WorldWideScience

Sample records for reactors neutron part

  1. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  2. NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  3. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  4. Nuclear reactor neutron shielding

    Science.gov (United States)

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

    2017-09-12

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

  5. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

    Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels

  6. Reactor Neutron Sources

    International Nuclear Information System (INIS)

    Aksenov, V.L.

    1994-01-01

    The present status and the prospects for development of reactor neutron sources for neutron scattering research in the world are considered. The fields of application of neutron scattering relative to synchrotron radiation, the creation stages of reactors (steady state and pulsed) and their position in comparison with spallation neutron sources at present and in the foreseen future are discussed. (author). 15 refs.; 8 figs.; 3 tabs

  7. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

    Najzer, M.; Pauko, M.; Glumac, B.; Acquah, I.N.; Moskon, F.

    1977-01-01

    An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

  8. Neutronics and thermohydraulics of the reactor C.E.N.E. Part II

    International Nuclear Information System (INIS)

    Caro, R.

    1976-01-01

    In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

  9. Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I

    International Nuclear Information System (INIS)

    Caro, R.; Ahnert, C.; Naudin, A. E.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-01-01

    In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

  10. Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte I

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R; Ahnert, C; Naudin, A E; Martinez Fanegas, R; Minguez, E; Rovira, A

    1976-07-01

    In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

  11. Neutronics and thermohydraulics of the reactor C.E.N.E. Part II; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte II

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R

    1976-07-01

    In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

  12. Measurements of neutron flux in the RA reactor

    International Nuclear Information System (INIS)

    Raisic, N.

    1961-12-01

    This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire [sr

  13. Semiconductor research with reactor neutrons

    International Nuclear Information System (INIS)

    Kimura, Itsuro

    1992-01-01

    Reactor neutrons play an important role for characterization of semiconductor materials as same as other advanced materials. On the other hand reactor neutrons bring about not only malignant irradiation effects called radiation damage, but also useful effects such as neutron transmutation doping and defect formation for opto-electronics. Research works on semiconductor materials with the reactor neutrons of the Kyoto University Reactor (KUR) are briefly reviewed. In this review, a stress is laid on the present author's works. (author)

  14. NEUTRONIC REACTOR STRUCTURE

    Science.gov (United States)

    Weinberg, A.M.; Vernon, H.C.

    1961-05-30

    A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

  15. Neutronic Analysis of the 3 MW TRIGA MARK II Research Reactor, Part I: Monte Carlo Simulation

    International Nuclear Information System (INIS)

    Huda, M.Q.; Chakrobortty, T.K.; Rahman, M.; Sarker, M.M.; Mahmood, M.S.

    2003-05-01

    This study deals with the neutronic analysis of the current core configuration of a 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Final Safety Analysis Report (FSAR) values. The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Continuous energy cross-section data from ENDF/B-VI and S(α, β) scattering functions from the ENDF/B-V library were used. The validation of the model against benchmark experimental results is presented. The MCNP predictions and the experimentally determined values are found to be in very good agreement, which indicates that the Monte Carlo model is correctly simulating the TRIGA reactor. (author)

  16. Design of emergency shutdown system for the Tehran Research Reactor; Part I: Neutronics investigation

    International Nuclear Information System (INIS)

    Safarinia, M.; Faghihi, F.; Mirvakili, S.M.; Fakhraei, A.

    2017-01-01

    Highlights: • An emergency shutdown system for the TRR is carried out based on a heavy water tank. • The performance of the heavy water tank are carried out based on “first and equilibrium cores”. • Heavy water discharging flow rate is also studied in the current research. • Thermal flux in the radioisotope channel with and without the heavy water tank are studied. • A core with and without the heavy water tank for the cases of 5 × 6, 5 × 5, 5 × 4, and 4 × 4 fuel assemblies are investigated (for two types of fuel loading—first and equilibrium cores). - Abstract: In this paper, a neutronics design of the secondary (i.e., emergency) shutdown system for the Tehran Research Reactor (TRR) is carried out based on a heavy water tank design. The heavy water tank in a cylindrical shape is around the core, and calculations for the optimized radius and height of the tank are performed. The performance of the heavy water tank calculations are carried out based on two types of fuel loading, which are called the “first and equilibrium cores” of the TRR. For both cases, neutronics and standard safety analysis are taken into account, benchmarked, and described herein. Heavy water discharging flow rate is also studied in the current research, and the results are compared with the IAEA criteria. Moreover, thermal flux in the radioisotope channel with and without the heavy water tank (as the reflector) are studied herein. Specifically, a core with and without the heavy water tank for the cases of 5 × 6, 5 × 5, 5 × 4, and 4 × 4 fuel assemblies are investigated (for two types of fuel loading—first and equilibrium cores). Based on our optimization, the 5 × 5 fuel assembly, which is called “B configuration,” has better performance and efficiency than that of the other described layouts.

  17. Split core experiments; Part I. Axial neutron flux distribution measurements in the reactor core with a central horizontal reflector

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P; Raisic, N; Obradovic, D; Jovanovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1965-05-01

    A series of critical experiments were performed on the RB reactor in order to determine the thermal neutron flux increase in the central horizontal reflector formed by a split reactor core. The objectives of these experiments were to study the possibilities of improving the thermal neutron flux characteristics of the neutron beam in the horizontal beam tube of the RA research reactor. The construction of RA reactor enables to split the core in two, to form a central horizontal reflector in front of the beam tube. This is achieved by replacing 2% enriched uranium slugs in the fuel channel by dummy aluminium slugs. The purpose of the first series of experiments was to study the gain in thermal neutron component inside the horizontal reflector and the loss of reactivity as a function of the lattice pitch and central reflector thickness.

  18. Reference Neutron Radiographs of Nuclear Reactor Fuel

    DEFF Research Database (Denmark)

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

  19. 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)

  20. EPRTM Reactor neutron instrumentation

    International Nuclear Information System (INIS)

    Pfeiffer, Maxime; SALA, Stephanie

    2013-06-01

    The core safety during operation is linked, in particular, to the respect of criteria related to the heat generated in fuel rods and to the heat exchange between the rods and the coolant. This local power information is linked to the power distribution in the core. In order to evaluate the core power distribution, the EPR TM reactor relies on several types of neutron detectors: - ionization chambers located outside the vessel and used for protection and monitoring - a fixed in-core instrumentation based on Cobalt Self Powered Neutron Detectors used for protection and monitoring - a mobile reference in-core instrumentation based on Vanadium aero-balls This document provides a description of this instrumentation and its use in core protection, limitation, monitoring and control functions. In particular, a description of the detectors and the principles of their signal generation is supplied as well as the description of the treatments related to these detectors in the EPR TM reactor I and C systems (including periodical calibration). (authors)

  1. Calculations of spetial and energy distribution of neutrons in reactor core and shielding. Part 1

    International Nuclear Information System (INIS)

    Rataj, J.

    1985-01-01

    The most important methods are briefly described of solving the Boltzmann equation describing neutron transport, such as are used at UJV Rez. These are the methods of P N -approximation, the S N method, further the Monte Carlo method and the moments method in which the solution of the Boltzmann equation describing the spatial and energy distribution of neutron flux density proceeds from the system of equations given by the P N -approximation. It is emphasized that the results are loaded with a big error or are totally erroneous when inaccurate cross sections or unsuitable libraries of cross sections are used. (J.P.)

  2. Analysis of different alternatives for a reactor aimed to produce Co60. Neutronic part

    International Nuclear Information System (INIS)

    Relloso, J.M.

    1990-01-01

    Different alternatives are presented for a reactor aimed to produce radioisotopes with a low enrichment fuel element geometrically equal to those of low enrichment. Different alternatives are presented which essentially differ in the core's configuration, fuel management and 'meat' fuel material: U3O8, U3Si or U3Si2 in aluminium matrix. Fluxes at different irradiation positions and power full peak factors are described for each case. The main aspects to be taken into account for cobalt irradiation and the method to be followed to measure the reactor as a function of a required production are presented. (Author) [es

  3. Neutron source for a reactor

    International Nuclear Information System (INIS)

    Kobayashi, Hiromasa.

    1975-01-01

    Object: To easily increase a start-up power of a reactor without irradiation in other reactors. Structure: A neutron source comprises Cf 252 , a natural antimony rod, a layer of beryllium, and a vessel of neutron source. On upper and lower portion of Cf 252 are arranged natural antimony rods, which are surrounded by the Be layer, the entirety being charged into the vessel. The Cf 252 may emit neutron, has a half life more than a period of operating cycle of the reactor and is less deteriorated even irradiated by radioactive rays while being left within the reactor. The natural antimony rod is radioactivated by neutron from Cf 252 and neutron as reactor power increases to emit γ rays. The Be absorbs γ rays to emit the neutron. The antimony rod is irradiated within the reactor. Further, since the Cf 252 is small in neutron absorption cross section, it is hard to be deteriorated even while being inserted within the reactor. (Kamimura, M.)

  4. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

  5. Safety aspects of neutron noise diagnostics and loose parts monitoring in WWER reactors

    International Nuclear Information System (INIS)

    Por, G.

    1997-01-01

    In this paper our aim is to give very short introduction on different types of well selected noise diagnostics methods and then mentioning their occurrence in WWER reactors we analyze what impact they might have to operational safety and for ageing (which also affects on safety of the installations). We do not deny, that one of our main aim is to call the attention of management staff of NPP, which deals with safety, safety culture, maintenance and operation proving, that such methods and system can give not only benefit to economy but also impact on safety of nuclear installations. 14 refs, 6 figs

  6. Neutron activation analysis of pollutants in human hair using research reactors. Part of a coordinated programme

    International Nuclear Information System (INIS)

    Al-Shahristani, H.I.

    1978-02-01

    Hair is a suitable indicator of man's exposure to trace element environmental pollutants. Bearing this in mind, two hundred and fifty human head hair samples were randomly collected from various regions of Iraq representing the general population. These elements were analyzed by thermal neutron activation analysis and the following elements were instrumentally determined: Cr, Fe, Co, Ni, Zn, As, Se, Br, Ag, Cd, Sb, La, Au, Hg, Th and U. The arithmetic mean, geometric mean, standard deviation and range of concentration of these elements and the frequency distributions within the population are given and compared with concentrations from other regions of the world. In general the concentrations determined for this population are similar to those reported for other areas except for Br and Fe. The possible causes of these anomalies are discussed. For certain population sub-groups, high levels of Au, Ag, Cr, Se and Hg have been measured and the reasons for these deviations are discussed

  7. Reactor utilization, Part 1

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1981-01-01

    The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

  8. Monitor for reactor neutron detector

    International Nuclear Information System (INIS)

    Shirakami, Hisayuki; Shibata, Masatoshi

    1992-01-01

    The device of the present invention judges as to whether a neutron detector is normal or not while considering the change of indication value depending on the power change of a reactor core. That is, the device of the present invention comprises a standard value setting device for setting the standard value for calibrating the neutron detector and an abnormality judging device for comparing the standard value with a measured value of the neutron detector and judging the abnormality when the difference is greater than a predetermined value. The measured value upon initialization of each of the neutron detectors is determined as a quasi-standard value. An average value of the difference between the measured value and the quasi-standard value of a plurality of effective neutron detectors at a same level for the height of the reactor core is multiplied to a power rate based on the reactor core power at a position where the neutron detector is disposed upon calibration. The value obtained by adding the multiplied value and the quasi-standard value is determined as a standard value. The abnormality judging device compares the standard value with the measured value of the neutron detector and, if the difference is greater than a predetermined value, the neutron detector is determined as abnormal. As a result, judgement can be conducted more accurately than conventional cases. (I.S.)

  9. Experimental study of the effect of neutron radiation on pressurised water reactor vessel steel resilience - First part

    International Nuclear Information System (INIS)

    Verdeau, Jean-Jacques

    1969-12-01

    After having outlined the importance of the embrittlement of vessel steels by neutrons during the exploitation of pressurised water reactors, the author reports a set of tests which aimed at determining the effect of neutron irradiation on vessel steel resilience for operated, under construction or projected pressurized water reactors. He also tries to highlight the influence of irradiation temperature and of initial thermal treatments, and to look for a restoration thermal treatment of neutron-induced damages which could be applied to the considered vessels. Tests were performed on V Charpy resilience samples. Some samples have been irradiated by the Pile Department of the Grenoble CEN and then broken by the Laboratory of very high activity, whereas other samples have been irradiated in a prototype vessel and broken by a Cadarache department. The author presents characteristics of the studied steels (chemical compositions, thermal treatments), describes sample irradiation conditions, and the method of assessment of the transition temperature after irradiation, presents experimental results, discusses their interpretation, and presents future tests to be performed [fr

  10. Reactor cold neutron source facility, the first in Japan

    International Nuclear Information System (INIS)

    Utsuro, Masahiko; Maeda, Yutaka; Kawai, Takeshi; Tashiro, Tameyoshi; Sakakibara, Shoji; Katada, Minoru.

    1986-01-01

    In the Research Reactor Institute, Kyoto University, the first cold neutron source facility for the reactor in Japan was installed, and various tests are carried out outside the reactor. Nippon Sanso K.K. had manufactured it. After the prescribed tests outside the reactor, this facility will be installed soon in the reactor, and its outline is described on this occasion. Cold neutrons are those having very small energy by being cooled to about-250 deg C. Since the wavelength of the material waves of cold neutrons is long, and their energy is small, they are very advantageous as an experimental means for clarifying the structure of living body molecules and polymers, the atom configuration in alloys, and atomic and molecular movements by neutron scattering and neutron diffraction. The basic principle of the cold neutron source facility is to irradiate thermal neutrons on a cold moderator kept around 20 K, and to moderate and cool the neutrons by nuclear scattering to convert to cold neutrons. The preparatory research on cold neutrons and hydrogen liquefaction, the basic design to put the cold neutron source facility in the graphite moderator facility, the safety countermeasures, the manufacture and quality control, the operation outside the reactor and the performance are reported. The cold neutron source facility comprises a cold moderator tank and other main parts, a deuterium gas tank, a helium refrigerator and instrumentation. (Kako, I.)

  11. Neutron noise in nuclear reactors

    International Nuclear Information System (INIS)

    Blaquiere, A.; Pachowska, R.

    1961-06-01

    The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [fr

  12. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

    International Nuclear Information System (INIS)

    Hernandez L, H.; Ortiz V, J.

    2003-01-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  13. Neutron excitation function guide for reactor dosimetry

    International Nuclear Information System (INIS)

    Gritzay, O.; Vlasov, M.; Chervonna, L.; Klimova, N.; Kolota, G.; Zerkin, V.

    2002-01-01

    Neutron Excitation Function Guide for Reactor Dosimetry (NEFGRD) has been prepared in the Ukrainian Nuclear Data Center (UKRNDC) using ZVV 9.2 code for graphical data presentation. The data can be retrieved through Web or obtained on CD-ROM or as hard copy report. NEFGRD contains graphical and text information for 56 nuclides (81 dosimetry reactions). Each reaction is provided by the information part and several graphical function blocks (from one to nine). (author)

  14. Research reactor operations for neutron activation analysis

    International Nuclear Information System (INIS)

    Tv'ehlov, Yu.

    2002-01-01

    The IAEA Special Manual devoted to quality control during neutron activation analysis (NAA) on research and test reactors is discussed. Three parts of the publication involve presentation of common rules for performance of NAA, quantitative and qualitative analyses, statistic and systematic errors, safety regulations and radioactive waste management. Besides, the publication contains practical manual for the performance of NAA, and examples of different NAA regulating registration forms are presented [ru

  15. Utilizing horizontal reactors channels for neutron therapy

    International Nuclear Information System (INIS)

    Stankovsky, E.Yu.; Kurachenko, Yu.A.

    2000-01-01

    Two experimental heterogeneous reactors have been considered. The reactors may be applied in neutron capture therapy and in a conventional manner. The channel out of the core serves as the neutron source. At each of these facilities, both fast and epithermal neutron fluxes for BNCT research, human clinical trials, and characterized common computational techniques have been evaluated. (authors)

  16. Neutronic reactor thermal shield

    International Nuclear Information System (INIS)

    Lowe, P.E.

    1976-01-01

    A shield for a nuclear reactor includes at least two layers of alternating wide and narrow rectangular blocks so arranged that the spaces between blocks in adjacent layers are out of registry, each block having an opening therein equally spaced from the sides of the blocks and nearer the top of the block than the bottom, the distance from the top of the block to the opening in one layer being different from this distance in adjacent layers, openings in blocks in adjacent layers being in registry. 1 claim, 7 drawing figures

  17. Parallel Monte Carlo reactor neutronics

    International Nuclear Information System (INIS)

    Blomquist, R.N.; Brown, F.B.

    1994-01-01

    The issues affecting implementation of parallel algorithms for large-scale engineering Monte Carlo neutron transport simulations are discussed. For nuclear reactor calculations, these include load balancing, recoding effort, reproducibility, domain decomposition techniques, I/O minimization, and strategies for different parallel architectures. Two codes were parallelized and tested for performance. The architectures employed include SIMD, MIMD-distributed memory, and workstation network with uneven interactive load. Speedups linear with the number of nodes were achieved

  18. Seminar on Heat-transfer fluids for fast neutron reactors

    International Nuclear Information System (INIS)

    Brechet, Yves; Dautray, Robert; Friedel, Jacques; Brezin, Edouard; Martin, Georges; Pineau, Andre; Carre, Francois; Gauche, Francois; Rodriguez, Guillaume; Latge, Christian; Cabet, Celine; Garnier, Jean-Claude; Bamberger, Yves; Sauvage, Jean-Francois; Buisine, Denis; Agostini, Pietro; Ulyanov, Vladimir; Auger, Thierry; Heuer, Daniel; Ghetta, Veronique; Bubelis, Evaldas; Charlaix, Elisabeth; Barrat, Jean-Louis; Boquet, Lyderic; Glickman, Evgueny; Escaravage, Claude

    2014-03-01

    This book reports the content of a two-day meeting held by the Academy of Sciences on the use of heat-transfer fluids in fast neutron reactors. After a first part which proposes an overview of scientific and technical problems related to these heat-transfer fluids (heat transfer process, nuclear properties, chemistry, materials, risks), a contribution proposes a return on experience on the use of heat-transfer fluids in the different design options of reactors of fourth generation: from mercury to NaK in the first fast neutron reactor projects, specific assets and constraints of sodium used as heat-transfer fluid, concepts of fast neutron reactors cooled by something else than sodium, perspectives for projects and research in fast neutron reactors. The next contribution discusses the specifications of future fast-neutron reactors: expectations for fourth-generation reactors, expectations in terms of performance and of safety, specific challenges. The last contribution addresses actions to be undertaken in the field of research and development: actions regarding all reactor types or specific types as sodium-cooled reactors, lead cooled reactors, molten salt reactors, and gas-cooled fast reactors

  19. General remarks on fast neutron reactor physics

    International Nuclear Information System (INIS)

    Barre, J.Y.

    1980-01-01

    The main aspects of fast reactor physics, presented in these lecture notes, are restricted to LMFBR's. The emphasis is placed on the core neutronic balance and the burn-up problems. After a brief description of the power reactor main components and of the fast reactor chronology, the fundamental parameters of the one-group neutronic balance are briefly reviewed. Then the neutronic burn-up problems related to the Pu production and to the doubling time are considered

  20. Application of reactors for testing neutron-induced upsets in commercial SRAMs

    International Nuclear Information System (INIS)

    Griffin, P.J.; Luera, T.F.; Sexton, F.W.; Cooper, P.J.; Karr, S.G.; Hash, G.L.; Fuller, E.

    1997-01-01

    Reactor neutron environments can be used to test/screen the sensitivity of unhardened commercial SRAMs to low-LET neutron-induced upset. Tests indicate both thermal/epithermal (< 1 keV) and fast neutrons can cause upsets in unhardened parts. Measured upset rates in reactor environments can be used to model the upset rate for arbitrary neutron spectra

  1. Opportunities for TRIGA reactors in neutron radiography

    International Nuclear Information System (INIS)

    Barton, John P.

    1978-01-01

    In this country the two most recent installations of TRIGA reactors have both been for neutron radiography, one at HEDL and the other at ANL. Meanwhile, a major portion of the commercial neutron radiography is performed on a TRIGA fueled reactor at Aerotest. Each of these installations has different primary objectives and some comparative observations can be drawn. Another interesting comparison is between the TRIGA reactors for neutron radiography and other small reactors that are being installed for this purpose such as the MIRENE slow pulse reactors in France, a U-233 fueled reactor for neutron radiography in India and the L88 solution reactor in Denmark. At Monsanto Laboratory, in Ohio, a subcritical reactor based on MTR-type fuel has recently been purchased for neutron radiography. Such systems, when driven by a Van de Graaff neutron source, will be compared with the standard TRIGA reactor. Future demands on TRIGA or competitive systems for neutron radiography are likely to include the pulsing capability of the reactor, and also the extraction of cold neutron beams and resonance energy beams. Experiments recently performed on the Oregon State TRIGA Reactor provide information in each of these categories. A point of particular current concern is a comparison made between the resonance energy beam intensity extracted from the edge of the TRIGA core and from a slot which penetrated to the center of the TREAT reactor. These results indicate that by using such slots on a TRIGA, resonance energy intensities could be extracted that are much higher than previously predicted. (author)

  2. A neutron radiography facility on the IRT-2000 reactor

    International Nuclear Information System (INIS)

    Khadduri, I.Y.

    1976-01-01

    A neutron radiography facility has been constructed on the thermal neutron channel of the IRT-2000 reactor. A collimated thermal neutron beam exposure area of 10 cm diameter is obtained with an L/D ratio of 48.8. The film used is cellulose nitrate coated with lithium tetraborate which is insensitive to gamma and beta radiation. Some pictures with good contrast and resolution have been obtained. Pictures of parts of an IRT-2000 reactor fuel pin have also been recorded. (orig) [de

  3. 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

  4. 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)

  5. Neutron inelastic scattering studies of crystal hydrates. Part of a coordinated programme on the application of research reactor neutron scattering techniques in the study of solids

    International Nuclear Information System (INIS)

    Amilius, Z.

    1977-12-01

    Using an inverted berillium filter spectrometer, neutrons inelastically scattered from the samples of hydrated and deuterated crystals of Cu Cl 2 .2H 2 O, Cu Cl 2 .2H 2 O and K 2 Cu Cl 2 .2D 2 O were measured, with the objective of determining the librational frequencies of the water molecule. The experiments were performed at room temperature only, permitting to draw only limited conclusions on the assignment of specific librational modes to the peak in the observed neutron spectra

  6. Neutron flux measuring system for nuclear reactor

    International Nuclear Information System (INIS)

    Aoki, Kazuo.

    1977-01-01

    Purpose: To avoid the generation of an undesired scram signal due to abrupt changes in the neutron level given to the detectors disposed near the boundary between the moderator and the atmosphere. Constitution: In a nuclear reactor adapted to conduct power control by the change of the level in the moderator such as heavy water, the outputs from the neutron detectors disposed vertically are averaged and the nuclear reactor is scramed corresponding to the averaged value. In this system, moderator level detectors are additionally provided to the nuclear reactor and their outputs, moderator level signal, are sent to a power averaging device where the output signals of the neutron detectors are judged if they are delivered from neutrons in the moderator or not depending on the magnitude of the level signal and the outputs of the detectors out of the moderator are substantially excluded. The reactor interlock signal from the device is utilized as a scram signal. (Seki, T.)

  7. Neutron Radiography at the RP-10 reactor

    International Nuclear Information System (INIS)

    Hinostrosa, H.; Ravello, Y.; Cornejo, N.; Mendoza, M.; Montoya, M.

    1992-01-01

    The facility of neutron radiography at the RP-10 peruvian research reactor is described. The factor of collimation L/D is 149; the Cadmium ratio for the gold in the inspection's area is 4.5, and the thermal neutrons flux on the sample is 3,14 x 10 6 n/cm 2 s (author). 5 refs. 5 fig

  8. 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

  9. Neutronic of heterogenous gas cooled reactors

    International Nuclear Information System (INIS)

    Maturana, Roberto Hernan

    2008-01-01

    At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design. [es

  10. Theoretical analysis of nuclear reactors (Phase II), I-V, Part V, Determining the fine neutron flux distribution by Pn method

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-10-01

    Expression for spherical harmonic moments were applied. They were obtained by spherical harmonics expansion of monoenergetic transport equation. This report presents the procedure for calculating the neutron flux distribution in the nine-zone reactor cell of the RA reactor in Vinca. The procedure was modelled for digital computer ZUSE Z-23 by expansion of the diagram of the automated P 3 code, which is adequate for P n code with minor changes. The needed subroutines were developed. The most important ones were those for modified first and second order Bessel functions of n-th order. Computer Z-23 was operating only 15 hours during three months, and thus only the subroutines for modified Bessel functions could be tested and the obtained results were excellent. For the mentioned reason the neutron flux distribution will be calculated in the forthcoming period [sr

  11. Commissioning of the Opal reactor cold neutron source

    International Nuclear Information System (INIS)

    Thiering, R.; Lu, W.; Ullah, R.

    2006-01-01

    Full text: At OPAL, Australia's first cold neutron facility will form an essential part of the reactor's research programs. Fast neutrons, born in the core of a reactor, interact with a cryogenic material, in this case liquid deuterium, to give them very low energies ( 1 0 m eV). A cold neutron flux of 1.4 1 0 E 1 4 n /cm 2/ s is expected, with a peak in the energy spectrum at 4.2m eV. The cold neutron source reached cryogenic conditions for the first time in late 2005. The cold neutron source operates with a sub-cooled liquid Deuterium moderator at 24 K. The moderator chamber, which contains the deuterium, has been constructed from AlMg 5. The thermosiphon and moderator chamber are cooled by helium gas, in a natural convection thermosiphon loop. The helium refrigeration system utilises the Brayton cycle, and is fully insulated within a high vacuum environment. Despite the proximity of the cold neutron source to the reactor core, it has been considered as effectively separate to the reactor system, due to the design of its special vacuum containment vessel. As OPAL is a multipurpose research reactor, used for beam research as well as radiopharmaceutical production and industrial irradiations, the cold neutron source has been designed with a stand-by mode, to maximise production. The stand-by mode is a warm operating mode using only gaseous deuterium at ambient temperatures (∼ 3 00 K ), allowing for continued reactor operations whilst parts of the cold source are unavailable or in maintenance. This is the first time such a stand-by feature has been incorporated into a cold source facility

  12. Calculation of fluences of fast neutrons hitting the pressure vessel of the Dukovany NPP WWER-440 reactor. Part I. Theory, calculations, comparison with the experiment

    International Nuclear Information System (INIS)

    Rataj, J.

    1993-10-01

    The method of calculating neutron spectra and integral flux densities of neutrons hitting the pressure vessel of the Dukovany NPP WWER-440 reactor is outlined. The one-dimensional and two-dimensional calculations were performed by means of the DORT code in R, R-Z, and R-Θ geometries using the cross sections from the ELXSIR library. In the R-Θ geometry, the coupled neutron flux densities were determined. The calculated values of the maximum activation of detectors differ less than 15% from the values measured in surveillance specimens, which is within the limit of uncertainty associated with the position of the detector in the casing. The differences between the calculated and observed data behind the pressure vessel were below 4%. 10 tabs., 3 figs., 41 refs

  13. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H; Ortiz V, J [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  14. Reference neutron radiations. Part 1: Characteristics and methods of production

    International Nuclear Information System (INIS)

    2001-01-01

    ISO 8529 consists of the following parts, under the general title Reference neutron radiations: Part 1: Characteristics and methods of production; Part 2: Calibration fundamentals of radiation protection devices related to the basic quantities characterizing the radiation field; Part 3: Calibration of area and personal dosimeters and determination of response as a function of energy and angle of incidence. This Part 1. of ISO 8529 specifies the reference neutron radiations, in the energy range from thermal up to 20 MeV, for calibrating neutron-measuring devices used for radiation protection purposes and for determining their response as a function of neutron energy. Reference radiations are given for neutron fluence rates of up to 1x10 9 m 2 s-1 , corresponding, at a neutron energy of 1 MeV, to dose-equivalent rates of up to 100 mSv h -1 . This part of ISO 8529 is concerned only with the methods of producing and characterizing the neutron reference radiations. The procedures for applying these radiations for calibrations are described in ISO 8529-2 and ISO 8529-3. The reference radiations specified are the following: neutrons from radionuclide sources, including neutrons from sources in a moderator; neutrons produced by nuclear reactions with charged particles from accelerators; neutrons from reactors. In view of the methods of production and use of them, these reference radiations are divided, for the purposes of this part of ISO 8529, into the following two separate sections. In clause 4, radionuclide neutron sources with wide spectra are specified for the calibration of neutron measuring devices. These sources should be used by laboratories engaged in the routine calibration of neutron-measuring devices, the particular design of which has already been type tested. In clause 5, accelerator-produced monoenergetic neutrons and reactor-produced neutrons with wide or quasi monoenergetic spectra are specified for determining the response of neutron-measuring devices

  15. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

    The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

  16. On the research activities in reactor and neutron physics using the first egyptian research reactor

    International Nuclear Information System (INIS)

    Hassan, A.M.

