green's function.Results. It was found that the simplest truss system reduces deflection and increases the strength of design. The upper crossbar is subjected not only to bending and shear and compression work due to tightening tension. Preliminary determination of the geometrical characteristics of the crane farms elements are offered to make a comparison with previous similar configuration of his farms, using a simple approximate calculation methods.Conclusion.The method of sequential movements (incrementally the two bridge cranes along the length of the upper crossbar truss beams is suggested. We give the corresponding formulas and conditions of safety.

  2. 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

  3. Beam propagation through a gaseous reactor: classical transport

    International Nuclear Information System (INIS)

    Yu, S.S.; Buchanan, H.L.; Lee, E.P.; Chambers, F.W.

    1979-01-01

    The present calculations are applicable to any beam geometry with cylindrical symmetry, including the converging beam geometry (large entrance port with radius > or approx. = 10 cm), as well as the pencil-shaped beam (small porthole with radius approx. mm). The small porthole is clearly advantageous from the reactor vessel design point of view. While the physics of the latter mode of propagation may be more complex, analyses up to this point have not revealed any detrimental instability effects that will inhibit propagation. In fact, the large perpendicular velocity v/sub perpendicular/ that the pinched mode can accommodate provides a mechanism for the quenching of filamentary instability. Furthermore, this mode of propagation can withstand more ion scattering and is not subject to the upper bound on pressure (p < 10 torr) which is imposed on the converging beam mode

  4. 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)

  5. 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

  6. 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

  7. 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

  8. HFBR handbook, 1992: High flux beam reactor

    International Nuclear Information System (INIS)

    Axe, J.D.; Greenberg, R.

    1992-10-01

    Welcome to the High Flux Beam Reactor (HFBR), one of the world premier neutron research facilities. This manual is intended primarily to acquaint outside users (and new Brookhaven staff members) with (almost) everything they need to know to work at the HFBR and to help make the stay at Brookhaven pleasant as well as profitable. Safety Training Programs to comply with US Department of Energy (DOE) mandates are in progress at BNL. There are several safety training requirements which must be met before users can obtain unescorted access to the HFBR. The Reactor Division has prepared specific safety training manuals which are to be sent to experimenters well in advance of their expected arrival at BNL to conduct experiments. Please familiarize yourself with this material and carefully pay strict attention to all the safety and security procedures that are in force at the HFBR. Not only your safety, but the continued operation of the facility, depends upon compliance

  9. Influence of core model parameters on the characteristics of neutron beams of the research reactor

    Directory of Open Access Journals (Sweden)

    N. A. Khafizova

    2013-12-01

    Full Text Available IRT MEPhI reactor is equipped with a number of facilities at horizontal experimental channels (HEC. Knowing of parameters influencing spatio-angular distribution of irradiation fields is essential for each application area. The research for neutron capture therapy (NCT facility at HEC of the reactor was made. Calculation methods have been used to estimate how the reactor core parameters influence neutron beam characteristics at the HEC output. The impact of neutron source model in Monte Carlo calculations by MCNP code on the parameters of neutron and secondary photon field at the output of irradiation beam tubes of research reactor is estimated. The study shows that specifying neutron source with fission reaction rate distribution in SDEF option gives almost the same results as criticality calculation considered the most accurate. Our calculations show that changes of the core operational parameters have insignificant influence on characteristics of neutron beams at HEC output.

  10. 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

  11. Reactor - and accelerator-based filtered beams

    International Nuclear Information System (INIS)

    Mill, A.J.; Harvey, J.R.

    1980-01-01

    The neutrons produced in high flux nuclear reactors and in accelerator, induced fission and spallation reactions, represent the most intense sources of neutrons available for research. However, the neutrons from these sources are not monoenergetic, covering the broad range extending from 10 -3 eV up to 10 7 eV or so. In order to make quantitative measurements of the effects of neutrons and their dependence on neutron energy it is desirable to have mono-energetic neutron sources. The paper describes briefly methods of obtaining mono-energetic neutrons and different methods of filtration. This is followed by more detailed discussion of neutron window filters and a summary of the filtered beam facilities using this technique. The review concludes with a discussion of the main applications of filtered beams and their present and future importance

  12. 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

  13. 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)

  14. 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

  15. 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

  16. A high intensity positron beam at the Brookhaven reactor

    International Nuclear Information System (INIS)

    Weber, M.; Lynn, K.G.; Roellig, L.O.; Mills, A.P. Jr.; Moodenbaugh, A.R.

    1987-01-01

    We describe a high intensity, low energy positron beam utilizing high specific activity /sup 64/Cu sources (870 Ci/g) produced in a reactor with high thermal neutron flux. Fast-to-slow moderation can be performed in a self moderation mode or with a transmission moderator. Slow positron rates up to 1.6 x 10/sup 8/ e/sup +//s with a half life of 12.8 h are calculated. Up to 1.0 x 10/sup 8/ e/sup +//s have been observed. New developments including a Ne moderator and an on-line isotope separation process are discussed. 21 refs., 9 figs

  17. Colliding beam fusion reactor space propulsion system

    International Nuclear Information System (INIS)

    Wessel, Frank J.; Binderbauer, Michl W.; Rostoker, Norman; Rahman, Hafiz Ur; O'Toole, Joseph

    2000-01-01

    We describe a space propulsion system based on the Colliding Beam Fusion Reactor (CBFR). The CBFR is a high-beta, field-reversed, magnetic configuration with ion energies in the range of hundreds of keV. Repetitively-pulsed ion beams sustain the plasma distribution and provide current drive. The confinement physics is based on the Vlasov-Maxwell equation, including a Fokker Planck collision operator and all sources and sinks for energy and particle flow. The mean azimuthal velocities and temperatures of the fuel ion species are equal and the plasma current is unneutralized by the electrons. The resulting distribution functions are thermal in a moving frame of reference. The ion gyro-orbit radius is comparable to the dimensions of the confinement system, hence classical transport of the particles and energy is expected and the device is scaleable. We have analyzed the design over a range of 10 6 -10 9 Watts of output power (0.15-150 Newtons thrust) with a specific impulse of, I sp ∼10 6 sec. A 50 MW propulsion system might involve the following parameters: 4-meters diameterx10-meters length, magnetic field ∼7 Tesla, ion beam current ∼10 A, and fuels of either D-He 3 ,P-B 11 ,P-Li 6 ,D-Li 6 , etc

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. 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

  3. 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

  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. 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

  6. 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

  7. Assessment of beam tube performance for the maple research reactor

    International Nuclear Information System (INIS)

    Lee, A.G.

    1986-06-01

    The MAPLE research reactor is a versatile new research facility that can be adapted to meet the requirements of a variety of reactor applications. A particular group of reactor applications involves the use of beams of radiation extracted from the reactor core via tubes that penetrate through the biological shield and terminate in the reflector surrounding the fuelled core. An assessment is given of the neutron and gamma radiation fields entering beam tubes that are located radially or tangentially with respect to the core

  8. 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.

  9. Calculation of ballistic focusing of ion beams

    International Nuclear Information System (INIS)

    Astrelin, V.T.; Syresin, E.M.

    1984-01-01

    The motion of ions passing from the homogeneous magnetic field into a conical one is treated analytically in paraxial approximation. Further ions transform into neutral particles at the recharging target which is placed in the conical area of field. The optimal conditions for maximum compression of the beams of neutral particles are investigated. An influence of the initial angular spread on the beam compression is analysed. The computation results together with the those of analytical treatment are presented

  10. Radiation dosimetry at the BNL High Flux Beam Reactor

    International Nuclear Information System (INIS)

    Holden, N.E.; Hu, J.P.; Reciniello, R.N.

    1998-02-01

    The HFBR is a heavy water, D 2 O, cooled and moderated reactor with twenty-eight fuel elements containing a maximum of 9.8 kilograms of 235 U. The core is 53 cm high and 48 cm in diameter and has an active volume of 97 liters. The HFBR, which was designed to operate at forty mega-watts, 40 NW, was upgraded to operate at 60 NW. Since 1991, it has operated at 30 MW. In a normal 30 MW operating cycle the HFBR operates 24 hours a day for thirty days, with a six to fourteen day shutdown period for refueling and maintenance work. While most reactors attempts to minimize the escape of neutrons from the core, the HFBR's D 2 O design allows the thermal neutron flux to peak in the reflector region and maximizes the number of thermal neutrons available to nine horizontal external beams, H-1 to H-9. The HFBR neutron dosimetry effort described here compares measured and calculated energy dependent neutron and gamma ray flux densities and/or dose rates at horizontal beam lines and vertical irradiation thimbles

  11. 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

  12. Beam dynamics calculations for fault-tolerance

    International Nuclear Information System (INIS)

    Biarrotte, J.L.; Uriot, D.

    2007-10-01

    The European Transmutation Demonstration requires a high-power proton accelerator operating in CW mode. This accelerator is also expected to have a very limited number of unexpected beam interruptions per year. To reach such an ambitious goal, it is clear that reliability-oriented design practices need to be followed from the early stage of components design and fault-tolerance capabilities have to be introduced to the maximum extent. The goal of this document is precisely to investigate in more details the fault-tolerance capability of the XT-ADS linac. From previous analysis, it appears that if nothing is done, a cavity's failure leads in nearly all the cases to a complete beam loss, due to the non-relativistic varying velocity of the particles. To avoid such a total beam loss, it is clear that some kind of retuning has to be performed to compensate the lack of acceleration due to the faulty cavity. We have to identify and develop fast failure recovery scenarios to ensure that such retuning can be performed in less than 1 second. 2 ways are investigated. The first way is to stop the beam to achieve the retuning (Scenario 1). The other way is to try to perform the retuning without stopping the beam (Scenario 2). The present analysis demonstrates on the beam dynamics point of view that a fast retuning procedure can be envisaged without stopping the beam (Scenario 2). Nevertheless, this Scenario 2 implies stringent specifications, especially on: - the fault detection time, that has to be extremely short (order of magnitude: 100 μs) and - the margins required on the accelerating field and RF power point of view, that are higher than in Scenario 1

  13. 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)

  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. 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)

  16. Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

    International Nuclear Information System (INIS)

    Ratanatongchai, Wichian

    2009-07-01

    Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

  17. 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.)

  18. 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.)

  19. Filtered epithermal quasi-monoenergetic neutron beams at research reactor facilities.

    Science.gov (United States)

    Mansy, M S; Bashter, I I; El-Mesiry, M S; Habib, N; Adib, M

    2015-03-01

    Filtered neutron techniques were applied to produce quasi-monoenergetic neutron beams in the energy range of 1.5-133keV at research reactors. A simulation study was performed to characterize the filter components and transmitted beam lines. The filtered beams were characterized in terms of the optimal thickness of the main and additive components. The filtered neutron beams had high purity and intensity, with low contamination from the accompanying thermal emission, fast neutrons and γ-rays. A computer code named "QMNB" was developed in the "MATLAB" programming language to perform the required calculations. Copyright © 2014 Elsevier Ltd. All rights reserved.

  20. 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 ...

  1. 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)

  2. Tokamak Fusion Test Reactor neutral beam injection system vacuum chamber

    International Nuclear Information System (INIS)

    Pedrotti, L.R.

    1977-01-01

    Most of the components of the Neutral Beam Lines of the Tokamak Fusion Test Reactor (TFTR) will be enclosed in a 50 cubic meter box-shaped vacuum chamber. The chamber will have a number of unorthodox features to accomodate both neutral beam and TFTR requirements. The design constraints, and the resulting chamber design, are presented

  3. 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)

  4. 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

  5. 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)

  6. Neutral-beam systems for magnetic-fusion reactors

    International Nuclear Information System (INIS)

    Fink, J.H.

    1981-01-01

    Neutral beams for magnetic fusion reactors are at an early stage of development, and require considerable effort to make them into the large, reliable, and efficient systems needed for future power plants. To optimize their performance to establish specific goals for component development, systematic analysis of the beamlines is essential. Three ion source characteristics are discussed: arc-cathode life, gas efficiency, and beam divergence, and their significance in a high-energy neutral-beam system is evaluated

  7. Methods and codes for neutronic calculations of the MARIA research reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.; Bretscher, M.M.; Hanan, N.A.; Matos, J.E.

    1998-01-01

    The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6x8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminium. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization. (author)

  8. Neutron beam facilities at the Australian Replacement Research Reactor

    International Nuclear Information System (INIS)

    Kennedy, Shane; Robinson, Robert; Hunter, Brett

    2001-01-01

    Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10 14 n/cm 2 /sec and a liquid D 2 cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

  9. 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)

  10. 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

  11. Beam-gas background calculation for DEAR

    International Nuclear Information System (INIS)

    Guiducci, S.

    1997-01-01

    The present paper describes the results obtained from a Monte Carlo simulation of the particles lost in the DAY-ONE Interaction Region (IR) of the DAFNE machine. Coulomb scattering on the nuclei of residual gas and beam-gas Bremsstrahlung, which are the main sources of background in the initial phase of machine operation, were considered. A ray-tracking program (TURTLE) was used to follow the trajectories of the particles in the rings and to evaluate the number of particles that hit the vacuum chamber in the interaction region

  12. Direct energy conversion and neutral beam injection for catalyzed D and D-3He tokamak reactors

    International Nuclear Information System (INIS)

    Blum, A.S.; Moir, R.W.

    1977-01-01

    The calculated performance of single stage and Venetian blind direct energy converters for Catalyzed D and D- 3 He Tokamak reactors are discussed. Preliminary results on He pumping are outlined. The efficiency of D and T neutral beam injection is reviewed

  13. Focused proton beams propagating in reactor of fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Niu, K [Teikyo Heisei Univ., Uruido, Ichihara, Chiba (Japan)

    1997-12-31

    One of the difficult tasks of light ion beam fusion is to propagate the beam in the reactor cavity and to focus the beam on the target. The light ion beam has a certain local divergence angle because there are several causes for divergence at the diode. The electrostatic force induced at the leading edge causes beam divergence during propagation. To confine the beam within a small radius during propagation, the magnetic field must be employed. Here the electron beam is proposed to be launched simultaneously with the launching of the proton beam. If the electron beam has the excess current, the beam induces a magnetic field in the negative azimuthal direction, which confines the ion beam within a small radius by the electrostatic field as well as the electron beam by the Lorentz force. The metal guide around the beam path helps the beam confinement and reduces the total amount of magnetic field energy induced by the electron current. (author). 2 figs., 15 refs.

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. Neutron beam facilities at the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    Kim, S.

    2003-01-01

    The exciting development for Australia is the construction of a modern state-of-the-art 20-MW Replacement Research Reactor which is currently under construction to replace the aging reactor (HIFAR) at ANSTO in 2006. To cater for advanced scientific applications, the replacement reactor will provide not only thermal neutron beams but also a modern cold-neutron source moderated by liquid deuterium at approximately -250 deg C, complete with provision for installation of a hot-neutron source at a later stage. The latest 'supermirror' guides will be used to transport the neutrons to the Reactor Hall and its adjoining Neutron Guide Hall where a suite of neutron beam instruments will be installed. These new facilities will expand and enhance ANSTO's capabilities and performance in neutron beam science compared with what is possible with the existing HIFAR facilities, and will make ANSTO/Australia competitive with the best neutron facilities in the world. Eight 'leading-edge' neutron beam instruments are planned for the Replacement Research Reactor when it goes critical in 2006, followed by more instruments by 2010 and beyond. Up to 18 neutron beam instruments can be accommodated at the Replacement Research Reactor, however, it has the capacity for further expansion, including potential for a second Neutron Guide Hall. The first batch of eight instruments has been carefully selected in conjunction with a user group representing various scientific interests in Australia. A team of scientists, engineers, drafting officers and technicians has been assembled to carry out the Neutron Beam Instrument Project to successful completion. Today, most of the planned instruments have conceptual designs and are now being engineered in detail prior to construction and procurement. A suite of ancillary equipment will also be provided to enable scientific experiments at different temperatures, pressures and magnetic fields. This paper describes the Neutron Beam Instrument Project and gives

  19. 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)

  20. Numerical Calculation of the Phase Space Density for the Strong-Strong Beam-Beam Interaction

    International Nuclear Information System (INIS)

    Sobol, A.; Ellison, J.A.

    2003-01-01

    We developed a parallel code to calculate the evolution of the 4D phase space density of two colliding beams, which are coupled via the collective strong-strong beam-beam interaction, in the absence of diffusion and damping, using the Perron-Frobenius (PF) operator technique

  1. Materials research with neutron beams from a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Root, J.; Banks, D. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, Ontario (Canada)

    2015-03-15

    Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

  2. 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)

  3. Three-dimensional electron-beam dose calculations

    International Nuclear Information System (INIS)

    Shiu, A.S.

    1988-01-01

    The MDAH pencil-beam algorithm developed by Hogstrom et al (1981) has been widely used in clinics for electron-beam dose calculations for radiotherapy treatment planning. The primary objective of this research was to address several deficiencies of that algorithm and to develop an enhanced version. Two enhancements were incorporated into the pencil-beam algorithm; one models fluence rather than planar fluence, and the other models the bremsstrahlung dose using measured beam data. Comparisons of the resulting calculated dose distributions with measured dose distributions for several test phantoms have been made. From these results it is concluded (1) that the fluence-based algorithm is more accurate to use for the dose calculation in an inhomogeneous slab phantom, and (2) the fluence-based calculation provides only a limited improvement to the accuracy the calculated dose in the region just downstream of the lateral edge of an inhomogeneity. A pencil-beam redefinition model was developed for the calculation of electron-beam dose distributions in three dimensions

  4. The neutron beam facility at the Australian replacement research reactor

    International Nuclear Information System (INIS)

    Hunter, B.; Kennedy, S.

    1999-01-01

    Full text: The Australian federal government gave ANSTO final approval to build a research reactor to replace HIFAR on August 25th 1999. The replacement reactor is to be a multipurpose reactor with a thermal neutron flux of 3 x 10 14 n.cm -2 .s -1 and having improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The replacement reactor will commence operation in 2005 and will cater for Australian scientific, industrial and medical needs well into the 21st century. The scientific capabilities of the neutron beams at the replacement reactor are being developed in consultation with representatives from academia, industry and government research laboratories to provide a facility for condensed matter research in physics, chemistry, materials science, life sciences, engineering and earth sciences. Cold, thermal and hot neutron sources are to be installed, and neutron guides will be used to position most of the neutron beam instruments in a neutron guide hall outside the reactor confinement building. Eight instruments are planned for 2005, with a further three to be developed by 2010. A conceptual layout for the neutron beam facility is presented including the location of the planned suite of neutron beam instruments. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by an accredited reactor builder in a turnkey contract. Tenders have been called for December 1999, with selection of contractor planned by June 2000. The neutron beam instruments will be developed by ANSTO and other contracted organisations in consultation with the user community and interested overseas scientists. The facility will be based, as far as possible, around a neutron guide hall that is be served by three thermal and three cold neutron guides. Efficient transportation of thermal and cold neutrons to the guide hall requires the use of modern super

  5. 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

  6. 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)

  7. Calculation of beam neutralization in the IPNS-Upgrade RCS

    International Nuclear Information System (INIS)

    Chae, Yong-Chul.

    1995-01-01

    The author calculated the neutralization of circulating beam in this report. In the calculation it is assumed that all electrons liberated from the background molecules due to the collisional processes are trapped in the potential well of the proton beam. Including the dependence of ionization cross sections on the kinetic energy of the incident particle, the author derived the empirical formula for beam neutralization as a function of time and baseline vacuum pressure, which is applicable to the one acceleration cycle of the IPNS-Upgrade RCS

  8. Utilizations of filtered neutron beams at Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    Hien, P.D.; Chau, L.N.; Tan, V.H.; Hiep, N.T.; Phuong, L.B.

    1992-01-01

    Neutron beam utilizations in basic and applied researches have been important activities at the Dalat nuclear reactor. The neutron filters with single crystal of silicon are used to produce thermal neutrons at the tangential horizontal channel and quasi-monoenergetic 144 KeV and 54 KeV neutrons at the piercing beam tube. The paper presents some relevant characteristics of the filtered neutron beams at the two horizontal channels. Applications of neutron beams in prompt gamma-ray activation analysis and in nuclear data measurements are briefly described. (author)

  9. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    International Nuclear Information System (INIS)

    Park, Jong-beom; Park, No-Cheol; Lee, Sang-Jeong; Park, Young-Pil; Choi, Youngin

    2017-01-01

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  10. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: pnch@yonsei.ac.kr [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)

    2017-03-15

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  11. 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

  12. Calculated intensity of high-energy neutron beams

    International Nuclear Information System (INIS)

    Mustapha, B.; Nolen, J.A.; Back, B.B.

    2004-01-01

    The flux, energy and angular distributions of high-energy neutrons produced by in-flight spallation and fission of a 400 MeV/A 238 U beam and by the break-up of a 400 MeV/A deuteron beam are calculated. In both cases very intense secondary neutron beams are produced, peaking at zero degrees, with a relatively narrow energy spread. Such secondary neutron beams can be produced with the primary beams from the proposed rare isotope accelerator driver linac. The break-up of a 400 kW deuteron beam on a liquid-lithium target can produce a neutron flux of >10 10 neutrons/cm 2 /s at a distance of 10 m from the target

  13. 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

  14. Filtered epithermal quasi-monoenergetic neutron beams at research reactor facilities

    International Nuclear Information System (INIS)

    Mansy, M.S.; Bashter, I.I.; El-Mesiry, M.S.; Habib, N.; Adib, M.

    2015-01-01

    Filtered neutron techniques were applied to produce quasi-monoenergetic neutron beams in the energy range of 1.5–133 keV at research reactors. A simulation study was performed to characterize the filter components and transmitted beam lines. The filtered beams were characterized in terms of the optimal thickness of the main and additive components. The filtered neutron beams had high purity and intensity, with low contamination from the accompanying thermal emission, fast neutrons and γ-rays. A computer code named “QMNB” was developed in the “MATLAB” programming language to perform the required calculations. - Highlights: • Quasi-monoenergetic neutron beams in energy range from (1.5–133) keV. • Interference between the resonance and potential scattering amplitudes. • Epithermal neutron beams used in BNCT

  15. 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)

  16. 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.)

  17. 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.

  18. 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)

  19. 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

  20. Automated Calculation of DIII-D Neutral Beam Availability

    International Nuclear Information System (INIS)

    Phillips, J.C.; Hong, R.M.; Scoville, B.G.

    1999-01-01

    The neutral beam systems for the DIII-D tokamak are an extremely reliable source of auxiliary plasma heating, capable of supplying up to 20 MW of injected power, from eight separate beam sources into each tokamak discharge. The high availability of these systems for tokamak operations is sustained by careful monitoring of performance and following up on failures. One of the metrics for this performance is the requested injected power profile as compared to the power profile delivered for a particular pulse. Calculating this was a relatively straightforward task, however innovations such as the ability to modulate the beams and more recently the ability to substitute an idle beam for one which has failed during a plasma discharge, have made the task very complex. For example, with this latest advance it is possible for one or more beams to have failed, yet the delivered power profile may appear perfect. Availability used to be manually calculated. This paper presents the methods and algorithms used to produce a system which performs the calculations based on information concerning the neutral beam and plasma current waveforms, along with post-discharge information from the Plasma Control System, which has the ability to issue commands for beams in real time. Plots representing both the requested and actual power profiles, along with statistics, are automatically displayed and updated each shot, on a web-based interface viewable both at DIII-D and by our remote collaborators using no-cost software

  1. Linear beam-beam tune shift calculations for the Tevatron Collider

    International Nuclear Information System (INIS)

    Johnson, D.

    1989-01-01

    A realistic estimate of the linear beam-beam tune shift is necessary for the selection of an optimum working point in the tune diagram. Estimates of the beam-beam tune shift using the ''Round Beam Approximation'' (RBA) have over estimated the tune shift for the Tevatron. For a hadron machine with unequal lattice functions and beam sizes, an explicit calculation using the beam size at the crossings is required. Calculations for various Tevatron lattices used in Collider operation are presented. Comparisons between the RBA and the explicit calculation, for elliptical beams, are presented. This paper discusses the calculation of the linear tune shift using the program SYNCH. Selection of a working point is discussed. The magnitude of the tune shift is influenced by the choice of crossing points in the lattice as determined by the pbar ''cogging effects''. Also discussed is current cogging procedures and presents results of calculations for tune shifts at various crossing points in the lattice. Finally, a comparison of early pbar tune measurements with the present linear tune shift calculations is presented. 17 refs., 13 figs., 3 tabs

  2. Intermediate-energy neutron beams from reactors for NCT

    International Nuclear Information System (INIS)

    Brugger, R.M.; Less, T.J.; Passmore, G.G.

    1986-01-01

    This paper discusses ways that a beam of intermediate-energy neutrons might be extracted from a nuclear reactor. The challenge is to suppress the fast-neutron component and the gamma-ray component of the flux while leaving enough of the intermediate-energy neutrons in the beam to be able to perform neutron capture therapy in less than an hour exposure time. Moderators, filters, and reflectors are considered. 11 references, 7 figures, 3 tables

  3. Calculation of the muon contamination in a π- meson beam

    International Nuclear Information System (INIS)

    Tran, A.H.

    1965-01-01

    We present here a method for calculating the μ contamination of a π-meson beam which is parallel and of cylindrical symmetry, and also the so-called 'CONTAMU' programme which makes it possible to carry out this calculation. An evaluation of the μ contamination is necessary for correcting the experimental values (gross) of the cross-sections of the various reactions using the π-meson beam as a beam of incident particles. The following two cases are dealt with: 1 - The beam is defined by an S 1 counter: the μ contamination is calculated when the beam passes through this counter. 2 - The beam is defined by 2 counters, S 1 and S 2 : the μ contamination is calculated when the beam passes through the 2 counters successively. After presenting the problem in the first introductory paragraph, we deal in detail in paragraph II with the calculation, following the order of the programme. At the end of paragraph II will be found definitions of a certain number of values which the programme calculates; these are the values of the contamination in one of the two preceding cases integrated in certain well-defined disintegration volumes. In paragraph III is given as an example a table of results for a few values of the parameters. The listing of the 'CONTAMU' programme is given in the appendix. This programme was established in 1963 for correcting the experimental values of the cross-sections obtained during an experiment carried out on the synchrotron Saturne by the Falk-Vairant group. (author) [fr

  4. 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.