    2000-01-01

    A review on the most important research activities in reactor and neutron physics using the first Egyptian Research Reactor (ET-RR-1) is given. An out look on: neutron cross-sections, neutron flux, neutron capture gamma-ray spectroscopy, neutron activation analysis, neutron diffraction and radiation shielding experiments, is presented

  17. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    Rousseau, A.; Genthon, J.P.

    1958-01-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  18. Neutron flux distribution forecasting device of reactor

    International Nuclear Information System (INIS)

    Uematsu, Hitoshi

    1991-01-01

    A neutron flux distribution is forecast by using current data obtained from a reactor. That is, the device of the present invention comprises (1) a neutron flux monitor disposed in various positions in the reactor, (2) a forecasting means for calculating and forecasting a one-dimensional neutron flux distribution relative to imaginable events by using data obtained from the neutron flux monitor and physical models, and (3) a display means for displaying the results forecast in the forecasting means to a reactor operation console. Since the forecast values for the one-dimensional neutron flux distribution relative to the imaginable events are calculated in the device of the present invention by using data obtained from the neutron flux monitor and the physical models, the data as a base of the calculation are new and the period for calculating the forecast values can be shortened. Accordingly, although there is a worry of providing some errors in the forecast values, they can be utilized sufficiently as reference data. As a result, the reactor can be operated more appropriately. (I.N.)

  19. Neutron personnel dosimetry considerations for fusion reactors

    International Nuclear Information System (INIS)

    Barton, T.P.; Easterly, C.E.

    1979-07-01

    The increasing development of fusion reactor technology warrants an evaluation of personnel neutron dosimetry systems to aid in the concurrent development of a radiation protection program. For this reason, current state of knowledge neutron dosimeters have been reviewed with emphasis placed on practical utilization and the problems inherent in each type of dosimetry system. Evaluations of salient parameters such as energy response, latent image instability, and minimum detectable dose equivalent are presented for nuclear emulsion films, track etch techniques, albedo and other thermoluminescent dosimetry techniques, electrical conductivity damage effects, lyoluminescence, thermocurrent, and thermally stimulated exoelectron emission. Brief summaries of dosimetry regulatory requirements and intercomparison study results help to establish compliance and recent trends, respectively. Spectrum modeling data generated by the Neutron Physics Division of Oak Ridge National Laboratory for the Princeton Tokamak Fusion Test Reactor (TFTR) Facility have been analyzed by both International Commission on Radiological Protection fluence to dose conversion factors and an adjoint technique of radiation dosimetry, in an attempt to determine the applicability of current neutron dosimetry systems to deuterium and tritium fusion reactor leakage spectra. Based on the modeling data, a wide range of neutron energies will probably be present in the leakage spectra of the TFTR facility, and no appreciable risk of somatic injury to occupationally exposed workers is expected. The relative dose contributions due to high energy and thermal neutrons indicate that neutron dosimetry will probably not be a serious limitation in the development of fusion power

  20. Neutronics methods for transient and safety analysis of fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marchetti, Marco

    2017-07-01

    Modeling the evolution of possible or postulated accidents in nuclear reactors is fundamental in designing safe systems. For the next generation of reactors, in particular fast reactors, fuel movement during an accident can, in principle, drive an energetic event. Such is the issue of recriticality. The thermal energy produced during these events will, possibly, be converted into mechanical energy by some mechanisms. For example, the nuclear heat deposited in the fuel could cause fuel vaporization and its subsequent expansion. This movement would accelerate the surrounding sodium: part of the initial energy in the fuel is thus converted into sodium kinetic energy. This mechanical energy will finally be absorbed, in some way or another, by the reactor vessel. Providing an accurate estimate for the maximum mechanical work that any accidental sequence can do onto the reactor vessel is an essential step in designing a reactor containment that would withstand any load generated by any accident. That would assure accident containment, without consequences for the general public. Fast reactor accident modeling is a complicated task. The outcome of an accident is determined by different physical phenomena, all acting at almost the same time. Safety analysts must track all these different phenomena. Multi-physics codes have been developed for this task. They must contain accurate models for fluid-dynamics, neutronics, and structures. This work has to do with neutronics modeling of such accidents. Past and recent analyses have been limited to the approximate description of the neutronic field, for example by using a rough description of the energy and/or of the angular dependence of the neutron flux. In this work, different neutronic solvers are selected and coupled into a general multi-physics code for fast reactor accident analysis. Performances of each of them is then assessed. Some emphasis has been put also in assessing the speed of these solvers for determining the

  1. Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

    International Nuclear Information System (INIS)

    Jammes, Ch.

    2010-07-01

    After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

  2. Nuclear reactor with scrammable part length rod

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

    A new part length rod is provided. It may be used to control xenon induced power oscillations but to contribute to shutdown reactivity when a rapid shutdown of the reactor is required. The part length rod consists of a control rod with three regions. The lower control region is a longer weaker active portion separated from an upper stronger shorter poison section by an intermediate section which is a relative non-absorber of neutrons. The combination of the longer weaker control section with the upper high worth poison section permits the part length rod of this to be scrammed into the core when a reactor shutdown is required but also permits the control rod to be used as a tool to control power distribution in both the axial and radial directions during normal operation

  3. Neutron flux measurements in PUSPATI Triga Reactor

    International Nuclear Information System (INIS)

    Gui Ah Auu; Mohamad Amin Sharifuldin Salleh; Mohamad Ali Sufi.

    1983-01-01

    Neutron flux measurement in the PUSPATI TRIGA Reactor (PTR) was initiated after its commissioning on 28 June 1982. Initial measured thermal neutron flux at the bottom of the rotary specimen rack (rotating) and in-core pneumatic terminus were 3.81E+11 n/cm 2 sec and 1.10E+12n/cm 2 sec respectively at 100KW. Work to complete the neutron flux data are still going on. The cadmium ratio, thermal and epithermal neutron flux are measured in the reactor core, rotary specimen rack, in-core pneumatic terminus and thermal column. Bare and Cadmium covered gold foils and wires are used for the above measurement. The activities of the irradiated gold foils and wires are determined using Ge(Li) and hyperpure germinium detectors. (author)

  4. Solving the uncommon reactor core neutronics problems

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1980-01-01

    The common reactor core neutronics problems have fundamental neutron space, energy spectrum solutions. Typically the most positive eigenvalue is associated with an all-positive flux for the pseudo-steady-state condition (k/sub eff/), or the critical state is to be effected by selective adjustment of some variable such as the fuel concentration. With sophistication in reactor analysis has come the demand for solutions of other, uncommon neutronics problems. Importance functionss are needed for sensitivity and uncertainty analyses, as for ratios of intergral reaction rates such as the fuel conversion (breeding) ratio. The dominant higher harmonic solution is needed in stability analysis. Typically the desired neutronics solution must contain negative values to qualify as a higher harmonic or to satisfy a fixed source containing negative values. Both regular and adjoint solutions are of interest as are special integrals of the solutions to support analysis

  5. Maximum neutron flux in thermal reactors

    International Nuclear Information System (INIS)

    Strugar, P.V.

    1968-12-01

    Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples

  6. Neutron noise measurement technique in a coupled reactor

    International Nuclear Information System (INIS)

    Genoud, J.P.

    1976-01-01

    Describes work carried out on the swimming pool reactor at the Physikalisch-Technische Bundesanstalt at Braunschweig. The reactor has two multiplying zones, is light water moderated, with 90% enriched 235 U fuel. There is a D 2 0 reservoir between the two parts of the reactor. Signal/noise ratio obtained by means of ionisation chamber type neutron detectors of 10 -13 amp/u.f. sensitivity is of the order of 40 dB and band frequency 1.5 kHz. Spectral density of the interzone interaction energy was obtained by use of Fourier transforms, previously corrected by a Hanning window. (S.W.)

  7. 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

  8. Neutron converter at reactor RB; Konvertor neutrona na reaktoru RB

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P; Sotic, P; Ninkovic, M; Pesic, M [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1977-07-01

    A neutron converter at Reactor RB in the 'Boris Kidric' Institute of Nuclear Sciences - Vinca has been constructed. Preliminary measurements have been shown that the converted neutron spectrum is very similar to the fission neutron spectrum. For the same integral reactor power, the measured neutron radiation dose has been for about ten times larger with the neutron converter. The neutron converter offers wide possibilities, as in investigations in the reactor physics, where the fission neutron spectra have been required, as well as in the field of neutron dosimetry and biological irradiations (author)

  9. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  10. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

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

  11. 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

  12. 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

  13. Neutronics analysis of Dalat Research Reactor

    International Nuclear Information System (INIS)

    Pham Van Lam; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong; Nguyen Manh Hung; Pham Hong Son; Tran Quoc Duong

    2006-01-01

    Many neutronics codes have been used to calculate for Dalat Research Reactor (DRR) from 1983 (the first critical of DRR in December, 1983). The purposes of all calculations are to know exactly many important parameters related to Reactor Physics and Neutron Physics in reactor core. The results from calculation play important role in core and fuel management for DRR. Especially basing on the results we can predict about fuel cycle, fuel burn up distribution and plan for using optimize remain fresh fuel assemblies of DRR. By using system neutronics code including transport codes, diffusion codes and Mote Carlo code, many characteristics of fuel assemblies and other parameters of whole core were received such as main features of VVR-M2 fuel assembly type, multiplication factor, neutron flux distribution, power distribution, burn up distribution, excess reactivity, control rods worth, neutron spectrum, temperature reactivity coefficient ect. In the paper, brief description all computer codes to being used in DRR and the calculation results from the codes above are presented. (author)

  14. maximum neutron flux at thermal nuclear reactors

    International Nuclear Information System (INIS)

    Strugar, P.

    1968-10-01

    Since actual research reactors are technically complicated and expensive facilities it is important to achieve savings by appropriate reactor lattice configurations. There is a number of papers, and practical examples of reactors with central reflector, dealing with spatial distribution of fuel elements which would result in higher neutron flux. Common disadvantage of all the solutions is that the choice of best solution is done starting from the anticipated spatial distributions of fuel elements. The weakness of these approaches is lack of defined optimization criteria. Direct approach is defined as follows: determine the spatial distribution of fuel concentration starting from the condition of maximum neutron flux by fulfilling the thermal constraints. Thus the problem of determining the maximum neutron flux is solving a variational problem which is beyond the possibilities of classical variational calculation. This variational problem has been successfully solved by applying the maximum principle of Pontrjagin. Optimum distribution of fuel concentration was obtained in explicit analytical form. Thus, spatial distribution of the neutron flux and critical dimensions of quite complex reactor system are calculated in a relatively simple way. In addition to the fact that the results are innovative this approach is interesting because of the optimization procedure itself [sr

  15. Neutronics in ICF reactor ''SENRI-I''

    International Nuclear Information System (INIS)

    Nakai, S.; Ido, S.; Yamanaka, C.

    1983-01-01

    The neutronic behavior of SENRI-I has been examined taking into account the effect of fuel rhoR and Pb tamper on the emitted neutron from micro-explosion. One dimensional neutron transport was calculated by ANISIN-JR code with the nuclear data GICX-40. The effect of beam ports on neutronics and neutron streaming was examined by the three dimensional Monte-Carlo calculation code MORSE-E with the same nuclear data. The emitted neutrons are softened noticeably by the increase of the compressed fuel rhoR and the thickness of Pb coating. The latter also multiplies the net neutron number from pellet. The energy deposition and temperature increase and its distribution in the blankets and structural elements were obtained as a function of neutron spectrum from pellet. As for the tritium breeding ratio, the softening of neutron has little effect because the decrease of breeding by 7 Li with softening is compensated by the increase of breeding by 6 Li. The breeding ratio was 1.678, 1.639 and 1.576 with 14 MeV neutron, rhoR=0.7, rhoR=3 and rhoR=6 respectively. Neutron shielding and streaming from beam ports were examined and the dose rate of final optical elements were calculated to estimate the life of mirror. All these results show the feasibility of SENRI-I as a long life, maintenance free ICF pulse reactor and motivate to go further investigation and design studies in detail. (author)

  16. 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)

  17. Determination of the Neutron Flux in the Reactor Zones with the Strong Neutron Absorption and Leakage

    International Nuclear Information System (INIS)

    Ljubenov, V.; Milosevic, M.

    2004-11-01

    The procedures for the numerical and experimental determination of the neutron flux in the zones with the strong neutron absorption and leakage are described in this paper. Numerical procedure is based on the application of the SCALE-4.4a code system where the Dancoff factors are determined by the VEGA2DAN code. Two main parts of the experimental methodology are measurement of the activity of irradiated foils and determination of the averaged neutron absorption cross-section in the foils by the SCALE-4.4a calculation procedure. The proposed procedures have been applied for the determination of the neutron flux in the internal neutron converter used with the RB reactor core configuration number 114. (author)

  18. Neutron measurements at nuclear power reactors [55

    CERN Document Server

    Scherpelz, R I

    2002-01-01

    Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D sub 2 O-moderated sup 2 sup 5 sup 2 Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participate...

  19. 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

  20. Activation analysis with reactor neutrons

    International Nuclear Information System (INIS)

    Gangadharan, S.

    1983-01-01

    The potentialities of neutron as an analytical probe are indicated, pointing out the need for development of other approaches, besides the conventional activation method. Development of instrumental approach to activation and applications, carried out at Analytical Chemistry Division are outlined. The role of, and the need for, the development and application of mathematical methods in enhancing the information content, and in turn the interpretation of the analytical results, is demonstrated. (author)

  1. Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

    OpenAIRE

    Verma, Vasudha

    2017-01-01

    Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

  2. Reactor internals vibration monitoring by neutron noise methods in PWRs

    International Nuclear Information System (INIS)

    Pazsit, I.; Por, G.; Lux, I.

    1983-01-01

    Certain elements of PWR cores such as control/fuel rods or cassettes, or other parts of reactor internals, often represent a vibration problem. Early analyses at operating PWR plant revealed that these vibrations can be detected by in-core neutron detectors, opening up the possibility of vibration monitoring and diagnostics by noise methods. Theoretical methods of calculating vibration induced neutron noise and its application to vibration diagnostics are summarized. Experiments to check theoretical conclusions are under way at the Central Research Institute for Physics, Budapest. (author)

  3. Quasielastic neutron scattering facility at Dhruva reactor

    International Nuclear Information System (INIS)

    Mukhopadhyay, R.; Mitra, S.; Paranjpe, S.K.; Dasannacharya, B.A.

    2001-01-01

    Quasi-elastic neutron scattering is a powerful experimental tool for studying the various dynamical motions in solids and liquids. In this paper, we have described the salient features of the quasi-elastic neutron spectrometer in operation at Dhruva reactor at Trombay, India. The design criteria have been such as to maximise the throughput by various means like closer approach to the source, focusing a larger beam on to a sample, and Multi-Angle Reflecting X-tal mode of energy analysis. Some results of molecular motions from recently studied systems using this spectrometer are also reported

  4. Prompt Neutron Lifetime for the NBSR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, A.L.; Diamond, D.

    2012-06-24

    In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

  5. Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire. [Serbo-Croat] Ovaj izvestaj sadrzi sledece referate: Fluks termalnih neutrona u eksperimentalnim kanalima reaktora RA; Fluks epitermalnih neutrona u eksperimentalnim kanalima reaktora RA; Fluks brzih neutrona u eksperimentalnim kanalima reaktora RA; Fluks termalnih neurona u termalnoj koloni i bioloskom eksperimentalnom kanalu; Neutronska merenja u elementarnoj celiji reaktora RA; Temperaturni koeficijent reaktivnosti reaktora RA; Projekat uredjaja za merenje radioaktivnosti zice.

  6. LIGHT WATER MODERATED NEUTRONIC REACTOR

    Science.gov (United States)

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  7. Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

    International Nuclear Information System (INIS)

    Jammes, Christian; Filliatre, Philippe; Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan

    2015-01-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

  8. Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, Christian; Filliatre, Philippe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan [Division of Applied Nuclear Physics, Uppsala University, SE-75120 Uppsala, (Sweden)

    2015-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

  9. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matrices. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  10. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matricies. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  11. 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.)

  12. Neutronic study of the two french heavy water reactors

    International Nuclear Information System (INIS)

    Horowitz, J.

    1955-01-01

    The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [fr

  13. Neutron importance and the generalized Green function for the conventionally critical reactor with normalized neutron distribution

    International Nuclear Information System (INIS)

    Khromov, V.V.

    1978-01-01

    The notion of neutron importance when applied to nuclear reactor statics problems described by time-independent homogeneous equations of neutron transport with provision for normalization of neutron distribution is considered. An equation has been obtained for the function of neutron importance in a conditionally critical reactor with respect to an arbitrary nons linear functional determined for the normalized neutron distribution. Relation between this function and the generalized Green function of the selfconjugated operator of the reactor equation is determined and the formula of small perturbations for the functionals of a conditionally critical reactor is deduced

  14. Neutron flux enhancement in the NRAD reactor

    International Nuclear Information System (INIS)

    Weeks, A.A.; Heidel, C.C.; Imel, G.R.

    1988-01-01

    In 1987 a series of experiments were conducted at the NRAD reactor facility at Argonne National Laboratory - West (ANL-W) to investigate the possibility of increasing the thermal neutron content at the end of the reactor's east beam tube through the use of hydrogenous flux traps. It was desired to increase the thermal flux for a series of experiments to be performed in the east radiography cell, in which the enhanced flux was required in a relatively small volume. Hence, it was feasible to attempt to focus the cross section of the beam to a smaller area. Two flux traps were constructed from unborated polypropylene and tested to determine their effectiveness. Both traps were open to the entire cross-sectional area of the neutron beam (as it emerges from the wall and enters the beam room). The sides then converged such that at the end of the trap the beam would be 'focused' to a greater intensity. The differences in the two flux traps were primarily in length, and hence angle to the beam as the inlet and outlet cross-sectional areas were held constant. It should be noted that merely placing a slab of polypropylene in the beam will not yield significant multiplication as neutrons are primarily scattered away

  15. Neutronics of a mixed-flow gas-core reactor

    International Nuclear Information System (INIS)

    Soran, P.D.; Hansen, G.E.

    1977-11-01

    The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

  16. Neutron physics computation of CERCA fuel elements for Maria Reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.J.; Kulikowska, T.; Marcinkowska, Z.

    2008-01-01

    Neutron physics parameters of CERCA design fuel elements were calculated in the framework of the RERTR (Reduced Enrichment for Research and Test Reactors) program for Maria reactor. The analysis comprises burnup of experimental CERCA design fuel elements for 4 cycles in Maria Reactor To predict the behavior of the mixed core the differences between the CERCA fuel (485 g U-235 as U 3 Si 2 , 5 fuel tubes, low enrichment 19.75 % - LEU) and the presently used MR-6 fuel (430 g as UO 2 , 6 fuel tubes, high enrichment 36 % - HEU) had to be taken into account. The basic tool used in neutron-physics analysis of Maria reactor is program REBUS using in its dedicated libraries of effective microscopic cross sections. The cross sections were prepared using WIMS-ANL code, taking into account the actual structure, temperature and material composition of the fuel elements required preparation of new libraries.The problem is described in the first part of the present paper. In the second part the applicability of the new library is shown on the basis of the fuel core computational analysis. (author)

  17. 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

  18. Determination of the kinetic parameters of the CALIBAN metallic core reactor from stochastic neutron measurements

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, P.; Authier, N.; Chapelle, A. [Commissariat a l' Energie Atomique et Aux Energies Alternatives, CEA, DAM, F-21120 Is sur Tille (France)

    2012-07-01

    Several experimental devices are operated by the Criticality and Neutron Science Research Dept. of the CEA Valduc Laboratory. One of these is the Caliban metallic core reactor. The purpose of this study is to develop and perform experiments allowing to determinate some of fundamental kinetic parameters of the reactor. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as Rossi-{alpha} and Feynman variance-to-mean methods. Subcritical, critical, and even supercritical experiments were performed. Fission chambers detectors were put nearby the core and measurements were analyzed with the Rossi-{alpha} technique. A new value of the prompt neutron decay constant at criticality was determined, which allows, using the Nelson number method, new evaluations of the effective delayed neutron fraction and the in core neutron lifetime. As an introduction of this paper, some motivations of this work are given in part 1. In part 2, principles of the noise measurements experiments performed at the CEA Valduc Laboratory are reminded. The Caliban reactor is described in part 3. Stochastic neutron measurements analysis techniques used in this study are then presented in part 4. Results of fission chamber experiments are summarized in part 5. Part 6 is devoted to the current work, improvement of the experimental device using He 3 neutron detectors and first results obtained with it. Finally, conclusions and perspectives are given in part 7. (authors)

  19. The application of semianalytic method for calculating the thickness of biological shields of nuclear reactors. Part 1. Theoretical basis of a semianalytic method. Attenuation of neutrons' radiation

    International Nuclear Information System (INIS)

    Lukaszek, W.; Kucypera, S.

    1982-01-01

    The basis of a semianalytic method for calculating attenuation of rays (neutron, gamma) in material medium is described. The method was applied in determining the neutrons' flux density in one dimensional Cartesian geometry of the reflector and the shield. (author)

  20. Fast-neutron nuclear reactor vessel

    International Nuclear Information System (INIS)

    Presciuttini, L.

    1984-01-01

    The reactor vessel comprises a cylindrical shell, of which axis is vertical, coupled at its lower part with a dished bottom. The reactor core rests on a support plate bearing on the bottom of the vessel. The above dished bottom comprises a spherical sector having the same radius and the same axis as the cylindrical shell and joining the lower part of the shell, and a spherical head of which radius is a little more important than the spherical sector one. A cylindrical support for the reactor core is attached to the vessel at the joint between the two dished sections. The invention applies more particularly to integrated type reactors cooled by liquid sodium [fr

  1. Neutron physics for nuclear reactors unpublished writings by Enrico Fermi

    CERN Document Server

    Fermi, Enrico; Pisanti, O

    2010-01-01

    This unique volume gives an accurate and very detailed description of the functioning and operation of basic nuclear reactors, as emerging from yet unpublished papers by Nobel Laureate Enrico Fermi. In the first part, the entire course of lectures on Neutron Physics delivered by Fermi at Los Alamos is reported, according to the version made by Anthony P French. Here, the fundamental physical phenomena are described very clearly and comprehensively, giving the appropriate physics grounds for the functioning of nuclear piles. In the second part, all the patents issued by Fermi (and coworkers) on

  2. Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This report includes results of the following measurements performed at the RA reactor: thermal neutron flux in the experimental channels, epithermal and fast neutron flux, neutron flux in the biological shield, neutron flux distribution in the reactor cell.

  3. Core of a fast neutron nuclear reactor

    International Nuclear Information System (INIS)

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

    1974-01-01

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

  4. Fast neutron dosimetry in research reactors

    International Nuclear Information System (INIS)

    Eckert, R.

    1960-01-01

    This work chiefly concerns the measurement of fast neutron fluxes by means of threshold detectors. It is shown first that the cross sections to use for measurements by threshold detectors depend largely on the neutron spectrum, that is the position in which the measurement is performed. The spectrum is determined by calculation for several positions in the piles EL2 and EL3; from this can be deduced the cross-sections to be used for the measurements carried out in these positions. In the last part of the report, possible methods for the experimental determination of the spectrum are indicated. (author) [fr

  5. Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

    International Nuclear Information System (INIS)

    Hirayama, Shoichi; Kawabe, Takaya

    1995-01-01

    The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

  6. Condensed matter and materials research using neutron diffraction and spectroscopy: reactor and pulsed neutron sources

    International Nuclear Information System (INIS)

    Bisanti, Paola; Lovesey, S.W.

    1987-05-01

    The paper provides a short, and partial view of the neutron scattering technique applied to condensed matter and materials research. Reactor and accelerator-based neutron spectrometers are discussed, together with examples of research projects that illustrate the puissance and modern applications of neutron scattering. Some examples are chosen to show the range of facilities available at the medium flux reactor operated by Casaccia ENEA, Roma and the advanced, pulsed spallation neutron source at the Rutherford Appleton Laboratory, Oxfordshire. (author)

  7. 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)

  8. Small Angle Neutron Scattering instrument at Malaysian TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mohd, Shukri; Kassim, Razali; Mahmood, Zal Uyun [Malaysian Inst. for Nuclear Technology Research (MINT), Bangi, Kajang (Malaysia); Radiman, Shahidan

    1998-10-01

    The TRIGA MARK II Research reactor at the Malaysian Institute for Nuclear Research (MINT) was commissioned in July 1982. Since then various works have been performed to utilise the neutrons produced from this steady state reactor. One of the project involved the Small Angle Neutron Scattering (SANS). (author)

  9. Neutron flux enhancement in the NRAD reactor

    International Nuclear Information System (INIS)

    Weeks, A.A.; Heidel, C.C.; Imel, G.R.

    1988-01-01

    In 1987 a series of experiments were conducted at the NRAD reactor facility at Argonne National Laboratory - West (ANL-W) to investigate the possibility of increasing the thermal neutron content at the end of the reactor's east beam tube through the use of hydrogenous flux traps. It was desired to increase the thermal flux for a series of experiments to be performed in the east radiography cell, in which the enhanced flux was required in a relatively small volume. Hence, it was feasible to attempt to focus the cross section of the beam to a smaller area. Two flux traps were constructed from unborated polypropylene and tested to determine their effectiveness. Both traps were open to the entire cross-sectional area of the neutron beam (as it emerges from the wall and enters the beam room). The sides then converged such that at the end of the trap the beam would be 'focused' to a greater intensity. The differences in the two flux traps were primarily in length, and hence angle to the beam as the inlet and outlet cross-sectional areas were held constant. The experiments have contributed to the design of a flux trap in which a thermal flux of nearly 10 9 was obtained, with an enhancement of 6.61

  10. 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.)

  11. Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

    International Nuclear Information System (INIS)

    Nguyen Hoang Hai; Do Quang Binh

    2011-01-01

    Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

  12. Experimental determination of neutron temperature distribution in reactor cell

    International Nuclear Information System (INIS)

    Bosevski, T.

    1965-12-01

    This paper describes theoretical preparation of the experiment for measuring neutron temperature distribution at the RB reactor by activation foils. Due to rather low neutron flux Cu and Lu foil were irradiated for 4 days. Special natural uranium fuel element was prepared to enable easy removal of foils after irradiation. Experimental device was placed in the reactor core at half height in order to measure directly the mean neutron density. Experimental data of neutron temperature distribution for square lattice pitch 16 cm are presented with mean values of neutron temperature in the moderator, in the fuel and on the fuel element surface

  13. Theory of neutron slowing down in nuclear reactors

    CERN Document Server

    Ferziger, Joel H; Dunworth, J V

    2013-01-01

    The Theory of Neutron Slowing Down in Nuclear Reactors focuses on one facet of nuclear reactor design: the slowing down (or moderation) of neutrons from the high energies with which they are born in fission to the energies at which they are ultimately absorbed. In conjunction with the study of neutron moderation, calculations of reactor criticality are presented. A mathematical description of the slowing-down process is given, with particular emphasis on the problems encountered in the design of thermal reactors. This volume is comprised of four chapters and begins by considering the problems

  14. Experimental determination of the neutron source for the Argonauta reactor subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Renke, Carlos A.C.; Furieri, Rosanne C.A.A.; Pereira, Joao C.S.; Voi, Dante L.; Barbosa, Andre L.N., E-mail: renke@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplier medium requires a well defined neutron source to carry out the experiments necessary for the acquisition of the desired data. The Argonauta research reactor installed at the Instituto de Engenharia Nuclear has a subcritical assembly, under development, to be coupled at the upper part of the reactor core that will provide the needed neutrons emerging from its internal thermal column made of graphite. In order to perform neutronic calculations to compare with the experimental results, it is necessary a precise knowledge of the emergent neutron flux that will be used as neutron source in the subcritical assembly. In this work, we present the thermal neutron flux profile determined experimentally via the technique of neutron activation analysis, using dysprosium wires uniformly distributed at the top of the internal thermal neutron column of the Argonauta reactor and later submitted to a detection system using Geiger-Mueller detector. These experimental data were then compared with those obtained through neutronic calculation using HAMMER and CITATION codes in order to validate this calculation system and to define a correct neutron source distribution to be used in the subcritical assembly. This procedure avoids a coupled neutronic calculation of the subcritical assembly and the reactor core. It has also been determined the dimension of the graphite pedestal to be used in the bottom of the subcritical assembly tank in order to smooth the emergent neutron flux at the reactor top. Finally, it is estimated the thermal neutron flux inside the assembly tank when filled with water. (author)

  15. 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

  16. Compact tokamak reactors. Part 1 (analytic results)

    International Nuclear Information System (INIS)

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1996-01-01

    We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model

  17. 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)

  18. AUS - the Australian modular scheme for reactor neutronics computations

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1975-12-01

    A general description is given of the AUS modular scheme for reactor neutronics calculations. The scheme currently includes modules which provide the capacity for lattice calculations, 1D transport calculations, 1 and 2D diffusion calculations (with feedback-free kinetics), and burnup calculations. Details are provided of all system aspects of AUS, but individual modules are only outlined. A complete specification is given of that part of user input which controls the calculation sequence. The report also provides sufficient details of the supervisor program and of the interface data sets to enable additional modules to be incorporated in the scheme. (author)

  19. Reactor neutron activation for multielemental analysis

    International Nuclear Information System (INIS)

    Reddy, A.V.R.

    1999-01-01

    Neutron Activation Analysis using single comparator (K 0 NAA method) has been used for obtaining multielemental profiles in a variety of matrices related to environment. Gold was used as the comparator. Neutron flux was characterised by determining f, the epithermal to thermal neutron flux ratio and cc, the deviation from ideal shape of the neutron spectrum. The f and a were determined in different irradiation positions in APSARA reactor, PCF position in CIRUS reactor and tray rod position in Dhruva reactor using both cadmium cut off and multi isotope detector methods. High resolution gamma ray spectrometry was used for radioactive assay of the activation products. This technique is being used for multielement analysis in a variety of matrices like lake sediments, sea nodules and crusts, minerals, leaves, cereals, pulses, leaves, water and soil. Elemental profiles of the sediments corresponding to different depths from Nainital lake were determined and used to understand the history of natural absorption/desorption pattern of the previous 160 years. Ferromanganese crusts from different locations of Indian Ocean were analysed with a view to studying the distribution of some trace elements along with Fe and Mn. Variation of Mn/Fe ratio was used to identify the nature of the crusts as hydrogenous or hydrothermal. Fe-rich and Fe-depleted nodules from Indian Ocean were analysed to understand the REE patterns and it is proposed that REE-Th associated minerals could be the potential Th contributors to the sea water and thus reached ferromanganese nodules. Dolomites (unaltered and altered), two types of serpentines and intrusive rock dolerite from the asbestos mines of Cuddapah basin were analysed for major, minor and trace elements. The elemental concentrations are used for distinguishing and characterising these minerals. From our investigations, it was concluded that both dolomite and dolerite contribute elements in the serpentinisation process. Chemical neutron

  20. Conceptual design of neutron diagnostic systems for fusion experimental reactor

    International Nuclear Information System (INIS)

    Iguchi, T.; Kaneko, J.; Nakazawa, M.

    1994-01-01

    Neutron measurement in fusion experimental reactors is very important for burning plasma diagnostics and control, monitoring of irradiation effects on device components, neutron source characterization for in-situ engineering tests, etc. A conceptual design of neutron diagnostic systems for an ITER-like fusion experimental reactor has been made, which consists of a neutron yield monitor, a neutron emission profile monitor and a 14-MeV spectrometer. Each of them is based on a unique idea to meet the required performances for full power conditions assumed at ITER operation. Micro-fission chambers of 235 U (and 238 U) placed at several poloidal angles near the first wall are adopted as a promising neutron yield monitor. A collimated long counter system using a 235 U fission chamber and graphite neutron moderators is also proposed to improve the calibration accuracy of absolute neutron yield determination

  1. Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

  2. Automation of reactor neutron activation analysis

    International Nuclear Information System (INIS)

    Pavlov, S.S.; Dmitriev, A.Yu.; Frontasyeva, M.V.

    2013-01-01

    The present status of the development of a software package designed for automation of NAA at the IBR-2 reactor of FLNP, JINR, Dubna, is reported. Following decisions adopted at the CRP Meeting in Delft, August 27-31, 2012, the missing tool - a sample changer - will be installed for NAA in compliance with the peculiar features of the radioanalytical laboratory REGATA at the IBR-2 reactor. The details of the design are presented. The software for operation with the sample changer consists of two parts. The first part is a user interface and the second one is a program to control the sample changer. The second part will be developed after installing the tool.

  3. 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

  4. Use of self powered neutron detectors in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Galo Rocha, F. del.

    1989-01-01

    A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

  5. 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)

  6. Neutronics characteristics of space power reactors

    International Nuclear Information System (INIS)

    Little, W.; Barner, J.

    1986-01-01

    The objective of the paper is to describe the neutronic characteristics of a range of possible space reactor designs, and indicate the relative advantages and disadvantages of the various designs. Fuel designs to be considered are cermets (i.e., ceramic particles embedded in a metal matrix) consisting of UO 2 or Nn ceramic particles in matrices of Nb, Mo, Ta, or W. These cermet fuels are compared to a UN pin-type design. UN was selected for the reference fuel material since it has a somewhat higher density than UO 2 (i.e., 14.32 versus 10.96 gm/cc), which allows a lower minimum critical mass for both ceramic and cermet designs

  7. Reprocessing of fast neutron reactor fuel

    International Nuclear Information System (INIS)

    Bourgeois, M.

    1981-05-01

    A PUREX process specially adapted to fast neutron reactor fuels is employed. The results obtained indicate that the aqueous process can be applied to this type of fuel: almost 10 years operation at the AT 1 plant which processes fuel from RAPSODIE; the good results obtained at the MARCOULE pilot plant on large batches of reference fuels. The CEA is continuing its work to transfer this technology onto an industrial scale. Industrial prototypes and the launching of the TOR (traitement d'oxydes rapides) project will facilitate this transfer. In 1984, it is expected that fast fuels will be able to be processed on a significant scale and that supplementary R and D facilities will be available [fr

  8. 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

  9. Nuclear reactor, fuel assembly and neutron measuring system

    International Nuclear Information System (INIS)

    Chaki, Masao; Murase, Michio; Zukeran, Atsushi; Moriya, Kimiaki

    1998-01-01

    The present invention provides a BWR type reactor improved with the efficiency of used fuels and fuel economy by increasing a rated power and reducing exchange fuels. Namely, in a BWR type reactor at present, a thermal limit value is determined by conducting nuclear calculation of the reactor core based on data of reactor flow rate measurement and data of neutron flux measurement. However, since the neutron calculation of the reactor core is based on fuel assemblies while the points for the neutron measurement are present at the outside of the fuel assemblies, errors are caused. A margin including the errors has been used as a thermal limit value during operation. In the present invention, neutron fluxes in the fuel assembly as a base of the nuclear calculation can be measured by the same number of neutron detector tubes, but the number of the measuring points is increased to four times. With such procedures, errors caused by the difference of the neutron calculation and values at neutron measuring points can be reduced. As a result, a margin of the thermal limit value is reduced to increase the degree of freedom of reactor operation. Then, the economical property of the reactor operation can be improved. (N.H.)

  10. 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)

  11. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.

    1983-04-01

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  12. Experimental determination of neutron lifetimes through macroscopic neutron noise in the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitaria, Sao Paulo (Brazil)

    2013-05-06

    The neutron lifetimes of the core, reflector, and global were experimentally obtained through macroscopic neutron noise in the IPEN/MB-01 reactor for five levels of subcriticality. The theoretical Auto Power Spectral Densities were derived by point kinetic equations taking the reflector effect into account, and one of the approaches consider an additional group of delayed neutrons.

  13. Intermediate and fast neutron absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1987-10-01

    The experimental fuel channel EFC is created as one of the fast neutron fields at the RB reactor. The intermediate and fast neutron spectra in EFC are measured by activation technique. The intermediate and fast neutron absorbed doses are computed on the basis of these experimental results. At the end the obtained doses are compared. (author)

  14. Limitations for qualitative and quantitative neutron activation analysis using reactor neutrons

    International Nuclear Information System (INIS)

    El-Abbady, W.H.; El-Tanahy, Z.H.; El-Hagg, A.A.; Hassan, A.M.

    1999-01-01

    In this work, the most important limitations for qualitative and quantitative analysis using reactor neutrons for activation are reviewed. Each limitation is discussed using different examples of activated samples. Photopeak estimation, nuclear reactions interference and neutron flux measurements are taken into consideration. Solutions for high accuracy evaluation in neutron activation analysis applications are given. (author)

  15. Absorbing rods for nuclear fast neutron reactor absorbing assembly

    International Nuclear Information System (INIS)

    Aji, M.; Ballagny, A.; Haze, R.

    1986-01-01

    The invention proposes a neutron absorber rod for neutron absorber assembly of a fast neutron reactor. The assembly comprises a bundle of vertical rods, each one comprising a stack of pellets made of a neutron absorber material contained in a long metallic casing with a certain radial play with regard to this casing; this casing includes traps for splinters from the pellets which may appear during reactor operation, at the level of contact between adjacent pellets. The present invention prevents the casing from rupture involved by the disintegration of the pellets producing pieces of boron carbide of high hardness [fr

  16. Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor

    International Nuclear Information System (INIS)

    Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung

    2012-01-01

    10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example

  17. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    Tavera D, L.; Camacho L, M.E.

    1991-01-01

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  18. Application of neutron noise analysis to a swimming pool research reactor

    International Nuclear Information System (INIS)

    Behringer, K.; Lescano, V.H.; Meier, F.; Phildius, J.; Winkler, H.

    1982-01-01

    This work is part of a programme of establishing practical applications of neutron noise techniques to a swimming pool research reactor and deals with two different items: (1) The identification of local boiling caused e.g. by a partial blockage of the coolant flow in a fuel element. Local boiling can easily lead to a burn-out situation. The onset of boiling can be detected by neutron noise analysis and a boiling detection system is presently under development. (2) The measurement of the time evolution of the reactivity induced by xenon after reactor shut-down by an on-line reactivity meter based on neutron noise analysis. From the data, the prompt neutron decay constant at delayed critical, the equilibrium xenon reactivity worth, and an estimate of the average steady-state power flux in the core before reactor shut-down were obtained. (author)

  19. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

    A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

  20. Utilization of RP-10 reactor for neutron therapy

    International Nuclear Information System (INIS)

    Paucar, R.; Nieto, M.; Parreno, F.; Vela, M.; Pozo, Z.

    1997-01-01

    In the Nuclear Energy Peruvian Institute, IPEN, a research area has established of Neutron Radiotherapy, know as NCT. This research joins the physics of particles (Neutrons and photons) and Medical Physics, and this one is an applied investigation where in considering the construction of a treatment hall in Huarangal (Peru) Reactor's irradiation facility, it can treat patients with brain tumors. In Neutron Therapy (NCT), it tries to use neutrons to destroy tumor cells where other therapeutic techniques are not effective. This process consist on to incise a neutrons beam of adequate characteristics over the tumor area of the patient. The neutrons used are of thermal energy and therefore irradiations are developed in experimental reactors. For this one, it is used horizontal channels prepared suitably. Before the irradiation, it is injected to the patient a substance which is absorbed by tumoral tissue. The substance components will be B-10, nuclide with an absorption cross section high to thermal neutrons (3837 b). The B-10 irradiate with thermal neutrons produce alpha particles of short reach (10 μm. on soft tissue) and with LET values (lineal energy transference) very high. The result is a cell preferential destruction which have absorbed the substance and it's next neighbors, like the cell size is 10 μm. This process as know as Boron Neutron Capture Therapy (BNCT). This work describes Peruvian RP-10 reactor and recently efforts to assess the design and feasibility of the medical neutron irradiation facility for NCT. (author). 22 refs., 6 tabs

  1. Neutron spectrometry for reactor applications: status, limitations, and future directions

    International Nuclear Information System (INIS)

    Gold, R.

    1975-08-01

    The ability of ''state-of-the-art'' reactor neutron spectrometry to provide definitive environmental results required for high fluence radiation damage experiments is reviewed. A formal definition of the neutron component is presented as well as general considerations which accrue from both this definition and the existence of the mixed radiation field generally encountered in reactors. A description of four selected methods of reactor neutron spectrometry is included, namely Proton Recoil (PR) methods, Time-Of-Flight (TOF) methods, the 6 Li(n,α) 3 H coincidence method, and Multiple Foil Activation (MFA) methods. These selected methods are compared. Future requirements and directions for reactor neutron spectrometry are discussed. In particular, the needs of future CTR research are stressed and the He 4 - recoil proportional counter spectroscopy method is advanced as a means of meeting these future requirements. 50 references. (auth)

  2. Neutron diagnostic investigations with a research reactor

    International Nuclear Information System (INIS)

    Harms, A.A.

    1977-01-01

    Some aspects of the use of neutron transmission analysis in applied research, as pursued at McMaster University (Canada), are examined. Examples considered are void measurements in two-phase flow, neutron conversion enhancement in neutron radiography, reconstruction of interior bulk heterogenities in solids and temperature sensing with neutrons. (author)

  3. Personal neutron dosimetry at a research reactor facility

    International Nuclear Information System (INIS)

    Kamenopoulou, V.; Carinou, E.; Stamatelatos, I.E.

    2001-01-01

    Individual neutron monitoring presents several difficulties due to the differences in energy response of the dosemeters. In the present study, an individual dosemeter (TLD) calibration approach is attempted for the personnel of a research reactor facility. The neutron energy response function of the dosemeter was derived using the MCNP code. The results were verified by measurements to three different neutron spectra and were found to be in good agreement. Three different calibration curves were defined for thermal, intermediate and fast neutrons. At the different working positions around the reactor, neutron spectra were defined using the Monte Carlo technique and ambient dose rate measurements were performed. An estimation of the neutrons energy is provided by the ratio of the different TLD pellets of each dosemeter in combination with the information concerning the worker's position; then the dose equivalent is deduced according to the appropriate calibration curve. (author)

  4. 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)

  5. 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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  7. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L; Camacho L, M E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  8. IBM-PC-based reactor neutronics analysis package

    International Nuclear Information System (INIS)

    Nigg, D.W.; Wessol, D.E.; Grimesey, R.A.; Parsons, D.K.; Wheeler, F.J.; Yoon, W.Y.; Lake, J.A.

    1985-01-01

    Technical advances over the past few years have led to a situation where a wide range of complex scientific computations can now be done on properly configured microcomputers such as the IBM-PC (personal computer). For a number of reasons, including security, economy, and user convenience, the development of a comprehensive system of reactor neutronics codes suitable for operation on the IBM-PC has been undertaken at the Idaho National Engineering Laboratory (INEL). It is anticipated that a PC-based code system could also have wide applicability in the nuclear engineering education community since conversion of software generated by national laboratories and others to college and university mainframe hardware has historically been a time-consuming process that has sometimes met with only limited success. This paper discusses the philosophy behind the INEL reactor neutronics PC code system and describes those parts of the system that are currently complete, those that are now under development, and those that are still in the planning stage

  9. Neutron and thermal dynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    van Dam, H.; Kuijper, J.C.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

    1989-01-01

    In this paper neutron kinetics and thermal dynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focused on the properties of the fuel gas, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

  10. Solving the uncommon nuclear reactor core neutronics problems

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1983-01-01

    Calculational procedures have been implemented for solving importance and higher harmonic neutronics problems. Solutions are obtained routinely to support analysis of reactor core performance, treating up to three space coordinates with the multigroup diffusion theory approximation to neutron transport. The techniques used and some of the calculational difficulties are discussed

  11. Determination of reactor fuel burnup using passive neutron assay

    International Nuclear Information System (INIS)

    Kodeli, I.; Trkov, A.; Najzer, M.; Ertek, C.

    1988-01-01

    Passive neutron assay (PNA) method was developed to verify the fissile inventory of the irradiated reactor fuels. The characteristics of the method were studied at 'Jozef Stefan' Institute. The dependence of neutron source in the fuel on burnup, cooling time, initial enrichment and specific power were investigated and the accuracy of the method, using available computer codes was estimated. (author)

  12. Thermogravimetric analysis of reactor-neutrons-irradiated LEXAN polycarbonate film

    International Nuclear Information System (INIS)

    Kalsi, P.C.

    2000-01-01

    The effects of reactor-neutrons irradiation on the thermogravimetric (TG) analysis of LEXAN polycarbonate film in air were studied. Irradiation enhances the degradation rate and the effect increases further with increasing neutron fluence. The kinetics of the different steps of degradation were also evaluated from the TG curves. The activation energy values calculated for all the degradation stages decrease on irradiation. (author)

  13. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Alvarez, I.; Herrera, E.; Lima, L.; Tores, J.; Lopez, M.C.; Ixquiac, M.

    1996-01-01

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K o neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott's formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented

  14. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Rizo, O; Alvarez, I; Herrera, E; Lima, L; Tores, J [Secretaria Ejecutiva para Asuntos Nucleares, Holguin (Cuba). Delegacion Territorial; Manso, M V [Centro de Isotopos, La Habana (Cuba); Lopez, M C [Instituto Nacional de Investigaciones Nucleares, Mexico City (Mexico); Ixquiac, M [Universidad de San Carlos de Guatemala, Guatemala City (Guatemala)

    1997-12-31

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K{sub o} neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott`s formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented.

  15. 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

  16. 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

  17. The neutron balance of the natural reactors at Oklo

    International Nuclear Information System (INIS)

    Naudet, R.; Filip, A.

    1975-01-01

    The authors discuss the main parameters determining criticality: the concentration of fissionable nuclei in the uranium; concentration of neutron-capturing nuclei in the gangue; concentration of uranium in the ore and rearrangement of the uranium to form ''critical masses''; amount of water present. Moderation was caused partly by the water of constitution of the clays in the gangue. Examination of the available data indicates that criticality could quite well have been achieved. A computer code (BINOCLE) was written for handling the neutron physics problems raised by the natural reactors. This very simple code, which can nevertheless handle the important points in sufficient detail, is well suited for describing the ores, providing a clear breakdown of the neutron balance and the quantities necessary for interpreting the analyses. It is designed to serve as a subprogram to a series of other codes: one-dimensional criticality; point evolution; spatial evolution; consideration of thermal transfers. Results showing the role of the main parameters are presented. The physical quantities measured by fission-product analysis are also found: proportions of fast fissions; conversion coefficient; spectral indices

  18. Neutronics analysis of Nigerian Research Reactor-1

    International Nuclear Information System (INIS)

    Azande, T.S.; Balogun, G.I.

    2010-01-01

    Feasibility studies for the conversion of the Nigerian Research Reactor-1 (NIRR-1) have been performed using WIMS and CITATION codes (Azande et al, 2009 and Balogun, 2003) at the Centre for Energy Research and Training (CERT), Ahmadu Bello University, Zaria Kaduna State. In this work, the neutronics analysis of NIRR-1 core concerning mass loading of U-235 in the core, shut down margin (SDM), safety reactivity factor (SRF), control rod worth, and control rod critical depth of insertion were investigated at low enrichment. Two fuel types (UAl 4 and UO 2 ) were considered and the uranium densities required for the conversion of NIRR-1 core to low enrichment were computed to be 1201g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1274 g/cc with 15% enrichment, 1448 g/cc with 10% enrichment for UAl 4 fuel type and 1141g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1216 g/cc with 15% enrichment, and 1389 g/cc with 10% enrichment for UO 2 fuel type. Signi ficantly, higher uranium densities are required to convert NIRR-1 from HEU to LEU - indicating a drastic review of the NIRR-1 core.

  19. 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

  20. Neutronics issues in fusion-fission hybrid reactor design

    International Nuclear Information System (INIS)

    Liu Chengan

    1995-01-01

    The coupled neutron and γ-ray transport equations and nuclear number density equations, and its computer program systems concerned in fusion-fission hybrid reactor design are briefly described. The current status and focal point for coming work of nuclear data used in fusion reactor design are explained

  1. Criteria design of the CAREM 25 reactor's core: neutronic aspects

    International Nuclear Information System (INIS)

    Lecot, C.A.

    1990-01-01

    The criteria that guided the design, from the neutronic point of view, of the CAREM reactor's core were presented. The minimum set of objectives and general criteria which permitted the design of the particular systems constituting the CAREM 25 reactor's core is detailed and stated. (Author) [es

  2. 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

  3. Graphs of neutron cross section data for fusion reactor development

    International Nuclear Information System (INIS)

    Asami, Tetsuo; Tanaka, Shigeya

    1979-03-01

    Graphs of neutron cross section data relevant to fusion reactor development are presented. Nuclides and reaction types in the present compilation are based on a WRENDA request list from Japan for fusion reactor development. The compilation contains various partial cross sections for 55 nuclides from 6 Li to 237 Np in the energy range up to 20 MeV. (author)

  4. Equipment for thermal neutron flux measurements in reactor R2

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, E; Nilsson, T; Claeson, S

    1960-04-15

    For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

  5. An approach to neutronics analysis of candu reactors

    International Nuclear Information System (INIS)

    Gul, S.; Arshad, M.

    1982-12-01

    An attempt is made to tackle the problem of neutronics analysis of CANDU reactors. Until now CANDU reactors have been analysed by the methods developed at AECL and CGE using mainly receipe methods. Relying on multigroup transport codes GAM-GATHER in combination with diffusion code CITATION a package of codes is established to use it for survey as well as production purposes. (authors)

  6. Neutronic analysis of the ford nuclear reactor leu core

    International Nuclear Information System (INIS)

    Raza, S.S.; Hayat, T.

    1989-08-01

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

  7. A new neutron noise technique for fast reactors

    International Nuclear Information System (INIS)

    Zhuo Fengguan; Jin Manyi; Yao Shigui; Su Zhuting

    1987-12-01

    This paper gives a new neutron noise technique for fast reactors, which is known as thermalization measurement technique of the neutron noise. The theoretical formulas of the technique were developed, and a digital delayed coincidence time analyzer consisted of TTL integrated circuits was constructed for the study of this technique. The technique has been tested and applied practically at Df-VI fast zero power reactor. It was shown that the provided technique in this work has a number of significant advantages in comparison with the conventional neutron noise method

  8. Modeling delayed neutron monitoring systems for fast breeder reactors

    International Nuclear Information System (INIS)

    Bunch, W.L.; Tang, E.L.

    1983-10-01

    The purpose of the present work was to develop a general expression relating the count rate of a delayed neutron monitoring system to the introduction rate of fission fragments into the sodium coolant of a fast breeder reactor. Most fast breeder reactors include a system for detecting the presence of breached fuel that permits contact between the sodium coolant and the mixed oxide fuel. These systems monitor for the presence of fission fragments in the sodium that emit delayed neutrons. For operational reasons, the goal is to relate the count rate of the delayed neutron monitor to the condition of the breach in order that appropriate action might be taken

  9. Neutron activation of building materials used in the reactor shield

    International Nuclear Information System (INIS)

    Hernandez, A.T.; Perez, G.; D'Alessandro, K.

    1993-01-01

    Cuban concretes and their main components (mineral aggregates and cement) were investigated through long-lived activation products induced by neutrons from a reactor. The multielemental content in the materials studied was obtained by neutron activation analysis in an IBR-2 reactor and gamma activation analysis in an MT-25 microtron from Join Institute of Nuclear Research of Dubna. After irradiation of building materials for 30 years by a neutron flow of unitary density, induced radioactivity was calculated according to experimental data. The comparative evaluation of different concretes aggregates and two types of cement related to the activation properties is discussed

  10. Kartini Research Reactor prospective studies for neutron scattering application

    International Nuclear Information System (INIS)

    Widarto

    1999-01-01

    The Kartini Research Reactor (KRR) is located in Yogyakarta Nuclear Research Center, Yogyakarta - Indonesia. The reactor is operated for 100 kW thermal power used for research, experiments and training of nuclear technology. There are 4 beam ports and 1 column thermal are available at the reactor. Those beam ports have thermal neutron flux around 10 7 n/cm 2 s each other and used for sub critical assembly, neutron radiography studies and Neutron Activation Analysis (NAA). Design of neutron collimator has been done for piercing radial beam port and the calculation result of collimated neutron flux is around 10 9 n/cm 2 s. This paper describes experiment facilities and parameters of the Kartini research reactor, and further more the prospective studies for neutron scattering application. The purpose of this paper is to optimize in utilization of the beam ports facilities and enhance the manpower specialty. The special characteristic of the beam ports and preliminary studies, pre activities regarding with neutron scattering studies for KKR is presented. (author)

  11. Device for detecting neutron flux in nuclear reactor. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Bessho, Y; Nishizawa, Y

    1976-07-30

    The object of the invention is to ensure accuracy in the operation of the nuclear reactor by reducing the difference that results between the readings of a Traversing Incore Probe (TIP) and a Local Power Range Monitor (LPRM) when the neutron flux distribution undergoes a change. In an apparatus for detecting neutrons in a nuclear reactor, an LPRM sensor comprising a layer containing a substance capable of nuclear fission, a section filled with argon gas and a collector is constructed so as to surround a TIP within a TIP guide tube at the height of the reactor axis. In this way, the LPRM detects the average value of neutron distribution in the region surrounding the TIP, so that no great difference between the readings of both the sensors is produced even if the neutron flux distribution is changed.

  12. Optimization of neutron flux distribution in Isotope Production Reactor

    International Nuclear Information System (INIS)

    Valladares, G.L.

    1988-01-01

    In order to optimize the thermal neutrons flux distribution in a Radioisotope Production and Research Reactor, the influence of two reactor parameters was studied, namely the Vmod / Vcomb ratio and the core volume. The reactor core is built with uranium oxide pellets (UO 2 ) mounted in rod clusters, with an enrichment level of ∼3 %, similar to LIGHT WATER POWER REATOR (LWR) fuel elements. (author) [pt

  13. Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5

    International Nuclear Information System (INIS)

    Pena V, J.D.

    2016-01-01

    The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

  14. General considerations for neutron capture therapy at a reactor facility

    International Nuclear Information System (INIS)

    Binney, S.E.

    2001-01-01

    In addition to neutron beam intensity and quality, there are also a number of other significant criteria related to a nuclear reactor that contribute to a successful neutron capture therapy (NCT) facility. These criteria are classified into four main categories: Nuclear design factors, facility management and operations factors, facility resources, and non-technical factors. Important factors to consider are given for each of these categories. In addition to an adequate neutron beam intensity and quality, key requirements for a successful neutron capture therapy facility include necessary finances to construct or convert a facility for NCT, a capable medical staff to perform the NCT, and the administrative support for the facility. The absence of any one of these four factors seriously jeopardizes the overall probability of success of the facility. Thus nuclear reactor facility management considering becoming involved in neutron capture therapy, should it be proven clinically successful, should take all these factors into consideration. (author)

  15. Neutron irradiation facilities for fission and fusion reactor materials studies

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

    The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

  16. Nuclear reactor ex-core startup neutron detector

    International Nuclear Information System (INIS)

    Wyvill, J.R.

    1980-01-01

    A sensitive ex-core neutron detector is needed to monitor the power level of reactors during startup. The neutron detector of this invention has a photomultiplier with window resistant to radiation darkening at the input end and an electrical connector at the output end. The photomultiplier receives light signals from a neutron-responsive scintillator medium, typically a cerium-doped lithium silicate glass, that responds to neutrons after they have been thermalized by a silicone resin moderator. Enclosing and shielding the photmultiplier, the scintillator medium and the moderator is a combined lead and borated silicone resin housing

  17. Neutron spectrum determination by activation method in fast neutron fields at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1994-01-01

    The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (author)

  18. Neutron spectrum determination by activation method in fast neutron fields at the RB reactors

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.S.; Pesic, M.P.; Antic, D.P.

    1994-01-01

    The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (authors). 7 refs., 3 tabs

  19. Extension of the AUS reactor neutronics system for application to fusion blanket neutronics

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1984-03-01

    The AUS modular code scheme for reactor neutronics computations has been extended to apply to fusion blanket neutronics. A new group cross-section library with 200 neutron groups, 37 photon groups and kerma factor data has been generated from ENDF/B-IV. The library includes neutron resonance subgroup parameters and temperature-dependent data for thermal neutron scattering matrices. The validity of the overall calculation system for fusion applications has been checked by comparison with a number of published conceptual design studies

  20. Neutron noise analysis for malfunction diagnosis at sodium cooled reactors

    International Nuclear Information System (INIS)

    Hoppe, P.