  5. 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

  6. 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

  7. 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

  8. Calculation of proton beam initial orbit at cyclotron central region

    International Nuclear Information System (INIS)

    Pramudita Anggraita

    2012-01-01

    A calculation of proton beam initial orbits at cyclotron central region was carried out using Scilab 5.2.0. The calculation was done in 2 dimensions in a homogeneous magnetic field of 1.66 tesla at frequency of fourth harmonics. The positions of ion source, dees, and dummy dees follow those of GE Minitrace cyclotron, peak dee voltage 30 kV. The calculation yields result comparable to those simulated at KIRAMS-13 cyclotron. (author)

  9. 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

  10. 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

  11. Calculation of the beam injector steering system using Helmholtz coils

    International Nuclear Information System (INIS)

    Passaro, A.; Sircilli Neto, F.; Migliano, A.C.C.

    1991-03-01

    In this work, a preliminary evaluation of the beam injector steering system of the IEAv electron linac is presented. From the existing injector configuration and with the assumptions of monoenergetic beam (100 keV) and uniform magnetic field, two pairs of Helmholtz coils were calculated for the steering system. Excitations of 105 A.turn and 37 A.turn were determined for the first and second coils, respectively. (author)

  12. Radiation transport calculations for the ANS [Advanced Neutron Source] beam tubes

    International Nuclear Information System (INIS)

    Engle, W.W. Jr.; Lillie, R.A.; Slater, C.O.

    1988-01-01

    The Advanced Neutron Source facility (ANS) will incorporate a large number of both radial and no-line-of-sight (NLS) beam tubes to provide very large thermal neutron fluxes to experimental facilities. The purpose of this work was to obtain comparisons for the ANS single- and split-core designs of the thermal and damage neutron and gamma-ray scalar fluxes in these beams tubes. For experimental locations far from the reactor cores, angular flux data are required; however, for close-in experimental locations, the scalar fluxes within each beam tube provide a credible estimate of the various signal to noise ratios. In this paper, the coupled two- and three-dimensional radiation transport calculations employed to estimate the scalar neutron and gamma-ray fluxes will be described and the results from these calculations will be discussed. 6 refs., 2 figs

  13. Neutron beam facilities at the replacement research reactor

    International Nuclear Information System (INIS)

    Kennedy, S.

    1999-01-01

    Full text: On September 3rd 1997 the Australian Federal Government announced their decision to replace the HIFAR research reactor by 2005. The proposed reactor will be a multipurpose reactor with improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The neutron beam facilities are intended to cater for Australian scientific needs well into the 21st century. In the first stage of planning the neutron Beam Facilities at the replacement reactor, a Consultative Group was formed (BFCG) to determine the scientific capabilities of the new facility. Members of the group were drawn from academia, industry and government research laboratories. The BFCG submitted their report in April 1998, outlining the scientific priorities to be addressed. Cold and hot neutron sources are to be included, and cold and thermal neutron guides will be used to position most of the instruments in a neutron guide hall outside the reactor confinement building. In 2005 it is planned to have eight instruments installed with a further three to be developed by 2010, and seven spare instrument positions for development of new instruments over the life of the reactor. A beam facilities technical group (BFTG) was then formed to prepare the engineering specifications for the tendering process. The group consisted of some members of the BFCG, several scientists and engineers from ANSTO, and scientists from leading neutron scattering centres in Europe, USA and Japan. The BFTG looked in detail at the key components of the facility such as the thermal, cold and hot neutron sources, neutron collimators, neutron beam guides and overall requirements for the neutron guide hall. The report of the BFTG, completed in August 1998, was incorporated into the draft specifications for the reactor project, which were distributed to potential reactor vendors. An assessment of the first stage of reactor vendor submissions was completed in

  14. 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)

  15. Beam heating requirements for a tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Bertoncini, P.J.; Brooks, J.N.; Fasolo, J.A.; Stacey, W.M. Jr.

    1976-01-01

    Typical beam heating requirements for effective tokamak experimental power reactor (TEPR) operation have been studied in connection with the Argonne preliminary conceptual TEPR design. For an ignition level plasma (approximately 100 MWt fusion power) for the nominal case envisioned, the neutral beam is only used to heat the plasma to ignition. This typically requires a beam power output of 40 MW at 180 keV for about 3 sec with a total energy of 114 MJ supplied to the plasma. The beam requirements for an ignition device are not very sensitive to changes in wall-sputtered impurity levels or plasma resistivity. For a plasma that must be driven due to poor confinement, the beam must remain on for most of the burn cycle. For representative cases, beam powers of approximately 23 MW are required for a total on-time of 20 to 50 sec. Reqirements on power level, beam energy, on-time, and beam-generation efficiency all represent considerable advances over present technology. For the Argonne TEPR design, a total of 16 to 32 beam injectors is envisioned. For a 40-MW, 180-keV, one-component beam, each injector supplies about 7 to 14 A of neutrals to the plasma. For positive ion sources, about 50 to 100 A of ions are required per injector and some form of particle and/or energy recycling appears to be essential in order to meet the power and efficiency requirements

  16. Alize 3 - first critical experiment for the franco-german high flux reactor - calculations

    International Nuclear Information System (INIS)

    Scharmer, K.

    1969-01-01

    The results of experiments in the light water cooled D 2 O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured k eff was smaller than 0.5 per cent δk/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D 2 O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author) [fr

  17. 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.)

  18. The High Flux Beam Reactor at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Shapiro, S.M.

    1994-01-01

    Brookhaven National Laboratory's High Flux Beam Reactor (HFBR) was built because of the need of the scientist to always want 'more'. In the mid-50's the Brookhaven Graphite reactor was churning away producing a number of new results when the current generation of scientists, led by Donald Hughes, realized the need for a high flux reactor and started down the political, scientific and engineering path that led to the BFBR. The effort was joined by a number of engineers and scientists among them, Chemick, Hastings, Kouts, and Hendrie, who came up with the novel design of the HFBR. The two innovative features that have been incorporated in nearly all other research reactors built since are: (i) an under moderated core arrangement which enables the thermal flux to peak outside the core region where beam tubes can be placed, and (ii) beam tubes that are tangential to the core which decrease the fast neutron background without affecting the thermal beam intensity. Construction began in the fall of 1961 and four years later, at a cost of $12 Million, criticality was achieved on Halloween Night, 1965. Thus began 30 years of scientific accomplishments

  19. Parametrisation of linear accelerator electron beam for computerised dosimetry calculations

    International Nuclear Information System (INIS)

    Millan, P.E.; Millan, S.; Hernandez, A.; Andreo, P.

    1979-01-01

    A previously published age-diffusion model has been adapted to obtain parameters for the Saggittaire linear accelerator electron beams. The calculations are shown and the results discussed. A comparison is presented between measured and predicted percentage depth doses for electron beams at various energies between 10 and 32 MeV. Theoretical isodose curves are compared, for an energy of 10 MeV, with experimental curves. The parameters obtained are used for computer electron isodose curve calculation in a program called FIJOE adapted from a previously published program. This program makes it possible to correct for irregular body contours, but not for internal inhomogeneities. (UK)

  20. Achievement and development of neutron beam utilization in research reactors

    International Nuclear Information System (INIS)

    Isshiki, Masahiko

    1996-01-01

    Especially regarding the neutron beam experiment in Japan, the basic research has been developed by utilizing the JRR-2 of Japan Atomic Energy Research Institute and the KUR of Kyoto University over long years. Now, the JRR-3M of JAERI was revived as a high performance, general purpose reactor, and bears important roles as the neutron beam experiment center in Japan. Thanks to one of the most powerful reactor neutron sources in the world and the cold neutron source, the environment of research was greatly improved, and the excellent results of researches began to be reported. The discovery of neutrons by Chadwick and the history of the related researches are described. As neutron sources, radioisotopes, accelerators and nuclear reactors are properly used corresponding to purposes. As the utilization of research reactors for neutron sources, the utilization for irradiation and neutron beam experiment are carried out. The outline of the research reactor JRR-3M is explained. The state of utilization in neutron scattering experiment, neutron radiography, prompt γ-ray analysis and the medical irradiation of neutrons is reported. (K.I.)

  1. Repetition rates in heavy ion beam driven fusion reactors

    Science.gov (United States)

    Peterson, Robert R.

    1986-01-01

    The limits on the cavity gas density required for beam propagation and condensation times for material vaporized by target explosions can determine the maximum repetition rate of Heavy Ion Beam (HIB) driven fusion reactors. If the ions are ballistically focused onto the target, the cavity gas must have a density below roughly 10-4 torr (3×1012 cm-3) at the time of propagation; other propagation schemes may allow densities as high as 1 torr or more. In some reactor designs, several kilograms of material may be vaporized off of the target chamber walls by the target generated x-rays, raising the average density in the cavity to 100 tor or more. A one-dimensional combined radiation hydrodynamics and vaporization and condensation computer code has been used to simulate the behavior of the vaporized material in the target chambers of HIB fusion reactors.

  2. Repetition rates in heavy ion beam driven fusion reactors

    International Nuclear Information System (INIS)

    Peterson, R.R.

    1986-01-01

    The limits on the cavity gas density required for beam propagation and condensation times for material vaporized by target explosions can determine the maximum repetition rate of Heavy Ion Beam (HIB) driven fusion reactors. If the ions are ballistically focused onto the target, the cavity gas must have a density below roughly 10 -4 torr (3 x 10 12 cm -3 ) at the time of propagation; other propagation schemes may allow densities as high as 1 torr or more. In some reactor designs, several kilograms of material may be vaporized off of the target chamber walls by the target generated x-rays, raising the average density in the cavity to 100 tor or more. A one-dimensional combined radiation hydrodynamics and vaporization and condensation computer code has been used to simulate the behavior of the vaporized material in the target chambers of HIB fusion reactors

  3. Beam transport calculations for BARC-TIFR 14UD pelletron

    International Nuclear Information System (INIS)

    Prasad, K.G.

    1993-01-01

    The 14UD pelletron tandem accelerator installed at Tata Institute of Fundamental Research (TIFR) as a joint BARC-TIFR project, is supplied by National Electrostatic Corporation (NEC), U.S.A. To optimise the parameters of various elements along the beam path, it is essential to work out the beam optics of the entire system. There are various computer codes in use for such calculations. All these codes, except the detailed ray tracing programs, use matrix formulation. Thus each ion optical element is characterised in terms of a transport matrix, whose elements are assumed to be independent of particle trajectory. We have performed only the first order calculations, meaning thereby that no aberrations are included. Further, all calculations are carried out assuming ideal conditions like axial beam injection, perfectly aligned beam line elements, etc. The main code that has been employed in our calculations is based on the one at the Australian National University, Canberra, suitably modified for use with CYBER 170/730 computer at TIFR. However, codes at NEC and Stony Brook were also used for the checking the results. The results of calculations are given and discussed. (author). 2 figs

  4. Visual beam tube inspection at the TRIGA reactor Vienna

    International Nuclear Information System (INIS)

    Boeck, H.; Musilek, A.; Villa, M.

    2006-01-01

    Of the four TRIGA beam tubes two have been visually inspected in 1985. Prior to the inspection the reactor was shut down for 3 weeks. The fuel elements around the beam tubes were removed. Stainless steel dummy elements were inserted in the fuel positions to shield the core radiation. The active part of the Fast Rabbit Tube was removed into the beam tube loading device and transferred to an interim storage: Front dose rate was ∼ 50 mSv/h. Generally the beam tube was very clean, after the last inspection about 30 years ago. A1 cm cut was observed at the beam tube front end. A rigid endoscope was used to check the beam tube's inner surface using a 90 degree deflection objective and photo- and video equipment. The direct dose rate in front of the beam tube was about 30 mSv/h. The beam tube was vacuum cleaned. A corroded shielding tank containing boric acid has leaked. A wooden collimator partially disintegrating due to extreme temperature was removed from beam tube D. Documentation of the inspection for visible defects is produced for later comparison

  5. 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

  6. 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

  7. Neutral-beam-injected tokamak fusion reactors: a review

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1976-08-01

    The theories of energetic-ion velocity distributions, stability, injection, and orbits were summarized. The many-faceted role of the energetic ions in plasma heating, fueling, and current maintenance, as well as in the direct enhancement of fusion power multiplication and power density, is discussed in detail for three reactor types. The relevant implications of recent experimental results on several beam-injected tokamaks are examined. The behavior of energetic ions is found to be in accordance with classical theory, large total ion energy densities are readily achieved, and plasma equilibrium and stability are maintained. The status of neutral-beam injectors and of conceptual design studies of beam-driven reactors are briefly reviewed. The principal plasma-engineering problems are those associated directly with achieving quasi-stationary operation

  8. Calculation of beam quality correction factor using Monte Carlo simulation

    International Nuclear Information System (INIS)

    Kawachi, T.; Saitoh, H.; Myojoyama, A.; Katayose, T.; Kojima, T.; Fukuda, K.; Inoue, M.

    2005-01-01

    In recent years, a number of the CyberKnife systems (Accuray C., U.S.) have been increasing significantly. However, the CyberKnife has unique treatment head structure and beam collimating system. Therefore, the global standard protocols can not be adopted for absolute absorbed dose dosimetry in CyberKnife beam. In this work, the energy spectrum of photon and electron from CyberKnife treatment head at 80 cm SSD and several depths in water are simulated with conscientious geometry using by the EGS Monte Carlo method. Furthermore, for calculation of the beam quality correction factor k Q , the mean restricted mass stopping power and the mass energy absorption coefficient of air, water and several chamber wall and waterproofing sleeve materials are calculated. As a result, the factors k Q CyberKnife beam for several ionization chambers are determined. And the relationship between the beam quality index PDD(10) x in CyberKnife beam and k Q is described in this report. (author)

  9. 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)

  10. 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)

  11. Experimental evaluation of analytical penumbra calculation model for wobbled beams

    International Nuclear Information System (INIS)

    Kohno, Ryosuke; Kanematsu, Nobuyuki; Yusa, Ken; Kanai, Tatsuaki

    2004-01-01

    The goal of radiotherapy is not only to apply a high radiation dose to a tumor, but also to avoid side effects in the surrounding healthy tissue. Therefore, it is important for carbon-ion treatment planning to calculate accurately the effects of the lateral penumbra. In this article, for wobbled beams under various irradiation conditions, we focus on the lateral penumbras at several aperture positions of one side leaf of the multileaf collimator. The penumbras predicted by an analytical penumbra calculation model were compared with the measured results. The results calculated by the model for various conditions agreed well with the experimental ones. In conclusion, we found that the analytical penumbra calculation model could predict accurately the measured results for wobbled beams and it was useful for carbon-ion treatment planning to apply the model

  12. 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

  13. 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)

  14. High-energy tritium beams as current drivers in tokamak reactors

    International Nuclear Information System (INIS)

    Mikkelsen, D.R.; Grisham, L.R.

    1983-04-01

    The effect on neutral-beam design and reactor performance of using high-energy (approx. 3-10 MeV) tritium neutral beams to drive steady-state tokamak reactors is considered. The lower current of such beams leads to several advantages over lower-energy neutral beams. The major disadvantage is the reduction of the reactor output caused by the lower current-drive efficiency of the high-energy beams

  15. 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.)

  16. 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

  17. LTRACK: Beam-transport calculation including wakefield effects

    International Nuclear Information System (INIS)

    Chan, K.C.D.; Cooper, R.K.

    1988-01-01

    LTRACK is a first-order beam-transport code that includes wakefield effects up to quadrupole modes. This paper will introduce the readers to this computer code by describing the history, the method of calculations, and a brief summary of the input/output information. Future plans for the code will also be described

  18. Calculated LET Spectrum from Antiproton Beams Stopping in Water

    DEFF Research Database (Denmark)

    Bassler, Niels; Holzscheiter, Michael

    2009-01-01

    significantly differ from unity, which seems to warrant closer inspection of the radiobiology in this region. Monte Carlo simulations using FLUKA were performed for calculating the entire particle spectrum of a beam of 126 MeV antiprotons hitting a water phantom. In the plateau region of the simulated...

  19. Beam brightness calculation for analytical and empirical distribution functions

    International Nuclear Information System (INIS)

    Myers, T.J.; Boulais, K.A.; O, Y.S.; Rhee, M.J.

    1992-01-01

    The beam brightness, a figure of merit for a beam quality useful for high-current low-emittance beams, was introduced by van Steenbergen as B = I/V 4 , where I is the beam current and V 4 is the hypervolume in the four-dimensional trace space occupied by the beam particles. Customarily, the brightness is expressed in terms of the product of emittances ε x ε y as B = ηI/(π 2 ε x ε y ), where η is a form factor of order unity which depends on the precise definition of emittance and hypervolume. Recently, a refined definition of the beam brightness based on the arithmetic mean value defined in statistics is proposed. The beam brightness is defined as B triple-bond 4 > = I -1 ∫ ρ 4 2 dxdydx'dy', where I is the beam current given by I ∫ ρ 4 dxdydx'dy'. Note that in this definition, neither the hypervolume V 4 nor the emittance, are explicitly used; the brightness is determined solely by the distribution function. Brightnesses are unambiguously calculated and expressed analytically in terms of the respective beam current and effective emittance for a few commonly used distribution functions, including Maxwellian and water-bag distributions. Other distributions of arbitrary shape frequently encountered in actual experiments are treated numerically. The resulting brightnesses are expressed in the form B = ηI/(π 2 ε x ε y ), and η is found to be weakly dependent on the form of velocity distribution as well as spatial distribution

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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

  5. Decommissioning of the High Flux Beam Reactor at Brookhaven Lab

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2011-05-27

    The High Flux Beam Reactor at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on October 31, 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shutdown in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor’s spent fuel pool. The reactor remained shutdown for almost three years for safety and environmental reviews. In November 1999 the United States Department of Energy decided to permanently shutdown the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR cleanup conducted during 1999-2009 will be described in the paper.

  6. 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.

  7. 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

  8. 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.)

  9. 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

  10. 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)

  11. 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)

  12. Seismic hazard studies for the high flux beam reactor at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Costantino, C.J.; Heymsfield, E.; Park, Y.J.; Hofmayer, C.H.

    1991-01-01

    This paper presents the results of a calculation to determine the site specific seismic hazard appropriate for the deep soil site at Brookhaven National Laboratory (BNL) which is to be used in the risk assessment studies being conducted for the High Flux Beam Reactor (HFBR). The calculations use as input the seismic hazard defined for the bedrock outcrop by a study conducted at Lawrence Livermore National Laboratory (LLNL). Variability in site soil properties were included in the calculations to obtain the seismic hazard at the ground surface and compare these results with those using the generic amplification factors from the LLNL study

  13. 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)

  14. 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

  15. Hydrodynamic calculations of 20-TeV beam interactions with the SSC beam dump

    International Nuclear Information System (INIS)

    Wilson, D.C.; Wingate, C.A.; Goldstein, J.C.; Godwin, R.P.; Mokhov, N.V.

    1993-01-01

    The 300μs, 400 MJ SSC proton beam must be contained when extracted to the external beam dump. The current design for the SSC beam dump can tolerate the beat load produced if the beam is deflected into a raster scan over the face of the dump. If the high frequency deflecting magnet were to fail, the beam would scan a single strip across the dump face resulting in higher local energy deposition. This could vaporize some material and lead to high pressures. Since the beam duration is comparable to the characteristic time of expected hydrodynamic motions, we have combined the static energy deposition capability of the MARS computer code with the two- and three-dimensional hydrodynamics of the MBA and SPHINX codes. EOS data suggest an energy deposition threshold of 15 kJ/g, below which hydrodynamic effects are minimal. Above this our 2D calculations show a hole boring rate of 7 cm/μs for the nominal beam, and pressures of a few kbar. Scanning the nominal beam faster than 0.08 cm/μs should minimize hydrodynamic effects. 3D calculations support this

  16. 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

  17. Investigation on the neutron beam characteristics for boron neutron capture therapy with 3D and 2D transport calculations

    International Nuclear Information System (INIS)

    Kodeli, I.; Diop, C.M.; Nimal, J.C.

    1994-01-01

    In the framework of future Boron Neutron Capture Therapy (BNCT) experiments, where cells and animals irradiations are planned at the research reactor of Strasbourg University, the feasibility to obtain a suitable epithermal neutron beam is investigated. The neutron fluence and spectra calculations in the reactor are performed using the 3D Monte Carlo code TRIPOLI-3 and the 2D SN code TWODANT. The preliminary analysis of Al 2 O 3 and Al-Al 2 O 3 filters configurations are carried out in an attempt to optimize the flux characteristics in the beam tube facility. 7 figs., 7 refs

  18. 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)

  19. 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

  20. Fluence-convolution broad-beam (FCBB) dose calculation

    Energy Technology Data Exchange (ETDEWEB)

    Lu Weiguo; Chen Mingli, E-mail: wlu@tomotherapy.co [TomoTherapy Inc., 1240 Deming Way, Madison, WI 53717 (United States)

    2010-12-07

    IMRT optimization requires a fast yet relatively accurate algorithm to calculate the iteration dose with small memory demand. In this paper, we present a dose calculation algorithm that approaches these goals. By decomposing the infinitesimal pencil beam (IPB) kernel into the central axis (CAX) component and lateral spread function (LSF) and taking the beam's eye view (BEV), we established a non-voxel and non-beamlet-based dose calculation formula. Both LSF and CAX are determined by a commissioning procedure using the collapsed-cone convolution/superposition (CCCS) method as the standard dose engine. The proposed dose calculation involves a 2D convolution of a fluence map with LSF followed by ray tracing based on the CAX lookup table with radiological distance and divergence correction, resulting in complexity of O(N{sup 3}) both spatially and temporally. This simple algorithm is orders of magnitude faster than the CCCS method. Without pre-calculation of beamlets, its implementation is also orders of magnitude smaller than the conventional voxel-based beamlet-superposition (VBS) approach. We compared the presented algorithm with the CCCS method using simulated and clinical cases. The agreement was generally within 3% for a homogeneous phantom and 5% for heterogeneous and clinical cases. Combined with the 'adaptive full dose correction', the algorithm is well suitable for calculating the iteration dose during IMRT optimization.

  1. Heavy ion beam transport through liquid lithium first wall ICF reactor cavities

    International Nuclear Information System (INIS)

    Stroud, P.D.

    1985-01-01

    This analysis addresses the critical issue of the final transport of a heavy ion beam in an inertial confinement fusion reactor. The beam must traverse the reaction chamber from the final focusing lens to the target without being disrupted. This requirement has a strong impact on the reactor design. It is essential to the development of ICF fusion reactor technology, that the restrictions placed on the reactor engineering parameters by final beam transport consideration be understood early on

  2. 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)

  3. Dose calculations algorithm for narrow heavy charged-particle beams

    Energy Technology Data Exchange (ETDEWEB)

    Barna, E A; Kappas, C [Department of Medical Physics, School of Medicine, University of Patras (Greece); Scarlat, F [National Institute for Laser and Plasma Physics, Bucharest (Romania)

    1999-12-31

    The dose distributional advantages of the heavy charged-particles can be fully exploited by using very efficient and accurate dose calculation algorithms, which can generate optimal three-dimensional scanning patterns. An inverse therapy planning algorithm for dynamically scanned, narrow heavy charged-particle beams is presented in this paper. The irradiation `start point` is defined at the distal end of the target volume, right-down, in a beam`s eye view. The peak-dose of the first elementary beam is set to be equal to the prescribed dose in the target volume, and is defined as the reference dose. The weighting factor of any Bragg-peak is determined by the residual dose at the point of irradiation, calculated as the difference between the reference dose and the cumulative dose delivered at that point of irradiation by all the previous Bragg-peaks. The final pattern consists of the weighted Bragg-peaks irradiation density. Dose distributions were computed using two different scanning steps equal to 0.5 mm, and 1 mm respectively. Very accurate and precise localized dose distributions, conform to the target volume, were obtained. (authors) 6 refs., 3 figs.

  4. Calculated LET spectrum from antiproton beams stopping in water

    CERN Document Server

    Bassler, Niels

    2009-01-01

    Antiprotons have been proposed as a potential modality for radiotherapy because the annihilation at the end of range leads to roughly a doubling of physical dose in the Bragg peak region. So far it has been anticipated that the radiobiology of antiproton beams is similar to that of protons in the entry region of the beam, but very different in the annihilation region, due to the expected high-LET components resulting from the annihilation. On closer inspection we find that calculations of dose averaged LET in the entry region may suggest that the RBE of antiprotons in the plateau region could significantly differ from unity, which seems to warrant closer inspection of the radiobiology in this region. Materials and Methods. Monte Carlo simulations using FLUKA were performed for calculating the entire particle spectrum of a beam of 126 MeV antiprotons hitting a water phantom. Results and Discussion. In the plateau region of the simulated antiproton beam we observe a dose-averaged unrestrict...

  5. 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).

  6. Muon catalyzed fusion - fission reactor driven by a recirculating beam

    International Nuclear Information System (INIS)

    Eliezer, S.; Tajima, T.; Rosenbluth, M.N.

    1986-01-01

    The recent experimentally inferred value of multiplicity of fusion of deuterium and tritium catalyzed by muons has rekindled interest in its application to reactors. Since the main energy expended is in pion (and consequent muon) productions, we try to minimize the pion loss by magnetically confining pions where they are created. Although it appears at this moment not possible to achieve energy gain by pure fusion, it is possible to gain energy by combining catalyzed fusion with fission blankets. We present two new ideas that improve the muon fusion reactor concept. The first idea is to combine the target, the converter of pions into muons, and the synthesizer into one (the synergetic concept). This is accomplished by injecting a tritium or deuterium beam of 1 GeV/nucleon into DT fuel contained in a magnetic mirror. The confined pions slow down and decay into muons, which are confined in the fuel causing little muon loss. The necessary quantity of tritium to keep the reactor viable has been derived. The second idea is that the beam passing through the target is collected for reuse and recirculated, while the strongly interacted portion of the beam is directed to electronuclear blankets. The present concepts are based on known technologies and on known physical processes and data. 29 refs., 6 figs., 4 tabs

  7. Calculations of beam dynamics in Sandia linear electron accelerators, 1984

    International Nuclear Information System (INIS)

    Poukey, J.W.; Coleman, P.D.