    1978-09-01

    For the investigation of the potential use of neutron noise analysis at sodium cooled power reactors, measurements have been performed at the KNK I reactor over a period of 18 month under different operational conditions. The signal fluctuations of the following tranducers have been recorded: In-core and Ex-core neutron detectors, temperature-, flow-, pressure-, vibration- and acoustic sensors. These extensive measurements have been analyzed in the frequency range from 0,001 Hz to 1000 Hz with all currently known methods for the identification of noise sources. The following results have been found: - Neutron noise for f 20 Hz the white detection noise prevails. In the region from 1 Hz to 20 Hz the vibrations of core components contribute to neutron noise. - Neutron noise is influenced by the state of the plant. - The contributions to neutron noise due to the fluctuations of coolant flow and inlet temperature are small compared to those produced by the movements of the control rod initiated by the reactor control system. The quantitatively unidentifiable amount of reactivity fluctuations (0,6 time-dependent thermal bowing of the core. With respect to these results and by calculation of the neutron noise patterns to be expected for the SNR 300, the following possible applications for neutron noise analysis have been found: By means of neutron noise analysis only reactivity fluctuations can be identified and supervised which are produced by time dependent changes of the core geometry. Furthermore neutron noise analysis is well suited for a sensitive detection of control rod vibrations and of local sodium boiling. Finally it can be used for the surveillance of the proper functioning of the reactor control system and of the control rod drive mechanism. (orig./HP) 891 HP [de

  1. Critical fluctuations of the number of neutrons in a reactor

    International Nuclear Information System (INIS)

    Ryazanov, V.V.; Lakoza, E.L.; Sysoev, V.M.

    1995-01-01

    The nuclear chain reaction is the most important physical process in a reactor. The theory of nuclear chain reaction fluctuations (neutron noise), developed in and other studies, has given results that are important for reactor physics and reactor practice (correlation analysis of neutron noise for measurement of the physical characteristics and reactor monitoring, stability of the critical state, etc.). Here we propose to study these problems by applying the methods of continuous phase transitions and synergetics and using the analogy with chemical chain reactions and the general laws of critical phenomena. The optimal reactor operating conditions are critical. To predict how a critical reactor will behave it is necessary to reveal those features of the neutron laws that are universal in some way, i.e., do not depend on the details of the individual acts of neutron motion and transformation that occur in reactors of different types. The similarity between chemical and nuclear chain reactions was noted long ago. Consequently, a universal theory of continuous phase transition was developed for systems of diverse physical nature

  2. A neutron amplifier: prospects for reactor-based waste transmutation

    International Nuclear Information System (INIS)

    Blanovsky, A.

    2004-01-01

    A design concept and characteristics for an epithermal breeder controlled by variable feedback and external neutron source intensity are presented. By replacing the control rods with neutron sources, we could maintain good power distribution and perform radioactive waste burning in high flux subcritical reactors (HFSR) that have primary system size, power density and cost comparable to a pressurized water reactor (PWR). Another approach for actinide transmutation is a molten salt subcritical reactor proposed by Russian scientists. To increase neutron source intensity the HFSR is divided into two zones: a booster and a blanket with solid and liquid fuels. A neutron gate (absorber and moderator) imposed between two zones permits fast neutrons from the booster to flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed in the absorber zone. In the HFSR, neptunium-plutonium fuel is circulated in the booster and blanket, and americium-curium in the absorber zone and outer reflector. Use of a liquid actinide fuel permits transport of the delayed-neutron emitters from the blanket to the booster, where they can provide additional neutrons (source-dominated mode) or all the necessary excitation without an external neutron source (self-amplifying mode). With a blanket neutron multiplication gain of 20 and a booster gain of 50, an external neutron source rate of at least 10 15 n/s (0.7 MW D-T or 2.5 MW electron beam power) is needed to control the HFSR that produces 300 MWt. Most of the power could be generated in the blanket that burns about 100 kg of actinides a year. The analysis takes into consideration a wide range of HFSR design aspects including the wave model of observed relativistic phenomena, plant seismic diagnostics, fission electric cells (FEC) with a multistage collector (anode) and layered cathode. (author)

  3. Determining of the intermediate neutron spectrum in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1987-01-01

    The activation method for intermediate neutron spectrum determination is given in this paper. The intermediate neutron spectrum in experimental fuel channel (EFC) at the RB reactor is determined om the basis of this method. The results of measurements are treated with PRAG code and will be treated with KRIFIT and TENET codes that are also developed. (author)

  4. Canadian Neutron Source (CNS): a research reactor solution for medical isotopes and neutrons for science

    International Nuclear Information System (INIS)

    Chapman, D.

    2009-01-01

    This presentation describes a dual purpose research facility at the University of Saskatchewan for Canada for the production of medical isotopes and neutrons for scientific research. The proposed research reactor is intended to supply most of Canada's medical isotope requirements and provide a neutron source for Canada's research community. Scientific research would include materials research, biomedical research and imaging.

  5. An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons

    International Nuclear Information System (INIS)

    Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.

    2013-01-01

    Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons

  6. Measurements at the RA Reactor related to the VISA-2 project - Part 2, Measurement of neutron flux in VISA-2 channels; Fizicka merenja na reaktoru RA u vezi projekta VISA-2 - II deo, Merenje fluksa neutrona u kanalima VISA - 2

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    This report describes the task concerned with measurements of neutron flux in four experimental channels, called VISA-2 channels. All the channels are made of aluminium tubes, one is sealed to prevent contact of foils with heavy water, and are placed in the regular reactor lattice next to the central experimental channel VK-5. Measuring results, i.e. absolute values of neutron flux and flux distribution are needed for realisation of the VISA-2 project. Measurements of neutron flux are done by activation method. Activation foils are placed in cylindrical aluminium tubes specially prepared for this purpose and placed in VISA-2 channels. Foils are irradiated simultaneously for 5 hours at reactor power of 150 kW. Neutron flux distribution is determined by measuring the relative activity of cobalt foils. [Serbo-Croat] Ovaj izvestaj opisuje merenja neutronskog fluksa u cetiri eksperimentalna kanala VISA-2. Svi VISA-2 kanali nacinjeni su od aluminijuma, jedan je hermeticki zatvoren da folije koje se ozacuju ne dodju ukontakt sa teskom vodom i svi su smesteni neposredno pored centralnog eksperimentalnog kanala VK-5. Rezultati merenja, odnosno apsolutne vrednosti neutronskog fluksa i reposdele fluksa potrebni su za realizaciju projekta VISA-2. Mrerenja neutronskog fluksa izvreseno je aktivacionom tehnikom. Aktivacione folije smestene su u prethodno napravljene aluminijumske cevcice. Folije su ozracivane istovremeno pri snazi od 150 kW pe casova. Raspodela neutronskog fluksa odredjena je merenjem relativne aktivnosti folija od kobalta.

  7. 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

  8. Neutron radiography applications in I.T.U. TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Tugrul, A. B.

    2002-01-01

    Neutron radiography is an important radiographic technique which is supplied different and advanced information according to the X or gamma ray radiography. However, it has a trouble for supplying the convenient neutron sources. Tangential beam tube of Istanbul Technical University (ITU) TRIGA Mark-II Training and Research Reactor has been arranged for using neutron radiography. The neutron radiography set defined as detailed for the application of the technique. Two different techniques for neutron radiography are defined as namely, transfer method and direct method. For the transfer method dysprosium and indium screens are used in the study. But, dysprosium generally was preferred in many studies in the point of view nuclear safety. Gadolinium was used for direct method. Two techniques are compared and explained the preferring of the transfer technique. Firstly, reference composition is prepared for seeing the differences between neutron and X-ray or gamma radiography. In addition of it, some radiograph samples are given neutron and X-ray radiography which shows the different image characters. Lastly, some examples are given from archaeometric studies. One of them the brass plates of Great Mosque door in Cizre. After the neutron radiography application, organic dye traces are noticed. Other study is on a sword that belong to Urartu period at the first millennium B.C. It is seen that some wooden part on it. Some different artefacts are examined with neutron radiography from the Ikiztepe excavation site, then some animal post parts are recognized on them. One of them is sword and sheath which are corroded together. After the neutron radiography application, it can be noticed that there are a cloth between the sword and its sheath. By using neutron radiography, many interesting and detailed results are observed in ITU TRIGA Mark-II Training and Research Reactor. Some of them shouldn't be recognised by using any other technique

  9. D-3He fuel cycles for neutron lean reactors

    International Nuclear Information System (INIS)

    Kernbichler, W.; Miley, G.H.; Heindler, M.

    1989-01-01

    The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

  10. Physical particularities of nuclear reactors using heavy moderators of neutrons

    International Nuclear Information System (INIS)

    Kulikov, G. G.; Shmelev, A. N.

    2016-01-01

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using "2"3"3U as a fissile nuclide and "2"3"2Th and "2"3"1Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  11. Physical particularities of nuclear reactors using heavy moderators of neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

    2016-12-15

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  12. Absolute measurement of neutron fluxes inside the reactor core

    International Nuclear Information System (INIS)

    Ajdacic, S. V.

    1964-10-01

    The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li 6 -semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li 6 -semiconductor spectrometer with plane geometry is given. A new type of Li 6 -semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li 6 -spectrometer made (author)

  13. Absolute measurement of neutron fluxes inside the reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Ajdacic, S V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1964-10-15

    The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li{sup 6}-semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li{sup 6}-semiconductor spectrometer with plane geometry is given. A new type of Li{sup 6}-semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li{sup 6}-spectrometer made (author)

  14. Fractional neutron point kinetics equations for nuclear reactor dynamics

    International Nuclear Information System (INIS)

    Espinosa-Paredes, Gilberto; Polo-Labarrios, Marco-A.; Espinosa-Martinez, Erick-G.; Valle-Gallegos, Edmundo del

    2011-01-01

    The fractional point-neutron kinetics model for the dynamic behavior in a nuclear reactor is derived and analyzed in this paper. The fractional model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutron flux contains a fractional order, acting as exponent of the relaxation time, to obtain the best representation of a nuclear reactor dynamics. The physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. The numerical approximation to the solution of the fractional neutron point kinetics model, which can be represented as a multi-term high-order linear fractional differential equation, is calculated by reducing the problem to a system of ordinary and fractional differential equations. The numerical stability of the fractional scheme is investigated in this work. Results for neutron dynamic behavior for both positive and negative reactivity and for different values of fractional order are shown and compared with the classic neutron point kinetic equations. Additionally, a related review with the neutron point kinetics equations is presented, which encompasses papers written in English about this research topic (as well as some books and technical reports) published since 1940 up to 2010.

  15. Prompt neutron decay constant for the Oak Ridge Research Reactor with 20 wt % 235U enriched fuel

    International Nuclear Information System (INIS)

    Ragan, G.E.; Mihalczo, J.T.

    1986-01-01

    This paper describes measurements of the prompt neutron decay constant at delayed criticality for the Oak Ridge Research Reactor (ORR) using 20 wt % 235 U enriched fuel and compares these measurements with similar measurements using 93.2 wt % 235 U enriched fuel. This reactor parameter is of interest because it affects the transient behavior of the reactor in prompt criticality accident situations. This experiment is part of a program to investigate the differences in the performance of research reactors fueled with highly enriched and low enriched uranium. The prompt neutron decay constants were obtained using noise analysis measurement techniques for a core with newly fabricated, unirradiated fuel elements

  16. The measurement of neutron and neutron induced photon spectra in fusion reactor related assemblies

    CERN Document Server

    Unholzer, S; Klein, H; Seidel, K

    2002-01-01

    The spectral neutron and photon fluence (or flux) measured outside and inside of assemblies related to fusion reactor constructions are basic quantities of fusion neutronics. The comparison of measured spectra with the results of MCNP neutron and photon transport calculations allows a crucial test of evaluated nuclear data as generally used in fusion applications to be carried out. The experiments concern mixed neutron/photon fields with about the same intensity of the two components. An NE-213 scintillation spectrometer, well described by response matrices for both neutrons and photons, is used as proton-recoil and Compton spectrometer. The experiments described here in more detail address the background problematic of two applications, an iron benchmark experiment with an ns-pulsed neutron source and a deep penetration mock-up experiment for the investigation of the ITER in-board shield system. The measured spectral neutron and photon fluences are compared with spectra calculated with the MCNP code on the b...

  17. The reactor and cold neutron research facility at NIST

    Energy Technology Data Exchange (ETDEWEB)

    Prask, H J; Rowe, J M [Reactor Radiation Division, National Institute of Standards and Technology, Gaithersburg, MD (United States)

    1992-07-01

    The NIST Reactor (NBSR) is a 20 MW research reactor located at the Gaithersburg, MD site, and has been in operation since 1969. It services 26 thermal neutron facilities which are used for materials science, chemical analysis, nondestructive evaluation, neutron standards work, and irradiations. In 1987 the Department of Commerce and NIST began development of the CNRF - a $30M National Facility for cold neutron research -which will provide fifteen new experimental stations with capabilities currently unavailable in this country. As of May 1992, four of the planned seven guides and a cold port were installed, eight cold neutron experimental stations were operational, and the Call for Proposals for the second cycle of formally-reviewed guest-researcher experiments had been sent out. Some details of the performance of instrumentation are described, along with the proposed design of the new hydrogen cold source which will replace the present D{sub 2}O/H{sub 2}O ice cold source. (author)

  18. Research reactor of the future: The advanced neutron source

    International Nuclear Information System (INIS)

    Appleton, B.; West, C.

    1994-01-01

    Agents for cancer detection and treatment, stronger materials, better electronic gadgets, and other consumer and industrial products - these are assured benefits of a research reactor project proposed for Oak Ridge. Just as American companies have again assumed world leadership in producing semiconductor chips as well as cars and trucks, the United States is poised to retake the lead in neutron science by building and operating the $2.9 billion Advanced Neutron Source (ANS) research reactor by the start of the next century. In 1985, the neutron community, led by ORNL researchers, proposed a pioneering project, later called the ANS. Scheduled to begin operation in 2003, the ANS is seen not only as a replacement for the aging HFIR and HFBR but also as the best laboratory in the world for conducting neutron-based research

  19. The reactor and cold neutron research facility at NIST

    International Nuclear Information System (INIS)

    Prask, H.J.; Rowe, J.M.

    1992-01-01

    The NIST Reactor (NBSR) is a 20 MW research reactor located at the Gaithersburg, MD site, and has been in operation since 1969. It services 26 thermal neutron facilities which are used for materials science, chemical analysis, nondestructive evaluation, neutron standards work, and irradiations. In 1987 the Department of Commerce and NIST began development of the CNRF - a $30M National Facility for cold neutron research -which will provide fifteen new experimental stations with capabilities currently unavailable in this country. As of May 1992, four of the planned seven guides and a cold port were installed, eight cold neutron experimental stations were operational, and the Call for Proposals for the second cycle of formally-reviewed guest-researcher experiments had been sent out. Some details of the performance of instrumentation are described, along with the proposed design of the new hydrogen cold source which will replace the present D 2 O/H 2 O ice cold source. (author)

  20. Status of neutron cross sections for reactor dosimetry

    International Nuclear Information System (INIS)

    Vlasov, M.F.; Fabry, A.; McElroy, W.N.

    1977-03-01

    The status of current international efforts to develop standardized sets of evaluated energy-dependent (differential) neutron cross sections for reactor dosimetry is reviewed. The status and availability of differential data are considered, some recent results of the data testing of the ENDF/B-IV dosimetry file using 252 Cf and 235 U benchmark reference neutron fields are presented, and a brief review is given of the current efforts to characterize and identify dosimetry benchmark radiation fields

  1. The neutron field perturbation effect in the Dalat Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Binh, Nguyen Duc; Tuan, Nguyen Minh; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The perturbation effect of the thermal neutron field of the Dalat reactor is investigated when a fuel element is replaced by a water column or a plexiglass rod. In consequence, it is possible to replace the measurement of the relative distribution of the thermal neutron field on the surface of fuel element by that in the water column or in the plexiglass rod. (author). 5 refs. 4 figs. 4 tabs.

  2. Research reactors for the social safety and prosperous neutron use

    International Nuclear Information System (INIS)

    Ito, Yasuo

    2000-01-01

    The present status of nuclear reactors in Japan and the world was briefly described in this report. Aiming to construct a background of stable future society dependent on nuclear energy, the necessity to establish an organization for research reactors in Japan was pointed out. There are a total of 468 reactors in the world, but only 248 of them are running at present and most of them are superannuated. In Japan, 15 research reactors are running and 8 of them are under collaborative utilization, but not a few of them have various problems. In the education of atomic energy, a reactor is dispensable for understanding its working principle through practice learning. Furthermore, a research reactor has important roles for development of power reactor in addition to various basic studies such as activation analysis, fission track, biological irradiation, neutron scattering, etc. Application of a reactor has been also progressing in industrial and medical fields. However, operation of the reactors has become more and more difficult in Japan because of a large running cost and a lack of residential consensus for nuclear reactor. Here, the author proposed an establishment of organization of research reactor in order to promote utilization of a reactor in the field of education, rearing of professionals and science and engineering. (M.N.)

  3. Standard Practice for Ensuring Test Consistency in Neutron-Induced Displacement Damage of Electronic Parts

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice sets forth requirements to ensure consistency in neutron-induced displacement damage testing of silicon and gallium arsenide electronic piece parts. This requires controls on facility, dosimetry, tester, and communications processes that affect the accuracy and reproducibility of these tests. It provides background information on the technical basis for the requirements and additional recommendations on neutron testing. In addition to neutrons, reactors are used to provide gamma-ray pulses of intensities and durations that are not achievable elsewhere. This practice also provides background information and recommendations on gamma-ray testing of electronics using nuclear reactors. 1.2 Methods are presented for ensuring and validating consistency in neutron displacement damage testing of electronic parts such as integrated circuits, transistors, and diodes. The issues identified and the controls set forth in this practice address the characterization and suitability of the radiation environm...

  4. Background determination for the neutron-neutron scattering experiment at the reactor YAGUAR

    Energy Technology Data Exchange (ETDEWEB)

    Muzichka, A.Yu. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Furman, W.I. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Lychagin, E.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Krylov, A.R. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Nekhaev, G.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Sharapov, E.I. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Shvetsov, V.N. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Strelkov, A.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Levakov, B.G. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Lyzhin, A.E. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Chernukhin, Yu.I. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Kandiev, Ya.Z. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Howell, C.R. [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Mitchell, G.E. [North Carolina State University, Raleigh, NC 27695-8202 (United States); Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Crawford, B.E. [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Stephenson, S.L. [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States)]. E-mail: sstephen@gettysburg.edu; Tornow, W. [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)

    2007-06-01

    The motivation and design is outlined for the experiment to measure the neutron-neutron singlet scattering length directly with thermal neutrons at the pulsed reactor YAGUAR. A statistical accuracy of 3% can be reached, though achieving the goal of an overall accuracy of 3-5% for the nn-scattering length depends on the background level. Possible sources of background are discussed in depth and the results of extensive modeling of the background are presented. Measurements performed at YAGUAR to test these background calculations are described. The experimental results indicate an anticipated background level up to 30% relative to the expected nn effect at the maximal energy burst of the reactor. The conclusion is made that the nn experiment at YAGUAR is feasible to produce the first directly measured value for the neutron-neutron scattering length.

  5. Updated neutron spectrum characterization of SNL baseline reactor environments

    International Nuclear Information System (INIS)

    Griffin, P.J.; Kelly, J.G.; Vehar, D.W.

    1994-04-01

    This document provides SAND-II and MANIPULATE output listings from calculations used to derive the new spectrum-averaged integral parameters that were reported in volume 1. When used in conjunction with volume 1, this document provides an audit trail for the neutron radiation field characterization and supports current quality assurance initiatives. This document provides detailed information on the neutron spectrum characteristics of the primary Sandia National Laboratories' (SNL) reactor environments. The information in this volume is not intended for the casual user of the SNL reactor facilities. This detailed characterization of the neutron and gamma environments at the Sandia Pulsed Reactor (SPR) and the Annular Core Research Reactor (ACRR) is provided to aid the users who wish to convert the information given in the Radiation Metrology Laboratory (RML) dosimetry reports into other (non-silicon) measures of neutron damage. The spectra provided in these appendices can be used as a source term for Monte Carlo radiation transport calculations to study the impact of experimenter's test package on the neutron environment

  6. Polarization of fast neutrons in VVR-M reactor

    International Nuclear Information System (INIS)

    Garusov, E.A.; Lifshits, E.P.; Petrov, Yu.V.

    1987-01-01

    Neutron polarization in the reactor leads to circular polarization of γ quanta emitted both in radiational capture of neutrons and in the transition of nuclei excited as a result of inelastic scattering to the ground state. This may be used to determine the polarization of reactor neutrons. The circular polarization of γ quanta at light-water and graphite targets at the center of the active zone of the VVR-M reactor at the B.P. Konstantinov Leningrad Institute of Nuclear Physics was recently measured. A simplified experimental scheme is shown. Fast neutrons leaving the active zone of the reactor were excited in the inelastic scattering at the target nuclei. The polarization of the γ quanta emitted by nuclei in transitions to the ground state was measured by a polarimeter positioned above the active zone. The reason for the circular polarization of γ quanta may also be nonconservation of P parity on account of weak interaction in the capture of a neutron by hydrogen

  7. DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

    Science.gov (United States)

    Snell, A.H.; Allison, S.K.

    1958-02-11

    This patent relates to water-cooled reactor systems and discloses a means to detect leaks in the jackets of jacketed fuel elements comprising a neutron detector located in the cooling water discharge pipe,the pipe being provided with an enlarged portion for housing the detector so that the latter is completely surrounded by the water in its passage through the pipe, said enlarged portion and detector being shielded from the reactor for the purpose of detecting only those delayed neutrons emitted in the cooling water and due to the latter picking up fission fragments from the defective fuel elements.

  8. Neutron and gamma characterization within the FFTF reactor cavity

    International Nuclear Information System (INIS)

    Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D.; Wood, M.R.

    1980-08-01

    Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables

  9. Neutron analysis of the fuel of high temperature nuclear reactors

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Francois L, J. L.

    2014-10-01

    In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

  10. Microstructural evolution in neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    English, C.A.; Phythian, W.J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

  11. 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

  12. Advanced neutron source reactor probabilistic flow blockage assessment

    International Nuclear Information System (INIS)

    Ramsey, C.T.

    1995-08-01

    The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool

  13. Reactor installation and maintenance for the Advanced Neutron Source

    International Nuclear Information System (INIS)

    Smith, B.R.

    1993-01-01

    Advanced Neutron Source (ANS) reactor assembly components have been modeled in great detail in IGRIP in order to realistically simulate preliminary installation and maintenance processes. Animation of these processes has been captured in a 15-minute video with narration. Approximately 90% of the parts were initially translated from CADAM (a two-dimensional drawing package) to IGRIP and then revolved or extruded. IGRIP's IGES translator greatly reduced the time required to perform this operation. The interfacing of devices in the work cell has identified numerous design inconsistencies. Most of the modeled reactor components are devices with a single degree of freedom (DOF) however, some of the slanted experiments required 6 DOF so that they could be removed at an angle in order to clear the reflector vessel flanges. IGRIP's collision detection feature proved to be extremely helpful in determining interferences when removing the experiments. The combination of three-dimensional visualization and collision detection allows engineers to clearly and easily visualize potential design problems before the construction phase of the project

  14. Study of the RP-10 reactor neutron beam applied to the neutron radiography

    International Nuclear Information System (INIS)

    Zegarra, Manuel; Lopez, Alcides

    2013-01-01

    We have studied the RP-10 reactor radial neutron beam No. 3, which is used for neutron radiographies, by comparing radiograph's with and without the inner duct, and neutron flux determination with in flakes along the external duct, being the presence of photons creating signals at comparable levels of neutron effects, which reduce the quality of the analysis, values around 10 6 and 10 4 n/cm 2 s for thermal and epithermal flux were obtained respectively. It is recommended evaluate the design of the internal duct which presents strong photon emission. (authors).

  15. 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.

  16. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Misao; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

  17. Neutronics comparative analysis between MNSR and slowpoke-II reactors

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.

    1999-01-01

    Neutronics analysis of both MNSR and Slowpoke reactors were made. Calculations including flux distribution, power estimation, excess and shutdown reactivity margins, flooding effects of irradiation sites, and initial investigation of fuel conversion from high to low enriched uranium were discussed. A neutronic 3-D model, dedicated mainly for the MNSR, has been developed to perform such neutronic calculations for both reactors. Well-known cell and core calculation codes such as WIMSD4 and CITATIONS have been used. It was found out that it is possible to lower the fuel enrichment of the Miniature Neutron Source Reactor (MNSR) to 20% using U O 2 as fuel instead of U Al 4 . The number of fuel elements required for the new core is 199. The use of double thickness of the bottom reflector in Slowpoke reactor made it possible to load the reactor with lower enriched fuel compared to MNSR. Values of reactivity flooding effects for single or combination of inner irradiation sites were obtained accurately. Results show good agreement with reported data for MNSR. (author)

  18. Change of neutron flow sensors effectiveness in the course of reactor experiments

    International Nuclear Information System (INIS)

    Kurpesheva, A.M.; Kotov, V.M.; Zhotabaev, Zh.R.

    2007-01-01

    Full text: IGR reactor is a reactor of thermal capacity type. During the operation, uranium-graphite core can be heated up to 1500 deg. C and reactivity can be changed considerably. Core dimensions are comparatively small. Amount of control rods, providing required reactivity, is not big as well. Increasing of core temperature leads to the rise of neutrons path length in its basic material - graphite. Change of temperature is not even. All this causes the non-conservation of neutron flows ratio in irradiated sample and in the place of reactor power sensors installation. Deviations in this ratio were registered during the number of reactor experiments. Empiric corrections can be introduced in order to decrease influence of change of neutron flow effectiveness upon provision of required parameters of investigated matters load. However, dependence of these corrections upon many factors can lead to the increasing of instability of process control. Previous experiment-calculated experiments showed inequality of neutron field in the place of sensors location (up to tens of percent), low effectiveness of experimental works, carried out without access to the individual reactor laying elements. Imperfection during the experiment was an idea of possibility to connect distribution of out of reactor neutron flow and control rods position. Subsequent analysis showed that for the development of representative phenomenon model it is necessary to take into account reactor operation dynamic subject to unevenness of heating of individual laying parts. Elemental calculations showed that temperature laying effects in the change of neutron outer field are great. Algorithm of calculations for the change of outer filed and field of investigated fabrication includes calculation of neutron-physic reactor characteristics interlacing with calculations of thermal-physic reactor characteristics, providing correlation of temperature fields for neutron-physic calculations. In the course of such

  19. Fast neutron reactors: the safety point of view

    International Nuclear Information System (INIS)

    Laverie, M.; Avenas, M.

    1984-01-01

    All versions of nuclear reactors present favourable and unfavourable characteristics from the point of view of safety. The safety of the installations is obtained by making efforts to utilize in the best possible way those which are favourable and by taking proper steps in the face of those which are unfavourable. The present article shows how this general principle has been applied as regards the fast neutron reactors of integrated design which have been developped in France, taking into account the specific features of this version. A qualitative method to compare the safety of this version with that of pressurized water reactors which has been widely put to the test commercially all over the world is presented. These analyses make, generally speaking, several positive characteristics stand out for these fast neutron reactors from the safety aspects [fr

  20. Preliminary design concepts for the advanced neutron source reactor systems

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1988-01-01

    This paper describes the initial design work to develop the reactor systems hardware concepts for the advanced neutron source (ANS) reactor. This project has not yet entered the conceptual design phase; thus, design efforts are quite preliminary. This paper presents the collective work of members of the Oak Ridge National Laboratory, Martin Marietta Energy Systems, Inc., Engineering Division, and other participating organizations. The primary purpose of this effort is to show that the ANS reactor concept is realistic from a hardware standpoint and to show that project objectives can be met. It also serves to generate physical models for use in neutronic and thermal-hydraulic core design efforts and defines the constraints and objectives for the design. Finally, this effort will develop the criteria for use in the conceptual design of the reactor

  1. Intense neutron irradiation facility for fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

  2. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    International Nuclear Information System (INIS)

    Cartier, J.; Casoli, P.; Chappert, F.

    2013-01-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  3. Neutron multiplier alternative for fusion reactor blankets

    International Nuclear Information System (INIS)

    Taczanowski, S.

    1980-01-01

    A proposal is given to replace neutron multiplier needed to enable low lithium and tritium inventories simultaneously assuring sufficient production of tritium, by an efficient moderator ( 7 LiH or 7 LiD). The advantageous effect of the intensified neutron energy degradation is due to the 1/v character of the main tritium producing reaction. The slowing-down medium is designed to be the source of moderated neutrons for the surrounding Li ( 6 Li enriched) region where the most of tritium is to be produced. The surplus tritium production remains stored in the moderator zone. Some preliminary calculations illustrating the above concept were carried out and the neutron flux and tritium production distributions are presented. The indications regarding further studies are also suggested. (author)

  4. Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

    International Nuclear Information System (INIS)

    Vaiana, F.

    2009-11-01

    This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

  5. Neutron scattering research at JAERI reactors - past, present and future -

    International Nuclear Information System (INIS)

    Funahashi, Satoru; Morii, Yukio; Minakawa, Nobuaki

    1992-01-01

    It was in 1961 that the first neutron scattering experiment was performed in Japan at JRR-2. The start of JRR-3 in 1964 accelerated the neutron scattering activities in Japan. The research in this field in Japan grew up by using these two research reactors. Among them JRR-2 has played an important role because its neutron flux was about seven times higher than that of the old JRR-3. The completion of the new JRR-3M in 1990 made an epoch to the neutron scattering activities in Japan. The long-waited JRR-3M came up to the expectations of the scientists of Japan. It is a realization of the ideal reactor with tangential beam holes, cold source and neutron guides in a large guide hall. The flux at the neutron scattering instruments is about five times higher than that of JRR-2. Utilization of JRR-3M has just started. Twelve neutron scattering machines are running there. The number will increase up to close twenty in a couple of years. (author)

  6. Neutrons down-under: Australia's research reactor review

    International Nuclear Information System (INIS)

    Murray, Allan

    1995-01-01

    Australian research reactor review commenced in September 1992, the Review had the following Terms of Reference: Whether, on review of the benefits and costs for scientific, commercial, industrial and national interest reasons, Australia has a need for a new reactor; a review of the present reactor, HIFAR, to include: an assessment of national and commercial benefits and costs of operations, its likely remaining useful life and its eventual closure and decommissioning; if Australia has a need for a new nuclear research reactor, the Review will consider: possible locations for a new reactor, its environmental impact at alternative locations, recommend a preferred location, and evaluate matters associated with regulation of the facility and organisational arrangements for reactor-based research. From the Review findings the following recommendations were stated: keep HIFAR going; commission a PRA to ascertain HIFAR's remaining life and refurbishment possibilities; identify and establish a HLW repository; accept that neither HIFAR nor a new reactor can be completely commercial; any decision on a new neutron source must rest primarily on benefits to science and Australia's national interest; make a decision on a new neutron source in about five years' time (1998). Design Proposals for a New Reactor are specified

  7. EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  8. Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

    Energy Technology Data Exchange (ETDEWEB)

    Gryzinski, M.A.; Wielgosz, M. [National Centre for Nuclear Research, Andrzeja Soltana 7, 05-400 Otwock-Swierk (Poland)

    2015-07-01

    The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel is additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with

  9. A conceptual design of neutron tumor therapy reactor facility with a YAYOI based fast neutron source reactor

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki; An, Shigehiro.

    1983-01-01

    Fast neutron is known as one of useful radiations for radiation therapy of tumors. Boron neutron capture therapy (BNCT) of tumors which makes use of 10 B(n, α) 7 Li reaction of 10 B compounds selectively attached to tumor cells with thermal and intermediate neutrons is another way of neutron based radiation therapy which is, above all, attractive enough to kill tumor cells selectively sparing normal tissue. In Japan, BNCT has already been applied and leaned to be effective. After more than a decade operational experiences and the specific experiments designed for therapeutical purposes, in this paper, a conceptual design of a special neutron therapy reactor facility based on YAYOI - fast neutron source reactor of Nuclear Engineering Research Laboratory, Faculty of Engineering, the University of Tokyo - modified to provide an upward beam of fast and intermediate neutrons is presented. Emphasis is placed on the in-house nature of facility and on the coordinating capability of biological and physical researches as well as maintenances of the facility. (author)

  10. Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

    International Nuclear Information System (INIS)

    Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

    2011-01-01

    Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

  11. Neutron density optimal control of A-1 reactor analoque model

    International Nuclear Information System (INIS)

    Grof, V.

    1975-01-01

    Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

  12. Dynamic modeling of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    March-Leuba, J.; Ibn-Khayat, M.

    1990-01-01

    The purpose of this paper is to provide a summary description and some applications of a computer model that has been developed to simulate the dynamic behavior of the advanced neutron source (ANS) reactor. The ANS dynamic model is coded in the advanced continuous simulation language (ACSL), and it represents the reactor core, vessel, primary cooling system, and secondary cooling systems. The use of a simple dynamic model in the early stages of the reactor design has proven very valuable not only in the development of the control and plant protection system but also of components such as pumps and heat exchangers that are usually sized based on steady-state calculations

  13. Evaluation of neutron exposure conditions for the Buffalo Reactor

    International Nuclear Information System (INIS)

    Lippincott, E.P.; Kellogg, L.S.; McElroy, W.N.; Baldwin, C.A.

    1984-04-01

    The light water test reactor at the Nuclear Science and Technology Facility of the State University of New York at Buffalo is currently being used to irradiate specimens in in-core positions for NRC-sponsored metallurgical tests. It is important that the neutron exposures for these Buffalo tests be consistent with those determined for related irradiations in the BSR and ORR reactor at ORNL. Therefore, HEDL National Reactor Dosimetry Center dosimetry procedures and ORNL calculational procedures were used for an evaluation of typical test conditions

  14. Neutron radiation characteristics of the IVth generation reactor spent fuel

    Science.gov (United States)

    Bedenko, Sergey; Shamanin, Igor; Grachev, Victor; Knyshev, Vladimir; Ukrainets, Olesya; Zorkin, Andrey

    2018-03-01

    Exploitation of nuclear power plants as well as construction of new generation reactors lead to great accumulation of spent fuel in interim storage facilities at nuclear power plants, and in spent fuel «wet» and «dry» long-term storages. Consequently, handling the fuel needs more attention. The paper is focused on the creation of an efficient computational model used for developing the procedures and regulations of spent nuclear fuel handling in nuclear fuel cycle of the new generation reactor. A Thorium High-temperature Gas-Cooled Reactor Unit (HGTRU, Russia) was used as an object for numerical research. Fuel isotopic composition of HGTRU was calculated using the verified code of the MCU-5 program. The analysis of alpha emitters and neutron radiation sources was made. The neutron yield resulting from (α,n)-reactions and at spontaneous fission was calculated. In this work it has been shown that contribution of (α,n)-neutrons is insignificant in case of such (Th,Pu)-fuel composition and HGTRU operation mode, and integral neutron yield can be approximated by the Watt spectral function. Spectral and standardized neutron distributions were achieved by approximation of the list of high-precision nuclear data. The distribution functions were prepared in group and continuous form for further use in calculations according to MNCP, MCU, and SCALE.

  15. Neutronic scoping studies for the tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Bettis, E.S.; McAlees, D.G.; Watts, H.L.; Williams, M.L.

    1976-02-01

    One-dimensional neutron and photon radiation transport methods have been used to investigate candidate blanket configurations and compositions for use in the Tokamak Experimental Power Reactor. Seven blanket designs are compared in terms of energy recovery, radiation attenuation, potential radiation damage, and, where applicable, tritium breeding

  16. Neutron-antineutron transition search at HFIR reactor

    International Nuclear Information System (INIS)

    Kamyshkov, Yuri A.

    1997-01-01

    A new experiment to search for neutron-antineutron transitions was recently proposed for High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). In this paper the physics motivation of a new search, the scheme and the discovery potential of the proposed HFIR-based experiment are discussed

  17. Neutron-antineutron transition search at HFIR Reactor

    International Nuclear Information System (INIS)

    Kamyshkov, Y.A.

    1997-01-01

    A new experiment to search for neutron-antineutron transitions was recently proposed for High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). In this paper the physics motivation of a new search, the scheme and the discovery potential of the proposed HFIR-based experiment are discussed

  18. A multi-group neutron noise simulator for fast reactors

    International Nuclear Information System (INIS)

    Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe

    2013-01-01

    Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring

  19. Neutron streaming evaluation for the DREAM fusion power reactor

    International Nuclear Information System (INIS)

    Seki, Yasushi; Nishio, Satoshi; Ueda, Shuzo; Kurihara, Ryoichi

    2000-01-01

    Aiming at high degree of safety and benign environmental effect, we have proposed a tokamak fusion reactor concept called DREAM, which stands for DRastically EAsy Maintenance Reactor. The blanket structure of the reactor is made from very low activation SiC/SiC composites and cooled by non-reactive helium gas. High net thermal efficiency of about 50% is realized by 900 C helium gas and high plant availability is possible with simple maintenance scheme. In the DREAM Reactor, neutron streaming is a big problem because cooling pipes with diameter larger than 80 cm are used for blanket heat removal. Neutron streaming through the cooling pipes could cause hot spots in the superconducting magnets adjacent to the cooling pipes to shorten the magnet lifetime or increase cryogenic cooling requirement. Neutron streaming could also activate components such as gas turbine further away from the fusion plasma. The effect of neutron streaming through the helium cooling pipes was evaluated for the two types of cooling pipe extraction scheme. The result of a preliminary calculation indicates the gas turbine activation prohibits personnel access in the case of inboard pipe extraction while with additional shielding measures, limited contact maintenance is possible in the case of outboard extraction. (author)

  20. Transient neutrons flux behaviour in a spherical reactor core

    International Nuclear Information System (INIS)

    Souza, A.W.A. de.

    1978-11-01

    This work studies the transient neutron flux in a fast reactor of spherical geometry. The burning of U 235 nuclei is equated and two kinds of reflector were studied. The numeric solutions are then compared with the results for those reflectors. (author) [pt

  1. Equipment for neutron measurements at VR-1 Sparrow training reactor

    International Nuclear Information System (INIS)

    Kolros, Antonin; Huml, Ondrej; Kos, Josef

    2008-01-01

    Full text: The VR-1 Sparrow training reactor is the experimental nuclear facility especially employed for education and teaching of students from different technical universities in the Czech Republic and other countries. Since 2005 the uniform all-purpose devices EMK310 have been used for measurement at reactor laboratory with different type of gas filled neutron detectors. The neutron detection system are employed for reactivity measurement, control rod calibration, critical experiment, study of delayed neutrons, study of nuclear reactor dynamics and study of detection systems dead time. The small dimension isotropic detectors are especially used for measurement of thermal neutron flux distribution inside the reactor core. The EMK-310 is a high performance, portable, three-channel fast amplitude analyzer designed for counting applications. It was developed for nuclear applications and made in close co-operation with firm TEMA Ltd. The precise rack eliminates electromagnetic disturbance and contains the control unit and four modules. The modules of high voltage supply and amplifier for gas filled detectors or scintillation probes are used in basic configuration. Software is tailored specifically to the reactor measurement and allows full online control. For applications involving the study of signals that may vary with the time, example study of delayed neutrons or nuclear reactor dynamics, the EMK-310 provides a Multichannel Scaling (MCS) acquisition mode. MCS dwell time can be set from 2 ms. Now, the new generation of digital multichannel analyzers DA310 is introduced. They have similarly attributes as EMK310 but the output information of unipolar signals from detector is more complete. The pipeline A/D converter with field programmable gate array (FPGA) is the hearth of the DA310 device. The resolution is 12 bits (4096 channels); the sample frequency is 80 MHz. The application for the neutron noise analysis is supposed. The correction method for non linearity

  2. Neutron detectors for nuclear reactor control

    International Nuclear Information System (INIS)

    Duchene, Jean; Verdant, Robert.

    1974-01-01

    In view of the importance of in-core measurements the distinction is made between detectors used outside and inside the core. In the former case proportional counters, fission chambers and boron chambers are reviewed in turn. The only in-core detectors considered are those giving a direct measurement, i.e. supplying an electric signal representative of the neutron fluence rate while in the measurement position at the point given. Two kinds of detectors are used for direct measurements: miniature fission chambers and collectors, known also as neutron-electron converters [fr

  3. 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)

  4. NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

    Science.gov (United States)

    Oakes, L.C.; Walker, C.S.

    1959-12-15

    ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

  5. Design of neutron 'fluse-trop' in Tehran Research Reactor

    International Nuclear Information System (INIS)

    Khalafi, H.

    1999-01-01

    In this project, a Neutron Flux-Trap intended for 5MW Tehran Research Reactor was designed. Fuel conversion from HEU to LEU in research reactors usually deprives the core from the high neutron flux. Therefore one has to look for a remedy in such situations otherwise radioisotope production, especially for those neutron demanding ones, falls down dramatically. The initiations of Neutron Flux-Trap idea comes true to face up this problem and provide an appropriate place inside or outside the core with sufficient neutron flux higher than the normal average level. To implement such a design, a number of codes and calculational tools have been used. At preliminary stage, WIMSD and EXTERMINATOR-II were used and then at later stages CITATION and MCNP codes were used for final design. Furthermore, SAND-II and ORIGEN were also employed for spectral analysis and radioisotope production calculations. Good agreements resulted with experiments and especially the case that Neutron Flux-Trap filled with ordinary water

  6. Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Soltes, Jaroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague, (Czech Republic); Viererbl, Ladislav; Lahodova, Zdena; Koleska, Michal; Vins, Miroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic)

    2015-07-01

    In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which has a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values

  7. Neutronic study of a nuclear reactor of fused salts

    International Nuclear Information System (INIS)

    Garcia B, F. B.; Francois L, J. L.

    2012-10-01

    The reactors of fused salts called Molten Salt Reactor have presented a resurgence of interest in the last decade, due to they have a versatility in particular to operate, either with a thermal or fast neutrons spectrum. The most active development was by the middle of 1950 and principles of 1970 in the Oak Ridge National Laboratory. In this work some developed models are presented particularly and studied with the help of the MCNPX code, for the development of the neutronic study of this reactor, starting of proposed models and from a simple and homogeneous geometry until other more complex models and approximate to more real cases. In particular the geometry conditions and criticality of each model were analyzed, the isotopic balance, as well as the concentrations of the salts and different assigned fuel types. (Author)

  8. Neutron radiography (NRAD) reactor 64-element core upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately ±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  9. Lowering the enrichment of the Syrian miniature neutron source reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.

    1999-01-01

    An investigation of lowering the fuel enrichment of MNSR was realized. A 3-D neutronic model was developed for the analysis of the reactor. It was found that lower number of fuel elements is needed when UO 2 is used with 5.45 g of 235 U content in each fuel rod. Sensitivity of the reactor to the purity of the beryllium reflector proved to be an important factor in determining the reactor neutronics as well as the weight of loaded fuel in the core. Inherent safety features of low excess reactivity and shutdown margins are preserved and enhanced. Average thermal fluxes over different zones of the core are kept very much unchanged

  10. Lowering the enrichment of the Syrian miniature neutron source reactor

    International Nuclear Information System (INIS)

    Kamis, I.; Khattab, K.

    1999-01-01

    An investigation of lowering the fuel enrichment of MNSR was realized. A 3-D neutronic model was developed for the analysis of the reactor. It was found that lower number of fuel elements is needed when UO 2 is used with 5.45 g of 235 U content in each fuel rod. sensitivity of the reactor to the purity of the beryllium reflector proved to be an important factor in determining the reactor neutronics as well as the weight of loaded fuel in the core. Inherent safety feature of low excess reactivity and shutdown margins are preserved and enhanced. average thermal fluxes over different zones of the core are kept very much unchanged. (author)

  11. Structural elements for fast-neutron reactors

    International Nuclear Information System (INIS)

    Blin, J.C.; Sainfort, Gerard; Silvent, Alain; Silvestres, Georges.

    1974-01-01

    These elements are characterized in that they are obtained from a nickel-alloy and at least a material M, selected from the group comprising iron and silicon, in proportions, by weight, such that irradiation by fast neutrons leads to the generation of Ni 3 -M with no noticeable swelling of said elements. This can be applied to fuel assembly cladding [fr

  12. Archaeometric studies by using neutron radiography in ITU TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Tugrul, A. Beril

    2008-01-01

    Archaeometric many studies have been done by using neutron radiography in ITU TRIGA Mark-II Training and Research Reactor for over 15 years. Tangential beam tube has been arranged for using neutron radiography. Generally, transfer technique has been preferred with using dysprosium screen, but indium screen also is used. Some studies are described which are all on the Anatolian artefacts. The first study from 13th century AD deals with Seljukian period from south-east Anatolia. It investigated a plate from Great Mosque door in Cizre. With means of the neutron radiography painting traces are investigated on the plates. Organic dye traces are noticed on some of plates, which have generally animal figures. Other studies from Urartu period at the first millennium B.C, investigates artefacts found at the vicinity of Van on east Anatolia. An important one is a sword that was found in a grave. It has some corrosion defects. The neutron radiography was applied and shown that wooden parts are there. Other studies referred to samples from the Ikiztepe Excavation site on north Anatolia. Many artefacts were examined by neutron radiography. Some of them evidenced animal parts are recognised as covering parts. An interesting result was obtained to a sword and its sheath that were corroded together. After the neutron radiography applications, it was noticed that there are a cloth between the sword and its sheath. Hence, it was the cause of corrosion of the artefact. By using neutron radiography, many interesting and detailed results were observed by means of the neutron beam from the ITU TRIGA Mark-II Training and Research Reactor. Some of them could not be evidenced by means of any other technique

  13. Comparison study on in-core neutron detector for online neutron flux mapping of research and power reactor

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Mohd Idris Taib; Izhar Abu Husin; Nurfarhana Ayuni

    2010-01-01

    This paper presents the comparison study on In-Core neutron detector using for online flux mapping of Research and Power reactor. Technical description of in-core neutron also taken into consideration to identify the different characterization of neutron detector and describe on Self Power neutron detector (SPND) for online neutron flux mapping. Able to provide information on the neutron flux distribution and understand how in-core neutron detector are being used in nuclear power plant including to enable to state the principles of neutron detector. (author)

  14. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    International Nuclear Information System (INIS)

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    Highlights: ► Kinetic parameters of Tehran research reactor mixed-core have been calculated. ► Burn-up effect on TRR kinetics parameters has been studied. ► Replacement of LEU-CFE with HEU-CFE in the TRR core has been investigated. ► Results of each mixed core were compared to the reference core. ► Calculation of kinetic parameters are necessary for reactivity and power excursion transient analysis. - Abstract: In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR P C package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change

  15. 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)

  16. 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)

  17. A new neutron reflectometer at Australia's HIFAR research reactor

    International Nuclear Information System (INIS)

    James, M.; Nelson, A.; Schulz, J.C.; Jones, M.J.; Studer, A.J.; Hathaway, P.

    2005-01-01

    A new neutron reflectometer has been built at Australia's 10MW HIFAR research reactor. The X172 reflectometer operates in a monochromatic, angular dispersive mode collecting reflectivity data as a function of angle. The incident neutron beam is monochromated by a pair of pyrolytic graphite crystals (λ=2.43A) before being collimated using a pair of motorised sintered B 4 C slits. Detection of the reflected neutron beam is via a 10-atmosphere, helium-3, linear position sensitive detector. Examples of data collected using the X172 reflectometer at air-solid and solid-liquid interfaces are given. Neutron reflectivity values as low as 10 -5 have been measured on this instrument

  18. EDF research on fast neutron reactors

    International Nuclear Information System (INIS)

    In order to make possible the calculation of the temperatures of the sodium, of the sheath and of the fuel in fast reactor assemblies, taking into account the mixing phenomena induced by the helicoidal wires, two design codes have been developed. Those codes have then been adapted for their integration in the Superalcyon system. This system shall constitute the reference tool for the development of those codes that shall manage Phenix, and other reactors of the family. Cooling accidents, thermohydraulic studies, and steam generator studies are also in progress

  19. Studying the effect of xenon poisoning on the power of the Syrian miniature neutron source reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.

    1999-07-01

    The uranium 235 is often used as a fuel to produce the energy in nuclear reactors. Uranium nuclei are fissioned with thermal neutrons and produce energy plus a number of neutrons. A fraction of such fission neutrons is involved in other fission with new nuclei to sustain the fission reactions. The remain fraction of the neutrons is lost from the reactor in two ways: escaped from the reactor, or absorbed with other nuclei that exist in the reactor before or produced from fission. Fission nuclei which absorb neutrons heavily are called p oison , such as Xe 135. Because Xe 135 absorbs neutrons heavily, it reduces the number of neutrons in the reactor. Hence, Xe 135 is studied explicitly in the MNSR reactor, and calculation of its negative reactivity is presented in this research during the operation, equilibrium, and after the shutting down of the reactor. (author)

  20. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  1. Online failed fuel identification using delayed neutron detector signals in pool type reactors

    International Nuclear Information System (INIS)

    Upadhyay, Chandra Kant; Sivaramakrishna, M.; Nagaraj, C.P.; Madhusoodanan, K.

    2011-01-01

    In todays world, nuclear reactors are at the forefront of modern day innovation and reactor designs are increasingly incorporating cutting edge technology. It is of utmost importance to detect failure or defects in any part of a nuclear reactor for healthy operation of reactor as well as the safety aspects of the environment. Despite careful fabrication and manufacturing of fuel pins, there is a chance of clad failure. After fuel pin clad rupture takes place, it allows fission products to enter in to sodium pool. There are some potential consequences due to this such as Total Instantaneous Blockage (TIB) of coolant and primary component contamination. At present, the failed fuel detection techniques such as cover gas monitoring (alarming the operator), delayed neutron detection (DND-automatic trip) and standalone failed fuel localization module (FFLM) are exercised in various reactors. The first technique is a quantitative measurement of increase in the cover gas activity background whereas DND system causes automatic trip on detecting certain level of activity during clad wet rupture. FFLM is subsequently used to identify the failed fuel subassembly. The later although accurate, but mainly suffers from downtime and reduction in power during identification process. The proposed scheme, reported in this paper, reduces the operation of FFLM by predicting the faulty sector and therefore reducing reactor down time and thermal shocks. The neutron evolution pattern gets modulated because fission products are the delay neutron precursors. When they travel along with coolant to Intermediate heat Exchangers, experienced three effects i.e. delay; decay and dilution which make the neutron pulse frequency vary depending on the location of failed fuel sub assembly. This paper discusses the method that is followed to study the frequency domain properties, so that it is possible to detect exact fuel subassembly failure online, before the reactor automatically trips. (author)

  2. Four energy group neutron flux distribution in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION code

    International Nuclear Information System (INIS)

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

    2009-01-01

    A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)

  3. A neutron tomography facility at a low power research reactor

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H

    2001-01-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at ...

  4. Reactor neutron activation analysis of industrial materials

    International Nuclear Information System (INIS)

    Niese, S.

    1983-01-01

    The specific application of neutron activation analysis (n.a.a.) for industrial materials is demonstrated by the determination of impurities in BeO, Al, Si, Cu, Ge, GaP, GaAs, steel, and irradiated uranium. A group scheme gives an orientation about the possibilities of n.a.a. The use of different standards, methods for the measurement of low radioactivities and errors caused by recoil reaction and radiation stimulated diffusion are discussed. (author)

  5. Higher order harmonics of reactor neutron equation

    International Nuclear Information System (INIS)

    Li Fu; Hu Yongming; Luo Zhengpei

    1996-01-01

    The flux mapping method using the higher order harmonics of the neutron equation is proposed. Based on the bi-orthogonality of the higher order harmonics, the process and formulas for higher order harmonics calculation are derived via the source iteration method with source correction. For the first time, not only any order harmonics for up-to-3-dimensional geometry are achieved, but also the preliminary verification to the capability for flux mapping have been carried out

  6. Multicomponent activation detector measurements of reactor neutron spectra

    International Nuclear Information System (INIS)

    Sandberg, J.; Aarnio, P. A.; Routti, J. T.

    1984-01-01

    Information on the neutron flux is required in many applications of research reactors, such as activation analysis or radiation damage measurements. Flux spectrum measurements are commonly carried out with activation foils. The reaction types used are threshold reactions in the fast energy region, resonance reactions in the intermediate region and neutron capture reactions with l/v-cross section in the thermal region. It has been shown that it is possible to combine several detector elements into homogeneous multicomponent detectors. The activities of all detector reaction products can be determined with a single gamma spectrum measurement. The multicomponent principle sets some restrictions on the choice of detector reactions, for example, each product nuclide may be produced in one reaction only. Separate multicomponent threshold and resonance detectors were designed for the fast and intermediate regions, respectively. The detectors were fabricated in polyethylene irradiation capsules or quartz glass ampoules, and they were irradiated in a cadmium cover. The detectors were succesfully used in the irradiation ring and in the core of a Triga reactor. The intermediate and fast neutron spectrum was unfolded with the least-squares unfolding program LOUHI. According to the preliminary results multicomponent activation detectors might constitute a convenient means for carrying out routine neutron spectrum measurements in research reactors. (orig.)

  7. Effective neutron temperature measurements in well moderated reactor by the reactivity coefficient method

    International Nuclear Information System (INIS)

    Raisic, N.; Klinc, T.

    1968-11-01

    The ratio of the reactivity changes of a nuclear reactor produced by successive introduction of two different neutron absorbers in the reactor core, has been measured and information on effective neutron temperature at a particular point obtained. Boron was used as a l/v absorber and cadmium as an absorber sensiti ve to neutron temperature. Effective neutron temperature distribution has been deduced by moving absorbers across the reactor core and observing the corresponding reactivity changes. (author)

  8. RA reactor operation and maintenance in 1992, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Tanaskovic, M.

    1992-01-01

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr

  9. Computational analysis of neutronic parameters of CENM TRIGA Mark II research reactor

    International Nuclear Information System (INIS)

    El Younoussi, C.; El Bakkari, B.; Boulaich, Y.; Riyach, D.; Otmani, S.; Marrhich, I.; Badri, H.; Htet, A.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Zoubair, M.; Ossama, M.; Chakir, E.

    2010-01-01

    The CENM TRIGA MARK II reactor is part of the National Center for Energy, Sciences and Nuclear Techniques (CNESTEN). It's a standard design 2MW, natural-convection-cooled reactor with a graphite reflector containing 4 beam tubes and a thermal column. The reactor has several applications in different fields as industry, agriculture, medicine, training and education. In the present work a computational study has been carried out in the framework of neutronic parameters studies of the reactor. A detailed MCNP model that include all elements of the core and surrounding structures has been developed to calculate different parameters of the core (The effective multiplication factor, reactivity experiments comprising control rods worth, excess reactivity and shutdown margin). Further calculations have been carried out to calculate the neutron flux profiles at different locations of the reactor core. The cross sections used are processed from the library provided with MCNP5 and based on the ENDF/B-VII with continuous dependence in energy and special treatment of thermal neutrons in lightweight materials. (author)

  10. Accelerator driven reactors, - the significance of the energy distribution of spallation neutrons on the neutron statistics

    Energy Technology Data Exchange (ETDEWEB)

    Fhager, V

    2000-01-01

    In order to make correct predictions of the second moment of statistical nuclear variables, such as the number of fissions and the number of thermalized neutrons, the dependence of the energy distribution of the source particles on their number should be considered. It has been pointed out recently that neglecting this number dependence in accelerator driven systems might result in bad estimates of the second moment, and this paper contains qualitative and quantitative estimates of the size of these efforts. We walk towards the requested results in two steps. First, models of the number dependent energy distributions of the neutrons that are ejected in the spallation reactions are constructed, both by simple assumptions and by extracting energy distributions of spallation neutrons from a high-energy particle transport code. Then, the second moment of nuclear variables in a sub-critical reactor, into which spallation neutrons are injected, is calculated. The results from second moment calculations using number dependent energy distributions for the source neutrons are compared to those where only the average energy distribution is used. Two physical models are employed to simulate the neutron transport in the reactor. One is analytical, treating only slowing down of neutrons by elastic scattering in the core material. For this model, equations are written down and solved for the second moment of thermalized neutrons that include the distribution of energy of the spallation neutrons. The other model utilizes Monte Carlo methods for tracking the source neutrons as they travel inside the reactor material. Fast and thermal fission reactions are considered, as well as neutron capture and elastic scattering, and the second moment of the number of fissions, the number of neutrons that leaked out of the system, etc. are calculated. Both models use a cylindrical core with a homogenous mixture of core material. Our results indicate that the number dependence of the energy

  11. Accelerator driven reactors, - the significance of the energy distribution of spallation neutrons on the neutron statistics

    International Nuclear Information System (INIS)

    Fhager, V.

    2000-01-01

    In order to make correct predictions of the second moment of statistical nuclear variables, such as the number of fissions and the number of thermalized neutrons, the dependence of the energy distribution of the source particles on their number should be considered. It has been pointed out recently that neglecting this number dependence in accelerator driven systems might result in bad estimates of the second moment, and this paper contains qualitative and quantitative estimates of the size of these efforts. We walk towards the requested results in two steps. First, models of the number dependent energy distributions of the neutrons that are ejected in the spallation reactions are constructed, both by simple assumptions and by extracting energy distributions of spallation neutrons from a high-energy particle transport code. Then, the second moment of nuclear variables in a sub-critical reactor, into which spallation neutrons are injected, is calculated. The results from second moment calculations using number dependent energy distributions for the source neutrons are compared to those where only the average energy distribution is used. Two physical models are employed to simulate the neutron transport in the reactor. One is analytical, treating only slowing down of neutrons by elastic scattering in the core material. For this model, equations are written down and solved for the second moment of thermalized neutrons that include the distribution of energy of the spallation neutrons. The other model utilizes Monte Carlo methods for tracking the source neutrons as they travel inside the reactor material. Fast and thermal fission reactions are considered, as well as neutron capture and elastic scattering, and the second moment of the number of fissions, the number of neutrons that leaked out of the system, etc. are calculated. Both models use a cylindrical core with a homogenous mixture of core material. Our results indicate that the number dependence of the energy

  12. Thermal hydraulic and neutronic interaction in the rotating bed reactor

    International Nuclear Information System (INIS)

    Lee, C.C.