    1985-03-01

    A number of code and analytic studies were made during 1984 which pertain to the Sandia linear accelerators MABE and RADLAC. In this report the authors summarize the important results of the calculations. New results include a better understanding of gap-induced radial oscillations, leakage currents in a typical MABE gas, emittance growth in a beam passing through a series of gaps, some new diocotron results, and the latest diode simulations for both accelerators. 23 references, 30 figures, 1 table

  8. 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)

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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

  14. Numerical calculation of beam coupling impedances in synchrotron accelerators

    International Nuclear Information System (INIS)

    Haenichen, Lukas

    2016-01-01

    Beams of charged particles are of interest in various fields of research including particle and nuclear physics, material and medical science and many more. In synchrotron accelerators the accelerating section is passed periodically. A closed loop trajectory is enforced, by increasing the frequency of the accelerating electric field and the magnitude of the dipolar magnetic guide field synchronously. A synchrotron therefore consists of a circular assembly of various beamline elements which serve the purposes of accelerating and guiding the particle beam. For the flawless operation of such a machine it has to be assured that the particles perform a controlled motion along predefined trajectories. Amongst others, the fulfillment of the corresponding stability criteria is in close conjuction with the so-called beam coupling impedances which are an important figure of merit for collective effects in synchrotron accelerators. This work focuses on analytical and numerical methods for the calculation of beam coupling impedances. One of the primary objectives is to gain a better understanding of the electrodynamics related to charged particle beams, furthermore to recapitulate the mathematical description of charged particle beams in both time and frequency domain and finally establish the links between actual physics and numerical modeling. Analytical methods are usually restricted to symmetrical geometry and may solely serve for the approximate determination of the field distribution in real geometries or to validate certain numerical methods. More accurate prognosis is only possible with three-dimensional simulation models. Numerical simulation techniques have been established in the second half of the last century accompanying the evolution of many particle accelerators. Classical time domain codes were the prevailing simulation tools where the actual process of the particle motion sequence is reproduced. For the present case of a heavy ion synchrotron accelerator

  15. Numerical calculation of beam coupling impedances in synchrotron accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Haenichen, Lukas

    2016-07-01

    Beams of charged particles are of interest in various fields of research including particle and nuclear physics, material and medical science and many more. In synchrotron accelerators the accelerating section is passed periodically. A closed loop trajectory is enforced, by increasing the frequency of the accelerating electric field and the magnitude of the dipolar magnetic guide field synchronously. A synchrotron therefore consists of a circular assembly of various beamline elements which serve the purposes of accelerating and guiding the particle beam. For the flawless operation of such a machine it has to be assured that the particles perform a controlled motion along predefined trajectories. Amongst others, the fulfillment of the corresponding stability criteria is in close conjuction with the so-called beam coupling impedances which are an important figure of merit for collective effects in synchrotron accelerators. This work focuses on analytical and numerical methods for the calculation of beam coupling impedances. One of the primary objectives is to gain a better understanding of the electrodynamics related to charged particle beams, furthermore to recapitulate the mathematical description of charged particle beams in both time and frequency domain and finally establish the links between actual physics and numerical modeling. Analytical methods are usually restricted to symmetrical geometry and may solely serve for the approximate determination of the field distribution in real geometries or to validate certain numerical methods. More accurate prognosis is only possible with three-dimensional simulation models. Numerical simulation techniques have been established in the second half of the last century accompanying the evolution of many particle accelerators. Classical time domain codes were the prevailing simulation tools where the actual process of the particle motion sequence is reproduced. For the present case of a heavy ion synchrotron accelerator

  16. 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.)

  17. 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)

  18. 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)

  19. 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

  20. Fast pencil beam dose calculation for proton therapy using a double-Gaussian beam model

    Directory of Open Access Journals (Sweden)

    Joakim eda Silva

    2015-12-01

    Full Text Available The highly conformal dose distributions produced by scanned proton pencil beams are more sensitive to motion and anatomical changes than those produced by conventional radiotherapy. The ability to calculate the dose in real time as it is being delivered would enable, for example, online dose monitoring, and is therefore highly desirable. We have previously described an implementation of a pencil beam algorithm running on graphics processing units (GPUs intended specifically for online dose calculation. Here we present an extension to the dose calculation engine employing a double-Gaussian beam model to better account for the low-dose halo. To the best of our knowledge, it is the first such pencil beam algorithm for proton therapy running on a GPU. We employ two different parametrizations for the halo dose, one describing the distribution of secondary particles from nuclear interactions found in the literature and one relying on directly fitting the model to Monte Carlo simulations of pencil beams in water. Despite the large width of the halo contribution, we show how in either case the second Gaussian can be included whilst prolonging the calculation of the investigated plans by no more than 16%, or the calculation of the most time-consuming energy layers by about 25%. Further, the calculation time is relatively unaffected by the parametrization used, which suggests that these results should hold also for different systems. Finally, since the implementation is based on an algorithm employed by a commercial treatment planning system, it is expected that with adequate tuning, it should be able to reproduce the halo dose from a general beam line with sufficient accuracy.

  1. Parallel processing of dose calculation for external photon beam therapy

    International Nuclear Information System (INIS)

    Kunieda, Etsuo; Ando, Yutaka; Tsukamoto, Nobuhiro; Ito, Hisao; Kubo, Atsushi

    1994-01-01

    We implemented external photon beam dose calculation programs into a parallel processor system consisting of Transputers, 32-bit processors especially suitable for multi-processor configuration. Two network conformations, binary-tree and pipeline, were evaluated for rectangular and irregular field dose calculation algorithms. Although computation speed increased in proportion to the number of CPU, substantial overhead caused by inter-processor communication occurred when a smaller computation load was delivered to each processor. On the other hand, for irregular field calculation, which requires more computation capability for each calculation point, the communication overhead was still less even when more than 50 processors were involved. Real-time responses could be expected for more complex algorithms by increasing the number of processors. (author)

  2. 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.

  3. 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)

  4. Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Bozic, M.; Zagar, T.; Ravnik, M.

    2001-01-01

    The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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

  10. 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)

  11. Laser Beam and Resonator Calculations on Desktop Computers.

    Science.gov (United States)

    Doumont, Jean-Luc

    There is a continuing interest in the design and calculation of laser resonators and optical beam propagation. In particular, recently, interest has increased in developing concepts such as one-sided unstable resonators, supergaussian reflectivity profiles, diode laser modes, beam quality concepts, mode competition, excess noise factors, and nonlinear Kerr lenses. To meet these calculation needs, I developed a general-purpose software package named PARAXIA ^{rm TM}, aimed at providing optical scientists and engineers with a set of powerful design and analysis tools that provide rapid and accurate results and are extremely easy to use. PARAXIA can handle separable paraxial optical systems in cartesian or cylindrical coordinates, including complex-valued and misaligned ray matrices, with full diffraction effects between apertures. It includes the following programs:. ABCD provides complex-valued ray-matrix and gaussian -mode analyses for arbitrary paraxial resonators and optical systems, including astigmatism and misalignment in each element. This program required that I generalize the theory of gaussian beam propagation to the case of an off-axis gaussian beam propagating through a misaligned, complex -valued ray matrix. FRESNEL uses FFT and FHT methods to propagate an arbitrary wavefront through an arbitrary paraxial optical system using Huygens' integral in rectangular or radial coordinates. The wavefront can be multiplied by an arbitrary mirror profile and/or saturable gain sheet on each successive propagation through the system. I used FRESNEL to design a one-sided negative-branch unstable resonator for a free -electron laser, and to show how a variable internal aperture influences the mode competition and beam quality in a stable cavity. VSOURCE implements the virtual source analysis to calculate eigenvalues and eigenmodes for unstable resonators with both circular and rectangular hard-edged mirrors (including misaligned rectangular systems). I used VSOURCE to

  12. Calculations of Neutral Beam Ion Confinement for the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Redi, M.H.; Darrow, D.S.; Egedal, J.; Kaye, S.M.; White, R.B.

    2002-01-01

    The spherical torus (ST) concept underlies several contemporary plasma physics experiments, in which relatively low magnetic fields, high plasma edge q, and low aspect ratio combine for potentially compact, high beta and high performance fusion reactors. An important issue for the ST is the calculation of energetic ion confinement, as large Larmor radius makes conventional guiding center codes of limited usefulness and efficient plasma heating by RF and neutral beam ion technology requires minimal fast ion losses. The National Spherical Torus Experiment (NSTX) is a medium-sized, low aspect ratio ST, with R=0.85 m, a=0.67 m, R/a=1.26, Ip*1.4 MA, Bt*0.6 T, 5 MW of neutral beam heating and 6 MW of RF heating. 80 keV neutral beam ions at tangency radii of 0.5, 0.6 and 0.7 m are routinely used to achieve plasma betas above 30%. Transport analyses for experiments on NSTX often exhibit a puzzling ion power balance. It will be necessary to have reliable beam ion calculations to distinguish among the source and loss channels, and to explore the possibilities for new physics phenomena, such as the recently proposed compressional Alfven eigenmode ion heating

  13. Utilization of cold neutron beams at intermediate flux reactors

    International Nuclear Information System (INIS)

    Clark, D.D.

    1992-01-01

    With the advent of cold neutron beam (CNB) facilities at U.S. reactors [National Institute of Standards and Technology (NIST) in 1991; Cornell University and the University of Texas at Austin, anticipated in 1992], it is appropriate to reexamine the types of research for which they are likely to be best suited or uniquely suited. With the exception of a small-angle neutron scattering facility at Brookhaven National Laboratory, there has been no prior experience in the United States with such beams, but they have been extensively used at European reactors where cold neutron sources and neutron guides were developed some years age. This paper does not discuss specialized cases such as ultracold neutrons or very high flux facilities such as the Institute Laue-Langevin ractor and the proposed advanced neutron source. Instead, it concentrates on potential utilization of CNBs at intermediate-flux reactors such as at Cornell and Texas, i.e., in the 1-MW range and operated <24 h a day

  14. Fast optimization and dose calculation in scanned ion beam therapy

    International Nuclear Information System (INIS)

    Hild, S.; Graeff, C.; Trautmann, J.; Kraemer, M.; Zink, K.; Durante, M.; Bert, C.

    2014-01-01

    Purpose: Particle therapy (PT) has advantages over photon irradiation on static tumors. An increased biological effectiveness and active target conformal dose shaping are strong arguments for PT. However, the sensitivity to changes of internal geometry complicates the use of PT for moving organs. In case of interfractionally moving objects adaptive radiotherapy (ART) concepts known from intensity modulated radiotherapy (IMRT) can be adopted for PT treatments. One ART strategy is to optimize a new treatment plan based on daily image data directly before a radiation fraction is delivered [treatment replanning (TRP)]. Optimizing treatment plans for PT using a scanned beam is a time consuming problem especially for particles other than protons where the biological effective dose has to be calculated. For the purpose of TRP, fast optimization and fast dose calculation have been implemented into the GSI in-house treatment planning system (TPS) TRiP98. Methods: This work reports about the outcome of a code analysis that resulted in optimization of the calculation processes as well as implementation of routines supporting parallel execution of the code. To benchmark the new features, the calculation time for therapy treatment planning has been studied. Results: Compared to the original version of the TPS, calculation times for treatment planning (optimization and dose calculation) have been improved by a factor of 10 with code optimization. The parallelization of the TPS resulted in a speedup factor of 12 and 5.5 for the original version and the code optimized version, respectively. Hence the total speedup of the new implementation of the authors' TPS yielded speedup factors up to 55. Conclusions: The improved TPS is capable of completing treatment planning for ion beam therapy of a prostate irradiation considering organs at risk in this has been overseen in the review process. Also see below 6 min

  15. 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.

  16. 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)

  17. Seismic upgrading of the Brookhaven High Flux Beam Research Reactor

    International Nuclear Information System (INIS)

    Subudhi, M.

    1985-01-01

    In recent years the High Flux Beam Research (HFBR) reactor facility at Brookhaven National Laboratory (BNL) was upgraded from 40 to 50 MW power level. The reactor plant was built in the early sixties to the seismic design requirements of the period, using the static load approach. While the plant power level was upgraded, the seismic design was also improved according to current design criteria. This included the development of new floor response spectra for the facility and an overall seismic analysis of those systems important to the safe shutdown of the reactor. Items included in the reanalysis are the containment building with its internal structure, the piping systems, tanks, equipment, and heat exchangers. This paper describes the procedure utilized in developing the floor response spectra for the existing facility. Also included in the paper are the findings and recommendations, based on the seismic analysis, regarding the seismic adequacy of structural and mechanical systems vital to achieving the safe shutdown of the reactor. 11 references, 4 figures, 1 table

  18. Monochromatic neutron beam production at Brazilian nuclear research reactors

    Science.gov (United States)

    Stasiulevicius, Roberto; Rodrigues, Claudio; Parente, Carlos B. R.; Voi, Dante L.; Rogers, John D.

    2000-12-01

    Monochomatic beams of neutrons are obtained form a nuclear reactor polychromatic beam by the diffraction process, suing a single crystal energy selector. In Brazil, two nuclear research reactors, the swimming pool model IEA-R1 and the Argonaut type IEN-R1 have been used to carry out measurements with this technique. Neutron spectra have been measured using crystal spectrometers installed on the main beam lines of each reactor. The performance of conventional- artificial and natural selected crystals has been verified by the multipurpose neutron diffractometers installed at IEA-R1 and simple crystal spectrometer in operator at IEN- R1. A practical figure of merit formula was introduced to evaluate the performance and relative reflectivity of the selected planes of a single crystal. The total of 16 natural crystals were selected for use in the neutron monochromator, including a total of 24 families of planes. Twelve of these natural crystal types and respective best family of planes were measured directly with the multipurpose neutron diffractometers. The neutron spectrometer installed at IEN- R1 was used to confirm test results of the better specimens. The usually conventional-artificial crystal spacing distance range is limited to 3.4 angstrom. The interplane distance range has now been increased to approximately 10 angstrom by use of naturally occurring crystals. The neutron diffraction technique with conventional and natural crystals for energy selection and filtering can be utilized to obtain monochromatic sub and thermal neutrons with energies in the range of 0.001 to 10 eV. The thermal neutron is considered a good tool or probe for general applications in various fields, such as condensed matter, chemistry, biology, industrial applications and others.

  19. 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)

  20. 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

  1. 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)

  2. 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

  3. First wall costs of an ion-beam fusion reactor

    International Nuclear Information System (INIS)

    Hovingh, J.

    1977-08-01

    This paper parametrically investigates the effects of microexplosion energy on the first wall costs of a 4000 MW/sub t/ ion-beam initiated, inertially confined fusion reactor for several first wall materials. The thermodynamic models and the results for microexplosion energies between 400 and 4000 MJ are presented. A solid stainless steel or a composite isotropic graphite over stainless steel first wall can operate for a year at a cost of 0.6 mills per kWh gross electric power output

  4. OPTIMIZATION OF THE EPITHERMAL NEUTRON BEAM FOR BORON NEUTRON CAPTURE THERAPY AT THE BROOKHAVEN MEDICAL RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HU,J.P.; RORER,D.C.; RECINIELLO,R.N.; HOLDEN,N.E.

    2002-08-18

    Clinical trials of Boron Neutron Capture Therapy for patients with malignant brain tumor had been carried out for half a decade, using an epithermal neutron beam at the Brookhaven's Medical Reactor. The decision to permanently close this reactor in 2000 cut short the efforts to implement a new conceptual design to optimize this beam in preparation for use with possible new protocols. Details of the conceptual design to produce a higher intensity, more forward-directed neutron beam with less contamination from gamma rays, fast and thermal neutrons are presented here for their potential applicability to other reactor facilities. Monte Carlo calculations were used to predict the flux and absorbed dose produced by the proposed design. The results were benchmarked by the dose rate and flux measurements taken at the facility then in use.

  5. Horizontal beams creation on a mean power reactor: Isis 700 kW

    International Nuclear Information System (INIS)

    Morin, C.

    1987-08-01

    To satisfy the requests of experimenters, two horizontal beams, tangential at core, have been created. After a brief recall of the Isis reactor the author describes the realized works and gives a summary description of the two beams equipment [fr

  6. 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)

  7. 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)

  8. 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

  9. ORBIT : BEAM DYNAMICS CALCULATIONS FOR HIGH - INTENSITY RINGS

    International Nuclear Information System (INIS)

    HOLMES, J.A.; DANILOV, V.; GALAMBOS, J.; SHISHLO, A.; COUSINEAU, S.; CHOU, W.; MICHELOTTI, L.; OSTIGUY, F.; WEI, J.

    2002-01-01

    We are developing a computer code, ORBIT, specifically for beam dynamics calculations in high-intensity rings. Our approach allows detailed simulation of realistic accelerator problems. ORBIT is a particle-in-cell tracking code that transports bunches of interacting particles through a series of nodes representing elements, effects, or diagnostics that occur in the accelerator lattice. At present, ORBIT contains detailed models for strip-foil injection including painting and foil scattering; rf focusing and acceleration; transport through various magnetic elements; longitudinal and transverse impedances; longitudinal, transverse, and three-dimensional space charge forces; collimation and limiting apertures; and the calculation of many useful diagnostic quantities. ORBIT is an object-oriented code, written in C++ and utilizing a scripting interface for the convenience of the user. Ongoing improvements include the addition of a library of accelerator maps, BEAMLINE/MXYZPTLK the introduction of a treatment magnet errors and fringe fields; the conversion of the scripting interface to the standard scripting language, Python; and the parallelization of the computations using MPI. The ORBIT code is an open source, powerful, and convenient tool for studying beam dynamics in high-intensity rings

  10. Performance analysis of the intense slow-positron beam at the NC State University PULSTAR reactor

    International Nuclear Information System (INIS)

    Moxom, J.; Hathaway, A.G.; Bodnaruk, E.W.; Hawari, A.I.; Xu, J.

    2007-01-01

    An intense positron beam, for application in nanophase characterization, is now under construction at the 1 MW PULSTAR nuclear reactor at North Carolina State University (NCSU). A tungsten converter/moderator is used, allowing positrons to be emitted from the surface with energies of a few electron volts. These slow positrons will be extracted from the moderator and formed into a beam by electrostatic lenses and then injected into a solenoidal magnetic field for transport to one of three experimental stations, via a beam switch. To optimize the performance of the beam and to predict the slow-positron intensity, a series of simulations were performed. A specialized Monte-Carlo routine was integrated into the charged-particle transport calculations to allow accounting for the probabilities of positron re-emission and backscattering from multiple-bank moderator/converter configurations. The results indicate that either a two-bank or a four-bank tungsten moderator/converter system is preferred for the final beam design. The predicted slow-positron beam intensities range from nearly 7x10 8 to 9x10 8 e + /s for the two-bank and the four-bank systems, respectively

  11. 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

  12. 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

  13. 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

  14. 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.)

  15. 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)

  16. Rebuilding the Brookhaven high flux beam reactor: A feasibility study

    International Nuclear Information System (INIS)

    Brynda, W.J.; Passell, L.; Rorer, D.C.

    1995-01-01

    After nearly thirty years of operation, Brookhaven's High Flux Beam Reactor (HFBR) is still one of the world's premier steady-state neutron sources. A major center for condensed matter studies, it currently supports fifteen separate beamlines conducting research in fields as diverse as crystallography, solid-state, nuclear and surface physics, polymer physics and structural biology and will very likely be able to do so for perhaps another decade. But beyond that point the HFBR will be running on borrowed time. Unless appropriate remedial action is taken, progressive radiation-induced embrittlement problems will eventually shut it down. Recognizing the HFBR's value as a national scientific resource, members of the Laboratory's scientific and reactor operations staffs began earlier this year to consider what could be done both to extend its useful life and to assure that it continues to provide state-of-the-art research facilities for the scientific community. This report summarizes the findings of that study. It addresses two basic issues: (i) identification and replacement of lifetime-limiting components and (ii) modifications and additions that could expand and enhance the reactor's research capabilities

  17. Conceptual design of light ion beam inertia nuclear fusion reactors

    International Nuclear Information System (INIS)

    1983-07-01

    Light ion beam, inertia nuclear fusion system drew attention recently as one of the nuclear fusion systems for power reactors in the history of the research on nuclear fusion. Its beginning seemed to be the judgement that the implosion of fusion fuel pellets with light ions can be realized with the light ions which can be obtained in view of accelerator techniques. Of course, in order to generate practically usable nuclear fusion reaction by this system and maintain it, many technical difficulties must be overcome. This research was carried out for the purpose of discovering such technical problems and searching for their solution. At the time of doing the works, the following policy was adopted. Though their is the difference of fine and rough, the design of a whole reactor system is performed conformably. In order to make comparison with other reactor types and nuclear fusion systems, the design is carried out as the power plant of about one million kWe output. As the extent of the design, the works at conceptual design stage are performed to present the concept of design which satisfies the required function. Basically, the design is made from conservative standpoint. This research of design was started in 1981, and in fiscal 1982, the mutual adjustment among the design of respective parts was performed on the basis of the results in 1981, and the possible revision and new proposal were investigated. (Kako, I.)

  18. A neutronic feasibility study for LEU conversion of the High Flux Beam Reactor (HFBR)

    International Nuclear Information System (INIS)

    Pond, R.B.; Hanan, N.A.; Matos, J.E.

    1997-01-01

    A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm 3 in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept. (author)

  19. Scattering calculation and image reconstruction using elevation-focused beams.

    Science.gov (United States)

    Duncan, David P; Astheimer, Jeffrey P; Waag, Robert C

    2009-05-01

    Pressure scattered by cylindrical and spherical objects with elevation-focused illumination and reception has been analytically calculated, and corresponding cross sections have been reconstructed with a two-dimensional algorithm. Elevation focusing was used to elucidate constraints on quantitative imaging of three-dimensional objects with two-dimensional algorithms. Focused illumination and reception are represented by angular spectra of plane waves that were efficiently computed using a Fourier interpolation method to maintain the same angles for all temporal frequencies. Reconstructions were formed using an eigenfunction method with multiple frequencies, phase compensation, and iteration. The results show that the scattered pressure reduces to a two-dimensional expression, and two-dimensional algorithms are applicable when the region of a three-dimensional object within an elevation-focused beam is approximately constant in elevation. The results also show that energy scattered out of the reception aperture by objects contained within the focused beam can result in the reconstructed values of attenuation slope being greater than true values at the boundary of the object. Reconstructed sound speed images, however, appear to be relatively unaffected by the loss in scattered energy. The broad conclusion that can be drawn from these results is that two-dimensional reconstructions require compensation to account for uncaptured three-dimensional scattering.

  20. Beam dynamics calculations and particle tracking using massively parallel processors

    International Nuclear Information System (INIS)

    Ryne, R.D.; Habib, S.

    1995-01-01

    During the past decade massively parallel processors (MPPs) have slowly gained acceptance within the scientific community. At present these machines typically contain a few hundred to one thousand off-the-shelf microprocessors and a total memory of up to 32 GBytes. The potential performance of these machines is illustrated by the fact that a month long job on a high end workstation might require only a few hours on an MPP. The acceptance of MPPs has been slow for a variety of reasons. For example, some algorithms are not easily parallelizable. Also, in the past these machines were difficult to program. But in recent years the development of Fortran-like languages such as CM Fortran and High Performance Fortran have made MPPs much easier to use. In the following we will describe how MPPs can be used for beam dynamics calculations and long term particle tracking

  1. 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.)

  2. 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

  3. 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)

  4. Radiation Dosimetry of the Pressure Vessel Internals of the High Flux Beam Reactor

    Science.gov (United States)

    Holden, Norman E.; Reciniello, Richard N.; Hu, Jih-Perng; Rorer, David C.

    2003-06-01

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. The measurements were made using Red Perspex™ polymethyl methacrylate high-level film dosimeters, a Radcal "peanut" ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rates, the Monte Carlo MCNP code and geometric progressive MicroShield code were used to model the gamma-ray transport and dose buildup.

  5. RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR

    International Nuclear Information System (INIS)

    HOLDEN, N.E.; RECINIELLO, R.N.; HU, J.P.; RORER, D.C.

    2002-01-01

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex(trademark) polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup

  6. Design of a tokamak fusion reactor first wall armor against neutral beam impingement

    International Nuclear Information System (INIS)

    Myers, R.A.

    1977-12-01

    The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiation damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  8. 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

  9. 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

  10. 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

  11. Cutting technique for reactor internals by laser beam

    International Nuclear Information System (INIS)

    Matsumoto, O.; Sugihara, M.; Matsuda, K.; Miya, K.

    1990-01-01

    At present in Japan the verification tests on the commercial nuclear power reactor decommissioning technology are being conducted as the project of The Ministry of International Trade and Industry by Nuclear Power Engineering Test Center. This paper summarizes the interim results of the verification test for the reactor core internals decommissioning technology, which is being conducted from 1986 as a theme of the above project. All core internals to be studied here are made of stainless steel, and the maximum wall thickness is about 500mm (the maximum one to be cut is about 300mm) for the PWR's, and about 100mm for the BWR's. Though the plasma cutting, arc saw cutting method, etc. have been studied u p to now as the cutting technology for decommissioning these core internals, the authors are carrying out the development and verification test of the cutting technology with the laser beam, which is expected to increase its power in future and can be applied to various materials

  12. Beam transport calculations for the EN tandem installation

    International Nuclear Information System (INIS)

    Sparks, R.J.

    1980-12-01

    Transport of a charged particle beam through the new EN tandem accelerator installation of the Institute of Nuclear Sciences has been analysed using simplified mathematical models. The purpose is to identify the factors affecting transmission of the beam, and to arrive at a design for the system to inject the beam into the accelerator

  13. 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

  14. Efficiency calculations for the direct energy conversion system of the Cadarache neutral beam injectors

    International Nuclear Information System (INIS)

    White, R.C.