    1986-01-01

    Power transient characteristics in a rotating fluidized bed reactor (RBR) are investigated theoretically. A propellant flow perturbation is assumed to occur in an initially equilibrium state of the core. Transfer functions representing quasi-one-dimensional mutual feedback between thermal hydraulics and neutronics are developed and analyzed in the frequency domain. Neutronic responses are determined by Fermi-age theory for slowing down of fast neutrons and diffusion theory for thermal neutron distribution. Neutron leakage through the exhaust nozzle is accounted for by applying diffuse view factors similar to those applied in radiative heat transfer. The bed expansion behavior is described by a kinematic wave equation derived from the continuity of the gas phase. The drift flux approach is used to determine the yield fractions in the equilibrium bed. Thermal responses of fuel are evaluated by dividing it into several volume-averaged zones to better account for the transient effects over single zone models. Sample calculations are undertaken for the various operation conditions and design parameters of the RBR based on 250 MW/sub t/, 1000 MW/sub t/, and 5000 MW/sub t/ power reactors. The results show that power transients are dependent on the parametric changes of optical thickness and view factors

  13. Cold neutron source conceptual designing for Tehran Research Reactor

    International Nuclear Information System (INIS)

    Khajvand, N.; Mirvakili, S.M.; Faghihi, F.

    2016-01-01

    Highlights: • Cold neutron source conceptual designing for Tehran research reactor is carried out. • Type and geometry of moderator and dimensions of cold neutron source are analyzed. • Liquid hydrogen with more ortho-concentration can be better option as moderator. - Abstract: A cold neutron source (CNS) conceptual designing for the Tehran Research Reactor (TRR) were carried out using MCNPX code. In this study, a horizontal beam tube of the core which has appropriate the highest thermal flux is selected and parametric analysis to choose the type and geometry of the moderator, and the required CNS dimensions for maximizing the cold neutron production was performed. In this design the moderator cell has a spherical annulus structure, and the cold neutron flux and its brightness are calculated together with the nuclear heat load of the CNS for a variety of materials including liquid hydrogen, liquid deuterium, and solid methane. Based on our study, liquid hydrogen with more ortho-concentration than para and solid methane are the best options.

  14. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    International Nuclear Information System (INIS)

    Rastogi, B.P.

    1989-01-01

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

  15. Reactor Noise: A study of Neutronic Fluctuations in Low-Power Nuclear Reactors, with Special Emphasis on Accurate Time-Domain Analysis. RCN Report

    Energy Technology Data Exchange (ETDEWEB)

    Dragt, J. B.

    1968-10-15

    Nuclear reactors can be considered as devices in which nuclear energy is produced as a result of neutron-induced fission reactions. Reactor physics is a branch of applied physics, and is concerned with the physical aspects of the design and study of nuclear reactors. The motivation is the achievement of configurations, which meet certain requirements regarding safety, reliability, economy, etc. The reactor physical method is to study neutron populations in a reactor. This study has two aspects : - the microscopic aspect: a study of the nuclear processes that take place. This aspect belongs to nuclear physics. Reaction probabilities can be expressed in cross sections,which are assumed to be known for the second part: - the macroscopic aspect, concerned with neutron migration and multiplication. All basic features may be traced back to a knowledge of neutron distribution functions. For most phenomena it is sufficient to study the singulet density, i.e. the mean number of neutrons per unit volume, unit velocity, moving in unit solid angle. For the subject of this thesis this singulet density will appear to be insufficient. The theory for the macroscopic aspect is part of statistical mechanics, and is closely related to other statistical theories, for phenomena like transfer of radiation in stellar atmosphere, penetration of radiation in scattering media, cosmic ray showers, etc.

  16. Reactor Noise: A study of Neutronic Fluctuations in Low-Power Nuclear Reactors, with Special Emphasis on Accurate Time-Domain Analysis. RCN Report

    International Nuclear Information System (INIS)

    Dragt, J.B.

    1968-10-01

    Nuclear reactors can be considered as devices in which nuclear energy is produced as a result of neutron-induced fission reactions. Reactor physics is a branch of applied physics, and is concerned with the physical aspects of the design and study of nuclear reactors. The motivation is the achievement of configurations, which meet certain requirements regarding safety, reliability, economy, etc. The reactor physical method is to study neutron populations in a reactor. This study has two aspects : - the microscopic aspect: a study of the nuclear processes that take place. This aspect belongs to nuclear physics. Reaction probabilities can be expressed in cross sections,which are assumed to be known for the second part: - the macroscopic aspect, concerned with neutron migration and multiplication. All basic features may be traced back to a knowledge of neutron distribution functions. For most phenomena it is sufficient to study the singulet density, i.e. the mean number of neutrons per unit volume, unit velocity, moving in unit solid angle. For the subject of this thesis this singulet density will appear to be insufficient. The theory for the macroscopic aspect is part of statistical mechanics, and is closely related to other statistical theories, for phenomena like transfer of radiation in stellar atmosphere, penetration of radiation in scattering media, cosmic ray showers, etc

  17. Computed phase equilibria for burnable neutron absorbing materials for advanced pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Corcoran, E.C. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, P.O. Box 17000, St. Forces, Kingston, Ont., K7K 7B4 (Canada)], E-mail: emily.corcoran@rmc.ca; Lewis, B.J.; Thompson, W.T. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, P.O. Box 17000, St. Forces, Kingston, Ont., K7K 7B4 (Canada); Hood, J. [Atomic Energy of Canada Ltd., Sheridan Park, 2251 Speakman Drive, Mississauga, Ont., L5K 1B2 (Canada); Akbari, F.; He, Z. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, Ont., K0J 1J0 (Canada); Reid, P. [Atomic Energy of Canada Ltd., Sheridan Park, 2251 Speakman Drive, Mississauga, Ont., L5K 1B2 (Canada)

    2009-03-31

    Burnable neutron absorbing materials are expected to be an integral part of the new fuel design for the Advanced CANDU [CANDU is as a registered trademark of Atomic Energy of Canada Limited.] Reactor. The neutron absorbing material is composed of gadolinia and dysprosia dissolved in an inert cubic-fluorite yttria-stabilized zirconia matrix. A thermodynamic model based on Gibbs energy minimization has been created to provide estimated phase equilibria as a function of composition and temperature. This work includes some supporting experimental studies involving X-ray diffraction.

  18. Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Rojas R, E.; Benitez R, J. S.; Segovia de los Rios, J. A.; Rivero G, T.

    2009-10-01

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

  19. 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

  20. The neutron radiography facility at Tehran Research Reactor (TRR)

    International Nuclear Information System (INIS)

    Ali Pazirandeh

    2009-01-01

    Full text: Non-destructive testing in many fields of industry including detection of explosives, at the airports, testing for micro-cracks on airplane wings and turbine blades cracks is badly needed. Thermal neutron beam is one of preferable method to detect the micro-cracks, reveals the internal structure of components and explosives. The purpose of this paper is to present the neutron radiography facility at Tehran Research Reactor (TRR), Science and Technology Research Institute, and in particular to emphasize the industrial applications in wood industry, automobile engine inspection, minerals composition identification, turbine blade cracks detection. (author)

  1. Cylindrical IEC neutron source design for driven research reactor operation

    International Nuclear Information System (INIS)

    Miley, G.H.; Ulmen, B.; Amadio, G.; Leon, H.; Hora, H.

    2009-01-01

    A resurgence in nuclear power use is now underway worldwide. However, due many university research reactors shutdown, they must rely on using subcritical assemblies which employs a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The source is inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory. (author)

  2. Upgrading of neutron radiography/tomography facility at research reactor

    International Nuclear Information System (INIS)

    Abd El Bar, Waleed; Mongy, Tarek

    2014-01-01

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  3. 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

  4. Optimization of the Neutronics of the Advanced High Temperature Reactor

    International Nuclear Information System (INIS)

    Zakova, Jitka; Talamo, Alberto

    2006-01-01

    In these studies, we have investigated the neutronic and safety performance of the Advanced High Temperature Reactor (AHTR) for plutonium and uranium fuels and we extended the analysis to five different coolants. The AHTR is a graphite-moderated and molten salt-cooled high temperature reactor, which takes advantage of the TRISO particles technology for the fuel utilization. The conceptual design of the core, proposed at the Oak Ridge National Laboratory, aims to provide an alternative to helium as coolant of high-temperature reactors for industrial applications like hydrogen production. We evaluated the influence of the radial reflector on the criticality of the core for the uranium and plutonium fuels and we focused on the void coefficient of 5 different molten salts; since the safety of the reactor is enhanced also by the large and negative coefficient of temperature, we completed our investigation by observing the keff changes when the graphite temperature varies from 300 to 1800 K. (authors)

  5. Neutronic Core Performance of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Villarino, Eduardo; Hergenreder, Daniel; Matzkin, S

    2000-01-01

    The actual design state of core of CAREM-25 reactor is presented.It is shown that the core design complains with the safety and operation established requirements.It is analyzed the behavior of the reactor safety and control systems (single failure of the fast shut down system, single failure of the shut down system, single failure of the second shut down system, reactivity worth of the adjust and control system in normal operation and hot shut down, reactivity worth of the adjust and control system and the scheme of movement of the control rod during the operation cycle).It is shown the burnup profile of fuel elements with the proposed scheme of refueling and the burnup and power density distribution at different moments of the operation cycle.The power peaking factor of the equilibrium core is 2.56, the minimum DNBR is 1.90 and its average is 2.09 during the operation cycle

  6. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verma, V., E-mail: vasudha.verma@physics.uu.se [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Barbot, L.; Filliatre, P. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Hellesen, C. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Jammes, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Svärd, S. Jacobsson [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden)

    2017-07-11

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment. - Highlights: • Studied possibility of using SPNDs as in-core detectors in SFRs. • Study done to detect local power profile changes when reactor is at nominal power. • SPND with a Pt-emitter gives measurable prompt current of the order of 600 nA/m. • Dominant proportion of prompt response is maintained throughout the operation. • Detector signal gives dynamic information on the power fluctuations.

  7. Report on neutron reflectometry for the Australian Replacement Reactor

    International Nuclear Information System (INIS)

    James, M.

    2001-01-01

    There is a clear need for at least one neutron reflectometer at the Australian Replacement Research Reactor when it commences operation in 2005. The participants at the reflectometry workshop have identified that the neutron reflectometer to be built at the Australian Replacement Research Reactor must be capable of the study of: 1. Specular scattering from air/solid, solid/liquid and in particular 'free liquid' samples; and 2. Off-specular' scattering from the above sample types. 3. Kinetics phenomena on a minute or slower time scale; 4. A range of samples of differing thicknesses, ranging from ultra-thin films to thousand angstrom thick films. In order to achieve this the reflectometer should have the capacity to vary its resolution. Interest was also expressed at the ability to conduct glancing-angle and wide-angle scattering studies for the investigation of short length scale, in-plane structures. There was little interest expressed by the workshop participants for polarised neutron reflectometry. This report contains a scientific case for a neutron reflectometer to be built at the Australian Replacement Research Reactor on a cold neutron guide, which is based on the areas of scientific research expressed by the workshop participants. In addition, trends in neutron reflectometry research conducted at major overseas neutron facilities are noted. The new neutron Reflectometer should: 1. Be based on the Time-of-Flight method; 2. Have a vertical scattering plane (i.e. operate for horizontal samples); 3. Be located on the end of a cold neutron guide, or be built off the guide axis using a bender, 4. Have a position sensitive area detector, 5. Be similar in spirit to the new D17 reflectometer at the ILL. Basic aspects of a reflectometer design are discussed which meet the above-stated scientific criteria and include a preliminary list of instrument specifications, capabilities and ancillary equipment requested by the workshop participants. A preliminary instrument

  8. Reactor-moderated intermediate-energy neutron beams for neutron-capture therapy

    International Nuclear Information System (INIS)

    Less, T.J.

    1987-01-01

    One approach to producing an intermediate energy beam is moderating fission neutrons escaping from a reactor core. The objective of this research is to evaluate materials that might produce an intermediate beam for NCT via moderation of fission neutrons. A second objective is to use the more promising moderator material in a preliminary design of an NCT facility at a research reactor. The evaluations showed that several materials or combinations of materials could produce a moderator source for an intermediate beam for NCT. The best neutron spectrum for use in NCT is produced by Al 2 O 3 , but mixtures of Al metal and D 2 O are also attractive. Using the best moderator materials, results were applied to the design of an NCT moderator at the Georgia Institute of Technology Research Reactor's bio-medical facility. The amount of photon shielding and thermal neutron absorber were optimized with respect to the desired photon dose rate and intermediate neutron flux at the patient position

  9. Study of neutron absorbing microspheres in research reactors - Neutronic analyse

    International Nuclear Information System (INIS)

    Gana Watkins, Ignacio A.; Prado, Miguel O.; Mazufri, Claudio; Tunon, Juan M

    2012-01-01

    Now-a-days, it is increasingly common for nuclear power plants, as well as research reactors, to be designed and built with an alternative safety system aside from control rods. The acids and/or salts in solution injection systems is most frequently used. However, these systems present several implementation and operation problems due to the physical and chemical properties of the used compounds. After analyzing these drawbacks, we developed a new alternative safety system that contains the absorbing element isolated from the aqueous medium. In this context, it's proposed the use of aluminum borosilicate microspheres. The current paper presents erosion wear experiments to determine under which conditions microspheres can be considered as a potential component of a secondary shut down system in a nuclear facility (author)

  10. Neutron dosimeters and survey meters in accelerators, reactors and other neutron environments

    International Nuclear Information System (INIS)

    1989-03-01

    Neutron fields in occupationally accessible areas around nuclear reactors, radioisotope sources and medical and high energy accelerators have been characterized using currently available information. Neutron, rem meters, such as the Leake detector, are the most suitable instruments available for conducting neutron dose rate surveys in the vicinity of radioisotope neutron sources, nuclear reactors and medical accelerators. However, these instruments have been shown to be deficient in that they overrespond by a factor of four to neutrons in the 0.1 to 1 MeV range and are insensitive to neutrons from about 1 eV up to about 10 keV. Also, they are insensitive to neutrons above 20 MeV and their use must be restricted near high energy accelerators where significant numbers of neutrons above 20 MeV are known to be present. The most suitable instrument of measure dose from neutrons above 20 MeV is the 12 C(n,2n) 11 C scintillation chamber. Commercially available rem meters frequently use BF 3 counters in the pulse mode to detect thermal neutrons. Therefore, measurements around pulsed accelerators must be made with caution to ensure that the detector is not saturated during each pulse and that the accelerator pulse period is greater than the response time of the detector. The personal neutron dosimeters currently available either are known to be insensitive to neutrons above 20 MeV or have not been tested. Also, all except the albedo dosimeter are insensitive to or have not been tested for neutron energies in the range 1 eV to 10 keV. Several dosimeter types respond reasonably well to neutrons in the energy range 0.1 to 15 MeV, for example, CR-39, bubble and superheated drop detectors. However, the first gas a lower limit of sensitivity of about 0.3 mSv. The bubble detector can be designed to measure doses as small as 1μSv and offers the additional benefit of direct-reading capability. The superheated drop detector is not suitable for use around pulsed accelerators because

  11. Neutronics and mass transport in a chemical reactor associated with controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, M.; Lazareth, O.W.; Powell, J.R.

    1976-05-01

    The formation of ozone from oxygen and the dissociation carbon dioxide to carbon monoxide and oxygen is studied in a gamma-neutron chemical process blanket associated with a controlled thermonuclear reactor. Materials used for reactor tube wall will affect the efficiency of the energy absorption by the reactants and consequently the yield of reaction products. Three kinds of materials, aluminum, stainless steel and fiber (Al 2 O 3 )-aluminium are investigated for the tube wall material in the study

  12. Kriging-based algorithm for nuclear reactor neutronic design optimization

    International Nuclear Information System (INIS)

    Kempf, Stephanie; Forget, Benoit; Hu, Lin-Wen

    2012-01-01

    Highlights: ► A Kriging-based algorithm was selected to guide research reactor optimization. ► We examined impacts of parameter values upon the algorithm. ► The best parameter values were incorporated into a set of best practices. ► Algorithm with best practices used to optimize thermal flux of concept. ► Final design produces thermal flux 30% higher than other 5 MW reactors. - Abstract: Kriging, a geospatial interpolation technique, has been used in the present work to drive a search-and-optimization algorithm which produces the optimum geometric parameters for a 5 MW research reactor design. The technique has been demonstrated to produce an optimal neutronic solution after a relatively small number of core calculations. It has additionally been successful in producing a design which significantly improves thermal neutron fluxes by 30% over existing reactors of the same power rating. Best practices for use of this algorithm in reactor design were identified and indicated the importance of selecting proper correlation functions.

  13. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Yoder, G.L.; Wendel, M.W.

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs

  14. Japanese list of requests for neutron nuclear data for fission reactors

    International Nuclear Information System (INIS)

    Igarasi, Sin-iti; Asami, Tetsuo

    1977-05-01

    Requests for neutron nuclear data for fission reactors are presented. These are screened by a WRENDA Working Group of Japanese Nuclear Data Committee and submitted to WRENDA 76/77. This report includes 163 requests of which 55 requests are newly registered in the WRENDA. Three requests of the previous list are withdrawn. This activity is a part of the international cooperation with CCDN, NEANDC and INDC. (auth.)

  15. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  16. Neutron physics of a high converting advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Berger, H.D.

    1985-01-01

    The neutron physics of an APWR are analysed by single pin-cell calculations as well as two-dimensional whole-reactor computations. The calculational methods of the two codes employed for this study, viz. the cell code SPEKTRA and the diffusion-burnup code DIBU, are presented in detail. The APWR-investigations carried out concentrate on the void coefficient characteristics of tight UO 2 /PuO 2 -lattices, control rod worths, burnup behaviour and spatial power distributions in APWR cores. The principal physics design differences between advanced pressurized water reactors and present-day PWRs are identified and discussed. (orig./HP) [de

  17. 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

  18. Neutron Transmutation Doping of Silicon at Research Reactors

    International Nuclear Information System (INIS)

    2012-05-01

    This publication details the processes and history of neutron transmutation doping of silicon, particularly its commercial pathway, followed by the requirements for a technologically modern and economically viable production scheme and the current trends in the global market for semiconductor products. It should serve as guidelines on the technical requirements, involved processes and required quality standards for the transmission of sound practices and advice for research reactor managers and operators planning commercial scale production of silicon. Furthermore, a detailed and specific database of most of the world's research reactor facilities in this domain is included, featuring their characteristics for irradiation capabilities, associated production capacities and processing.

  19. Neutron activation measurements in research reactor concrete shield

    International Nuclear Information System (INIS)

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

    2001-01-01

    The results of activation measurement inside TRIGA research reactor concrete shielding are given. Samples made of ordinary and barytes concrete together with gold and nickel foils were irradiated in the reactor body. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active longlived radioactive nuclides in ordinary concrete samples were found to be 60 Co and 152 Eu and in barytes concrete samples 60 Co, 152 Eu and 133 Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale.(author)

  20. Uses of reactor neutrons for studying the microcomposition of materials

    International Nuclear Information System (INIS)

    Jervis, R.E.

    1993-01-01

    Reactor neutrons constitute excellents 'probes' for exploring and measuring a wide range both of minor and trace constituents in solids and liquids with high sensitivity because of their transparency in materials. Nondestructive neutron prompt-gamma analysis (PGA) utilizing either cold or thermal neutrons, such as at JRR-3M, is compared and contrasted to the more common (delayed) instrumental neutron activation analysis (INAA) and epithermal NAA. Clearly PGA offers high sensitivity for selected elements: B, H, Cd and REE's in suitable matrices, and is therefore, complementary to INAA which is not as useful for them, or for Ni, Sn, Fe, C or N. Recent INAA applications in our laboratory that demonstrate some of the uniqueness of neutron methods include use of epithermal neutrons for small biological specimens to measure Cd, K, As, Zn and, multielemental INAA for environmental pollution studies. The latter involves large data sets of multielemental concentrations which are subjected to statistical multivariant factor analysis to reveal unknown or unsuspected quantitative relationships among groups of trace constituents. These patterns, or 'factors' are shown to be uniquely related to pollution sources and can be utilized to compute the relative source contributions at a given receptor site. (author)

  1. Neutron radiographic findings in light water reactor fuel

    International Nuclear Information System (INIS)

    Domanus, J.C.

    1979-06-01

    The assessment of neutron radiographs of nuclear fuel elements can be much easier, faster and simpler if reference can be made to typical defects, which can be revealed by neutron radiography. In other fields of industrial radiography collections of reference radiographs, showing typical defects in welding, or casting have been completed and published long ago. Since 1974 neutron radiography is routinely used at Risoe National Laboratory, Denmark, for the quality and performance control of nuclear fuel. About 2000 neutron radiographs were taken, mainly during the post irradiation examination of light water reactor fuel. During assessment of neutron radiographs some typical defects of the fuel were found and it was felt that a classification of such defects will help to speed up the assessment procedure. Therefore an attempt was made to establish such a classification, which is currently used at Risoe now. This classification is presented in this atlas, which contains 36 neutron radiographs reproduced on film (in original size) and on paper (twice enlarged). (author)

  2. Non-equilibrium statistical thermodynamics of neutron gas in reactor

    International Nuclear Information System (INIS)

    Hayasaka, Hideo

    1977-01-01

    The thermodynamic structures of non-equilibrium steady states of highly rarefied neutron gas in various media are considered for the irreversible processes owing to creative and destructive reactions of neutrons with nuclei of these media and supply from the external sources. Under the so-called clean and cold condition in reactor, the medium is regarded virtually as offering the different chemical potential fields for each subsystem of a steady neutron gas system. The fluctuations around a steady state are considered in a Markovian-Gaussian process. The generalized Einstein relations are derived for stationary neutron gas systems. The forces and flows of neutron gases in a medium are defined upon the general stationary solution of the Fokker-Planck equation. There exist the symmetry of the kinetic coefficients, and the minimum entropy production upon neutron-nuclear reactions. The distribution functions in various media are determined by each corresponding extremum condition under the vanishing of changes of the respective total entropies in the Gibbs equation. (auth.)

  3. Measuring neutron flux density in near-vessel space of a commercial WWER-1000 reactor

    International Nuclear Information System (INIS)

    Borodkin, G.I.; Eremin, A.N.; Lomakin, S.S.; Morozov, A.G.

    1987-01-01

    Distribution of neutron flux density in two experimental channels on the reactor vessel external surface and in ionization chamber channel of a commercial WWER-1000 reactor, is measured by the activation detector technique. Azimuthal distributions of fast and thermal neutron fluxes and height distributions of fast neutron flux density within energy range >1.2 and 2.3 MeV are obtained. Conclusion is made, that reactor core state and its structural peculiarities in the measurement range essentially affect space and energy distribution of neutron field near the vessel. It should be taken into account when determining permissible neutron fluence for the reactor vessel

  4. Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stages 14 and 15

    Energy Technology Data Exchange (ETDEWEB)

    Pazsit, Imre; Wihlstrand, Gustav; Tambouratzis, Tatiana; Jonsson, Anders; Dahl, Berit (Chalmers Univ. of Technology, Dept. of Nuclear Engineering, SE-412 96 Goeteborg (Sweden))

    2009-12-15

    This report constitutes Stages 14 and 15 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Stage 14 was a full one-year project, whereas Stage 15 consisted of a half-year project. The program executed in Stages 14 and 15 consists of the following three parts: - Study of criticality, neutron kinetics and neutron noise in molten salt reactors (MSR) (Stages 14 and 15); - An overview and introduction to fuzzy logics (Stage 14), and an application to two-phase flow identification (Stage 15) - Preparations for and execution of an IAEA-ICTP workshop on Neutron fluctuations, reactor noise and their applications in nuclear reactors (Stage 14)

  5. Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stages 14 and 15

    International Nuclear Information System (INIS)

    Pazsit, Imre; Wihlstrand, Gustav; Tambouratzis, Tatiana; Jonsson, Anders; Dahl, Berit

    2009-12-01

    This report constitutes Stages 14 and 15 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Stage 14 was a full one-year project, whereas Stage 15 consisted of a half-year project. The program executed in Stages 14 and 15 consists of the following three parts: - Study of criticality, neutron kinetics and neutron noise in molten salt reactors (MSR) (Stages 14 and 15); - An overview and introduction to fuzzy logics (Stage 14), and an application to two-phase flow identification (Stage 15) - Preparations for and execution of an IAEA-ICTP workshop on Neutron fluctuations, reactor noise and their applications in nuclear reactors (Stage 14)

  6. Fuelling study of CANDU reactors using neutron absorber poisoned fuel

    Energy Technology Data Exchange (ETDEWEB)

    Song, J.J.; Chan, P.K.; Bonin, H.W., E-mail: s25815@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

    2014-07-01

    A comparative fuelling study is conducted to determine the potential gain in operating margin for CANDU reactors incurred by implementing a change to the design of the conventional 37-element natural uranium (NU) fuel. The change involves insertion of minute quantities of neutron absorbers, Gd{sub 2}O{sub 3} and Eu{sub 2}O{sub 3}, into the fuel pellets. The Reactor Fuelling Simulation Program (RFSP) is used to conduct core-following simulations, for the regular 37-element NU fuel, which is to be used as control for comparison. Preliminary results are presented for fuelling with the regular 37-element NU fuel, which indicate constraints on fuelling that may be relaxed with addition of neutron absorbers. (author)

  7. Joint evaluated file qualification for thermal neutron reactors

    International Nuclear Information System (INIS)

    Tellier, H.; Van der Gucht, C.; Vanuxeem, J.

    1986-09-01

    The neutron and nuclear data which are needed by reactor physicists to perform core calculations are brought together in the evaluated files. The files are processed to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data, which is sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. We show, in this paper, how the use of these integral experiments and the application of a tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the joint evaluated file. For this purpose, 56 buckling measurements and 42 isotopic analysis of irradiated fuel were used. Small modifications of the initial data are proposed. The final values are compared with recent recommended values or microscopic data. 8 refs

  8. Joint evaluated file qualification for thermal neutron reactors

    International Nuclear Information System (INIS)

    Tellier, H.; van der Gucht, C.; Vanuxeem, J.

    1986-01-01

    The neutron and nuclear data which are needed by reactor physicists to perform core calculations are brought together in the evaluated files. The files are processes to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data, which is sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. The authors show, in this paper, how the use of these integral experiments and the application of a tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the Joint evaluated file. For this purpose, 56 buckling measurements and 42 isotopic analysis of irradiated fuel were used. Small modifications of the initial data are proposed. The final values are compared with recent recommended values or microscopic data

  9. Neutronics design for a spherical tokamak fusion-transmutation reactor

    International Nuclear Information System (INIS)

    Deng Meigen; Feng Kaiming; Yang Bangchao

    2002-01-01

    Based on studies of the spherical tokamak fusion reactors, a concept of fusion-transmutation reactor is put forward. By using the one-dimension transport and burn-up code BISON3.0 to process optimized design, a set of plasma parameters and blanket configuration suitable for the transmutation of MA (Minor Actinides) nuclear waste is selected. Based on the one-dimension calculation, two-dimension calculation has been carried out by using two-dimension neutronics code TWODANT. Combined with the neutron flux given by TWODANT calculation, burn-up calculation has been processed by using the one-dimension radioactivity calculation code FDKR and some useful and reasonable results are obtained

  10. Microstructural evolution in reactor pressure vessel steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

  11. Problem Oriented Neutron-Gamma Cross Sections Libraries for WWER-440 and WWER-1000 Shielding and Reactor Vessel Dosimetry Application

    International Nuclear Information System (INIS)

    Belousov, S.; Antonov, S.; Ilieva, K.