    1988-01-01

    A prototype energy conversion system is presently in operation at Cadarache, France. Such a device is planned for installation on each six neutral beam injectors for use in the Tore Supra experiment in 1989. We present calculations of beam performance that may influence design considerations. The calculations are performed with the DART charged particle beam code. We investigate the effects of cold plasma, direct energy conversion and neutral beam production. 4 refs., 6 figs., 4 tabs

  15. Shielding calculation of slow extracted beam facility at KEK proton synchrotron

    International Nuclear Information System (INIS)

    Hirabayashi, Hiromi; Katoh, Kazuaki

    1978-01-01

    The KEK proton synchrotron has two external beam lines, i.e. a fast extracted beam line for a bubble chamber and a slow extracted beam line for counter experiments. The maximum total intensity of the slow beam is estimated as 5 x 10 12 protons per sec. For beam losses along the line, shielding calculation was made, and on the basis of these results, adequacy of the current shielding construction plans was discussed. (Mori, K.)

  16. Prompt-gamma spectrometry for the optimization of reactor neutron beams in biomedical research

    International Nuclear Information System (INIS)

    Borisov, G.I.; Komkov, M.M.; Leonov, V.F.

    1988-01-01

    In order to select the optimal spectral composition and size for the reactor neutron beams applied to in vivo analysis and therapy in biomedical research it is necessary to determine the spatial slow-neutron flux distributions produced by the beam in the irradiated object and to calculate or measure the neutron dose equivalents of both the original spectrum and the moderated neutrons. In this study the maximum neutron dose equivalents are found by spectrometry of the prompt-γ emission from the interaction of neutrons with atomic nuclei in the irradiated object. Different spectral distributions were produced by using an unfiltered beam together with filters of quartz, cadmium, 10 B, iron, aluminum, and sulfur. The phantom used was a tank filled with an aqueous solution of urea. Cadmium-containing organs were simulated. For in vivo neutron-activation analysis of human tissues at a depth of 2-5 cm it was found advisable to use neutrons of 20-40 keV mean energy with a beam area of at least 45 cm 2

  17. Status of neutron beam utilization at the Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    Dien, Nguyen Nhi; Hai, Nguyen Canh

    2003-01-01

    The 500-kW Dalat nuclear research reactor was reconstructed from the USA-made 250-kW TRIGA Mark II reactor. After completion of renovation and upgrading, the reactor has been operating at its nominal power since 1984. The reactor is used mainly for radioisotope production, neutron activation analysis, neutron beam researches and reactor physics study. In the framework of the reconstruction and renovation project of the 1982-1984 period, the reactor core, the control and instrumentation system, the primary and secondary cooling systems, as well as other associated systems were newly designed and installed by the former Soviet Union. Some structures of the reactor, such as the reactor aluminum tank, the graphite reflector, the thermal column, horizontal beam tubes and the radiation concrete shielding have been remained from the previous TRIGA reactor. As a typical configuration of the TRIGA reactor, there are four neutron beam ports, including three radial and one tangential. Besides, there is a large thermal column. Until now only two-neutron beam ports and the thermal column have been utilized. Effective utilization of horizontal experimental channels is one of the important research objectives at the Dalat reactor. The research program on effective utilization of these experimental channels was conducted from 1984. For this purpose, investigations on physical characteristics of the reactor, neutron spectra and fluxes at these channels, safety conditions in their exploitation, etc. have been carried out. The neutron beams, however, have been used only since 1988. The filtered thermal neutron beams at the tangential channel have been extracted using a single crystal silicon filter and mainly used for prompt gamma neutron activation analysis (PGNAA), neutron radiography (NR) and transmission experiments (TE). The filtered quasi-monoenergetic keV neutron beams using neutron filters at the piercing channel have been used for nuclear data measurements, study on

  18. 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)

  19. 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)

  20. 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)

  1. 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

  2. Optimization study for an epithermal neutron beam for boron neutron capture therapy at the University of Virginia Research Reactor

    International Nuclear Information System (INIS)

    Burns, T.D. Jr.

    1995-05-01

    The non-surgical brain cancer treatment modality, Boron Neutron Capture Therapy (BNCT), requires the use of an epithermal neutron beam. This purpose of this thesis was to design an epithermal neutron beam at the University of Virginia Research Reactor (UVAR) suitable for BNCT applications. A suitable epithermal neutron beam for BNCT must have minimal fast neutron and gamma radiation contamination, and yet retain an appreciable intensity. The low power of the UVAR core makes reaching a balance between beam quality and intensity a very challenging design endeavor. The MCNP monte carlo neutron transport code was used to develop an equivalent core radiation source, and to perform the subsequent neutron transport calculations necessary for beam model analysis and development. The code accuracy was validated by benchmarking output against experimental criticality measurements. An epithermal beam was designed for the UVAR, with performance characteristics comparable to beams at facilities with cores of higher power. The epithermal neutron intensity of this beam is 2.2 x 10 8 n/cm 2 · s. The fast neutron and gamma radiation KERMA factors are 10 x 10 -11 cGy·cm 2 /n epi and 20 x 10 -11 cGy·cm 2 /n epi , respectively, and the current-to-flux ratio is 0.85. This thesis has shown that the UVAR has the capability to provide BNCT treatments, however the performance characteristics of the final beam of this study were limited by the low core power

  3. Vietnam Project For Production Of Radioactive Beam Based On ISOL Technique With The Dalat Reactor

    International Nuclear Information System (INIS)

    Le Hong Khiem; Phan Viet Cuong; Fadi Ibrahim

    2011-01-01

    The presence in Vietnam of Dalat nuclear reactor dedicated to fundamental studies is a unique opportunity to produce Radioactive Ion (RI) Beams with the fission of a 235 U induced by the thermal neutrons produced by the reactor. We propose to produce RI beams at the Dalat nuclear reactor using ISOL (Isotope Separation On-Line) technique. This project should be a unique opportunity for Vietnamese nuclear physics community to use its own facilities to produce RI beams for studying nuclear physics at an international level. (author)

  4. Scaling of heavy ion beam probes for reactor-size devices

    International Nuclear Information System (INIS)

    Hickok, R.L.; Jennings, W.C.; Connor, K.A.; Schoch, P.M.

    1984-01-01

    Heavy ion beam probes for reactor-size plasma devices will require beam energies of approximately 10 MeV. Although accelerator technology appears to be available, beam deflection systems and parallel plate energy analyzers present severe difficulties if existing technology is scaled in a straightforward manner. We propose a different operating mode which will use a fixed beam trajectory and multiple cylindrical energy analyzers. Development effort will still be necessary, but we believe the basic technology is available

  5. Effective and efficient method of calculating Bessel beam fields

    CSIR Research Space (South Africa)

    Litvin, IA

    2005-01-01

    Full Text Available Bessel beams have gathered much interest of late due to their properties of near diffraction free propagation and self reconstruction after obstacles. Such laser beams have already found applications in fields such as optical tweezers and as pump...

  6. Calculation And Design Of A New Configuration For Radiation Shielding At Neutron Beam No.3 For Fundamental And Applied Researches

    International Nuclear Information System (INIS)

    Vuong Huu Tan; Tran Tuan Anh; Nguyen Kien Cuong; Nguyen Canh Hai; Nguyen Xuan Hai; Pham Ngoc Son; Ho Huu Thang

    2011-01-01

    The tangential horizontal channel of No. 3 of the Dalat Research Reactor has been opened and used during the 1990s. The utilizations of the thermal neutron beam at this channel were the Neutron Radiography and the Prompt Gamma Neutron Activation Analysis method (PGNAA). At present, the neutron beam used for nuclear structure data researches based on the Summing of Amplitude Coincident Pulses system (SACP). Beside, several related research equipments have been set up and operated for the research purposes. A renovation of the neutron channel, therefore, will play an important role in safe and effective utilizations of the neutron beam in fields of nuclear physic training and researches. A new configuration for radiation shielding has been simulated by MCNP code. The calculated results of dose rates for neutron and gamma at working positions are in range of dose rate limit. (author)

  7. Do existing research reactors teach us all about beam tube optimisation?

    International Nuclear Information System (INIS)

    Roegler, Hans-Joachim; Feltes, Wolfgang

    1998-01-01

    The contribution makes the attempt to analyse the data base available in the literature and in Siemens' own projects and to find out potential systematics from the existing research reactors with beam tubes, separated into reactors with different reflectors and distinguished for tangential and radial tubes and cold neutron sources, respectively some generic calculations serve as gauging data. The contribution is not meant as critics on any design. The results might serve supporting designers and operators when evaluating the pros and cons of existing or planned design in terms of the optimum beam tubes. Existing lacks of systematics are evaluated in view of suitable explanations and constraints, which do not allow optimisation. Examples of such constraints are the different material layers between fuel zone and reflector zone which have various reasons. The limited data in the literature plus the numerous lacks of precision of the representation of those data should be an incentive to improve the performed analysis by collecting more exact data and re-doing the evaluation before answering the title question really

  8. Do existing research reactors teach us all about beam tube optimization?

    International Nuclear Information System (INIS)

    Roegler, Hans Joachim; Feltes, Wolfgang

    1998-01-01

    The contribution makes the attempt to analyse the data base available in the literature and in Siemens' own projects and to find out potential systematics from the existing research reactor with beam tubes, separated into reactors with different reflectors and distinguished for tangential and radial tubes and cold neutron sources, resp. Some generic calculations serve as gauging data. The contribution is not meant as critics on any design.The results might serve supporting designers and operators when evaluating the pros and cons of existing or planned design in terms of the optimum beam tubes. Existing lacks of systematics are evaluated in view of suitable explanations and constraints, which do not allow optimisation. Examples pf such constraints are the different material layers between fuel zone and reflector zone which have various reasons. The limited data in the literature plus the numerous lacks of precision of the representation of those data should be an incentive to improve the performed analysis by collecting more exact data and re-doing the evaluation before answering the title-question really. (author)

  9. 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

  10. Fusion reactor development using high power particle beams

    International Nuclear Information System (INIS)

    Ohara, Y.

    1990-01-01

    The present paper outlines major applications of the ion source/accelerator to fusion research and also addresses the present status and future plans for accelerator development. Applications of ion sources/accelerators for fusion research are discussed first, focusing on plasma heating, plasma current drive, plasma current profile control, and plasma diagnostics. The present status and future plan of ion sources/accelerators development are then described focusing on the features of existing and future tokamak equipment. Positive-ion-based NBI systems of 100 keV class have contributed to obtaining high temperature plasmas whose parameters are close to the fusion break-even condition. For the next tokamak fusion devices, a MeV class high power neutral beam injector, which will be used to obtain a steady state burning plasma, is considered to become the primary heating and current drive system. Development of such a system is a key to realize nuclear fusion reactor. It will be entirely indebted to the development of a MeV class high current negative deuterium ion source/accelerator. (N.K.)

  11. LIBRA - a light ion beam fusion conceptual reactor design

    International Nuclear Information System (INIS)

    Badger, B.; Moses, G.A.; Engelstad, R.L.; Kulcinski, G.L.; Lovell, E.; MacFarlane, J.; Peterson, R.R.; Sawan, M.E.; Sviatovslavsky, I.N.; Wittenberg, L.J.; Cook, D.L.; Olson, R.E.; Stinnett, R.W.; Ehrhardt, J.; Kessler, G.; Stein, E.

    1990-08-01

    The LIBRA light ion beam fusion commercial reactor study is a self-consistent conceptual design of a 330 MWe power plant with an accompanying economic analysis. Fusion targets are imploded by 4 MJ shaped pulses of 30 MeV Li ions at a rate of 3 Hz. The target gain is 80, leading to a yield of 320 MJ. The high intensity part of the ion pulse is delivered by 16 diodes through 16 separate z-pinch plasma channels formed in 100 torr of helium with trace amounts of lithium. The blanket is an array of porous flexible silicon carbind tubes with Li 17 Pb 83 flowing downward through them. These tubes (INPORT units) shield the target chamber wall from both neutron damage and the shock overpressure of the target explosion. The target chamber is 'self-pumped' by the target explosion generated overpressure into a surge tank partially filled with Li 17 Pb 83 that surrounds the target chamber. This scheme refreshes the chamber at the desired 3 Hz frequently without excessive pumping demands. The blanket multiplication is 1.2 and the tritium breeding ratio is 1.4. The direct capital cost of a 331 MWe LIBRA design is estimated to be 2843 Dollar/kWe while a 1200 MWe LIBRA design will cost approximately 1300 Dollar/kWe. (orig.) [de

  12. Apparatus for servicing a jet pump hold down beam in a nuclear reactor

    International Nuclear Information System (INIS)

    Howell, D.A.; Hydeman, J.E.; Slater, J.L.; Bodnar, R.J.; Golick, L.R.; Sckera, R.S.; Roth, C.H. Jr.

    1991-01-01

    This patent describes an apparatus for replacing the hold down beam of a fluid circulating jet pump mounted in a nuclear reactor, the hold down beam having a beam body, a pair of opposed beam tabs and a pair of opposed beam positioning trunnions extending outwardly from the beam body. It comprises a housing having a lower surface configured to be positionable over the body of the hold down beam; means coupled to the housing for engaging the beam trunnions and securing the beam body against the lower surface of the housing; means coupled to the housing for depressing the beam tabs while the beam body is secured against the lower surface of the housing; means coupled to the trunnion engaging means and the beam tab depressing means for selectively actuating the trunnion engaging means and the beam tab depressing means from a position remote from the nuclear reactor; and means connectable to the housing for selectively changing the directional orientation of the beam

  13. 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

  14. 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)

  15. 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

  16. Calculation of neutral beam deposition accounting for excited states

    International Nuclear Information System (INIS)

    Gianakon, T.A.

    1992-09-01

    Large-scale neutral-beam auxillary heating of plasmas has led to new plasma operational regimes which are often dominated by fast ions injected via the absorption of an energetic beam of hydrogen neutrals. An accurate simulation of the slowing down and transport of these fast ions requires an intimate knowledge of the hydrogenic neutral deposition on each flux surface of the plasma. As a refinement to the present generation of transport codes, which base their beam deposition on ground-state reaction rates, a new set of routines, based on the excited states of hydrogen, is presented as mechanism for computing the attenuation and deposition of a beam of energetic neutrals. Additionally, the numerical formulations for the underlying atomic physics for hydrogen impacting on the constiuent plasma species is developed and compiled as a numerical database. Sample results based on this excited state model are compared with the ground-state model for simple plasma configurations

  17. 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

  18. Construction of the Neutron Beam Facility at Australia's OPAL Research Reactor

    International Nuclear Information System (INIS)

    Kennedy, J.S.

    2005-01-01

    Full text: Australia's new research reactor, OPAL, has been designed for high quality neutron beam science and radioisotope production. It has a capacity for eighteen neutron beam instruments to be located at the reactor face and in a neutron guide hall. The new neutron beam facility features a 20 litre liquid deuterium cold neutron source and supermirror neutron reflecting guides for intense cold and thermal neutron beams. Nine neutron beam instruments are under development, of which seven are scheduled for completion in early 2007. The project is approaching the hot-commissioning stage, where criticality will be demonstrated. Installation of the neutron beam transport system and neutron beam instruments in the neutron guide hall and at the reactor face is underway, and the path to completion of this project is relatively clear. The lecture will outline Australia's aspirations for neutron science at the OPAL reactor, and describe the neutron beam facility under construction. The status of this project and a forecast of the program to completion, including commissioning and commencement of routine operation in 2007 will also be discussed. This project is the culmination of almost a decade of effort. We now eagerly anticipate catapulting Australia's neutron beam science capability to meet the best in the world today. (author)

  19. Radiological characterization of the pressure vessel internals of the BNL High Flux Beam Reactor.

    Science.gov (United States)

    Holden, Norman E; Reciniello, Richard N; Hu, Jih-Perng

    2004-08-01

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, measurements and calculations of the decay gamma-ray dose-rate were performed in the reactor pressure vessel and on vessel internal structures such as the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. Measurements of gamma-ray dose rates were made using Red Perspex polymethyl methacrylate high-dose film, a Radcal "peanut" ion chamber, and Eberline's RO-7 high-range ion chamber. As a comparison, the Monte Carlo MCNP code and MicroShield code were used to model the gamma-ray transport and dose buildup. The gamma-ray dose rate at 8 cm above the center of the Transition Plate was measured to be 160 Gy h (using an RO-7) and 88 Gy h at 8 cm above and about 5 cm lateral to the Transition Plate (using Red Perspex film). This compares with a calculated dose rate of 172 Gy h using Micro-Shield. The gamma-ray dose rate was 16.2 Gy h measured at 76 cm from the reactor core (using the "peanut" ion chamber) and 16.3 Gy h at 87 cm from the core (using Red Perspex film). The similarity of dose rates measured with different instruments indicates that using different methods and instruments is acceptable if the measurement (and calculation) parameters are well defined. Different measurement techniques may be necessary due to constraints such as size restrictions.

  20. 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

  1. 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)

  2. 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

  3. 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.

  4. X-ray and pressure conditions on the first wall of a particle beam inertial confinement reactor

    International Nuclear Information System (INIS)

    Magelssen, G.R.

    1979-01-01

    Because of the presence of a chamber gas in a particle beam reactor cavity, nonneutron target debris created from thermonuclear burn will be modified or stopped before it reaches the first reactor wall. The resulting modified spectra and pulse lengths of the debris need to be calculated to determine first wall effects. Further, the cavity overpressure created by the momentum and energy exchange between the debris and gas must also be calculated to determine its effect. The purpose of this paper is to present results of the debris-background gas problem obtained with a one fluid, two temperature plasma hydrodynamic computer code model which includes multifrequency radiation transport. Spherical symmetry, ideal gas equation of state, and LTE for each radiation frequency group were assumed. The transport of debris ions was not included and all the debris energy was assumed to be in radiation. The calculated x-ray spectra and pulse lengths and the background overpressure are presented

  5. A calculation technique of passing of a powerful relativistic beam through substance

    International Nuclear Information System (INIS)

    Pobitko, A.I.; Sal'nikov, L.I.; Sukhovitskij, E.Sh.

    1995-01-01

    The calculation algorithm of passing powerful relativistic beam through substance is developed. Algorithm of calculation is separated on the following problems: 1) a trial charge movement in electromagnetic field of the cylindrical geometry; 2) a computing of own electromagnetic field arising at movement of a particle heavy-current beam in a target; 3) accounting of an interaction of a beam with target atoms; 4) accounting of change of the target properties in a time; 5) geometry and construction of an iterative procedure of calculation. The calculation of passing heavy-current beams of charged particles for transient case is carried out by Monte Carlo method. A conclusion of equations of movement trial charge and technique of calculation own electromagnetic field of the powerful relativistic beam at passing through substance are resulted. 6 refs

  6. Refined Calculation of Beam Dynamics During UMER Injection

    CERN Document Server

    Bai, Gang; Godlove, Terry; Haber, Irving; Kishek, Rami A; Quinn, Bryan; Reiser, Martin; Thangaraj, Jayakar C T; Walter, Mark

    2005-01-01

    The University of Maryland Electron Ring (UMER) is built as a low-cost testbed for intense beam physics for benefit of larger ion accelerators. The beam intensity is designed to be variable, spanning the entire range from low current operation to highly space-charge-dominated transport. The ring has recently been closed and multi-turn commissioning has begun. Although we have conducted many experiments at high space charge during UMER construction, lower-current beams have become quite useful in this commissioning stage for assisting us with beam steering, measurement of phase advance, etc. One of the biggest challenges of multi-turn operation of UMER is correctly operating the Y-shaped injection section, hence called the Y-section, which is specially designed for UMER multi-turn operation. It is a challenge because the system requires several quadrupoles and dipoles in a very stringent space, resulting in mechanical, electrical, and beam control complexities. This paper presents a simulation study of the bea...

  7. Beam-Induced Damage Mechanisms and their Calculation

    CERN Document Server

    Bertarelli, A

    2016-01-01

    The rapid interaction of highly energetic particle beams with matter induces dynamic responses in the impacted component. If the beam pulse is sufficiently intense, extreme conditions can be reached, such as very high pressures, changes of material density, phase transitions, intense stress waves, material fragmentation and explosions. Even at lower intensities and longer time-scales, significant effects may be induced, such as vibrations, large oscillations, and permanent deformation of the impacted components. These lectures provide an introduction to the mechanisms that govern the thermomechanical phenomena induced by the interaction between particle beams and solids and to the analytical and numerical methods that are available for assessing the response of impacted components. An overview of the design principles of such devices is also provided, along with descriptions of material selection guidelines and the experimental tests that are required to validate materials and components exposed to interactio...

  8. Calculation of dynamic stresses in viscoelastic sandwich beams using oma

    DEFF Research Database (Denmark)

    Pelayo, F.; Aenlle, M. L.; Ismael, G.

    2017-01-01

    The mechanical response of sandwich elements with viscoelastic core is time and temperature dependent. Laminated glass is a sandwich element where the mechanical behavior of the glass layers is usually considered linear-elastic material whereas the core is made of an amorphous thermoplastic which...... data. In simple structures, analytical mode shapes can be used alternatively to the numerical ones. In this paper, the dynamic stresses on the glass layers of a laminated glass beam have estimated using the experimental acceleration responses measured at 7 points of the beam, and the experimental mode...

  9. 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

  10. About the possibility of use of different types of targets as a neutron source for subcritical nuclear reactor driven by particle beam accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Avdeev, E.F.; Dorokhovich, S.L.; Chusov, I.A. [Obninsk Institute of Nuclear Power Engineering (Russian Federation)

    1995-10-01

    The schemes of jet gas and liquid targets as well as the gastargets with a solid phase dispersion are introduced to use to receive the neutrons admitted to a subcritical reactor core. The possible variants of target position in the reactor are considered, target characteristics are calculated. The authors pay a great attention to the estimation of radioactive products yield receiving due to the interaction of the beam with the target.

  11. 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.)

  12. 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

  13. 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

  14. Calculation of integrated luminosity for beams stored in the Tevatron collider

    International Nuclear Information System (INIS)

    Finley, D.A.

    1989-01-01

    A model for calculating the integrated luminosity of beams stored in the Tevatron collider will be presented. The model determines the instantaneous luminosity by calculating the overlap integral of bunched beams passing through the interaction region. The calculation accounts for the variation in beam size due to the beta functions and also for effects due to finite longitudinal emittance and non-zero dispersion in the interaction region. The integrated luminosity is calculated for the beams as they evolve due to processes including collisions and intrabeam scattering. The model has been applied to both the extant and upgraded Tevatron collider, but is not limited to them. The original motivation for developing the computer model was to determine the reduction in luminosity due to beams with non-zero longitudinal emittances. There are two effects: the transverse beam size is increased where the dispersion is non-zero; the finite length of the beam bunch combined with an increasing β function results in an increased transverse beam size at the ends of the bunch. The derivation of a sufficiently useful analytic expression for the luminosity proved to be intractable. Instead, a numerical integration computer program was developed to calculate the luminosity in the presence of a finite longitudinal emittance. The program was then expanded into a model which allows the luminosity to vary due to changes in emittances and reduction in bunch intensities. At that point, it was not difficult to calculate the integrated luminosity. 5 refs., 2 figs., 4 tabs

  15. Calculated neutron spectrum from 800-MeV protons incident on a copper beam stop

    International Nuclear Information System (INIS)

    Perry, D.G.

    1975-10-01

    A Monte Carlo calculation was performed to obtain the neutron spectrum generated by 800-MeV protons incident on the LAMPF main copper beam stop. The total flux is calculated to be of the order of 10 13 n/cm 2 -sec-mA at full-beam intensity of 1 mA, with flux spectra calculated for angles of 20 0 , 30 0 , 60 0 , 90 0 , 120 0 , and 150 0 . (auth)

  16. 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

  17. 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.)

  18. 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

  19. Temperature calculations of heat loads in rotating target wheels exposed to high beam currents

    International Nuclear Information System (INIS)

    Greene, John P.; Gabor, Rachel; Neubauer, Janelle

    2001-01-01

    In heavy-ion physics, high beam currents can eventually melt or destroy the target. Tightly focused beams on stationary targets of modest melting point will exhibit short lifetimes. Defocused or 'wobbled' beams are employed to enhance target survival. Rotating targets using large diameter wheels can help overcome target melting and allow for higher beam currents to be used in experiments. The purpose of the calculations in this work is to try and predict the safe maximum beam currents which produce heat loads below the melting point of the target material

  20. Temperature calculations of heat loads in rotating target wheels exposed to high beam currents

    International Nuclear Information System (INIS)

    Greene, J. P.; Gabor, R.; Neubauer, J.

    2000-01-01

    In heavy-ion physics, high beam currents can eventually melt or destroy the target. Tightly focused beams on stationary targets of modest melting point will exhibit short lifetimes. Defocused or wobbled beams are employed to enhance target survival. Rotating targets using large diameter wheels can help overcome target melting and allow for higher beam currents to be used in experiments. The purpose of the calculations in this work is to try and predict the safe maximum beam currents which produce heat loads below the melting point of the target material

  1. 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.)

  2. Dependence of neutron rate production with accelerator beam profile and energy range in an ADS-TRIGA RC1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Firoozabadi, M.M.; Karimi, J. [Birjand Univ. (Iran, Islamic Republic of). Dept. of Physics; Zangian, M. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Nuclear Engineering Dept.

    2016-12-15

    Lead, mercury, tantalum and tungsten were used as target material for calculation of spallation processes in an ADS-TRIGA RC1 reactor. The results show that tungsten has the highest neutron production rate. Therefore it was selected as target material for further calculations. The sensitivity of neutron parameters of the ADS reactor core relative to a change of beam profile and proton energy was determined. The core assembly and parameters of the TRIGA RC1 demonstration facility were used for the calculation model. By changing the proton energy from 115 to 1 400 MeV by using the intra-nuclear cascade model of Bertini (INC-Bertini), the quantity of the relative difference in % for energy gain (G) and spallation neutron yield (Y{sub n/p}), increases to 289.99 % and 5199.15 % respectively. These changes also reduce the amount of relative difference for the proton beam current (I{sub p}) and accelerator power (P{sub acc}), 99.81 % and 81.28 % respectively. In addition, the use of a Gaussian distribution instead of a uniform distribution in the accelerator beam profile increases the quantity of relative difference for energy gain (G), net neutron multiplication (M) and spallation neutron yield (Y{sub n/p}), up to 4.93 %, 4.9 % and 5.55 % respectively.

  3. Status report of the program on neutron beam utilization at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Vuong Huu Tan

    1996-08-01

    The thermal reactor is an intense source not only of thermal neutron, but also intermediate as well as fast neutrons. Using the filtered neutron beam technique at steady state atomic reactor allows receiving the neutrons in the intermediate energy region with the most available intense flux at present. In the near time at the Dalat reactor the filtered neutron beam technique has been applied. Utilization of the filtered neutron beams in basic and applied researches has been a important activity of the Dalat Nuclear Research Institute (DNRI). This report presents some relevant characteristics of the filtered neutron beams and their utilization in nuclear data measurements, neutron capture gamma ray spectroscopy, neutron radiography, neutron dose calibration and other applications. (author). 3 refs, 2 figs

  4. 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

  5. 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

  6. 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

  7. 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)

  8. Construction of the neutron beam facility at Australia's OPAL research reactor

    International Nuclear Information System (INIS)

    Kennedy, Shane J.

    2006-01-01

    Australia's new research reactor, OPAL, has been designed principally for neutron beam science and radioisotope production. It has a capacity for 18 neutron beam instruments, located at the reactor face and in a neutron guide hall. The neutron beam facility features a 20 l liquid deuterium cold neutron source and cold and thermal supermirror neutron guides. Nine neutron beam instruments are under development, of which seven are scheduled for completion in early 2007. The project is approaching the hot-commissioning stage, when criticality will be demonstrated. Installation of the neutron beam transport system and neutron beam instruments in the neutron guide hall and at the reactor face is underway, and the path to completion of this project is relatively clear. This paper will outline the key features of the OPAL reactor, and will describe the neutron beam facility in particular. The status of the construction and a forecast of the program to completion, including commissioning and commencement of routine operation in 2007 will also be discussed

  9. 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

  10. Neutron capture therapy beams at the MIT Research Reactor

    International Nuclear Information System (INIS)

    Choi, J.R.; Clement, S.D.; Harling, O.K.; Zamenhof, R.G.

    1990-01-01

    Several neutron beams that could be used for neutron capture therapy at MITR-II are dosimetrically characterized and their suitability for the treatment of glioblastoma multiforme and other types of tumors are described. The types of neutron beams studied are: (1) those filtered by various thicknesses of cadmium, D2O, 6Li, and bismuth; and (2) epithermal beams achieved by filtration with aluminum, sulfur, cadmium, 6Li, and bismuth. Measured dose vs. depth data are presented in polyethylene phantom with references to what can be expected in brain. The results indicate that both types of neutron beams are useful for neutron capture therapy. The first type of neutron beams have good therapeutic advantage depths (approximately 5 cm) and excellent in-phantom ratios of therapeutic dose to background dose. Such beams would be useful for treating tumors located at relatively shallow depths in the brain. On the other hand, the second type of neutron beams have superior therapeutic advantage depths (greater than 6 cm) and good in-phantom therapeutic advantage ratios. Such beams, when used along with bilateral irradiation schemes, would be able to treat tumors at any depth in the brain. Numerical examples of what could be achieved with these beams, using RBEs, fractionated-dose delivery, unilateral, and bilateral irradiation are presented in the paper. Finally, additional plans for further neutron beam development at MITR-II are discussed

  11. 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)

  12. Study of dose calculation and beam parameters optimization with genetic algorithm in IMRT

    International Nuclear Information System (INIS)

    Chen Chaomin; Tang Mutao; Zhou Linghong; Lv Qingwen; Wang Zhuoyu; Chen Guangjie

    2006-01-01

    Objective: To study the construction of dose calculation model and the method of automatic beam parameters selection in IMRT. Methods: The three-dimension convolution dose calculation model of photon was constructed with the methods of Fast Fourier Transform. The objective function based on dose constrain was used to evaluate the fitness of individuals. The beam weights were optimized with genetic algorithm. Results: After 100 iterative analyses, the treatment planning system produced highly conformal and homogeneous dose distributions. Conclusion: the throe-dimension convolution dose calculation model of photon gave more accurate results than the conventional models; genetic algorithm is valid and efficient in IMRT beam parameters optimization. (authors)

  13. 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

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. 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)

  19. 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)

  20. 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)

  1. 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

  2. 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

  3. 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

  4. 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)

  5. Calculation of microplanar beam dose profiles in a tissue/lung/tissue phantom

    International Nuclear Information System (INIS)

    Company, F.Z.; Allen, B.J.

    1998-01-01

    Recent advances in synchrotron generated x-ray beams with a high fluence rate permit investigation of the application of an array of closely spaced, parallel or converging microplanar beams in radiotherapy. The proposed technique takes advantage of the hypothesized repair mechanism of capillary cells between alternate microbeam zones, which regenerates the lethally irradiated endothelial cells. The lateral and depth doses of 100 keV microplanar beams are investigated for different beam dimensions and spacings in a tissue, lung and tissue/lung/tissue phantom. The EGS4 Monte Carlo code is used to calculate dose profiles at different depths and bundles of beams (up to 20x20cm square cross section). The maximum dose on the beam axis (peak) and the minimum interbeam dose (valley) are compared at different depths, bundles, heights, widths and beam spacings. (author)

  6. Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

    International Nuclear Information System (INIS)

    Vega C, H. R.; Hernandez D, V. M.; Paredes G, L.; Aguilar, F.

    2012-10-01

    Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131±11 and 124±10 p Sv-cm 2 for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55±4 p Sv-cm 2 for 10 W. (Author)

  7. Optics calculations and beam line design for the JANNuS facility in Orsay

    International Nuclear Information System (INIS)

    Chauvin, N.; Henry, S.; Flocard, H.; Fortuna, F.; Kaitasov, O.; Pariset, P.; Pellegrino, S.; Ruault, M.O.; Serruys, Y.; Trocelier, P.

    2007-01-01

    JANNuS (Joint Accelerators for Nano-Science and Nuclear Simulation) will be a unique user facility in Europe dedicated to material modification by ion beam implantation and irradiation. The main originality of the project is that it will be possible to perform implantation and irradiation with simultaneous multiple ions beams and in situ characterization by transmission electron microscopy (TEM) observation or ion beam analysis. This facility will be composed of two experimental platforms located in two sites: the CEA-SRMP in Saclay and the CNRS-CSNSM in Orsay. This paper will focus on the design of two new transport beam lines for the Orsay site. One of the most challenging parts of the JANNuS project (Orsay site) is to design two new beam lines in order to inject, into a 200 kV TEM, two different ion beams (low and medium energy) coming from two existing pieces of equipment: a 2 MV Tandem accelerator and a 190 kV ion implanter. For these new beam lines, first order beam calculations have been done using transfer matrix formalism. A genetic algorithm has been written and adapted to perform the optimization of the beam line parameters. Then, using the SIMION code, field maps of the electrostatic elements (quadrupoles, spherical sectors) have been calculated and ion trajectories have been simulated. We studied specifically the optical aberrations induced by the electrostatic spherical deflectors. Finally, the results of the first order calculations and the field map simulations show a good agreement

  8. Installation and testing of an optimized epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR)

    Energy Technology Data Exchange (ETDEWEB)

    Fairchild, R.G.; Kalef-Ezra, J.; Saraf, S.K.; Fiarman, S.; Ramsey, E.; Wielopolski, L.; Laster, B.; Wheeler, F. (Brookhaven National Lab., Upton, NY (USA); Ioannina Univ. (Greece); Brookhaven National Lab., Upton, NY (USA); State Univ. of New York, Stony Brook, NY (USA). Health Science Center; Brookhaven National Lab., Upton, NY (USA); EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1989-01-01

    Various calculations indicate that an optimized epithermal neutron beam can be produced by moderating fission neutrons either with a combination of Al and D{sub 2}O, or with Al{sub 2}O{sub 3}. We have designed, installed and tested an Al{sub 2}O{sub 3} moderated epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR). The epithermal neutron fluence rate of 1.8 {times} 10{sup 9} n/cm{sup 2}-sec produces a peak thermal neutron fluence rate of 1.9 to 2.8 {times} 10{sup 9} n/cm{sup 2}-sec in a tissue equivalent (TE) phantom head, depending on the configuration. Thus a single therapy treatment of 5 {times} 10{sup 12} n/cm{sup 2} can be delivered in 30--45 minutes. All irradiation times are given for a BMRR power of 3 MW, which is the highest power which can be delivered continuously. 18 refs., 8 figs., 4 tabs.

  9. 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

  10. 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)

  11. Radiation protection commissioning of neutron beam instruments at the OPAL research reactor

    International Nuclear Information System (INIS)

    Parkes, Alison; Saratsopoulos, John; Deura, Michael; Kenny, Pat

    2008-01-01

    The neutron beam facilities at the 20 MW OPAL Research Reactor were commissioned in 2007 and 2008. The initial suite of eight neutron beam instruments on two thermal neutron guides, two cold neutron guides and one thermal beam port located at the reactor face, together with their associated shielding were progressively installed and commissioned according to their individual project plans. Radiation surveys were systematically conducted as reactor power was raised in a step-wise manner to 20 MW in order to validate instrument shielding design and performance. The performance of each neutron guide was assessed by neutron energy spectrum and flux measurements. The activation of beam line components, decay times assessments and access procedures for Bragg Institute beam instrument scientists were established. The multiple configurations for each instrument and the influence of operating more than one instrument or beamline simultaneously were also tested. Areas of interest were the shielding around the secondary shutters, guide shield and bunker shield interfaces and monochromator doors. The shielding performance, safety interlock checks, improvements, radiation exposures and related radiation protection challenges are discussed. This paper discusses the health physics experience of commissioning the OPAL Research Reactor neutron beam facilities and describes health physics results, actions taken and lessons learned during commissioning. (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. 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)

  15. 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)

  16. 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.)

  17. 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

  18. 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.

  19. 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

  20. 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.)

  1. 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

  2. Benchmarking of Touschek Beam Lifetime Calculations for the Advanced Photon Source

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, A.; Yang, B.

    2017-06-25

    Particle loss from Touschek scattering is one of the most significant issues faced by present and future synchrotron light source storage rings. For example, the predicted, Touschek-dominated beam lifetime for the Advanced Photon Source (APS) Upgrade lattice in 48-bunch, 200-mA timing mode is only ~ 2 h. In order to understand the reliability of the predicted lifetime, a series of measurements with various beam parameters was performed on the present APS storage ring. This paper first describes the entire process of beam lifetime measurement, then compares measured lifetime with the calculated one by applying the measured beam parameters. The results show very good agreement.

  3. 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

  4. 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.

  5. 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)

  6. 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.

  7. 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.)

  8. 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

  9. A computer code 'BEAM' for the ion optics calculation of the JAERI tandem accelerator system

    International Nuclear Information System (INIS)

    Kikuchi, Shiroh; Takeuchi, Suehiro

    1987-11-01

    The computer code BEAM is described, together with an outline of the formalism used for the ion optics calculation. The purpose of the code is to obtain the optimum parameters of devices, with which the ion beam is transported through the system without losses. The procedures of the calculation, especially those of searching for the parameters of quadrupole lenses, are discussed in detail. The flow of the code is illustrated as a whole and its constituent subroutines are explained individually. A few resultant beam trajectories and the parameters used to obtain them are shown as examples. (author)

  10. Stress analysis of neutral beam pivot point bellows for tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Johnson, J.J.; Benda, B.J.; Tiong, L.W.

    1983-01-01

    The neutral beam pivot point bellows serves as an airtight flexible linkage between the torus duct and the neutral beam transition duct in Princeton University's Tokamak Fusion Test Reactor. The bellows considered here is basically rectangular in cross section with rounded corners; a unique shape. Its overall external dimensions are about 28 in. (about 711 mm) X about 35 in. (about 889 mm). The bellows is formed from 18 convolutions and is of the nested ripple type. It is about 11 in. (about 43.3 mm) in length, composed of Inconel 718, and each leaf has a thickness of 0.034 in. (.86 mm). The bellows is subjected to a series of design loading conditions -- vacuum, vacuum + 2 psi (.12 MPa), vacuum + stroke (10,000 cycles), vacuum + temperature increase + extension, extension to a stress of 120 ksi (838 MPa), and a series of rotational loading conditions induced in the bellows by alignment of the neutral beam injector. A stress analysis of the bellows was performed by the finite element method -- locations and magnitude of maximum stresses were calculated for all of the design loading conditions to check with allowable values and help guide placement of strain gauges during proof testing. A typical center convolution and end convolution were analyzed. Loading conditions were separated into symmetric and antisymmetric cases about the planes of symmetry of the cross-section. Iterative linear analyses were performed, i.e. compressive loading conditions led to predicted overlap of the leaves from linear analysis and restraints were added to prevent such overlap. This effect was found to be substantial in stress predicition and necessary to be taken into account. A total of eleven loading conditions and seven models were analyzed. The results showed peak stresses to be within allowable limits and the number of allowable cycles to be greater than the design condition

  11. 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

  12. Program for calculating multi-component high-intense ion beam transport

    International Nuclear Information System (INIS)

    Kazarinov, N.Yu.; Prejzendorf, V.A.

    1985-01-01

    The CANAL program for calculating transport of high-intense beams containing ions with different charges in a channel consisting of dipole magnets and quadrupole lenses is described. The equations determined by the method of distribution function momenta and describing coordinate variations of the local mass centres and r.m.s. transverse sizes of beams with different charges form the basis of the calculation. The program is adapted for the CDC-6500 and SM-4 computers. The program functioning is organized in the interactive mode permitting to vary the parameters of any channel element and quickly choose the optimum version in the course of calculation. The calculation time for the CDC-6500 computer for the 30-40 m channel at the integration step of 1 cm is about 1 min. The program is used for calculating the channel for the uranium ion beam injection from the collective accelerator into the heavy-ion synchrotron

  13. International Thermonuclear Experimental Reactor (ITER) neutral beam design

    International Nuclear Information System (INIS)

    Myers, T.J.; Brook, J.W.; Spampinato, P.T.; Mueller, J.P.; Luzzi, T.E.; Sedgley, D.W.

    1990-10-01

    This report discusses the following topics on ITER neutral beam design: ion dump; neutralizer and module gas flow analysis; vacuum system; cryogenic system; maintainability; power distribution; and system cost

  14. Studies on the instrumentation of a beam-tube medium flux reactor

    International Nuclear Information System (INIS)

    Axmann, A.; Pollet, J.L.; Queudot, J.

    1979-01-01

    In the years 1977/78, the ad hoc commitee for medium-flux reactor development of the Federal Ministry for Research and Technology developed constructional concepts for a medium-flux reactor to be utilized by beam tube experiments. The HMI has elaborated contributions for discussions of the subject of instrumentation, in particular for experiments in solid state physics. These contributions are contained in the report. (orig./RW) [de

  15. Structural biology facilities at Brookhaven National Laboratory`s high flux beam reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korszun, Z.R.; Saxena, A.M.; Schneider, D.K. [Brookhaven National Laboratory, Upton, NY (United States)

    1994-12-31

    The techniques for determining the structure of biological molecules and larger biological assemblies depend on the extent of order in the particular system. At the High Flux Beam Reactor at the Brookhaven National Laboratory, the Biology Department operates three beam lines dedicated to biological structure studies. These beam lines span the resolution range from approximately 700{Angstrom} to approximately 1.5{Angstrom} and are designed to perform structural studies on a wide range of biological systems. Beam line H3A is dedicated to single crystal diffraction studies of macromolecules, while beam line H3B is designed to study diffraction from partially ordered systems such as biological membranes. Beam line H9B is located on the cold source and is designed for small angle scattering experiments on oligomeric biological systems.

  16. Neutron beam utilization at the TRIGA Mark II reactor Vienna

    International Nuclear Information System (INIS)

    Villa, M.; Boeck, H.; Ismail, S.; Koerner, S.; Baron, M.; Hainbuchner, M.; Badurek, G.; Buchelt, R.J.

    1999-01-01

    A review is given about the research activities around the 250 kw TRIGA reactor Vienna, which are adequate to other neutron sources of comparable or bigger size. The topics selected for presentation range from neutron radiography, materials irradiation, neutron small-angle scattering, neutron activation analysis, neutron polarization to neutron interferometry. It is the aim of this presentation to stimulate programs for more efficient use around TRIGA research reactors with neutron flux densities of 1013 cm-2a-1 at the center of the reactor core. We briefly describe the experimental facilities installed at the 250 kw TRIGA reactor of the Austrian Universities in Vienna and present a great part of the current research activities performed with them. We believe that most of the techniques and experiments presented here are adequate for implementation to other reactors of similar or even higher power. Those technologies which require extremely specialized know-how not generally available at every research Inst.e will not be treated here or are just mentioned without any further details.(author)

  17. 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

  18. 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)

  19. 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.

  20. Requirements of a proton beam accelerator for an accelerator-driven reactor

    International Nuclear Information System (INIS)

    Takahashi, H.; Zhao, Y.; Tsoupas, N.; An, Y.; Yamazaki, Y.

    1997-01-01

    When the authors first proposed an accelerator-driven reactor, the concept was opposed by physicists who had earlier used the accelerator for their physics experiments. This opposition arose because they had nuisance experiences in that the accelerator was not reliable, and very often disrupted their work as the accelerator shut down due to electric tripping. This paper discusses the requirements for the proton beam accelerator. It addresses how to solve the tripping problem and how to shape the proton beam

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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)

  7. Measurements and calculations for determination of discharge of 41Ar from IFEs research reactor JEEP II at Kjeller, Norway

    International Nuclear Information System (INIS)

    Raaum, A.; Straelberg, E.

    2003-01-01

    41 Ar is formed by neutron irradiation of 40 Ar, which occurs naturally in air with a concentration of 9300 ppm. The discharge of 41 Ar from IFEs research reactor Jeep II is yearly reported to the Norwegian Radiation Protection Authority (NRPA). Until year 2000 the reported values were based on theoretical calculations of produced 41 Ar per operating hour of 6.8 GBq/h. During 2000 and 2001 the reactor was upgraded to increase the irradiation capacity and to meet the markets demand for irradiation of 5'Si-crystalls. After the upgrading, measurements and calculations were initiated to determine the new discharge rate for 41 Ar. During reactor operation an approximately constant discharge of 41 Ar is expected, mainly due to irradiation of air in open beam channels. In addition 41 Ar is released from irradiation pockets when they are opened to transfer samples in and out during reactor stop. The new value for discharge rate was determined from measurements of air samples from the discharge channel during operation and theoretical calculations of the release from the irradiation pockets. The new discharge rate was determined to 5.9 ± 0.5 GBq/h, which is a small reduction compared to the former value of 6.8 GBq/h. A small reduction in discharge rate was expected because the number of air-filled irradiation pockets was reduced after the upgrading. In a normal year the discharge of 41 Ar will be about 2 % of the Institutes discharge permission. (orig.)

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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.)

  13. 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

  14. 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)

  15. 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

  16. 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.)

  17. 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

  18. Beam shape coefficients calculation for an elliptical Gaussian beam with 1-dimensional quadrature and localized approximation methods

    Science.gov (United States)

    Wang, Wei; Shen, Jianqi

    2018-06-01

    The use of a shaped beam for applications relying on light scattering depends much on the ability to evaluate the beam shape coefficients (BSC) effectively. Numerical techniques for evaluating the BSCs of a shaped beam, such as the quadrature, the localized approximation (LA), the integral localized approximation (ILA) methods, have been developed within the framework of generalized Lorenz-Mie theory (GLMT). The quadrature methods usually employ the 2-/3-dimensional integrations. In this work, the expressions of the BSCs for an elliptical Gaussian beam (EGB) are simplified into the 1-dimensional integral so as to speed up the numerical computation. Numerical results of BSCs are used to reconstruct the beam field and the fidelity of the reconstructed field to the given beam field is estimated. It is demonstrated that the proposed method is much faster than the 2-dimensional integrations and it can acquire more accurate results than the LA method. Limitations of the quadrature method and also the LA method in the numerical calculation are analyzed in detail.

  19. The Intense Slow Positron Beam Facility at the NC State University PULSTAR Reactor

    International Nuclear Information System (INIS)

    Hawari, Ayman I.; Moxom, Jeremy; Hathaway, Alfred G.; Brown, Benjamin; Gidley, David W.; Vallery, Richard; Xu, Jun

    2009-01-01

    An intense slow positron beam is in its early stages of operation at the 1-MW open-pool PULSTAR research reactor at North Carolina State University. The positron beam line is installed in a beam port that has a 30-cmx30-cm cross sectional view of the core. The positrons are created in a tungsten converter/moderator by pair-production using gamma rays produced in the reactor core and by neutron capture reactions in cadmium cladding surrounding the tungsten. Upon moderation, slow (∼3 eV) positrons that are emitted from the moderator are electrostatically extracted, focused and magnetically guided until they exit the reactor biological shield with 1-keV energy, approximately 3-cm beam diameter and an intensity exceeding 6x10 8 positrons per second. A magnetic beam switch and transport system has been installed and tested that directs the beam into one of two spectrometers. The spectrometers are designed to implement state-of-the-art PALS and DBS techniques to perform positron and positronium annihilation studies of nanophases in matter.

  20. Underwater laser beam welding technology for reactor vessel nozzles of PWRs

    International Nuclear Information System (INIS)

    Yoda, Masaki; Tamura, Masataka; Tamura, Masataka

    2010-01-01

    Toshiba has developed an underwater laser beam welding technology for the maintenance of reactor vessel nozzles of pressurized water reactors (PWRs), which eliminates the need for the drainage of water from the reactor vessel. The new welding system makes it possible to both reduce the work period and minimize the radiation exposure of workers compared with conventional technologies for welding in ambient air. We have confirmed the effectiveness of this technology through experiments in which stress corrosion cracking (SCC) was mitigated on the inner surfaces of nozzles. We are promoting its practical application in Japan and overseas in cooperation with Westinghouse Electric Company, a group company of Toshiba. (author)

  1. Comparison of calculated neutral beam shine through with measured shine-through in DIII-D

    International Nuclear Information System (INIS)

    Chiu, H.K.; Hong, R.

    1997-11-01

    A comparison of the calculated shine through of neutral particle beams in the DIII-D plasma to measured values inferred from the target temperature rise is reported. This provides an opportunity to verify the shine through calculations and makes them more reliable in those cases where the shine through can not be measured. The DIII-D centerpost neutral beam target tiles are safe-guarded against excessive beam shine-through by pyrometry and thermocouple (TC) arrays on the tiles. Shine-through beam power is calculated from the measured temperature changes reported by the target tile TC array. These measurements are performed at the beginning of each operational year at DIII-D. Theoretically, the beam energy deposited into the plasma can be expressed as a function of the change in beam density. Neutral beam energy deposition in plasma (of known density) is inferred by comparing the results of a series of shine-through measurements for the 1997 campaign at DIII-D to the expected shine-through given by theory

  2. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    Energy Technology Data Exchange (ETDEWEB)

    Zamani, M. [National Radiation Protection Department - NRPD, Atomic Energy Organization of Iran - AEOI, Tehran (Iran, Islamic Republic of); End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Kasesaz, Y.; Khalafi, H.; Shayesteh, M. [Radiation Application School, Nuclear Science and Technology Research Institute, AEOI, Tehran (Iran, Islamic Republic of)

    2015-07-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  3. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    International Nuclear Information System (INIS)

    Zamani, M.; Kasesaz, Y.; Khalafi, H.; Shayesteh, M.

    2015-01-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  4. 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

  5. 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

  6. 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

  7. 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)

  8. 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)

  9. 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)

  10. 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

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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)

  16. Calculational investigations and analysis of characteristics of research reactor WWR-M as a source of neutrons for solution of scientific and applied tasks

    International Nuclear Information System (INIS)

    Vorona, P.M.; Razbudej, V.F.

    2010-01-01

    Calculational studies and analysis of the neutron fields of WWR-M research reactor of the Institute for Nuclear Research, National Academy of Sciences of Ukraine, as a basic nuclear facility for performing the fundamental and applied investigations and for experimentalindustrial production of radioisotope products for various spheres of application are carried out. The calculations are carried out by the method of statistic tests (Monte Carlo) applying the computer program MCNP-4C. The data on the spectra and the neutron flux density values at the 10 MW reactor power for all technological facilities designed for the works with neutrons: 19 vertical experimental channels for irradiation of specimens and 10 horizontal channels for beams extraction from the reactor are obtained. The effect of the neutron traps (water cavities) mounted in the core on the characteristics of the extracted from the reactor beams is demonstrated. Recommendations associated with optimization of the reactor core are adduced for amplification of its capabilities as a neutron source in experimental researches.

  17. 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.

  18. 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.

  19. Single-crystal filters for attenuating epithermal neutrons and gamma rays in reactor beams

    DEFF Research Database (Denmark)

    Rustad, B.M.; Als-Nielsen, Jens Aage; Bahnsen, A.

    1965-01-01

    Cross section of representative samples of bismuth and quartz were measured at room and liquid nitrogen temperatures over neutron energy range of 0.0007 to 2.0 ev to obtain data for design of single-crystal 32-cm bismuth filters for attenuating fast neutrons and γ-rays in reactor beams; filters may...

  20. Calculating the Carrying Capacity of Flexural Prestressed Concrete Beams with Non-Metallic Reinforcement

    Directory of Open Access Journals (Sweden)

    Mantas Atutis

    2011-04-01

    Full Text Available The article reviews moment resistance design methods of prestressed concrete beams with fibre-reinforced polymer (FRP reinforcement. FRP tendons exhibit linear elastic response to rupture without yielding and thus failure is expected to be brittle. The structural behaviour of beams prestressed with FRP tendons is different from beams with traditional steel reinforcement. Depending on the reinforcement ratio, the flexural behaviour of the beam can be divided into several groups. The numerical results show that depending on the nature of the element failure, moment resistance calculation results are different by using reviewed methods. It was found, that the use of non-metallic reinforcement in prestressed concrete structures is effective: moment capacity is about 5% higher than that of the beams with conventional steel reinforcement.Article in Lithuanian

  1. Sub 100 nm proton beam micromachining: theoretical calculations on resolution limits

    International Nuclear Information System (INIS)

    Kan, J.A. van; Sum, T.C.; Osipowicz, T.; Watt, F.