    1997-01-01

    The 47 neutron and 20 gamma group libraries BGL-440 and BGL-1000 for the shielding and reactor vessel dosimetry application have been generated for WWER-440 and WWER-1000 by collapsing the VITAMIN-B6 library (199 neutron and 42 gamma groups on the base of ENDF/B-6). The first parts of the libraries for neutron-gamma transport calculation, BGL-440-1 (150 nuclides) and BGL-1000-1 (140 nuclides), have been generated by a modified version of SAS1X control module of the SCALE system. The appropriate zone-average neutron flux had been used for these sub-libraries collapsing. The BGL-440-2 and BGL-1000-2 sub-libraries consist of cross sections for all 120 nuclides of VITAMIN-B6, for calculation of the transport through non-reactor materials of dosimeters, capsules, specimens which may be placed in the cavity behind the reactor vessel. The neutron spectrum just beyond the RPV had been used for this collapsing. As the first test the comparative calculations of the neutron flux on/behind the WWER-1000 reactor vessel have been realised using the libraries BGL-1000 and BUGLE, intended for the American PWR reactors. The integral neutron flux values by BGL-1000 and BUGLE differ by 3% onto the vessel, and 5% behind the vessel. This result shows that the calculations of the neutron flux responses for the WWER vessel surveillance, especially in locations behind the WWER vessel have to be done by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries

  12. Neutron Field Characterization of Irradiation Locations Applied to the Slovenian TRIGA Reactor

    International Nuclear Information System (INIS)

    Barbot, Loic; Domergue, Christophe; Breaud, Stephane; Destouches, Christophe; Villard, Jean-Francois; Snoj, Luka; Stancar, Ziga; Radulovic, Vladimir; Trkov, Andrej

    2013-06-01

    This work deals with several neutron flux measurement instruments and particle transport calculations combined in a method to assess the neutron field in experimental locations in nuclear reactor core or reflector. First test of this method in the TRIGA Mark II of Slovenia led to the assessment of three energy groups neutron fluxes in central irradiation locations within reactor core. (authors)

  13. Nuclear fuel assembly for fast neutron reactors

    International Nuclear Information System (INIS)

    Ilyunin, V.G.; Murogov, V.M.; Troyanov, M.F.; Rinejskij, A.A.; Ustinov, G.G.; Shmelev, A.N.

    1982-01-01

    The fuel assembly of a fast reactor consists of fuel elements comprising sections with fissionable and breeding material and tubes with hollows designed for entrapping gaseous fission products. Tubes joining up to the said sections are divided in a middle and a peripheral group such that at least one of the tube groups is placed in the space behind the coolant inlet ports. The configuration above allows reducing internal overpressure in the fuel assembly, thus reducing the volume of necessary structural elements in the core. (J.B.)

  14. The proposed cold neutron irradiation facility at the Breazeale reactor

    International Nuclear Information System (INIS)

    Dimeo, R. M.; Sokol, P. E.; Carpenter, J. M.

    1997-01-01

    We discuss the design considerations of a Cold Neutron Irradiation Facility (CNIF) originally to have been installed at the Penn State Breazeale Reactor (PSBR). The goal of this project was to study the effects of radiation-induced damage to cryogenic moderators and, in particular, solid methane. This work evolved through the design stage undergoing a full safety analysis and received tentative approval from the PSBR Safeguards Committee but was discontinued due to budgetary constraints. (auth)

  15. Neutron disadvantage factors in heavy water and light water reactors

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1966-01-01

    A number od heavy water and light water reactor cells are analyzed in this paper by applying analytical methods of neutron thermalization. Calculations done according to the one-group Amouyal-Benoist method are included in addition. Computer codes for ZUSE Z-23 computer were written by applying both methods. The obtained results of disadvantage factors are then compared to results obtained by one-group P 3 approximation and by multigroup K7-THERMOS code [sr

  16. 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 stuneutron 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.

  17. A methodology of neutronic-thermodynamics simulation for fast reactor

    International Nuclear Information System (INIS)

    Waintraub, M.

    1986-01-01

    Aiming at a general optimization of the project, controlled fuel depletion and management, this paper develop a neutronic thermodynamics simulator, SIRZ, which besides being sufficiently precise, is also economic. That results in a 75% reduction in CPU time, for a startup calculation, when compared with the same calculation at the CITATION code. The simulation system by perturbation calculations, applied to fast reactors, which produce errors smaller than 1% in all components of the reference state given by the CITATION code was tested. (author)

  18. 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

  19. 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

  20. The interpretation of neutron noise in boiling water reactors

    International Nuclear Information System (INIS)

    John, T.M.; Singh, O.P.

    1985-01-01

    Some qualitative results of neutron noise in a boiling water reactor (BWR) are reported. By using one-group theory, it has been shown that the neutron flux fluctuations caused by a distributed source in space, representative of the coolant boiling noise in BWRs, can be considered as made up of two components: The first one, having a global character, is a quickly varying function of frequency and follows the fundamental mode solution in space; the second, called nonglobal (local), follows the spatial variation of noise-source intensity distribution and is independent of frequency for ω γΣ, this component decreases with increasing frequency. The formulation indicates that the global component is quite sensitive to the neutron multiplication factor of the system and, for the local component, the medium behaves like a nonmultiplying one. The global effect is dominant at lower frequencies in a critical system, and the local effect is dominant at higher fre quencies

  1. Consolidation of the neutron spectrum in the RA-6 reactor

    International Nuclear Information System (INIS)

    Bazzana, S.; Chiaraviglio, N.

    2013-01-01

    Unfolding procedures can be used to determine the neutron or gamma spectrum in a multigroup structure from experimental and calculation results. In this way, it is possible to adjust with high reliability magnitudes that cannot be directly measured. For neutron unfolding it is necessary the use of a set of detectors with different energetic response. In this work we describe two unfolding experiences in different positions of the RA-6 reactor of the Bariloche Atomic Centre. One of them consisted in the unfolding in an incore position and the other one in the BNCT facility beam.Experimental techniques and neutron detectors for each experience are described along with the correction factors that must be taken into account for each experience. In both cases there is good agreement between measured and adjusted quantities. (author) [es

  2. Study of the neutron flux distribution in acylindrical reactor

    Directory of Open Access Journals (Sweden)

    A. Vidal-Ferràndiz

    2017-08-01

    Full Text Available In the Energy Engineering Degree of the Universitat Politècnica de València, the students attend to the Nuclear Technology course, in which the basic knowledge of this technology is presented. A main objective of this technology is to obtain neutron population distribution inside a reactor core, in order to maintain the fission reaction chain. As this activity cannot be experimentally developed, mathematical modelling is of great importance to achieve such objective.  One of the computer laboratories proposed consists in the neutron flux determination analytically and numerically in a cylindrical geometry. The analytical solution makes use of the Bessel functions and is a good example of their applications. Alternatively, a numerical solution based on finite differences is used to obtain an approximate solution of the neutron flux. In this work, different discretizations of the cylindrical geometry are implemented and their results are compared.

  3. Introduction to neutron metrology for reactor radiation damage

    International Nuclear Information System (INIS)

    Alberman, A.; Genthon, J.P.; Wright, S.B.; Zijp, W.L.

    1977-01-01

    This document, prepared by members of the Irradiation Damage Subgroup of the Euratom Working Group on Reactor Dosimetry (EWGRD) describes the background of the procedures for determining irradiation parameters which are of interest in radiation damage experiments. The first two chapters outline the concept of damage functions and damge models. The next two chapters give information on methods to determine neutron fluences and neutron spectra. The fifth chapter gives a review of correlation data available for graphite and steels. The last chapter gives guidance how to report the relevant irradiation parameters. Attention is given to the role of the neutron spectrum in deriving values for damage fluence, energy transferred to the lattice, and number of displacements. A suggested list to report data relevant to the irradiation, the instrumentation and the testing of material is included

  4. Need for nuclear data for thermal neutron reactors

    International Nuclear Information System (INIS)

    Bouchard, J.; Golinelli, C.; Tellier, H.

    1983-01-01

    The need for nuclear data for thermal neutron reactors is conditioned by the persisting lack of agreement between the calculation and measurement of certain parameters, by the benefit that can be drawn from reduction of the marginal areas and by envisaged modifications. Three particular fields are delineated. Reduction of the deviation in temperature coefficients by modification of the shape of the effective capture cross sections of uranium-238 and -235 in the thermal range. The increase in precision of kinetic measurements by a better knowledge of data connected to slowed-down neutrons. Improvement in predicting the neutron activity of the fuels used in measuring the effective capture cross sections of plutonium-242 and americium-243. (Auth.)

  5. Theoretical analysis of nuclear reactors (Phase II), I-V, Part V, Determining the fine neutron flux distribution by Pn method; Razrada metoda teorijske analize nuklearnih reaktora (II faza) I-V, V Deo, Odredjivanje fine raspodele fluksa Pn metodom

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-10-15

    Expression for spherical harmonic moments were applied. They were obtained by spherical harmonics expansion of monoenergetic transport equation. This report presents the procedure for calculating the neutron flux distribution in the nine-zone reactor cell of the RA reactor in Vinca. The procedure was modelled for digital computer ZUSE Z-23 by expansion of the diagram of the automated P{sub 3} code, which is adequate for P{sub n} code with minor changes. The needed subroutines were developed. The most important ones were those for modified first and second order Bessel functions of n-th order. Computer Z-23 was operating only 15 hours during three months, and thus only the subroutines for modified Bessel functions could be tested and the obtained results were excellent. For the mentioned reason the neutron flux distribution will be calculated in the forthcoming period. Koriscen je izraz za svernoharmonicne momente koji se dobija razvijanjem monoenergetske transportne jednacine u sverne harmonike. Dat je postupak za odredjivanje raspodele neutronskog fluksa u devet medijalnoj celiji reaktora RA u Vinci. Gornji postupak je logicki organizovan na digitalnoj masini ZUSE Z-23 razvijanjem tekuceg dijagrama automatskog P{sub 3} koda, koji sa malim izmenama odgovara Pn kodu. Razvijene su potrebne osnovne subrutine, medju kojima su najznacajnije one za modificirane Beselove funkcije prve i druge vrste n - tog reda. Masina ZUSE Z-23 bila je u radu svega 15 zasova u toku tri meseca, te sam stigao da testiram samo subrutine za modificirane Beselove funkcije koje su dale odlicne rezultate. Iz ovoga razloga nije se mogla dobiti trazena raspodela fluksa. To ce biti ucinjeno u narednom periodu posto cemo moci koristiti masinu Z-23 (author)

  6. Tests for validation of fast neutron reactors safety

    International Nuclear Information System (INIS)

    Nagata, T.; Yamashita, H.

    2001-01-01

    Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

  7. Startup data for the Stationary Neutron Radiography Reactor (SNRS)

    International Nuclear Information System (INIS)

    Heidel, C.C.; Richards, W.J.

    1990-01-01

    Moisture and corrosion in the F-lll aluminum honeycomb structure is a significant maintenance problem for these 1960 vintage aircraft. Corrosion and moisture are not readily detectable with conventional nondestructive testing techniques such as X-ray. However, by using neutron radiography, high quality images of the corrosion and moisture within the panels can be obtained. The SNRS-1 TRIGA was developed to provide the high intensity neutron source needed to support systems that enable neutron radiography to be performed on these aircraft parts

  8. 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

  9. Operation and maintenance of the RA Reactor in 1985, Part 1, Annex A - Reactor applications

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1985-01-01

    This document describes reactor operation from 1981 to 1985, including data about short term (shorter than 24 hours) and long term operation interruptions, as well as safety shutdown and reactor applications. During 1982, 1983 until July 1984 reactor was operated at 2 MW power according to the plan. Plan was not fulfilled in 1983 because deposits were noticed again, at the end of 1982, on the surface of fuel elements. Reactor was mainly used for neutron activation purposes and isotope production as source of neutrons for experimental purposes [sr

  10. New applications of neutron noise theory in power reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, Vasiliy

    2000-04-01

    The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

  11. New applications of neutron noise theory in power reactor physics

    International Nuclear Information System (INIS)

    Arzhanov, Vasiliy

    2000-04-01

    The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

  12. Multipurpose epithermal neutron beam on new research station at MARIA research reactor in Swierk-Poland

    Energy Technology Data Exchange (ETDEWEB)

    Gryzinski, M.A.; Maciak, M. [National Centre for Nuclear Research, Andrzeja Soltana 7, 05-400 Otwock-Swierk (Poland)

    2015-07-01

    MARIA reactor is an open-pool research reactor what gives the chance to install uranium fission converter on the periphery of the core. It could be installed far enough not to induce reactivity of the core but close enough to produce high flux of fast neutrons. Special design of the converter is now under construction. It is planned to set the research stand based on such uranium converter in the near future: in 2015 MARIA reactor infrastructure should be ready (preparation started in 2013), in 2016 the neutron beam starts and in 2017 opening the stand for material and biological research or for medical training concerning BNCT. Unused for many years, horizontal channel number H2 at MARIA research rector in Poland, is going to be prepared as a part of unique stand. The characteristics of the neutron beam will be significant advantage of the facility. High flux of neutrons at the level of 2x10{sup 9} cm{sup -2}s{sup -1} will be obtainable by uranium neutron converter located 90 cm far from the reactor core fuel elements (still inside reactor core basket between so called core reflectors). Due to reaction of core neutrons with converter U{sub 3}Si{sub 2} material it will produce high flux of fast neutrons. After conversion neutrons will be collimated and moderated in the channel by special set of filters and moderators. At the end of H2 channel i.e. at the entrance to the research room neutron energy will be in the epithermal energy range with neutron intensity at least at the level required for BNCT (2x10{sup 9} cm{sup -2}s{sup -1}). For other purposes density of the neutron flux could be smaller. The possibility to change type and amount of installed filters/moderators which enables getting different properties of the beam (neutron energy spectrum, neutron-gamma ratio and beam profile and shape) is taken into account. H2 channel is located in separate room which is adjacent to two other empty rooms under the preparation for research laboratories (200 m2). It is

  13. Development of Neutron Imaging System for Neutron Tomography at Thai Research Reactor TRR-1/M1

    Science.gov (United States)

    Wonglee, S.; Khaweerat, S.; Channuie, J.; Picha, R.; Liamsuwan, T.; Ratanatongchai, W.

    2017-09-01

    The neutron imaging is a powerful non-destructive technique to investigate the internal structure and provides the information which is different from the conventional X-ray/Gamma radiography. By reconstruction of the obtained 2-dimentional (2D) images from the taken different angle around the specimen, the tomographic image can be obtained and it can provide the information in more detail. The neutron imaging system at Thai Research Reactor TRR-1/M1 of Thailand Institute of Nuclear Technology (Public Organization) has been developed to conduct the neutron tomography since 2014. The primary goal of this work is to serve the investigation of archeological samples, however, this technique can also be applied to various fields, such as investigation of industrial specimen and others. This research paper presents the performance study of a compact neutron camera manufactured by Neutron Optics such as speed and sensitivity. Furthermore, the 3-dimentional (3D) neutron image was successfully reconstructed at the developed neutron imaging system of TRR-1/M1.

  14. Neutron spectrum measurements from a neutron guide tube facility at the ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maayouf, R M.A.; El-Sayed, L A.A.; El-Kady, A S.I. [Reactor and Neutron Physics Dept., NRC, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The present work deals with measurements of the neutron spectrum emitted from a neutron guide tube (NGT) recently installed at one of the ETRR-1 reactor horizontal channels designed to deliver thermal neutrons, free from fast neutrons and gamma ray background, to a fourier reverse-time-of-flight (RTOF) diffractometer. The measurements were performed using a {sup 6} Li glass scintillation detector combined with a multichannel analyzer set at channel width 4 M sec and installed at 3.4 m from a disc Fermi chopper. Also a theoretical model was specially developed for the neutron spectrum calculations. According to the model developed, the spectrum calculated was found to be in good agreement with the measured one. It was found, both from measurements and calculations, that the spectrum emitted from the NGT covers, after transmission through a fourier chopper, neutron wavelengths from 1-4 A adequate for neutron diffraction measurements at D values between 0.71-2.9 A respectively. 6 FIGS.

  15. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cartier, J. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France); Casoli, P. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, Valduc, F-21120 Is sur Tille (France); Chappert, F. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France)

    2013-07-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  16. Neutron mean annihilation time and inverse of the mean annihilation rate in nuclear reactor

    International Nuclear Information System (INIS)

    Hayashi, Masatoshi

    1999-01-01

    There is a dogma in nuclear reactor theory that in a critical reactor the mean annihilation time of neutron is equal to the mean generation time. The author insists that this is a dogma from the basic reexamination of the mean annihilation time of neutron. There are two kinds of neutrons, one participating in chain reactions and the other not participating in chain reactions. The mean annihilation time of neutron is the mean time of the time to annihilation of all neutrons generated in the reactor. The 'prompt neutron life' as a dynamic characteristic parameter proper to nuclear reactor can not be understood as the mean time of neutron to annihilation. The author explains the logic quantitatively with two kinds of nuclear reactors, a bare reactor and an infinite reactor, for which two different mean neutron annihilation times can be defined. Thus, (1) the inverse of the annihilation rate can not simply be considered as the mean annihilation time, (2) the mean annihilation time of a critical reactor is not necessarily equal to the mean generation time, and (3) the prompt neutron life used as a dynamic characteristic parameter of a nuclear reactor can not be understood as the mean time of neutron to annihilation. (M.M.)

  17. Preliminary neutronic design of TRIGA Mark II Reactor

    International Nuclear Information System (INIS)

    Sarikaya, B.; Tombakoglu, M.; Cecen, Y.; Kadiroglu, O. K.

    2001-01-01

    It is very important to analyse the behaviour of the research reactors, since, they play a key role in developing the power reactor technology and radiation applications such as isotope generation for medical treatments. In this study, the neutronic behaviour of the TRIGA MARK II reactor, owned and operated by Istanbul Technical University is analysed by using the SCALE code system. In the analysis, in order to overcome the disadvantages of special TRIGA codes, such as TRIGAP, the SCALE code system is chosen to perform the calculations. TRIGAP and similar codes have limited geometrical (one-dimensional geometry) and cross sectional options (two-group calculations), however, SCALE has the capability of wider range of geometrical modelling capability (three-dimensional modelling is possible) and multi-group calculations are possible

  18. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

    2017-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

  19. Neutron scattering at the high-flux isotope reactor

    International Nuclear Information System (INIS)

    Cable, J.W. Chakoumakos, B.C.; Dai, P.

    1995-01-01

    The title facilities offer the brightest source of neutrons in the national user program. Neutron scattering experiments probe the structure and dynamics of materials in unique and complementary ways as compared to x-ray scattering methods and provide fundamental data on materials of interest to solid state physicists, chemists, biologists, polymer scientists, colloid scientists, mineralogists, and metallurgists. Instrumentation at the High- Flux Isotope Reactor includes triple-axis spectrometers for inelastic scattering experiments, a single-crystal four diffractometer for crystal structural studies, a high-resolution powder diffractometer for nuclear and magnetic structure studies, a wide-angle diffractometer for dynamic powder studies and measurements of diffuse scattering in crystals, a small-angle neutron scattering (SANS) instrument used primarily to study structure-function relationships in polymers and biological macromolecules, a neutron reflectometer for studies of surface and thin-film structures, and residual stress instrumentation for determining macro- and micro-stresses in structural metals and ceramics. Research highlights of these areas will illustrate the current state of neutron science to study the physical properties of materials

  20. Fast-neutron dosimetry in the seed-irradiation facility, ASTRA reactor, Seibersdorf

    International Nuclear Information System (INIS)

    Ahnstroem, G.; Burtscher, A.; Casta, J.

    1967-01-01

    An important part of the co-ordinated programme on the neutron irradiation of seeds has been the construction of a fast-neutron irradiation facility for swimming-pool reactors. This facility was installed around 70 cm from the core in the ASTRA reactor swimming-pool at the end of December, 1966. Also, for this programme a pair of constant potential ionization chambers have been constructed at the Institute of Biochemistry, Stockholm University. These chambes are of the type described in the technical annex and are the same size as the seed-irradiation vials to be used in the seed-irradiation container (diam. =15 mm, length = 60 mm). Some preliminary dosimetry experiments were undertaken to test the irradiation facility and the ionization chambers, and to investigate the usefulness of the dosimetry instructions in the Technical Annex. The results of these experiments are discussed in this paper. 3 refs, 6 figs, 7 tabs

  1. A pulsed fast reactor; Un reacteur pulse a neutrons rapides; Impul'snyj reaktor na bystrykh nejtronakh; Reactor rapido pulsado

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, G. E.; Blokhintsev, D. I.; Blyumkina, Yu. A.; Bondarenko, I. I.; Deryagin, B. N.; Zajmovskij, A. S.; Zinov' ev, V. P.; Kazachkovskij, O. D.; Krasnoyarov, N. V.; Lejpunskij, A. I.; Malykh, V. A.; Nazarov, P. M.; Nikolaev, S. K.; Stavisskij, Yu. Ya.; Ukraintsev, F. I.; Frank, I. M.; Shapiro, F. Ji.; Yazvitskij, Yu. S. [Akademiya Nauk, Moscow, SSSR (Russian Federation)

    1962-03-15

    A pulsed fast reactor (IBR) has been operating at rated capacity since December 1960 in the Joint Institute for Nuclear Research. This reactor is used as a pulsed neutron source for physical experiments carried out by the time-of-flight method. It is used for total cross-section and intermediate neutron capture cross- section measurements, for studying the interaction between slow neutrons and solids and liquids, and for measuring neutron spectra produced in various media. The paper describes the basic structural features of the reactor and the results of the experiments for which it has been used. The reactor's operating system is based on recurrent pulses. Power pulses are produced when the mobile part of the reactor core moves swiftly through the stationary part of the core. The mobile part of the core is fastened to a rotating disc and travels at a speed of 230 m/s. The frequency of power pulses can be altered by means of an auxiliary mobile zone which has a range of 2.3-88 pulses per second. The mean power of the reactor is 1 kW, and the half-width of the power pulse in 36 {mu}s. The reactor is provided with a control and safety system which ensures automatic maintenance of mean power and swift shutdown in the event of any operational irregularity. It is fitted with a system of evacuated-neutron-flight tubes used in time-of-flight experiments. The main tube is 1000 m in length. In the start-up process and during physical experiments carried out on the reactor, the influence on reactivity of displacing the controls and the mobile parts of the core was studied ; the length of the pulse was measured under various operating conditions, and power pulse amplitude fluctuations were studied. Further measurements were made to establish the lifetime of prompt neutrons, the effective fraction of delayed neutrons, and coefficients of reactivity. (author) [French] L'Institut unifie de recherches nucleaires dispose d'un reacteur puise a neutrons rapides (IBR), qui

  2. A design study on hyper-thermal neutron irradiation field for neutron capture therapy at Kyoto University Reactor

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kobayashi, T.

    2000-01-01

    A study about the installation of a hyper-thermal neutron converter to a clinical collimator was performed, as a series of the design study on a hyper-thermal neutron irradiation field at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. From the parametric-surveys by Monte Carlo calculation, it was confirmed that the practical irradiation field of hyper-thermal neutrons would be feasible by the modifications of the clinical collimator and the bismuth-layer structure. (author)

  3. Neutron dose rate in the upper part of a PWR containment. Comparison between measurements and TRIPOLI-2 calculations

    International Nuclear Information System (INIS)

    Vergnaud, T.; Bourdet, L.; Gonnord, J.; Nimal, J.C.; Champion, G.

    1984-01-01

    Conception of a reactor building requires large openings in the primary concrete shield for a postulated loss-of-coolant accident. Through these openings neutrons escape and produce dose rates in several parts of the reactor building. Some calculations using ANISN, DOT and essentially TRIPOLI-2 codes allow to compute the neutron dose rates at several places such as reactor containment operating floor and containment annulus. Some complementary shields are provided and the instrumentations are placed in area where the dose rate is lower. Comparisons are presented between measurements and calculations

  4. Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

    Science.gov (United States)

    Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

    2018-02-01

    In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

  5. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

    1982-12-01

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  6. 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

  7. Neutronic and thermal hydraulic assessment of fast reactor cooling by water of super critical parameters

    International Nuclear Information System (INIS)

    Baranaev, Yu. D.; Glebov, A. P.; Ukraintsev, V. F.; Kolesov, V. V.

    2007-01-01

    Necessity of essential improvement of competitiveness for reactors on light water determines development of new generation power reactors on water of super critical parameters. The main objective of these projects is reaching of high efficiency coefficients while decreasing of investment to NPP and simplification of thermal scheme and high safety level. International programme of IV generation in which super critical reactors present is already started. In the frame of this concept specific Super Critical Fast Reactor with tight lattice of pitch is developing by collaboration of the FEI and IATE. In present article neutronic and thermal hydraulic assessment of fast reactor with plutonium MOX fuel and a core with a double-path of super critical water cooling is presented (SCFR-2X). The scheme of double path of coolant via the core in which the core is divided by radius on central and periphery parts with approximately equal number of fuel assemblies is suggested. Periferia part is cooling while down coming coolant movement. At the down part of core into the mix chamber flows from the periphery assemblies joining and come to the inlet of the central part which is cooling by upcoming flow. Eight zone of different content of MOX fuel are used (4 in down coming and 4 in upcoming) sub zones. Calculation of fuel burn-up and approximate scheme of refueling is evaluated. Calculation results are presented and discussed

  8. CONIFERS: a neutronics code for reactors with channels

    International Nuclear Information System (INIS)

    Davis, R.S.

    1977-04-01

    CONIFERS is a neutronics code for nuclear reactors whose fuel is in channels that are separated from each other by several neutron mean-free-path lengths of moderator. It can treat accurately situations in which the usual homogenized-cell diffusion equation becomes inaccurate, but is more economical than other advanced methods such as response-matrix and source-sink formalisms. CONIFERS uses exact solutions of the neutron diffusion equation within each cell. It allows for the breakdown of this equation near a channel by means of data that almost any cell code can supply. It uses the results of these cell analyses in a reactor equations set that is as readily solvable as the familiar finite-difference equations set. CONIFERS can model almost any configuration of channels and other structures in two or three dimensions. It can use any number of energy groups and any reactivity scales, including scales based on control operations. It is also flexible from a programming point of view, and has convenient input and output provisions. (author)

  9. Neutron dosimetry for radiation damage in fission and fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1979-01-01

    The properties of materials subjected to the intense neutron radiation fields characteristic of fission power reactors or proposed fusion energy devices is a field of extensive current research. These investigations seek important information relevant to the safety and economics of nuclear energy. In high-level radiation environments, neutron metrology is accomplished predominantly with passive techniques which require detailed knowledge about many nuclear reactions. The quality of neutron dosimetry has increased noticeably during the past decade owing to the availability of new data and evaluations for both integral and differential cross sections, better quantitative understanding of radioactive decay processes, improvements in radiation detection technology, and the development of reliable spectrum unfolding procedures. However, there are problems caused by the persistence of serious integral-differential discrepancies for several important reactions. There is a need to further develop the data base for exothermic and low-threshold reactions needed in thermal and fast-fission dosimetry, and for high-threshold reactions needed in fusion-energy dosimetry. The unsatisfied data requirements for fission reactor dosimetry appear to be relatively modest and well defined, while the needs for fusion are extensive and less well defined because of the immature state of fusion technology. These various data requirements are examined with the goal of providing suggestions for continued dosimetry-related nuclear data research

  10. Calculation of the inventory and near-field release rates of radioactivity from neutron-activated metal parts discharged from the high flux isotope reactor and emplaced in solid waste storage area 6 at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kelmers, A.D.; Hightower, J.R.

    1987-05-01

    Emplacement of contaminated reactor components involves disposal in lined and unlined auger holes in soil above the water table. The radionuclide inventory of disposed components was calculated. Information on the composition and weight of the components, as well as reasonable assumptions for the neutron flux fueling use, the time of neutron exposure, and radioactive decay after discharge, were employed in the inventory calculation. Near-field release rates of /sup 152/Eu, /sup 154/Eu, and /sup 155/Eu from control plates and cylinders were calculated for 50 years after emplacement. Release rates of the europium isotopes were uncertain. Two release-rate-limiting models were considered and a range of reasonable values were assumed for the time-to-failure of the auger-hole linear and aluminum cladding and europium solubility in SWSA-6 groundwater. The bounding europium radionuclide near-field release rates peaked at about 1.3 Ci/year total for /sup 152,154,155/Eu in 1987 for the lower bound, and at about 420 Ci/year in 1992 for the upper bound. The near-field release rates of /sup 55/Fe, /sup 59/Ni, /sup 60/Co, and /sup 63/Ni from stainless steel and cobalt alloy components, as well as of /sup 10/Be, /sup 41/Ca, and /sup 55/Fe from beryllium reflectors, were calculated for the next 100 years, assuming bulk waste corrosion was the release-rate-limiting step. Under the most conservative assumptions for the reflectors, the current (1986) total radionuclide release rate was calculated to be about 1.2 x 10/sup -4/ Ci/year, decreasing by 1992 to a steady release of about 1.5 x 10/sup -5/ Ci/year due primarily to /sup 41/Ca. 50 refs., 13 figs., 8 tabs.