    2000-01-01

    Proton beam micromachining is a novel direct-write process for the production of three-dimensional (3D) microstructures. A focused beam of MeV protons is scanned in a pre-determined pattern over a suitable resist material (e.g. PMMA or SU-8) and the latent image formed is subsequently developed chemically. In this paper calculations on theoretical resolution limits of proton beam micromachined three-dimensional microstructures are presented. Neglecting the finite beam size, a Monte Carlo ion transport code was used in combination with a theoretical model describing the delta-ray (δ-ray) energy deposition to determine the lateral energy deposition distribution in PMMA resist material. The energy deposition distribution of ion induced secondary electrons (δ-rays) has been parameterized using analytical models. It is assumed that the attainable resolution is limited by a convolution of the spread of the ion beam and energy deposition of the δ-rays

  2. 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)

  3. New north beam tube for the neutron radiography reactor

    International Nuclear Information System (INIS)

    Pruett, D.P.; Richards, W.J.; Heidel, C.C.

    1982-01-01

    Neutron radiography of the fuel undergoing examination in the argon cell is performed in the NRAD Facility and is one of many examinations performed on the fuel. The reactor and examination procedure are described. The new radiography system, developed to expand the present radiography capabilities to radiograph both irradiated and unirradiated specimens and to provide for the development of new radiography techniques without interfering with the argon cell production schedule is presented

  4. 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

  5. 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)

  6. 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

  7. The nuclear heating calculation scheme for material testing in the future Jules Horowitz Reactor

    International Nuclear Information System (INIS)

    Huot, N.; Aggery, A.; Blanchet, D.; Courcelle, A.; Czernecki, S.; Di-Salvo, J.; Doederlein, C.; Serviere, H.; Willermoz, G.

    2004-01-01

    An innovative nuclear heating calculation scheme for materials testing carried out in in the future Jules Horowitz reactor (JHR) is described. A heterogeneous gamma source calculation is first performed at assembly level using the deterministic code APOLLO2. This is followed by a Monte Carlo gamma transport calculation in the whole core using the TRIPOLI4 code. The calculated gamma sources at the assembly level are applied in the whole core simulation using a weighting based on power distribution obtained from the neutronic core calculation. (authors)

  8. Calculation of ion storage in electron beams with account of ion-ion interactions

    International Nuclear Information System (INIS)

    Perel'shtejn, Eh.A.; Shirkov, G.D.

    1979-01-01

    Ion storage in relativistic electron beams was calculated taking account of ion-ion charge exchange and ionization. The calculations were made for nitrogen ion storage from residual gas during the compression of electron rings in the adhezator of the JINR heavy ion accelerator. The calculations were made for rings of various parameters and for various pressures of the residual gas. The results are compared with analogous calculations made without account of ion-ion processes. It is shown that at heavy loading of a ring by ions ion-ion collisions play a significant part, and they should be taken into account while calculating ion storage

  9. Calculations of the beam transport through the low energy side of the Lund Pelletron accelerator

    International Nuclear Information System (INIS)

    Dymnikov, A.; Hellborg, R.; Pallon, J.; Skog, G.; Yang, C.

    1993-01-01

    A new recursive technique has been used to solve the equations of motion of charged particles in electric and magnetic fields taking into account the effect of space charge. Based on this technique a computer code has been written and calculations have been carried out for the beam optics, from the ion-source to the terminal, stripper of the Lund Pelletron tandem accelerator. The code has been found capable of describing the beam-optics of the existing setup and will in future be used together with a library of typical field descriptions to design new beam lines. (orig.)

  10. Telescope-based cavity for negative ion beam neutralization in future fusion reactors.

    Science.gov (United States)

    Fiorucci, Donatella; Hreibi, Ali; Chaibi, Walid

    2018-03-01

    In future fusion reactors, heating system efficiency is of the utmost importance. Photo-neutralization substantially increases the neutral beam injector (NBI) efficiency with respect to the foreseen system in the International Thermonuclear Experimental Reactor (ITER) based on a gaseous target. In this paper, we propose a telescope-based configuration to be used in the NBI photo-neutralizer cavity of the demonstration power plant (DEMO) project. This configuration greatly reduces the total length of the cavity, which likely solves overcrowding issues in a fusion reactor environment. Brought to a tabletop experiment, this cavity configuration is tested: a 4 mm beam width is obtained within a ≃1.5  m length cavity. The equivalent cavity g factor is measured to be 0.038(3), thus confirming the cavity stability.

  11. Analytical and numerical calculations of resistive wall impedances for thin beam pipe structures at low frequencies

    Energy Technology Data Exchange (ETDEWEB)

    Niedermayer, U., E-mail: u.niedermayer@gsi.de [Technische Universitaet Darmstadt, Institut fuer Theorie Elektromagnetischer Felder, Schlossgartenstrasse 8, 64289 Darmstadt (Germany); Boine-Frankenheim, O. [Technische Universitaet Darmstadt, Institut fuer Theorie Elektromagnetischer Felder, Schlossgartenstrasse 8, 64289 Darmstadt (Germany)

    2012-09-21

    The resistive wall impedance is one of the main sources for beam instabilities in synchrotrons and storage rings. The fast ramped SIS18 synchrotron at GSI and the projected SIS100 synchrotron for FAIR both employ thin (0.3 mm) stainless steel beam pipes in order to reduce eddy current effects. The lowest betatron sidebands are at about 100 kHz, which demands accurate impedance predictions in the low frequency (LF) range where the beam pipe and possibly also the structures behind the pipe are the dominating impedance sources. The longitudinal and transverse resistive wall impedances of a circular multi-layer pipe are calculated analytically using the field matching technique. We compare the impedances obtained from a radial wave model, which corresponds to the setup used in bench measurements, with the axial wave model, which corresponds to an actual beam moving with relativistic velocity. For thin beam pipes the induced wall current and the corresponding shielding properties of the pipe are important. In both models the wall current is obtained analytically. The characteristic frequencies for the onset of the wall current are calculated from equivalent lumped element circuits corresponding to the radial model. For more complex structures, like the SIS100 beam pipe, we use a numerical method, in which the impedance is obtained from the total power loss. The method is validated by the analytic expressions for circular beam pipes.

  12. 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.

    1980-01-01

    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: (a) 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. (b) Theoretical studies of some effects, including the following: (i) The effects of extrapolating to fuel operating temperature; (ii) Fuel-cycle and burnup effects, including the gradual replacement through a fuel cycle of control-rod absorption by fission product absorption, the loss of fissile material and the change in fuel nuclide relative composition; (iii) The heterogeneity effects of large fuelled subassemblies in pin geometry. (c) Theoretical studies of approximations in the calculational methods, including the following: (i) The importance in the whole reactor calculation of the energy group structure and the spatial mesh, including comparisons of calculations in two (RZ) and three-dimensional geometry; (ii) The importance of reactor material boundaries in the calculation of resonance shielding effects; (iii) The use of neutron fluxes calculated using neutron diffusion theory rather than transport theory. (author)

  13. Calculation of the effectiveness of manual control rods for the reactor of Ignalina NPP Unit 2

    International Nuclear Information System (INIS)

    Bubelis, E.; Pabarcius, R.

    2001-01-01

    On the basis of one of the recent databases of the reactor of Ignalina NPP Unit 2, calculations of the effectiveness of separate manual control rods, groups of manual control rods and axial characteristic of effectiveness of separate manual control rods were performed. The results of the calculations indicated, that all analyzed separate manual control rods have approximately the same effectiveness, which doesn't depend on the location of a control rod in the reactor core layout Manual control rod of the new design has about 10% greater effectiveness than manual control rod of the old design. (author)

  14. Comparison of KANEXT and SERPENT for fuel depletion calculations of a sodium fast reactor

    International Nuclear Information System (INIS)

    Lopez-Solis, R.C.; Francois, J.L.; Becker, M.; Sanchez-Espinoza, V.H.

    2014-01-01

    As most of Generation-IV systems are in development, efficient and reliable computational tools are needed to obtain accurate results in reasonably computer time. In this study, KANEXT code system is presented and validated against the well-known Monte Carlo SERPENT code, for fuel depletion calculations of a sodium fast reactor (SFR). The KArlsruhe Neutronic EXtended Tool (KANEXT) is a modular code system for deterministic reactor calculations, consisting of one kernel and several modules. Results obtained with KANEXT for the SFR core are in good agreement with the ones of SERPENT, e.g. the neutron multiplication factor and the isotopes evolution with burn-up. (author)

  15. The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

    International Nuclear Information System (INIS)

    Atkinson, C.A.; Collins, P.J.

    1987-01-01

    Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

  16. Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

    Science.gov (United States)

    Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

    2012-08-01

    The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

  17. Methods for calculating energy and current requirements for industrial electron beam processing

    International Nuclear Information System (INIS)

    Cleland, M.R.; Farrell, J.P.

    1976-01-01

    The practical problems of determining electron beam parameters for industrial irradiation processes are discussed. To assist the radiation engineer in this task, the physical aspects of electron beam absorption are briefly described. Formulas are derived for calculating the surface dose in the treated material using the electron energy, beam current and the area thruput rate of the conveyor. For thick absorbers electron transport results are used to obtain the depth-dose distributions. From these the average dose in the material, anti D, and the beam power utilization efficiency, F/sub p/, can be found by integration over the distributions. These concepts can be used to relate the electron beam power to the mass thruput rate. Qualitatively, the thickness of the material determines the beam energy, the area thruput rate and surface dose determine the beam current while the mass thruput rate and average depth-dose determine the beam power requirements. Graphs are presented showing these relationships as a function of electron energy from 0.2 to 4.0 MeV for polystyrene. With this information, the determination of electron energy and current requirements is a relatively simple procedure

  18. A pulsed neutron monochromatic beam at the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Adib, M.; Abdel-Kawy, A.; Eid, Y.; Maayouf, R.M.A.

    1985-01-01

    A pulsed neutron monochromatic beam, at the ET-RR-1 reactor, is produced by two 32 cm diameter rotors suspended in magnetic fields, whose centres are 126 cm apart rotating at speeds up to 16,000 rev/min. Each of the rotors has two slots, which are of constant cross-section in area - 7x10mm 2 , and are curved so that they have a maximum transmission for neutrons whose speed is 8.2 times that of the rotor tip. The jitter of the phase between the rotors at different rotation rates is found not to exceed +-1 μs. It has been found that both the observed time distribution and the TOF distribution of the neutrons at different rotation rates are in good agreement with the calculated ones. The observed intensity of the monochromatic neutrons of wavelength 2.74+-0.09 A, obtained by the rotors rotating at a speed of 10,500 rev/min with 864+-1 μs difference in phase between them, is 66.8 n/s. This value is found to be less than the predicted one by a factor of 5.5. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    domains will be expanded and the validation base of commonlyused calculation methods will be expanded to cover a new range of research reactor types. From a practical perspective, CROCUS and the UFTR will have fully validated reactor dynamic and transient models for dynamic and accident analysis. With these validated models, both facilities will have improved capabilities and flexibility for extended operations in the future. CROCUS and the UFTR will be able to make future reactor modifications with reduced regulatory resistance. A feasibility analysis of future power uprates at these facilities will also result.

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

    International Nuclear Information System (INIS)

    Jordan, K. A.; Schubring, D.; Girardin, G.; Pautz, A.

    2013-01-01

    domains will be expanded and the validation base of commonlyused calculation methods will be expanded to cover a new range of research reactor types. From a practical perspective, CROCUS and the UFTR will have fully validated reactor dynamic and transient models for dynamic and accident analysis. With these validated models, both facilities will have improved capabilities and flexibility for extended operations in the future. CROCUS and the UFTR will be able to make future reactor modifications with reduced regulatory resistance. A feasibility analysis of future power uprates at these facilities will also result

  1. Yield calculations for a facility for short-lived nuclear beams

    CERN Document Server

    Jiang, C L; Gomes, I; Heinz, A M; Nolen, Jerry A; Rehm, K E; Savard, G; Schiffer, J P

    2002-01-01

    Yields for a broad range of radioactive nuclei produced by spallation reactions, neutron-induced fission, in-flight projectile fragmentation and in-flight fission have been calculated for beams of stable nuclei at energies of 100-1000 MeV/u. Calculations of cross-sections and yields, attenuation effects due to absorption, production from secondary reactions, and transport efficiencies for mass selection are discussed. Rare isotope yields as functions of bombarding energies for both reaccelerated and directly produced fast-fragmentation beams are presented. This information provides a foundation for a cost-effective design of a next generation rare isotope accelerator.

  2. Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)

  3. Upgrades of the epithermal neutron beam at the Brookhaven Medical Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Hungyuan B.; Brugger, R.M.; Rorer, D.C.

    1994-12-31

    The first epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR) was installed in 1988 and produced a neutron beam that was satisfactory for the development of NCT with epithermal neutrons. This beam was used routinely until 1992 when the beam was upgraded by rearranging fuel elements in the reactor core to achieve a 50% increase in usable flux. Next, after computer modeling studies, it was proposed that the Al and Al{sub 2}O{sub 3} moderator material in the shutter that produced the epithermal neutrons could be rearranged to enhance the beam further. However, this modification was not started because a better option appeared, namely to use fission plates to move the source of fission neutrons closer to the moderator and the patient irradiation position to achieve more efficient moderation and production of epithermal neutrons. A fission plate converter (FPC) source has been designed recently and, to test the concept, implementation of this upgrade has started. The predicted beam parameters will be 12 x 10{sup 9} n{sub epi}/cm{sup 2}sec accompanying with doses from fast neutrons and gamma rays per epithermal neutron of 2.8 x 10{sup -11} and < 1 x 10{sup -11} cGycm{sup 2}/n, respectively, and a current-to-flux ratio of epithermal neutrons of 0.78. This conversion could be completed by late 1996.

  4. Comparison of Several Thermal Conductivity Constants for Thermal Hydraulic Calculation of Pebble Bed Reactor

    Science.gov (United States)

    Irwanto, Dwi; Setiadipura, Topan; Pramutadi, Asril

    2017-07-01

    There are two type of High Temperature Gas Reactor (HTGR), prismatic and pebble bed. Pebble Bed type has unique configuration because the fuels are randomly distributed inside the reactor core. In term of safety features, Pebble Bed Reactor (PBR) is one of the most promising reactor type in avoiding severe nuclear accidents. In order to analyze heat transfer and safety of this reactor type, a computer code is now under development. As a first step, calculation method proposed by Stroh [1] is adopted. An approach has been made to treat randomly distributed pebble balls contains fissile material inside the reactor core as a porous medium. Helium gas act as coolant on the reactor system are carrying heat flowing in the area between the pebble balls. Several parameters and constants are taken into account in the new developed code. Progress of the development of the code especially comparison of several thermal conductivity constants for a certain PBR-case are reported in the present study.

  5. Total Absorption Spectroscopy of Fission Fragments Relevant for Reactor Antineutrino Spectra and Decay Heat Calculations

    Directory of Open Access Journals (Sweden)

    Porta A.

    2016-01-01

    Full Text Available Beta decay of fission products is at the origin of decay heat and antineutrino emission in nuclear reactors. Decay heat represents about 7% of the reactor power during operation and strongly impacts reactor safety. Reactor antineutrino detection is used in several fundamental neutrino physics experiments and it can also be used for reactor monitoring and non-proliferation purposes. 92,93Rb are two fission products of importance in reactor antineutrino spectra and decay heat, but their β-decay properties are not well known. New measurements of 92,93Rb β-decay properties have been performed at the IGISOL facility (Jyväskylä, Finland using Total Absorption Spectroscopy (TAS. TAS is complementary to techniques based on Germanium detectors. It implies the use of a calorimeter to measure the total gamma intensity de-exciting each level in the daughter nucleus providing a direct measurement of the beta feeding. In these proceedings we present preliminary results for 93Rb, our measured beta feedings for 92Rb and we show the impact of these results on reactor antineutrino spectra and decay heat calculations.

  6. Selection method and device for reactor core performance calculation input indication

    International Nuclear Information System (INIS)

    Yuto, Yoshihiro.

    1994-01-01

    The position of a reactor core component on a reactor core map, which is previously designated and optionally changeable, is displayed by different colors on a CRT screen by using data of a data file incorporating results of a calculation for reactor core performance, such as incore thermal limit values. That is, an operator specifies the kind of the incore component to be sampled on a menu screen, to display the position of the incore component which satisfies a predetermined condition on the CRT screen by different colors in the form of a reactor core map. The position for the reactor core component displayed on the CRT screen by different colors is selected and designated on the screen by a touch panel, a mouse or a light pen, thereby automatically outputting detailed data of evaluation for the reactor core performance of the reactor core component at the indicated position. Retrieval of coordinates of fuel assemblies to be data sampled and input of the coordinates and demand for data sampling can be conducted at once by one menu screen. (N.H.)

  7. Neutronics calculations for denatured molten salt reactors: Assessing resource requirements and proliferation-risk attributes

    International Nuclear Information System (INIS)

    Ahmad, Ali; McClamrock, Edward B.; Glaser, Alexander

    2015-01-01

    Highlights: • We study the proliferation-risk and resource attributes of denatured MSRs. • MSRs offer significantly better resource efficiency compared to light-water reactors. • Denatured single-fluid MSRs reactors offer promising non-proliferation attributes. - Abstract: Molten salt reactors (MSRs) are often advocated as a radical but worthwhile alternative to traditional reactor concepts based on solid fuels. This article builds upon the existing research into MSRs to model and simulate the operation of thorium-fueled single-fluid and two-fluid reactors. The analysis is based on neutronics calculations and focuses on denatured MSR systems. Resource utilization and basic proliferation-risk attributes are compared to those of standard light-water reactors. Depending on specific design choices, even fully denatured reactors could reduce uranium and enrichment requirements by a factor of 3–4. Overall, denatured single-fluid designs appear as the most promising candidate technology minimizing both design complexity and overall proliferation risks despite being somewhat less attractive from the perspective of resource utilization

  8. Activities of research-reactor-technology project in FNCA from FY2005 to FY2007. Sharing neutronics calculation technique for core management and utilization of research reactors

    International Nuclear Information System (INIS)

    2010-07-01

    RRT project (Research-Reactor-Technology Project) was carried out with the theme of 'sharing neutronics calculation technique for core management and utilization of research reactors' in the framework of FNCA (Forum for Nuclear Cooperation in Asia) from FY2005 to FY2007. The objective of the project was to improve and equalize the level of neutronics calculation technique for the reactor core management among participating countries to assure the safe and stable operation of research reactors and the promotion of the effective utilization. Neutronics calculation codes, namely SRAC code system and MVP code, were adopted as common codes. Participating countries succeeded in applying the common codes to analyzing the core of each domestic research reactor. Some participating countries succeeded in applying the common codes to analyzing for utilization of own research reactors. Activities of RRT project have improved and equalized the level of neutronics calculation technique among participating countries. (author)

  9. Neutronics calculations for the Oak Ridge National Laboratory Tokamak Reactor Studies

    International Nuclear Information System (INIS)

    Santoro, R.T.; Baker, V.C.; Barnes, J.M.

    1976-01-01

    Neutronics calculations have been carried out to analyze the nuclear performance of conceptual blanket and shield designs for the Tokamak Experimental Power Reactor (EPR) and the Tokamak Demonstration Reactor Plant (DRP) being considered at the Oak Ridge National Laboratory. These reactor designs represent a sequence in the commercialization of fusion-generated electrical power. All of the calculations were carried out using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV coupled neutron-gamma-ray transport cross-section data, fluence-to-kerma conversion factors, and radiation damage cross-section data. The calculations include spatial and integral heating-rate estimates in the reactor with emphasis on the recovery of fusion neutron energy in the blanket and limiting the heat-deposition rate in the superconducting toroidal field coils. Radiation damage due to atomic displacements and gas production produced in the reactor structural material and in the toroidal field coil windings were also estimated. The tritium-breeding ratio when natural lithium is used as the fertile material in the DRP blanket and in the experimental breeding modules in the EPR is also given

  10. Calculations of steady-state and reactivity insertion transients in a research reactor simulating the PWR

    International Nuclear Information System (INIS)

    Mladin, Mirea; Mladin, Daniela; Prodea, Ilie

    2010-01-01

    In 2008, IAEA started a Coordinated Research Project for benchmarking the thermalhydraulic and neutronic computer codes for research reactor analysis against the experimental data. In this framework, for the first year of research contract, the Institute for Nuclear Research engaged in steady-state analysis of SPERT-III reactor and also in the simulation of the reactivity insertion tests performed in this reactor during mid sixties. In the first part, the paper describes a Monte Carlo input model of the oxide core selected for investigation and the results of the steady-state neutronic calculations with respect to hot and cold core reactivity excess and control rods worth. Also, prompt neutron life and reactivity feed-back coefficients were examined. These results were compared with the data provided in the reactor specification document concerning neutronic design calculated data. The second part of the paper is dedicated to calculation of the reactivity insertion transients with RELAP5 and CATHARE2 thermalhydraulic codes, both including point reactor kinetics models, and to comparison with experimental data. (authors)

  11. The neutron beam intensity increase by in-core fuel management enhancement in multipurpose research reactors

    International Nuclear Information System (INIS)

    Martinc, R.; Vukadin, Z.; Konstantinovic, J.

    1986-01-01

    The exploitation characteristics of an existing multipurpose research reactor can be increased not only by great reconstruction, but also, to the considerable extent, by the in-core fuel management sophistication. The optimisation of the in-core fuel management procedure in such reactors is governed (among others) by the identified reactor utilisation goals, i.e. by weighting factors dedicated to different utilisation goals, which are often (regarding the in-core fuel management procedure) highly controversial. In this work the best solution for in-core fuel management is sought, with the highest weighting factor dedicated to the neutron beam usage, rather than sample irradiation in the reactor core. The term in-core fuel management includes: the core configuration, the locations of the fresh fuel inflow zone and spent fuel excite zone, and the fuel transfers between these two zones (author)

  12. A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2017-06-26

    A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.

  13. Investigation for calculation methods used in analyzing the physics characteristics of nuclear power reactor

    International Nuclear Information System (INIS)

    Nguyen Tuan Khai; Hoang Van Khanh; Phan Quoc Vuong; Tran Viet Phu; Tran Vinh Thanh; Nguyen Thi Mai Huong; Nguyen Thi Dung; Le Tran Chung; Nguyen Minh Tuan; Tran Quoc Duong

    2014-01-01

    The project aims at nuclear human resource development and enhancement in research capability in reactor physics and kinetics at Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat). The main research items of the project can be summarized as follows: i) Considering possibility on using modern calculation techniques and methods in investigating neutronic characteristics and neutronics-thermal hydraulics coupling. This item is proposed to carry out based on international collaboration with Prof. Le Trong Thuy, San Jose University, US; ii) Carrying out the collaborative activities in research and training between Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat); iii) Opening two-week training course on nuclear reactor engineering (25 Nov - 12 Dec 2013) in collaboration with Japan Atomic Energy Agency (JAEA). (author)

  14. Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

    International Nuclear Information System (INIS)

    Aydogdu, K.

    1998-01-01

    Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

  15. Sub-second pencil beam dose calculation on GPU for adaptive proton therapy.

    Science.gov (United States)

    da Silva, Joakim; Ansorge, Richard; Jena, Rajesh

    2015-06-21

    Although proton therapy delivered using scanned pencil beams has the potential to produce better dose conformity than conventional radiotherapy, the created dose distributions are more sensitive to anatomical changes and patient motion. Therefore, the introduction of adaptive treatment techniques where the dose can be monitored as it is being delivered is highly desirable. We present a GPU-based dose calculation engine relying on the widely used pencil beam algorithm, developed for on-line dose calculation. The calculation engine was implemented from scratch, with each step of the algorithm parallelized and adapted to run efficiently on the GPU architecture. To ensure fast calculation, it employs several application-specific modifications and simplifications, and a fast scatter-based implementation of the computationally expensive kernel superposition step. The calculation time for a skull base treatment plan using two beam directions was 0.22 s on an Nvidia Tesla K40 GPU, whereas a test case of a cubic target in water from the literature took 0.14 s to calculate. The accuracy of the patient dose distributions was assessed by calculating the γ-index with respect to a gold standard Monte Carlo simulation. The passing rates were 99.2% and 96.7%, respectively, for the 3%/3 mm and 2%/2 mm criteria, matching those produced by a clinical treatment planning system.

  16. Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

    International Nuclear Information System (INIS)

    Koch, R.; Findeisen, A.

    1986-04-01

    The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

  17. Calculation of the flow distribution for the new core of the RA-6 reactor

    International Nuclear Information System (INIS)

    Garcia, J.C.; Delmastro, Dario F.

    2007-01-01

    In this work the pressure drop, the flow distribution, effective cooling flow rate and the velocity in the subchannels that cool fuel plates for the new core of RA-6 research reactor were calculated. These calculations were performed for a flow of 340 m 3 /hr and water temperatures of 12 C degrees, of 35 C degrees and 42 C degrees. The flow distribution was calculated without considering either safety factors or geometric changes. All the calculations were performed considering the flow as isothermal. (author) [es

  18. A functional method for estimating DPA tallies in Monte Carlo calculations of Light Water Reactors

    International Nuclear Information System (INIS)

    Read, Edward A.; Oliveira, Cassiano R.E. de

    2011-01-01

    There has been a growing need in recent years for the development of methodology to calculate radiation damage factors, namely displacements per atom (dpa), of structural components for Light Water Reactors (LWRs). The aim of this paper is to discuss the development and implementation of a dpa method using Monte Carlo method for transport calculations. The capabilities of the Monte Carlo code Serpent such as Woodcock tracking and fuel depletion are assessed for radiation damage calculations and its capability demonstrated and compared to those of the Monte Carlo code MCNP for radiation damage calculations of a typical LWR configuration. (author)

  19. Calculation of beam injection and modes of acceleration for the JINR phasotron

    International Nuclear Information System (INIS)

    Vorozhtsov, S.B.; Dmitrievsky, V.P.

    1981-01-01

    On the basis of computer simulation of particles motion from the injection region up to the final radius of the accelerated proton beam behaviour together with different modes of the JINR high current synchrocyclotron operation is investigated. The THOUR modified computer code is used for calculations. The calculations have been performed with allowance for particle radial-phase motion and particle axial motion and although with beam collective effects. Beam dynamics during first turns of particles has been considered by integrating equations of motion. Tolerances for magnetic field structure in the region of first phase oscillation are obtained. Verifications of time dependences of accelerated voltage amplitude are performed. Time dependences of beam intensity (with and without account for space charge effect) and of mean magnetic field disturbance and the dependence of the separatrice dimension on the orbit radius of the accelerated beam are given. The conclusion is drawn on the correctness of the earlier appreciation of beam intensity equaling 40-45 mkA

  20. Monte Carlo calculations of kQ, the beam quality conversion factor

    International Nuclear Information System (INIS)

    Muir, B. R.; Rogers, D. W. O.

    2010-01-01

    Purpose: To use EGSnrc Monte Carlo simulations to directly calculate beam quality conversion factors, k Q , for 32 cylindrical ionization chambers over a range of beam qualities and to quantify the effect of systematic uncertainties on Monte Carlo calculations of k Q . These factors are required to use the TG-51 or TRS-398 clinical dosimetry protocols for calibrating external radiotherapy beams. Methods: Ionization chambers are modeled either from blueprints or manufacturers' user's manuals. The dose-to-air in the chamber is calculated using the EGSnrc user-code egs c hamber using 11 different tabulated clinical photon spectra for the incident beams. The dose to a small volume of water is also calculated in the absence of the chamber at the midpoint of the chamber on its central axis. Using a simple equation, k Q is calculated from these quantities under the assumption that W/e is constant with energy and compared to TG-51 protocol and measured values. Results: Polynomial fits to the Monte Carlo calculated k Q factors as a function of beam quality expressed as %dd(10) x and TPR 10 20 are given for each ionization chamber. Differences are explained between Monte Carlo calculated values and values from the TG-51 protocol or calculated using the computer program used for TG-51 calculations. Systematic uncertainties in calculated k Q values are analyzed and amount to a maximum of one standard deviation uncertainty of 0.99% if one assumes that photon cross-section uncertainties are uncorrelated and 0.63% if they are assumed correlated. The largest components of the uncertainty are the constancy of W/e and the uncertainty in the cross-section for photons in water. Conclusions: It is now possible to calculate k Q directly using Monte Carlo simulations. Monte Carlo calculations for most ionization chambers give results which are comparable to TG-51 values. Discrepancies can be explained using individual Monte Carlo calculations of various correction factors which are more

  1. Modeling of water lighting process and calculation of the reactor-clarifier to improve energy efficiency

    Science.gov (United States)

    Skolubovich, Yuriy; Skolubovich, Aleksandr; Voitov, Evgeniy; Soppa, Mikhail; Chirkunov, Yuriy

    2017-10-01

    The article considers the current questions of technological modeling and calculation of the new facility for cleaning natural waters, the clarifier reactor for the optimal operating mode, which was developed in Novosibirsk State University of Architecture and Civil Engineering (SibSTRIN). A calculation technique based on well-known dependences of hydraulics is presented. A calculation example of a structure on experimental data is considered. The maximum possible rate of ascending flow of purified water was determined, based on the 24 hour clarification cycle. The fractional composition of the contact mass was determined with minimal expansion of contact mass layer, which ensured the elimination of stagnant zones. The clarification cycle duration was clarified by the parameters of technological modeling by recalculating maximum possible upward flow rate of clarified water. The thickness of the contact mass layer was determined. Likewise, clarification reactors can be calculated for any other lightening conditions.

  2. Neutronic calculation of the next fuel elements for the Argonaut reactor

    International Nuclear Information System (INIS)

    Oliveira, C.R.E.; Brito Aghina, L.O. de

    1981-01-01

    The best parameters of the next fuel elements of the Argonaut reactor, at IEN (Instituto de Engenharia Nuclear - Brazil), were determined. The next fuel elements will be rods of an uranium mixture (19.98% enriched), graphite and bakelite. The parameters to be determined are: mixture density, percentage of uranium in the mixture, pellet radius, rod material and elements arrangement (step). The calculations routines consisted in the analysis of several steps, using the LEOPARD computer code for cell calculations and RMAT1D for one dimensional spatial calculations (criticality) with four energy groups. Finally a neutronic study of the Argounat reactors present configuration was done, using the HAMMER computer code (cell), the EXTERMINATOR computer code (two-dimensional calculations) and RAMAT1D. (Author) [pt

  3. Core design calculations of IRIS reactor using modified CORD-2 code package

    International Nuclear Information System (INIS)

    Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

    2002-01-01

    Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

  4. Modelling of electron contamination in clinical photon beams for Monte Carlo dose calculation

    International Nuclear Information System (INIS)

    Yang, J; Li, J S; Qin, L; Xiong, W; Ma, C-M

    2004-01-01

    The purpose of this work is to model electron contamination in clinical photon beams and to commission the source model using measured data for Monte Carlo treatment planning. In this work, a planar source is used to represent the contaminant electrons at a plane above the upper jaws. The source size depends on the dimensions of the field size at the isocentre. The energy spectra of the contaminant electrons are predetermined using Monte Carlo simulations for photon beams from different clinical accelerators. A 'random creep' method is employed to derive the weight of the electron contamination source by matching Monte Carlo calculated monoenergetic photon and electron percent depth-dose (PDD) curves with measured PDD curves. We have integrated this electron contamination source into a previously developed multiple source model and validated the model for photon beams from Siemens PRIMUS accelerators. The EGS4 based Monte Carlo user code BEAM and MCSIM were used for linac head simulation and dose calculation. The Monte Carlo calculated dose distributions were compared with measured data. Our results showed good agreement (less than 2% or 2 mm) for 6, 10 and 18 MV photon beams

  5. The influence of the beam charge state on the analytical calculation of RBS and ERDA spectra

    Energy Technology Data Exchange (ETDEWEB)

    Barradas, Nuno P., E-mail: nunoni@ctn.ist.utl.pt [Centro de Ciências e Tecnologias Nucleares, Instituto Superior Técnico, Universidade de Lisboa, E.N. 10 ao km 139,7, 2695-066 Bobadela LRS (Portugal); Kosmata, Marcel [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany); Globalfoundries, Wilschdorfer Landstraße 101, 01109 Dresden (Germany); Hanf, Daniel; Munnik, Frans [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany)

    2016-03-15

    Analytical codes dedicated to the analysis of Ion Beam Analysis data rely on the accuracy of both the calculations and of basic data such as scattering cross sections and stopping powers. So far, the effect of the beam charge state of the incoming beam has been disregard by general purpose analytical codes such as NDF. In fact, the codes implicitly assume that the beam always has the equilibrium charge state distribution, by using tabulated stopping power values e.g. from SRIM, which are in principle valid for the effective charge state. The dependence of the stopping power with the changing charge state distribution is ignored. This assumption is reasonable in most cases, but for high resolution studies the actual change of the charge state distribution from the initial beam charge state towards equilibrium as it enters and traverses the sample must be taken into account, as it influences the shape of the observed data. In this work, we present an analytical calculation, implemented in NDF, that takes this effect into account. For elastic recoil detection analysis (ERDA), the changing charge state distribution of the recoils can also be taken into account. We apply the calculation to the analysis of experimental high depth resolution ERDA data for various oxide layers collected using a magnetic spectrometer.

  6. A Pearson VII distribution function for fast calculation of dechanneling and angular dispersion of beams

    International Nuclear Information System (INIS)

    Shao Lin; Peng Luohan

    2009-01-01

    Although multiple scattering theories have been well developed, numerical calculation is complicated and only tabulated values have been available, which has caused inconvenience in practical use. We have found that a Pearson VII distribution function can be used to fit Lugujjo and Mayer's probability curves in describing the dechanneling phenomenon in backscattering analysis, over a wide range of disorder levels. Differentiation of the obtained function gives another function to calculate angular dispersion of the beam in the frameworks by Sigmund and Winterbon. The present work provides an easy calculation of both dechanneling probability and angular dispersion for any arbitrary combination of beam and target having a reduced thickness ≥0.6, which can be implemented in modeling of channeling spectra. Furthermore, we used a Monte Carlo simulation program to calculate the deflection probability and compared them with previously tabulated data. A good agreement was reached.

  7. Calculations of neutron source at the KYIV research reactor for the boron neutron capture therapy aims

    International Nuclear Information System (INIS)

    Gritzay, O.; Kalchenko, O.; Klimova, N.; Razbudey, V.; Sanzhur, A.

    2006-01-01

    Calculation results of an epithermal neutron source which can be created at the Kyiv Research Reactor (KRR) by means of placing of specially selected moderators, filters, collimators, and shielding into the 10-th horizontal experimental tube (so-called thermal column) are presented. The general Monte-Carlo radiation transport code MCNP4C [1], the Oak Ridge isotope generation code ORIGEN2 [2] and the NJOY99 [3] nuclear data processing system have been used for these calculations

  8. Development, application and also modern condition of the calculated program Imitator of a reactor

    International Nuclear Information System (INIS)

    Aver'yanova, S.P.; Kovel', A.I.; Mamichev, V.V.; Filimonov, P.E.

    2008-01-01

    Features of the calculated program Imitator of a reactor (IR) for WWER-1000 operation simulation are discussed. It is noted that IR application at NPP provides for the project program (BIPR-7) on-line working. This offers a new means, on the one hand, for the efficient prediction and information support of operator, on the other hand, for the verification and development of calculated scheme and neutron-physical model of the WWER-1000 projection program [ru

  9. A model development for a thermohydraulic calculation material convection of MTR (Materials Testing Reactors)

    International Nuclear Information System (INIS)

    Abbate, P.

    1990-01-01

    The CONVEC program developed for the thermohydraulic calculation under a natural convection regime for MTR type reactors is presented. The program is based on a stationary, one dimensional model of finite differences that allow to calculate the temperatures of cooler, cladding and fuel as well as the flow for a power level specified by the user. This model has been satisfactorily validated by a water cooling (liquid phase) and air system. (Author) [es

  10. LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory

    International Nuclear Information System (INIS)

    Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.

    1976-04-01

    The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented

  11. Calculation of criticality of the AP600 reactor with KENO V.a code

    Energy Technology Data Exchange (ETDEWEB)

    Krumbein, A; Caner, M; Shapira, M [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center

    1996-12-01

    The Westinghouse AP600 PWR has been modeled using the KENO V.a three dimensional Monte Carlo criticality program of the SCALE-PC code system. These calculations and the use of a Monte Carlo neutron transport code such as KENO will provide us with an independent check on our WIMS/CITATION calculations for the AP600 as well as for other reactors. It will also enable us to model more complicated geometries. (authors).

  12. Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications

    International Nuclear Information System (INIS)

    Carluccio, Thiago

    2011-01-01

    This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R and D has been done about these subcritical concepts, mainly due to Minor Actinides (MA) and Long Lived Fission Products (LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (1) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (2) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN / MB-01 reactor, (3) to compare different nuclear data libraries calculation of integral parameters, such as k eff and k src , and differential distributions, such as spectrum and flux, and nuclides inventories and (4) apply the develop methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files. (author)

  13. 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

  14. Measurements and calculations of reactivity for the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Ferreira, P.S.B.; Maiorino, J.R.; Yamaguchi, M.

    1988-01-01

    This work shows a measurement of reactivity parameters, such as integral and diferential control rod worth, local void coefficient, and moderator temperature coefficient for the research reactor IEA-R1. The measured values were compared with those calculated through HAMMER-CITATION codes, having shown good agreement. (author) [pt

  15. Criticality calculation of the deposits for the fuel elements in RP-10 nuclear research reactor

    International Nuclear Information System (INIS)

    Aguirre, Alvaro; Bruna, Ruben

    2013-01-01

    This paper shows the results of the criticality calculation of the deposits for irradiated and non-irradiated fuel elements in the RP-10 research reactor with MCNP5 code. In all cases and for normal and incidental conditions, the effective multiplication factor (K eff ) results less than 0,90 according to the acceptance criterion. (authors).

  16. The utilization of Quabox/Cubox computer program for calculating Angra 1 Reactor core

    International Nuclear Information System (INIS)

    Pina, C.M. de.

    1981-01-01

    The utilization of Quabox/Cubox computer codes for calculating Angra 1 reactor core is studied. The results shows a dependency between the spent CPU time and the curacy of thermal power distribution in function of the polinomial expansion used. Comparison were mode between Citation code and some results from Westinghouse [pt

  17. Burn-up calculation of fusion-fission hybrid reactor using thorium cycle

    International Nuclear Information System (INIS)

    Shido, S.; Matsunaka, M.; Kondo, K.; Murata, I.; Yamamoto, Y.

    2006-01-01

    A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated nuclear data. The 3-dimensional ITER model was used as a base fusion reactor. The nuclear fuel (reprocessed plutonium as the fission materials mixed with thorium as the fertile materials), transmutation materials (minor actinides and long-lived fission products) and tritium breeder were loaded into the blanket. Performances of gas-cooled and water-cooled blankets were compared with each other. As a result, the proposed reactor can meet the requirement for TBP and power density. As far as nuclear waste incineration is concerned, the gas-cooled blanket has advantages. On the other hand, the water cooled-blanket is suited to energy production. (author)

  18. Calculation scheme for boiling water reactors cores; Methode de calcul des coeurs de reacteurs a eau bouillante par le systeme saphyr

    Energy Technology Data Exchange (ETDEWEB)

    Marsault, Ph [CEA Cadarache, Dept. d' Etudes des Reacteurs (DER/SERSI), 13 - Saint-Paul-lez-Durance (France); Nicolas, A; Lenain, R; Richebois, E; Royer, E; Caruge, D [CEA Saclay, Dept. de Mecanique et de Technologie (DMT/SERMA), 91 - Gif-sur-Yvette (France); Blaise, P [CEA Cadarache, Dept. d' Etudes des Reacteurs (DER/SPEX), 13 - Saint-Paul-lez-Durance (France); Gastaldi, B; Delpech, M [CEA Cadarache, Dept. d' Etudes des Reacteurs (DER/SPRC), 13 - Saint-Paul-lez-Durance (France)

    1999-07-01

    Boiling Water Reactors represent one third of the world's reactors. They are presently evolving towards greater simplification, allowing a reduction in the costs of operation, improved safety and a relative flexibility in their capacity to accommodate 100% MOX cores. The CEA, in a combined effort with its partners, the COGEMA and the EDF, would like to assess the interest of this reactor type, especially on this last point. A definition program and subsequent qualification of the calculation scheme have been undertaken. We are presenting here the specific features inherent in the calculation of these reactors, in comparison to PWRs, as well as the first results of the program. (authors)

  19. Ion trajectories calculation in a three dimensional beam subjected to a space charge

    International Nuclear Information System (INIS)

    Tauth, T.

    1978-04-01

    Physical and geometrical conditions allowing a first approximation of necessary sizes to numerical integration of the ions movement equations subjected to electrical and magnetic crossed fields and space charge action are investigated here. To take into consideration the effect of the last one, two artifices are put forward: replacing charged particles by equivalent particles in calculating the coulomb force, electrical field calculation produced in different points situated on the beam envelope by the uniform charges distribution [fr

  20. Calculation of the residual bearing capacity of reinforced concrete beams by the rigidity (deflection) criterion

    OpenAIRE

    V.S. Utkin; S.A. Solovyov

    2015-01-01

    The article proposes the method of calculating the bearing capacity of reinforced concrete beams at the operational stage by the rigidity (deflection) criterion. The methods, which were used in the article, are integral test and probabilistic methods for describing random variables. The author offers a new technique of calculating a deflection limit by a criterion of residual deformations. The article exemplifies the usage of the evidence theory for statistical information processing in the f...

  1. A Technique for Temperature and Ultimate Load Calculations of Thin Targets in a Pulsed Electron Beam

    DEFF Research Database (Denmark)

    Hansen, Jørgen-Walther; Lundsager, Per

    1979-01-01

    A technique is presented for the calculation of transient temperature distributions and ultimate load of rotationally symmetric thin membranes with uniform lateral load and exposed to a pulsed electron beam from a linear accelerator. Heat transfer by conduction is considered the only transfer...... mechanism. The ultimate load is calculated on the basis of large plastic strain analysis. Analysis of one aluminum and one titanium membrane is shown....

  2. Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

    Science.gov (United States)

    Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

    2018-02-01

    The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

  3. Design a computational program to calculate the composition variations of nuclear materials in the reactor operations

    International Nuclear Information System (INIS)

    Mohmmadnia, Meysam; Pazirandeh, Ali; Sedighi, Mostafa; Bahabadi, Mohammad Hassan Jalili; Tayefi, Shima

    2013-01-01

    Highlights: ► The atomic densities of light and heavy materials are calculated. ► The solution is obtained using Runge–Kutta–Fehlberg method. ► The material depletion is calculated for constant flux and constant power condition. - Abstract: The present work investigates an appropriate way to calculate the variations of nuclides composition in the reactor core during operations. Specific Software has been designed for this purpose using C#. The mathematical approach is based on the solution of Bateman differential equations using a Runge–Kutta–Fehlberg method. Material depletion at constant flux and constant power can be calculated with this software. The inputs include reactor power, time step, initial and final times, order of Taylor Series to calculate time dependent flux, time unit, core material composition at initial condition (consists of light and heavy radioactive materials), acceptable error criterion, decay constants library, cross sections database and calculation type (constant flux or constant power). The atomic density of light and heavy fission products during reactor operation is obtained with high accuracy as the program outputs. The results from this method compared with analytical solution show good agreements

  4. Design data for calculating neutral beam penetration into Z/sub eff/ > 1 plasmas

    International Nuclear Information System (INIS)

    Olson, R.E.; Berkner, K.H.; Graham, W.G.; Pyle, R.V.; Schlachter, A.S.; Stearns, J.W.

    1978-01-01

    Impurities such as C, N, O, Fe, and Mo in a confined plasma reduce the penetration of the energetic neutral deuterium or hydrogen beam injected for heating or fueling the plasma, thus affecting the energy- and fuel-deposition profiles. New calculations, confirmed by recent experimental results, show that previous estimates of the reduction of neutral beam penetration due to impurities in the plasma were overly pessimistic. Until recently, the cross sections used to calculate beam attenuation had been assumed to be q 2 times the cross section for H + + H obtained from the Born approximation, where q is the charge state of the ion. This led to very large cross sections for large values of q, and thus to very stringent requirements on the acceptable level of impurity ions in the plasma

  5. Tests and calculations of reinforced concrete beams subject to dynamic reversed loads

    International Nuclear Information System (INIS)

    Livolant, M.; Hoffmann, A.; Gauvain, J.

    1978-01-01

    This study presents the tests of a reinforced concrete beam conducted by the Department of Mechanical and Thermal Studies at the Centre d'Etudes Nucleaires, Saclay, France. The actual behavior of nuclear power plant buildings submitted to seismic loads is generally non linear even for moderate seismic levels. The non linearity is specially important for reinforced concrete beams type buildings. To estimate the safety factors when the building is designed by standard methods, accurate non linear calculations are necessary. For such calculations one of the most difficult point is to define a correct model for the behavior of a reinforced beam subject to reversed loads. For that purpose, static and dynamic experimental tests on a shaking table have been carried out and a model reasonably accurate has been established and checked on the tests results

  6. Non-perturbative calculation of equilibrium polarization of stored electron beams

    International Nuclear Information System (INIS)

    Yokoya, Kaoru.

    1992-05-01

    Stored electron/positron beams polarize spontaneously owing to the spin-flip synchrotron radiation. In the existing computer codes, the degree of the equilibrium polarization has been calculated using perturbation expansions in terms of the orbital oscillation amplitudes. In this paper a new numerical method is presented which does not employ the perturbation expansion. (author)

  7. A pencil beam dose calculation model for CyberKnife system

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Bin; Li, Yongbao; Liu, Bo; Zhou, Fugen [Image Processing Center, Beihang University, Beijing 100191 (China); Xu, Shouping [Department of Radiation Oncology, PLA General Hospital, Beijing 100853 (China); Wu, Qiuwen, E-mail: Qiuwen.Wu@Duke.edu [Department of Radiation Oncology, Duke University Medical Center, Durham, North Carolina 27710 (United States)

    2016-10-15

    Purpose: CyberKnife system is initially equipped with fixed circular cones for stereotactic radiosurgery. Two dose calculation algorithms, Ray-Tracing and Monte Carlo, are available in the supplied treatment planning system. A multileaf collimator system was recently introduced in the latest generation of system, capable of arbitrarily shaped treatment field. The purpose of this study is to develop a model based dose calculation algorithm to better handle the lateral scatter in an irregularly shaped small field for the CyberKnife system. Methods: A pencil beam dose calculation algorithm widely used in linac based treatment planning system was modified. The kernel parameters and intensity profile were systematically determined by fitting to the commissioning data. The model was tuned using only a subset of measured data (4 out of 12 cones) and applied to all fixed circular cones for evaluation. The root mean square (RMS) of the difference between the measured and calculated tissue-phantom-ratios (TPRs) and off-center-ratio (OCR) was compared. Three cone size correction techniques were developed to better fit the OCRs at the penumbra region, which are further evaluated by the output factors (OFs). The pencil beam model was further validated against measurement data on the variable dodecagon-shaped Iris collimators and a half-beam blocked field. Comparison with Ray-Tracing and Monte Carlo methods was also performed on a lung SBRT case. Results: The RMS between the measured and calculated TPRs is 0.7% averaged for all cones, with the descending region at 0.5%. The RMSs of OCR at infield and outfield regions are both at 0.5%. The distance to agreement (DTA) at the OCR penumbra region is 0.2 mm. All three cone size correction models achieve the same improvement in OCR agreement, with the effective source shift model (SSM) preferred, due to their ability to predict more accurately the OF variations with the source to axis distance (SAD). In noncircular field validation

  8. Linac beam dynamics calculations for low-current large-emittance beams

    International Nuclear Information System (INIS)

    Swain, G.R.; Butler, H.S.

    1992-01-01

    The beam in PILAC, a superconducting linac for pions proposed at LAUFF, will have a lager momentum spread (7% dp/p) and occupy a larger transverse space (13 cm dia. bore) than is usual in high-beta linacs. To find the effects of this large phase space, a cavity element is being added to the MOTER code. With this addition, pions and other particles may be tracked through the injection line and the PILAC linac. In one option, the particles may be cell by cell through a multicell cavity using formulas. The formulas are derived by integrating the energy gain and transverse impulse from the fields in a cell along the path of the particle. What is new in this analysis is that the transverse momentum is considered to be a significant part of the total momentum. The effect of a difference in velocity from the design velocity of the structure is considered. In another option still under development, field information is specified, and the particles may be tracked by stepwise integration

  9. Three-dimensional static and dynamic reactor calculations by the nodal expansion method

    International Nuclear Information System (INIS)

    Christensen, B.

    1985-05-01

    This report reviews various method for the calculation of the neutron-flux- and power distribution in an nuclear reactor. The nodal expansion method (NEM) is especially described in much detail. The nodal expansion method solves the diffusion equation. In this method the reactor core is divided into nodes, typically 10 to 20 cm in each direction, and the average flux in each node is calculated. To obtain the coupling between the nodes the local flux inside each node is expressed by use of a polynomial expansion. The expansion is one-dimensional, so inside each node such three expansions occur. To calculate the expansion coefficients it is necessary that the polynomial expansion is a solution to the one-dimensional diffusion equation. When the one-dimensional diffusion equation is established a term with the transversal leakage occur, and this term is expanded after the same polynomials. The resulting equation system with the expansion coefficients as the unknowns is solved with weigthed residual technique. The nodal expansion method is built into a computer program (also called NEM), which is divided into two parts, one part for steady-state calculations and one part for dynamic calculations. It is possible to take advantage of symmetry properties of the reactor core. The program is very flexible with regard to the number of energy groups, the node size, the flux expansion order and the transverse leakage expansion order. The boundary of the core is described by albedos. The program and input to it are described. The program is tested on a number of examples extending from small theoretical one up to realistic reactor cores. Many calculations are done on the wellknown IAEA benchmark case. The calculations have tested the accuracy and the computing time for various node sizes and polynomial expansions. In the dynamic examples various strategies for variation of the time step-length have been tested. (author)

  10. Effect of core configuration on the burnup calculations of MTR research reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Amin, E.H.; Sakr, A.M.

    2014-01-01

    Highlights: • 3D burn-up calculations of MTR-type research reactor were performed. Examination of the effect of control rod pattern on power density and neutron flux distributions is presented. • The calculations are performed using the MTR P C package and the programs (WIMS and CITVAP). • An empirical formula was generated for every fuel element type, to correlate irradiation to burn-up. - Abstract: In the present paper, three-dimensional burn-up calculations were performed using different patterns of control rods, in order to examine their effect on power density and neutron flux distributions through out the entire core and hence on the local burn-up distribution. These different cores burn-up calculations are carried out for an operating cycle equivalent to 15 Full Power Days (FPDs), with a power rating of 22 MW. Calculations were performed using an example of a typical research reactor of MTR-type using the internationally known computer codes’ package “MTR P C system”, using the cell calculation transport code WIMS-D4 with 12 energy groups and the core calculation diffusion code CITVAP with 5 energy groups. A depletion study was done and the effects on the research reactor fuel (U-235) were performed. The burn-up percentage (B.U.%) curves for every fuel element type were drawn versus irradiation (MWD/TE). Then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Charts of power density and neutron flux distribution for each core were plotted at different sections of each fuel element of the reactor core. Then a complete discussion and analysis of these curves are performed with comparison between the different core configurations, illustrating the effect of insertion or extraction of either of the four control rods directly on the neutron flux and consequently on the power distribution and burn-up. A detailed study of fuel burn-up gives detailed insight on the different B.U.% calculations

  11. Method of neutronic calculations for a spherical cell equivalent to cylindrical one for using computer codes in light water reactors in the fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Borges, V.; Sefidvash, F.; Rastogi, E.P.; Huria, H.C.; Krishnani, P.D.

    1989-01-01

    In order to use the existing light water reactor cell calculation codes for fluidized bed nuclear reactor having spherical fuel cells, an equivalence method has been developed. This method is shown to be adequate in calculation of the Dancoff factor. This method also was applicable in LEOPARD code and the results obtained in calculation of K ∞ was compared with the obtained using the DTF IV code, the results showed that the method is adequate for the calculations neutronics of the fluidized bed nuclear reactor. (author) [pt

  12. Dose calculation methods in photon beam therapy using energy deposition kernels

    International Nuclear Information System (INIS)

    Ahnesjoe, A.

    1991-01-01

    The problem of calculating accurate dose distributions in treatment planning of megavoltage photon radiation therapy has been studied. New dose calculation algorithms using energy deposition kernels have been developed. The kernels describe the transfer of energy by secondary particles from a primary photon interaction site to its surroundings. Monte Carlo simulations of particle transport have been used for derivation of kernels for primary photon energies form 0.1 MeV to 50 MeV. The trade off between accuracy and calculational speed has been addressed by the development of two algorithms; one point oriented with low computional overhead for interactive use and one for fast and accurate calculation of dose distributions in a 3-dimensional lattice. The latter algorithm models secondary particle transport in heterogeneous tissue by scaling energy deposition kernels with the electron density of the tissue. The accuracy of the methods has been tested using full Monte Carlo simulations for different geometries, and found to be superior to conventional algorithms based on scaling of broad beam dose distributions. Methods have also been developed for characterization of clinical photon beams in entities appropriate for kernel based calculation models. By approximating the spectrum as laterally invariant, an effective spectrum and dose distribution for contaminating charge particles are derived form depth dose distributions measured in water, using analytical constraints. The spectrum is used to calculate kernels by superposition of monoenergetic kernels. The lateral energy fluence distribution is determined by deconvolving measured lateral dose distributions by a corresponding pencil beam kernel. Dose distributions for contaminating photons are described using two different methods, one for estimation of the dose outside of the collimated beam, and the other for calibration of output factors derived from kernel based dose calculations. (au)

  13. A Monte Carlo simulation framework for electron beam dose calculations using Varian phase space files for TrueBeam Linacs.

    Science.gov (United States)

    Rodrigues, Anna; Sawkey, Daren; Yin, Fang-Fang; Wu, Qiuwen

    2015-05-01

    To develop a framework for accurate electron Monte Carlo dose calculation. In this study, comprehensive validations of vendor provided electron beam phase space files for Varian TrueBeam Linacs against measurement data are presented. In this framework, the Monte Carlo generated phase space files were provided by the vendor and used as input to the downstream plan-specific simulations including jaws, electron applicators, and water phantom computed in the EGSnrc environment. The phase space files were generated based on open field commissioning data. A subset of electron energies of 6, 9, 12, 16, and 20 MeV and open and collimated field sizes 3 × 3, 4 × 4, 5 × 5, 6 × 6, 10 × 10, 15 × 15, 20 × 20, and 25 × 25 cm(2) were evaluated. Measurements acquired with a CC13 cylindrical ionization chamber and electron diode detector and simulations from this framework were compared for a water phantom geometry. The evaluation metrics include percent depth dose, orthogonal and diagonal profiles at depths R100, R50, Rp, and Rp+ for standard and extended source-to-surface distances (SSD), as well as cone and cut-out output factors. Agreement for the percent depth dose and orthogonal profiles between measurement and Monte Carlo was generally within 2% or 1 mm. The largest discrepancies were observed within depths of 5 mm from phantom surface. Differences in field size, penumbra, and flatness for the orthogonal profiles at depths R100, R50, and Rp were within 1 mm, 1 mm, and 2%, respectively. Orthogonal profiles at SSDs of 100 and 120 cm showed the same level of agreement. Cone and cut-out output factors agreed well with maximum differences within 2.5% for 6 MeV and 1% for all other energies. Cone output factors at extended SSDs of 105, 110, 115, and 120 cm exhibited similar levels of agreement. We have presented a Monte Carlo simulation framework for electron beam dose calculations for Varian TrueBeam Linacs. Electron beam energies of 6 to 20 MeV for open and collimated

  14. Calculation of DPA in the Reactor Internal Structural Materials of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kim, Yong Deong; Lee, Hwan Soo

    2014-01-01

    The embrittlement is mainly caused by atomic displacement damage due to irradiations with neutrons, especially fast neutrons. The integrity of the reactor internal structural materials has to be ensured over the reactor life time, threatened by the irradiation induced displacement damage. Accurate modeling and prediction of the displacement damage is a first step to evaluate the integrity of the reactor internal structural materials. Traditional approaches for analyzing the displacement damage of the materials have relied on tradition model, developed initially for simple metals, Kinchin and Pease (K-P), and the standard formulation of it by Norgett et al. , often referred to as the 'NRT' model. An alternative and complementary strategy for calculating the displacement damage is to use MCNP code. MCNP uses detailed physics and continuous-energy cross-section data in its simulations. In this paper, we have performed the evaluation of the displacement damage of the reactor internal structural materials in Kori NPP unit 1 using detailed Monte Carlo modeling and compared with predictions results of displacement damage using the classical NRT model. The evaluation of the displacement damage of the reactor internal structural materials in Kori NPP unit 1 using detailed Monte Carlo modeling has been performed. The maximum value of the DPA rate was occurred at the baffle side of the reactor internal where the node has the maximum neutron flux

  15. A low background pulsed neutron polyenergetic beam at the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Adib, M.; Abdel-Kawy, A.; Habib, N.; Abu-El-Ela, M.; Wahba, M.; Kilany, M.

    1991-12-01

    A low background pulsed neutron polyenergetic thermal beam at ET-RR-1 is produced by a rotor and rotating collimator suspended in magnetic fields. Each of them is mounted on its mobile platform and whose centres are 66 cm apart, rotating synchronously at speeds up to 16000 rpm. It was found that the neutron burst produced by the rotor with almost 100% transmission passes through the collimator, when the rotation phase between them is 28.8 deg. Moreover the background level achieved at the detector position is low, constant and free from peaks due to gamma rays and fast neutrons accompanying the reactor thermal beam. (author). 12 refs, 3 figs

  16. Investigation of bulk electron densities for dose calculations on cone-beam CT images

    International Nuclear Information System (INIS)

    Lambert, J.; Parker, J.; Gupta, S.; Hatton, J.; Tang, C.; Capp, A.; Denham, J.W.; Wright, P.

    2010-01-01

    Full text: If cone-beam CT images are to be used for dose calculations, then the images must be able to provide accurate electron density information. Twelve patients underwent twice weekly cone-beam CT scans in addition to the planning CT scan. A standardised 5-field treatment plan was applied to 169 of the CBCT images. Doses were calculated using the original electron density values in the CBCT and with bulk electron densities applied. Bone was assigned a density of 288 HU, and all other tissue was assigned to be water equivalent (0 HU). The doses were compared to the dose calculated on the original planning CT image. Using the original HU values in the cone-beam images, the average dose del i vered by the plans from all 12 patients was I. I % lower than the intended 200 cOy delivered on the original CT plans (standard devia tion 0.7%, maximum difference -2.93%). When bulk electron densities were applied to the cone-beam images, the average dose was 0.3% lower than the original CT plans (standard deviation 0.8%, maximum difference -2.22%). Compared to using the original HU values, applying bulk electron densities to the CBCT images improved the dose calculations by almost I %. Some variation due to natural changes in anatomy should be expected. The application of bulk elec tron densities to cone beam CT images has the potential to improve the accuracy of dose calculations due to inaccurate H U values. Acknowledgements This work was partially funded by Cancer Council NSW Grant Number RG 07-06.

  17. Evaluation of a new commercial Monte Carlo dose calculation algorithm for electron beams.

    Science.gov (United States)

    Vandervoort, Eric J; Tchistiakova, Ekaterina; La Russa, Daniel J; Cygler, Joanna E

    2014-02-01

    In this report the authors present the validation of a Monte Carlo dose calculation algorithm (XiO EMC from Elekta Software) for electron beams. Calculated and measured dose distributions were compared for homogeneous water phantoms and for a 3D heterogeneous phantom meant to approximate the geometry of a trachea and spine. Comparisons of measurements and calculated data were performed using 2D and 3D gamma index dose comparison metrics. Measured outputs agree with calculated values within estimated uncertainties for standard and extended SSDs for open applicators, and for cutouts, with the exception of the 17 MeV electron beam at extended SSD for cutout sizes smaller than 5 × 5 cm(2). Good agreement was obtained between calculated and experimental depth dose curves and dose profiles (minimum number of measurements that pass a 2%/2 mm agreement 2D gamma index criteria for any applicator or energy was 97%). Dose calculations in a heterogeneous phantom agree with radiochromic film measurements (>98% of pixels pass a 3 dimensional 3%/2 mm γ-criteria) provided that the steep dose gradient in the depth direction is considered. Clinically acceptable agreement (at the 2%/2 mm level) between the measurements and calculated data for measurements in water are obtained for this dose calculation algorithm. Radiochromic film is a useful tool to evaluate the accuracy of electron MC treatment planning systems in heterogeneous media.

  18. Modelling lateral beam quality variations in pencil kernel based photon dose calculations

    International Nuclear Information System (INIS)

    Nyholm, T; Olofsson, J; Ahnesjoe, A; Karlsson, M

    2006-01-01

    Standard treatment machines for external radiotherapy are designed to yield flat dose distributions at a representative treatment depth. The common method to reach this goal is to use a flattening filter to decrease the fluence in the centre of the beam. A side effect of this filtering is that the average energy of the beam is generally lower at a distance from the central axis, a phenomenon commonly referred to as off-axis softening. The off-axis softening results in a relative change in beam quality that is almost independent of machine brand and model. Central axis dose calculations using pencil beam kernels show no drastic loss in accuracy when the off-axis beam quality variations are neglected. However, for dose calculated at off-axis positions the effect should be considered, otherwise errors of several per cent can be introduced. This work proposes a method to explicitly include the effect of off-axis softening in pencil kernel based photon dose calculations for arbitrary positions in a radiation field. Variations of pencil kernel values are modelled through a generic relation between half value layer (HVL) thickness and off-axis position for standard treatment machines. The pencil kernel integration for dose calculation is performed through sampling of energy fluence and beam quality in sectors of concentric circles around the calculation point. The method is fully based on generic data and therefore does not require any specific measurements for characterization of the off-axis softening effect, provided that the machine performance is in agreement with the assumed HVL variations. The model is verified versus profile measurements at different depths and through a model self-consistency check, using the dose calculation model to estimate HVL values at off-axis positions. A comparison between calculated and measured profiles at different depths showed a maximum relative error of 4% without explicit modelling of off-axis softening. The maximum relative error

  19. Calculation of fission product behavior in a multiple reactor barriers in case of an accident

    International Nuclear Information System (INIS)

    Ezzedin, A. A.; Dakhil, A. S.; Elbaden, S. E.

    2012-12-01

    Radiation protection of the population in case of a reactor accident utilizes reference levels which are based on doses values. Therefore, adequate provisions for effective and timely dose assessment for population in case of accidents at nuclear power plant (NPP) are important. Developing the background for such provisions is the objective of this study. In particular, an exponential model has been developed and utilized to calculate the release rate of the most volatile gaseous materials from different reactor barriers. Calculation has been performed for noble gases (1 33X e, 1 35X e, 1 38X e, 8 5K r, 8 7K r, 8 8K r) and the halogens(1'3 1I , 1 32I , 1 33I , 1'3 4I , 1 35I ). The effective dose rate equivalent is calculations in the nearly stage of a reactor accident. Calculations are performed using the MCNP-4C code. The results are comparable with the final analysis report which utilizes different codes. Results of our calculation shows no excessive dose in populated regions and it is recommended to use secondary containment barrier for highly reduction of the release rate to the environment. (Author)

  20. Calorific energy deposited by gamma radiations in a test reactor. Calorimetric measurements and calculations

    International Nuclear Information System (INIS)

    Mecheri, K.-F.

    1977-01-01

    The purpose of this work was to determine the calorific energy deposited by gamma radiations in the experimental devices irradiated in the test reactors of the Grenoble Nuclear Study Centre. A theoretical study briefly recalls to mind the various sorts of nuclear reactions that occur in a reactor, from the special angle of their ability to deposit calorific energy in the materials. A special study with the help of a graphite calorimeter made it possible to show the possible effect of the various parameters intervening in this energy absorption: the nature of the materials, their geometry, the spectrum of the incident gamma rays and the fact that the variation of this spectrum is due to the position of the measuring point with respect to the reactor core or to the presence of structures around the measuring instrument. The results of the calculations made with the help of the Mercury IV and ANISN codes are compared with those of the determinations in order to ascertain that very are adapted to the forecasts of energy deposition in the various materials. The conclusion was reached that in order to calculate with accuracy the depositifs of gamma energy in the experimental devices, it is necessary either to introduce the build-up calculation for the low energy photons, in the Mercury IV calculation code or to associate the DOT code to the ANISN calculation code [fr

  1. Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.

    1981-01-01

    The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru

  2. 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)

  3. 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)

  4. Analysis of kyoto university reactor physics critical experiments using NCNSRC calculation methodology

    International Nuclear Information System (INIS)

    Amin, E.; Hathout, A.M.; Shouman, S.

    1997-01-01

    The kyoto university reactor physics experiments on the university critical assembly is used to benchmark validate the NCNSRC calculations methodology. This methodology has two lines, diffusion and Monte Carlo. The diffusion line includes the codes WIMSD4 for cell calculations and the two dimensional diffusion code DIXY2 for core calculations. The transport line uses the MULTIKENO-Code vax Version. Analysis is performed for the criticality, and the temperature coefficients of reactivity (TCR) for the light water moderated and reflected cores, of the different cores utilized in the experiments. The results of both Eigen value and TCR approximately reproduced the experimental and theoretical Kyoto results. However, some conclusions are drawn about the adequacy of the standard wimsd4 library. This paper is an extension of the NCNSRC efforts to assess and validate computer tools and methods for both Et-R R-1 and Et-MMpr-2 research reactors. 7 figs., 1 tab

  5. Experimental and calculational works on characteristics of the Dalat Nuclear Research Reactor. Second edition

    International Nuclear Information System (INIS)

    Pham Ngoc Khoi; Nguyen Kim Dung

    2016-03-01

    Recognizing the significant value and necessity of publishing the scientific document of experimental and calculational works on the Dalat Nuclear Research Reactor (DNRR) physics and engineering for research, operation, training activities as well as for international scientific exchange, Vietnam Atomic Energy Agency (VAEA) and Vietnam Atomic Energy Institute have completed editing to publish the “Experimental and Calculational Works on Characteristics of THE DALAT NUCLEAR RESEARCH REACTOR” which consists of 26 typical papers representing the most important experimental and calculational results of the DNRR physics and engineering obtained during 30 years of operation and exploitation with the contribution of Vietnamese and former USSR’s experts, especially scientists and engineers working at the Reactor Center of the NRI

  6. 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

  7. Use of heterogeneous finite elements generated by collision probability solutions to calculate a pool reactor core

    International Nuclear Information System (INIS)

    Calabrese, C.R.; Grant, C.R.

    1990-01-01

    This work presents comparisons between measured fluxes obtained by activation of Manganese foils in the light water, enriched uranium research pool reactor RA-2 MTR (Materials Testing Reactors) fuel element) and fluxes calculated by the finite element method FEM using DELFIN code, and describes the heterogeneus finite elements by a set of solutions of the transport equations for several different configurations obtained using the collision probability code HUEMUL. The agreement between calculated and measured fluxes is good, and the advantage of using FEM is showed because to obtain the flux distribution with same detail using an usual diffusion calculation it would be necessary 12000 mesh points against the 2000 points that FEM uses, hence the processing time is reduced in a factor ten. An interesting alternative to use in MTR fuel management is presented. (Author) [es

  8. Nuclear performance calculations for the ELMO Bumpy Torus Reactor (EBTR) reference design

    International Nuclear Information System (INIS)

    Santoro, R.T.; Barnes, J.M.

    1977-12-01

    The nuclear performance of the ELMO Bumpy Torus Reactor reference design has been calculated using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV transport cross-section data and nuclear response functions. The calculated results include estimates of the spatial and integral heating rate with emphasis on the recovery of fusion neutron energy in the blanket assembly and minimization of the energy deposition rates in the cryogenic magnet coil assemblies. The tritium breeding ratio in the natural lithium-laden blanket was calculated to be 1.29 tritium nuclei per incident neutron. The radiation damage in the reactor structural material and in the magnet assembly is also given

  9. Intact and Degraded Component Criticality Calculations of N Reactor Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    L. Angers

    2001-01-01

    The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k eff ) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR

  10. The use of the codes from MCU family for calculations of WWER type reactors

    International Nuclear Information System (INIS)

    Abagijan, L.P.; Alexeyev, N.I.; Bryzgalov, V.I.; Gomin, E.A.; Glushkov, A.E.; Gorodkov, S.S.; Gurevich, M.I.; Kalugin, M.A.; Marin, S.V.; Shkarovsky, D.A.; Yudkevich, M.S.

    2000-01-01

    The MCU-RFFI/A and MCU-REA codes developed within the framework of the long term MCU project are widely used for calculations of neutron physic characteristics of WWER type reactors. Complete descriptions of the codes are available in both Russian and English. The codes are verified and validated by means of the comparison of calculated results with experimental data and mathematical benchmarks. The codes are licensed by Russian Nuclear and Criticality Safety Regulatory Body (Gosatomnadzor RF) (Code Passports: N 61 of 17.10.1966 and N 115 of 02.03.2000 accordingly)). The report gives examples of WWER reactor physic tasks important for practice solved using the codes from the MCU family. Some calculational results are given too. (Authors)

  11. Analysis of the Rossendorf SEG experiments using the JNC route for reactor calculation

    International Nuclear Information System (INIS)

    Dietze, Klaus

    1999-11-01

    The integral experiments performed at the Rossendorf fast-thermal coupled reactor RRR/SEG have been reanalyzed using the JNC route for reactor calculation JENDL3.2/SLAROM/CITATION/JOINT/PERKY. The Rossendorf experiments comprise sample reactivity measurements with pure fission products and structural material in five configurations with different neutron and adjoint spectra. The shapes of the adjoint spectra have been designed to get high sensitivities to neutron capture or the scattering effect. The calculated neutron and adjoint spectra are in good agreement with former results obtained with the European route JEF2.2/ECCO/ERANOS. The C/E-values of the central reactivity worths of samples under investigation are given. Deviations in the results of both routes are due to the different libraries, codes, and self-shielding treatments used in the calculations. Results exceeding the experimental error are discussed. (author)

  12. Use of the Streaming Matrix Hybrid Method for discrete-ordinates fusion reactor calculations

    International Nuclear Information System (INIS)

    Battat, M.E.; Davidson, J.W.; Dudziak, D.J.; Thayer, G.R.

    1984-01-01

    The use of the discrete-ordinates method for solving two-dimensional, neutral-particle transport in fusion reactor blankets and shields is often limited by inherent inaccuracies due to the ray-effect. This effect presents a particular problem in the case of neutron streaming in the large internal void regions of a fusion reactor. A deterministic streaming technique called the Streaming Matrix Hybrid Method (SMHM) has been incorporated in the two-dimensional discrete-ordinates code TRIDENT-CTR. Calculations have been performed for an actual inertial-confinement fusion (ICF) reactor design using TRIDENT-CTR both with and without the SMHM. Comparisons of the calculated fluxes indicate that substantial mitigation of the ray effect can be achieved with the SMHM. Calculations were performed for the Los Alamos FIRST STEP hybrid ICF reactor designed for tritium production. Conventional 238 U fuel rod assemblies surround the spherical steel target chamber to form an annular cylindrical blanket. An axial fuel region is included to complete the blanket

  13. Monte Carlo calculations of electron beam quality conversion factors for several ion chamber types

    Energy Technology Data Exchange (ETDEWEB)

    Muir, B. R., E-mail: Bryan.Muir@nrc-cnrc.gc.ca [Measurement Science and Standards, National Research Council Canada, 1200 Montreal Road, Ottawa, Ontario K1A 0R6 (Canada); Rogers, D. W. O., E-mail: drogers@physics.carleton.ca [Carleton Laboratory for Radiotherapy Physics, Physics Department, Carleton University, 1125 ColonelBy Drive, Ottawa, Ontario K1S 5B6 (Canada)

    2014-11-01

    Purpose: To provide a comprehensive investigation of electron beam reference dosimetry using Monte Carlo simulations of the response of 10 plane-parallel and 18 cylindrical ion chamber types. Specific emphasis is placed on the determination of the optimal shift of the chambers’ effective point of measurement (EPOM) and beam quality conversion factors. Methods: The EGSnrc system is used for calculations of the absorbed dose to gas in ion chamber models and the absorbed dose to water as a function of depth in a water phantom on which cobalt-60 and several electron beam source models are incident. The optimal EPOM shifts of the ion chambers are determined by comparing calculations of R{sub 50} converted from I{sub 50} (calculated using ion chamber simulations in phantom) to R{sub 50} calculated using simulations of the absorbed dose to water vs depth in water. Beam quality conversion factors are determined as the calculated ratio of the absorbed dose to water to the absorbed dose to air in the ion chamber at the reference depth in a cobalt-60 beam to that in electron beams. Results: For most plane-parallel chambers, the optimal EPOM shift is inside of the active cavity but different from the shift determined with water-equivalent scaling of the front window of the chamber. These optimal shifts for plane-parallel chambers also reduce the scatter of beam quality conversion factors, k{sub Q}, as a function of R{sub 50}. The optimal shift of cylindrical chambers is found to be less than the 0.5 r{sub cav} recommended by current dosimetry protocols. In most cases, the values of the optimal shift are close to 0.3 r{sub cav}. Values of k{sub ecal} are calculated and compared to those from the TG-51 protocol and differences are explained using accurate individual correction factors for a subset of ion chambers investigated. High-precision fits to beam quality conversion factors normalized to unity in a beam with R{sub 50} = 7.5 cm (k{sub Q}{sup ′}) are provided. These

  14. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

    1983-02-01

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

  15. TRAC calculations of a loss-of-coolant accident in a reactor scale model

    International Nuclear Information System (INIS)

    Pyun, J.J.

    1981-01-01

    The TRAC (Transient Reactor Analysis Code) is being developed at the Los Alamos National Laboratory as an advanced best-estimate computer program for analysis of postulated hypothetical accidents in pressurized water reactors. As a part of the TRAC developmental verification efforts, a TRAC posttest analysis of Semiscale Mod-3 Test S-07-6 was conducted. The results of this analysis show that the agreement between TRAC calculations and experimental data is not very good. In particular, TRAC does not predict the long term doncomer and core liquid level oscillations during the reflood phase

  16. Utilization of a statistical procedure for DNBR calculation and in the survey of reactor protection limits

    International Nuclear Information System (INIS)

    Pontedeiro, A.C.; Camargo, C.T.M.; Galetti, M.R. da Silva.

    1987-01-01

    A new procedure is applied to Angra 1 NPP, which is related to DNBR calculations, considering the design parameters statistically: Improved Thermal Design Procedure (ITDP). The ITDP application leads to the determination of uncertainties in the input parameters, the sensitivity factors on DNBR. The DNBR limit and new reactor protection limits. This was done to Angra 1 with the subchannel code COBRA-IIIP. The analysis of limiting accident in terms of DNB confirmed a gain in DNBR margin, and greater operation flexibility of the plant, decreasing unnecessary trips of the reactor. (author) [pt

  17. 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

  18. Application of CO2 laser beam weld for repair of fuel element of nuclear reactor 'YAYOI'

    International Nuclear Information System (INIS)

    Hashimoto, Mitsuo; Yanagi, Hideharu; Sukegawa, Toshio; Saito, Isao; Sasuga, Norihiko; Aizawa, Nagaaki; Miya, Kenzo

    1986-01-01

    The present studies are to develop CO 2 laser beam welding techniques in order to apply for repoint of nuclear reactor fuel of Fast Neutron Source Reactor YAYOI. For that purpos, many experiments were conduted to obtain various effects of laser welding variables with use of SUS 304 plates, pipes and simulated dumy fuels. These experiments provided us an optimal welding condition through metallurgical observations, non-destructive and mechanical tests. It was found that the laser welds exhibited properties equivalent to those of the base metal, in addition they provided us a favorable system than that of electron beam welds against a cladding of radioactive nuclear fuel in a hot cell. The present paper reports on the characteristics of laser welds, structural analysis of fuel element and a system design of remotely operated devices setting in a hot cell. (author)

  19. Dosimetry issues for an ultra-high flux beam and multipurpose research reactor design

    International Nuclear Information System (INIS)

    West, C.D.

    1993-01-01

    The Advanced Neutron Source is a new user facility for all fields of neutron research, including neutron beam experiments, materials analysis, materials testing, and isotope production. The complement and layout of the experimental facilities have been determined sufficiently, at a conceptual design level, to make reliable cost and schedule estimates. The source of neutrons will be a heavy water reactor, constructed largely of aluminum, with an available thermal neutron flux 5--10 times higher than existing research reactors. Among the dosimetry issues to be faced are damage prediction and surveillance for component life attainment; measurement of fluence and spectra in regions where both change substantially over a distance of a few centimeters; and characterization and measurement of the radiation field in the research areas around the neutron beam experiments

  20. Calculation of electron-beam induced displacement in thin films by using parameter-reduced formulas

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Qiang [College of Nuclear Science and Technology, Harbin Engineering University, Harbin 150001 (China); Chen, Di [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77843 (United States); Wang, Qingyu; Li, Zhongyu [College of Nuclear Science and Technology, Harbin Engineering University, Harbin 150001 (China); Shao, Lin, E-mail: lshao@tamu.edu [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77843 (United States)

    2017-03-01

    Based on the Mott cross sections of relativistic electron collisions with atoms, we calculate displacement creation by electron beams of arbitrary energies (up to 100 MeV) in thin films of arbitrary atomic numbers (up to Z = 90). In a comparison with Mont Carlo full damage cascade simulations, we find that total number of displacements in a film can be accurately estimated as the product of average displacements created per collision and average collision numbers in the film. To calculate average displacements per electron-atom collision, energy transfer from Mott cross section is combined with NRT model. To calculate collision numbers, mean deflection angles and multi-scattering theory are combined to extract collision number dependence on film thickness. For each key parameter, parameter-reduced formulas are obtained from data fitting. The fitting formulas provide a quick and accurate method to estimate radiation damage caused by electron beams.