  11. Neutronic design of a traveling wave reactor core

    International Nuclear Information System (INIS)

    Lopez S, R. C.; Francois L, J. L.

    2010-10-01

    The traveling wave reactor is an innovative kind of fast breeder reactor, capable of operate for decades without refueling and whose operation requires only a small amount of enriched fuel for the ignition. Also, one of its advantages is its versatility; it can be designed as small modules of about 100 M We or large scale units of 1000 M We. In this paper the behaviour of the traveling wave reactor core is studied in order to determine whether the traveling breeding/burning wave moves (as theoretically predicted) or not. To achieve this, we consider a two pieces cylinder, the first one, the ignition zone, containing highly enriched fuel and the second, the breeding zone, which is the larger, containing natural or depleted uranium or thorium. We consider that both zones are homogeneous mixtures of fuel, sodium as coolant and iron as structural material. We also include a reflector material outside the cylinder to reduce the neutron leakages. Simulations were run with MCNPX version 2.6 code. We observed that the wave does move as time passes as predicted by theory, and reactor remains supercritical in the time we have simulated (3000 days). Also, we found that thorium does not perform as well as uranium for breeding in this type of reactor. Further test with different reflectors are planned for both U-Pu and Th-U fuel cycles. (Author)

  12. Proposal for a new method of reactor neutron flux distribution determination

    Energy Technology Data Exchange (ETDEWEB)

    Popic, V R [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1964-01-15

    A method, based on the measurements of the activity produced in a medium flowing with variable velocity through a reactor, for the determination of the neutron flux distribution inside a reactor is considered theoretically (author)

  13. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  14. The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

    Energy Technology Data Exchange (ETDEWEB)

    Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

    2008-07-15

    This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

  15. Neutronics Study of the KANUTER Space Propulsion Reactor

    International Nuclear Information System (INIS)

    Venneri, Paolo; Nam, Seung Hyun; Kim, Yonghee

    2014-01-01

    The Korea Advanced Nuclear Thermal Engine Rocket (KANUTER) has been developed at the Korea Advanced Institute of Science and Technology (KAIST). This space propulsion system is unique in that it implements a HEU fuel with a thermal spectrum system. This allows the system to be designed with a minimal amount of fissile material and an incredibly small and light system. This then allows the implementation of the system in a cluster format which enables redundancy and easy scalability for different mission requirements. This combination of low fissile content, compact size, and thermalized spectrum contribute to an interesting and novel behavior of the reactor system. The two codes were both used for the burn up calculations in order to verify their validity while the static calculations and characterization of the core were done principally with MCNPX. The KANUTER space propulsion reactor is in the process of being characterized and improved. Its basic neutronic characteristics have been studied, and its behavior over time has been identified. It has been shown that this reactor will have difficulty operating as hoped in a bimodal configuration where it is able to provide both propulsion and power throughout mission to Mars. The reason for this has been identified as Xe 135 , and it is believed that a possible solution to this issue does exist, either in the form of an appropriately designed neutron spectrum or the building in of sufficient excess reactivity

  16. Neutronics Study of the KANUTER Space Propulsion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Nam, Seung Hyun; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    The Korea Advanced Nuclear Thermal Engine Rocket (KANUTER) has been developed at the Korea Advanced Institute of Science and Technology (KAIST). This space propulsion system is unique in that it implements a HEU fuel with a thermal spectrum system. This allows the system to be designed with a minimal amount of fissile material and an incredibly small and light system. This then allows the implementation of the system in a cluster format which enables redundancy and easy scalability for different mission requirements. This combination of low fissile content, compact size, and thermalized spectrum contribute to an interesting and novel behavior of the reactor system. The two codes were both used for the burn up calculations in order to verify their validity while the static calculations and characterization of the core were done principally with MCNPX. The KANUTER space propulsion reactor is in the process of being characterized and improved. Its basic neutronic characteristics have been studied, and its behavior over time has been identified. It has been shown that this reactor will have difficulty operating as hoped in a bimodal configuration where it is able to provide both propulsion and power throughout mission to Mars. The reason for this has been identified as Xe{sup 135}, and it is believed that a possible solution to this issue does exist, either in the form of an appropriately designed neutron spectrum or the building in of sufficient excess reactivity.

  17. Current status of neutron activation analysis in HANARO Research Reactor

    International Nuclear Information System (INIS)

    Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min

    2003-01-01

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10 13 - 1 x 10 14 n/cm 2 ·s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  18. Current status of neutron activation analysis in HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min [Korea Atomic Energy Research Institute, Daejeon (Korea)

    2003-03-01

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10{sup 13} - 1 x 10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  19. 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

  20. Reactor oscillator - I - III, Part I

    International Nuclear Information System (INIS)

    Lolic, B.

    1961-12-01

    Project 'Reactor oscillator' covers the following activities: designing reactor oscillators for reactors RA and RB with detailed engineering drawings; constructing and mounting of the oscillator; designing and constructing the appropriate electronic equipment for the oscillator; measurements at the RA and RB reactors needed for completing the oscillator construction

  1. New Measurements and Calculations to Characterize the Caliban Pulsed Reactor Cavity Neutron Spectrum by the Foil Activation Method

    Energy Technology Data Exchange (ETDEWEB)

    Jacquet, X.; Casoli, P.; Authier, N.; Rousseau, G. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Barsu, C. [Pl. de la fontaine, 25410 Corcelles-Ferrieres (France)

    2011-07-01

    Caliban is a cylindrical metallic core reactor mainly composed of uranium 235. It is operated by the Criticality and Neutron Science Research Laboratory located at the French Atomic Energy Commission research center in Valduc. As with other fast burst reactors, Caliban is used extensively for determining the responses of electronic parts or other objects and materials to neutron-induced displacements. Therefore, Caliban's irradiation characteristics, and especially its central cavity neutron spectrum, have to be very accurately evaluated. The foil activation method has been used in the past by the Criticality and Neutron Science Research Laboratory to evaluate the neutron spectrum of the different facilities it operated, and in particular to characterize the Caliban cavity spectrum. In order to strengthen and to improve our knowledge of the Caliban cavity neutron spectrum and to reduce the uncertainties associated with the available evaluations, new measurements have been performed on the reactor and interpreted by the foil activation method. A sensor set has been selected to sample adequately the studied spectrum. Experimental measured reaction rates have been compared to the results from UMG spectrum unfolding software and to values obtained with the activation code Fispact. Experimental and simulation results are overall in good agreement, although gaps exist for some sensors. UMG software has also been used to rebuild the Caliban cavity neutron spectrum from activation measurements. For this purpose, a default spectrum is needed, and one has been calculated with the Monte-Carlo transport code Tripoli 4 using the benchmarked Caliban description. (authors)

  2. Neutron flux measurement in the thermal column of the Malaysian TRIGA mark II reactor with MCNP verification

    International Nuclear Information System (INIS)

    Abdel Munem, E.; Shukri, A.; Tajuddin, A.A.

    2006-01-01

    A study of the thermal column of the Malaysian TRIGA Mark II reactor, forming part of a feasibility study for BNCT was proposed in 2001. In the current study, pure metals were used to measure the neutron flux at selected points in the thermal column and the neutron flux determined using SAND-II. Monte Carlo simulation of the thermal column was also carried out. The reactor core was homogenized and calculations of the neutron flux through the graphite stringers performed using MCNP5. The results show good agreement between the measured flux and the MCNP calculated flux. An obvious extension from this is that the MCNP neutron flux output can be utilized as an input spectrum for SAND-II for the flux iteration. (author)

  3. Neutron standard cross sections in reactor physics - Need and status

    International Nuclear Information System (INIS)

    Carlson, A.D.

    1990-01-01

    The design and improvement of nuclear reactors require detailed neutronics calculations. These calculations depend on comprehensive libraries of evaluated nuclear cross sections. Most of the cross sections that form the data base for these evaluations have been measured relative to neutron cross-section standards. The use of these standards can often simplify the measurement process by eliminating the need for a direct measurement of the neutron fluence. The standards are not known perfectly, however; thus the accuracy of a cross-section measurement is limited by the uncertainty in the standard cross section relative to which it is measured. Improvements in a standard cause all cross sections measured relative to that standard to be improved. This is the reason for the emphasis on improving the neutron cross-section standards. The continual process of measurement and evaluation has led to improvements in the accuracy and range of applicability of the standards. Though these improvements have been substantial, this process must continue in order to obtain the high-quality standards needed by the user community

  4. KAMCCO, a reactor physics Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.

    1976-06-01

    KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de

  5. Californium Multiplier. Part I. Design for neutron radiography

    International Nuclear Information System (INIS)

    Crosbie, K.L.; Preskitt, C.A.; John, J.; Hastings, J.D.

    1982-01-01

    The Californium Multiplier (CFX) is a subcritical assembly of enriched uranium surrounding a californium-252 neutron source. The function of the CFX is to multiply the neutrons emitted by the source to a number sufficient for neutron radiography. The CFX is designed to provide a collimated beam of thermal neutrons from which the gamma radiation is filtered, and the scattered neutrons are reduced to make it suitable for high resolution radiography. The entire system has inherent safety features, which provide for system and personnel safety, and it operates at moderate cost. In Part I, the CFX and the theory of its operation are described in detail. Part II covers the performance of the Mound Facility CFX

  6. The study on neutron and photon distribution of AP1000 reactor by MCNP code

    International Nuclear Information System (INIS)

    Chen Defeng; Shen Mingqi

    2014-01-01

    The core and reactor structural of AP1000 was modeled by the MCNP calculation program which is based on the Monte Carlo method in this paper, the neutron and photon distribution of AP1000 reactor core was calculated by the conditions of reactor critical. The results show that the AP1000 reactor neutron and photon distribution is in accordance with the critical design of PWR. (authors)

  7. Developments and application of neutron noise diagnostics of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Zylbersztejn, F.

    2013-01-01

    are the key points limiting the capacities of interpretation of noise measurements. The collaboration with the Chalmers team has allowed the improvement of a calculation code solving the neutron noise equations (CORESIM). The work has started with the use of an earlier version of CORESIM code for thermal reactors and the study of the noise induced by the statistical fluctuations of the coolant temperature. That work led to a publication in Annals of Nuclear Energy. I took part in the adaptation of the CORESIM code to the specificities of fast reactors and its application to a working version of a SFR. The modeling of the core flowering phenomenon and the direct application of the code on the CP-ESFR core case were carried out. The reactivity impact specific to the CP-ESFR core was calculated for two models of core deformations. The neutron noise induced by the modeled deformation has been then calculated. The energy, space and frequency dependence of the neutron noise has been analyzed and will contribute to the instrumentation positioning question. It comes out that such phenomena could be monitored by placing several detectors outside of the core along the same axial channel at several heights. It would also be able to identify the noise signature by the axial noise profile. One can note that the relative noise is significantly higher at the top fuel height than in the lower fuel height. This work could be continued by designing a neutron instrumentation dedicated to the core monitoring using the proposed neutron noise technique. (author)

  8. Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Salt Fast Reactor (MSFR)

    International Nuclear Information System (INIS)

    Laureau, A.; Rubiolo, P.R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

    2013-01-01

    Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor (MSFR) are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated. (authors)

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

    International Nuclear Information System (INIS)

    Matausek, M.V.

    1972-04-01

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

  10. Neutron spectrometric evaluations in the Argentine research reactor RA-1

    International Nuclear Information System (INIS)

    Kunst, J.J.; Papadopulos, S.B.; Gregori, B.N.; Cruzate, J.A.

    1998-01-01

    Full text: The determination of the quantities dose equivalent H * (10) and personal dose equivalent Hp(10) in mixed field (n,γ) needs the knowledge of the related spectrum. In order to fulfill this aim spectrometer system has been built based on the combination of polyethylene spheres of different diameters (Bonner Spheres System-BSS) and a He 3 proportional counter detector sensitive to thermal neutrons. The detector is located in the geometrical centre of each of the spheres and has an associated electronics with a charge preamplifier, an amplifier and a multichannel system that allows the outgoing spectrum analysis. In order to determine the neutron spectrum a deconvolution method is applied based on the LOUHI82 code. In this work are shown the spectra and the related values of H * (10) that have been got in five places of the reactor and in the command room with the BSS. (author) [es

  11. 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)

  12. Identification of neutron noise sources in a boiling water reactor

    International Nuclear Information System (INIS)

    Sides, W.H. Jr.; Mathis, M.V.; Smith, C.M.

    1977-01-01

    Measurements were made at units 2 and 3 of the Browns Ferry Nuclear Power Plant in order to characterize the noise signatures of the neutron and process signals and to determine the usefulness of such signatures for anomaly detection in BWR-4s. Previous measurements and theoretical analyses of BWR noise by others were concerned with the determination of steam velocity and void fraction (using the local component of neutron noise) and with the sources of global noise. The work described is under a five-part program to develop a complete and systematic analysis and representation of BWR neutron and process noise through complementary measurements and stochastic model developments. The parts are: (1) recording as many neutron detector and process noise signals as are available in a BWR-4; (2) reducing these data to noise signatures in order to perform an empirical analysis of these signatures, and documenting the relationships between the signals from spatially separated neutron detectors and between neutron and process variables; (3) developing spatially dependent neutronic models coupled with thermal-hydraulic models to aid in interpreting the observed relationships among the measured noise signatures, (4) comparing measured noise signatures with model predictions to obtain additional insight into BWR-4 dynamic behavior and to validate the models; and (5) using these models to predict the sensitivity of noise monitoring for detection, surveillance, and diagnosis of postulated in-core anomalies in BWRs. The paper describes the procedures used to obtain the noise recordings and presents initial empirical analysis and observations pertaining to the noise signatures and the relationships between several noise variables in the 0.01- to 1-Hz range. The mathematical models have not been developed sufficiently to report theoretical results or to compare measured spectra with model predictions at this time

  13. Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

    International Nuclear Information System (INIS)

    Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

    1988-04-01

    As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

  14. Research program in reactor core diagnostics with neutron noise methods: Stage 3. Final report

    International Nuclear Information System (INIS)

    Pazsit, I.; Garis, N.S.; Karlsson, J.; Racz, A.

    1997-09-01

    Stage 3 of the program has been executed 96-04-12. The long term goal is to develop noise methods for identification and localization of perturbations in reactor cores. The main parts of the program consist of modelling the noise source, calculation of the space- and frequency dependent transfer function, calculation of the neutron noise via a convolution of the transfer function of the system and the noise source, i.e. the perturbation, and finally finding an inversion or unfolding procedure to determine noise source parameters from the neutron noise. Most previous work is based on very simple (analytical) reactor models for the calculation of the transfer function as well as analytical unfolding methods. The purpose of this project is to calculate the transfer function in a more realistic model as well as elaborating powerful inversion methods that do not require analytical transfer functions. The work in stage 3 is described under the following headlines: Further investigation of simplified models for the calculation of the neutron noise; Further investigation of methods based on neural networks; Further investigation of methods for detecting the vibrations and impacting of detectors; Application of static codes for determination of the neutron noise using the adiabatic approximation

  15. 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

  16. Research and development of the software for visualizing nuclear reactor and neutronics analysis

    International Nuclear Information System (INIS)

    Okui, Shota; Sekimoto, Hiroshi

    2009-01-01

    It is not easy to image three-dimensional construct of a nuclear reactor with only its two-dimensional figure because it contains a number of structures and its construction is very complicated. Several visualization softwares for the nuclear reactor or some other plant exist, but require high skills and their operation is not simple. In this study, we developed nuclear reactor visualization software, called 'Visual Reactor (VR)', which does not require specific skills. We added the neutronics analysis code to that software. This code executes cell calculation, neutron diffusion calculation and nuclide burnup calculation by itself without any other codes. We tried to treat simple physics model in order to perform these calculation in a short time. Neutronics characteristics, such as neutron flux and power density distribution, are visualized on structure of nuclear reactor. Target operating system is Microsoft Windows XP or Vista. VR is utilized to figure out the structure of nuclear reactor and whole picture of neutronics characteristics. (author)

  17. Description and performance characteristics for the neutron Coincidence Collar for the verification of reactor fuel assemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1981-08-01

    An active neutron interrogation method has been developed for the measurement of 235 U content in fresh fuel assemblies. The neutron Coincidence Collar uses neutron interrogation with an AmLi neutron source and coincidence counting the induced fission reaction neutrons from the 235 U. This manual describes the system components, operation, and performance characteristics. Applications of the Coincidence Collar to PWR and BWR types of reactor fuel assemblies are described

  18. Measurement and analysis of leakage neutron energy spectra around the Kinki University Reactor, UTR-KINKI

    CERN Document Server

    Ogawa, Y; Sagawa, H; Tsujimoto, T

    2002-01-01

    The highly sensitive cylindrical multi-moderator type neutron spectrometer was constructed for measurement of low level environmental neutrons. This neutron spectrometer was applied for the determination of leakage neutron energy spectra around the Kinki University Reactor. The analysis of the leakage neutron energy spectra was performed by MCNP Monte Carlo code. From the obtained results, the agreement between the MCNP predictions and the experimentally determined values is fairly good, which indicates the MCNP model is correctly simulating the UTR-KINKI.

  19. 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

  20. Defects investigation in neutron irradiated reactor steels by positron annihilation

    International Nuclear Information System (INIS)

    Slugen, V.

    2003-01-01

    Positron annihilation spectroscopy (PAS) based on positron lifetime measurements using the Pulsed Low Energy Positron System (PLEPS) was applied to the investigation of defects of irradiated and thermally treated reactor pressure vessel (RPV) steels. PLEPS results showed that the changes in microstructure of the RPV-steel properties caused by neutron irradiation and post-irradiation heat treatment can be well detected. From the lifetime measurements in the near-surface region (20-550 nm) the defect density in Russian types of RPV-steels was calculated using the diffusion trapping model. The post-irradiation heat treatment studies performed on non-irradiated specimens are also presented. (author)

  1. Swelling of structural materials in fast neutron reactors

    International Nuclear Information System (INIS)

    Seran, J.L.

    1983-06-01

    The physical origin of swelling in irradiated materials and the main parameters acting on swelling of SS 316 are examined: temperature, neutron dose, dose rate, chemical composition, strain hardening. Results obtained, in Rapsodie and Phenix reactors, with fuel cans and with the hexagonal tube containing the fuel pins are analyzed and compared with results found in litterature. In conclusion hot swelling of SS 316 is too important at high doses and is will be replaced by austenitic steels stabilized by Ti and ferritic steels or high nickel steels with structural hardening [fr

  2. Review of the neutron capture process in fission reactors

    International Nuclear Information System (INIS)

    Poenitz, W.P.

    1981-07-01

    The importance of the neutron capture process and the status of the more important cross section data are reviewed. The capture in fertile and fissile nuclei is considered. For thermal reactors the thermal to epithermal capture ratio for 238 U and 232 Th remains a problem though some improvements were made with more recent measurements. The capture cross section of 238 U in the fast energy range remains quite uncertain and a long standing discrepancy for the calculated versus experimental central reaction rate ratio C28/F49 persists. Capture in structural materials, fission product nuclei and the higher actinides is also considered

  3. Effect of neutron anisotropic scattering in fast reactor analysis

    International Nuclear Information System (INIS)

    Chiba, Gou

    2004-01-01

    Numerical tests were performed about an effect of a neutron anisotropic scattering on criticality in the Sn transport calculation. The simplest approximation, the consistent P approximation and the extended transport approximation were compared with each other in one-dimensional slab fast reactor models. JAERI fast set which has been used for fast reactor analyses is inadequate to evaluate the effect because it doesn't include the scattering matrices and the self-shielding factors to calculate the group-averaged cross sections weighted by the higher-order moment of angular flux. In the present study, the sub-group method was used to evaluate the group-averaged cross sections. Results showed that the simplest approximation is inadequate and the transport approximation is effective for evaluating the anisotropic scattering. (author)

  4. Analytical electron microscopy of neutron-irradiated reactor alloys

    International Nuclear Information System (INIS)

    Thomas, L.E.

    1982-01-01

    Exposure to the high neutron fluxes and temperatures from 400 to 650 0 C in the core region of a fast breeder reactor profoundly alters the microstructure and properties of structural steels and superalloys. The development of irradiation-induced voids, dislocations and precipitates, as well as segregation of alloying elements on a microscopic scale has been related to macroscopic swelling, creep, hardening and embrittlement which occur during prolonged exposures in reactor. Microanalytical studies using TEM/STEM methods, primarily energy dispersive x-ray (EDX) microanalysis, have greatly aided understanding of alloy behavior under irradiation. The main uses of analytical electron microscopy in studying irradiated alloys have been the identification of irradiation-induced precipitates and determination of the changes in local composition due to irradiation-induced solute segregation

  5. Excitation of neutron flux waves in reactor core transients

    International Nuclear Information System (INIS)

    Carew, J.F.; Neogy, P.

    1983-01-01

    An analysis of the excitation of neutron flux waves in reactor core transients has been performed. A perturbation theory solution has been developed for the time-dependent thermal diffusion equation in which the absorption cross section undergoes a rapid change, as in a PWR rod ejection accident (REA). In this analysis the unperturbed reactor flux states provide the basis for the spatial representation of the flux solution. Using a simplified space-time representation for the cross section change, the temporal integrations have been carried out and analytic expressions for the modal flux amplitudes determined. The first order modal excitation strength is determined by the spatial overlap between the initial and final flux states, and the cross section perturbation. The flux wave amplitudes are found to be largest for rapid transients involving large reactivity perturbations

  6. Conversion of the RB reactor neutrons by highly enriched uranium fuel and lithium deuteride

    International Nuclear Information System (INIS)

    Strugar, P.; Sotic, O.; Ninkovic, M.; Pesic, M.; Altiparmakov, D.

    1981-01-01

    A thermal-to-fast-neutron converter has been constructed at the RB reactor. The material used for the conversion of thermal neutrons is highly enriched uranium fuel of Soviet production applied in Yugoslav heavy water experimental reactors RA and RB. Calculations and preliminary measurements show that the spectrum of converted neutrons only slightly differs from that of fission neutrons. The basic characteristics of converted neutrons can be expressed by the neutron radiation dose of 800 rad (8 Gy) for 1 h of reactor operation at a power level of 1 kW. This dose is approximately 10 times higher than the neutron dose at the same place without converter. At the same time, thermal neutron and gamma radiation doses are negligible. The constructed neutron converter offers wide possibilities for applications in reactor and nuclear physics and similar disciplines, where neutron spectra of high energies are required, as well as in the domain of neutron dosimetry and biological irradiations in homogeneous fields of larger dimensions. The possibility of converting thermal reactor neutrons with energies of about 14 MeV with the aid of lithium deuteride from natural lithium has been considered too. (author)

  7. Programmed elimination of neutronic poisons in nuclear reactors

    International Nuclear Information System (INIS)

    Perriere, G. de la

    1967-11-01

    This work deals with the use of salts of elements having a large neutron capture cross-section, so-called 'soluble poisons' which are dissolved in the moderating water to control the reactivity of heavy-water reactors, and more particularly to compensate the xenon effect in the reactor EL 4. The report describes the controlled elimination of these poisons by fixation on ion-exchange resins. The poisons considered are lithium-6, cadmium and gadolinium in the sulphate form, and boron as boric acid. The thermodynamic and kinetic constants of the ion-exchange reactions were first determined and a study was then made of the fixation of these compounds in beds of small-calibre resins placed in columns. Lithium-6 is the poison which is most easily applicable to compensate the xenon effect in the reactor EL 4. It can be eliminated rapidly and completely from heavy water, and its use does not lead to supplementary problems of protection against the gamma radiation of the reactor circuits. (author) [fr

  8. High Flux Isotope Reactor cold neutron source reference design concept

    International Nuclear Information System (INIS)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R.

    1998-05-01

    In February 1995, Oak Ridge National Laboratory's (ORNL's) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH 2 ) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH 2 cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept

  9. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  10. 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

  11. Neutron energy spectrum flux profile of Ghana's miniature neutron source reactor core

    International Nuclear Information System (INIS)

    Sogbadji, R.B.M.; Abrefah, R.G.; Ampomah-Amoako, E.; Agbemava, S.E.; Nyarko, B.J.B.

    2011-01-01

    Highlights: → The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was studied. → Using 20,484 energy grids, the thermal, slowing down and fast neutron energy regions were studied. - Abstract: The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) x 10 12 n/cm 2 s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) x 10 11 n/cm 2 s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) x 10 11 n/cm 2 s. The peak values of the thermal energy range occurred in the energy range (1.8939-3.7880) x 10 -08 MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) x 10 09 n/cm 2 s at the lower energy end of the slowing down region between 8.2491 x 10 -01 MeV and 8.2680 x 10 -01 MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) x 10 08 n/cm 2 s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.

  12. Neutronic analysis of two-fluid thorium molten salt reactor

    International Nuclear Information System (INIS)

    Frybort, Jan; Vocka, Radim

    2009-01-01

    The aim of this paper is to evaluate features of the two-fluid MSBR through a parametric study and compare its properties to one-fluid MSBR concepts. The starting point of the analysis is the original ORNL 1000 MWe reactor design, although simplified to some extent. We studied the influence of dimensions of distinct reactor parts - fuel and fertile channels radius, plenum height, design etc. - on fundamental reactor properties: breeding ratio and doubling time, reactor inventory, graphite lifetime, and temperature feedback coefficients. The calculations were carried out using MCNP5 code. Based on obtained results we proposed an improved reactor design. Our results show clear advantages of the concept with two separate fluoride salts if compared to the one fluid concept in breading, doubling time, and temperature feedback coefficients. Limitations of the two-fluid concept - particularly the graphite lifetime - are also pointed out. The reactor design can be a subject of further optimizations, namely from the viewpoint of reactor safety. (author)

  13. Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U 235 chain, analytical expressions for the concentrations of U 235 , U 236 and Np 237 as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer

  14. A neutronic assessment of the new Spherical Cermets Fuel concept for the BWR-PB reactor

    International Nuclear Information System (INIS)

    Benchrif, A.; Chetaine, A.; Amsil, H.; Bounakhla, M.

    2010-01-01

    The tri-structural-isotopic (TRISO) fuel directly cooled by boiling light water is used in the boiling water reactor with pebble-bed coated particles (BWR-PB). At the lower coolant temperature, the TRISO fuel particles demonstrate an unacceptable irradiation swelling in the silicon carbide coating layer during a fuel cycle. So, the objectives of this paper, on the one hand is to evaluate some neutronic parameters of a new fuel concept, Spherical Cermets Fuel (SCF), for a BWR-PB reactor. On the other hand, to assess the fact of SCF fuel concept on the fuel assembly lifetime and the burn-up characteristic. All the parameters as well as Infinite Multiplication Factor, Spectrum Index, Instantaneous Conversion Ratio and Neutron Energy Spectrum was calculated then compared for the TRISO and the SCF fuel concept. It can be seen from the assessment of fuel assembly burn-up characteristics that the normalised neutron spectra of all the assembly's parts pointed out a thermal spectrum for the SCF fuel assembly's parts than the TRISO one. The SCF fuel element increase the assembly life time about 6.1 EFPY corresponding 8000 MWd/t. So, the fuel assembly can be operated for a reasonably long period without outside refuelling. The difference in the assembly lifetime might leads to SCF fuel concept adopted, because the geometry and concept of TRISO fuel particles are wholly different to SCF ones. (author)

  15. Results of research and development activities in 1989 of the Institute for Neutron Physics and Reactor Technology

    International Nuclear Information System (INIS)

    1990-03-01

    The Institute for Neutron Physics and Reactor Technology treats research problems of nuclear engineering, mainly those that are related to the development of sodium-cooled fast breeder reactors and fusion reactor technology. The activities are in approximately equal parts of an experimental and theoretical nature. A great part of the research activities is performed in co-operation with other institutes and industrial groups in the framework of projects. For the Fast Breeder Reactor Project the Institute works on reactor physical design and safety problems by the core of large-scale fast breeder reactors. Questions concerning the consequences of accidents in light water reactors upon the environment and the population are treated as part of the Nuclear Safety Project. The Institute contributes to the Reprocessing Project with theoretical investigations on the physics of the fuel cycle and by developing control devices for a reprocessing plant. In the framework of the Fusion Project the Institute is concerned with neutron physical and technological questions of the breeder blanket. (orig.) [de

  16. Application of the neutron noise technique for measurement of reactivity for subcritical reactor RA-4

    International Nuclear Information System (INIS)

    Orso, J; Marenzana, A

    2012-01-01

    Reactor core RA-4 is divided into two parts that come together to start reactor. The reactor with core separate has the largest subcritical condition, this condition is more secure and therefore the reactor shutdown. In this paper measurements are made of the decay constant of the neutron prompt ' P ', using the α-Rossi and α-Feynman methods to calculate the reactivity of the reactor core for different positions. Both techniques are compared and reactivity is obtained for several position of the reactor core using the α-Rossi technical which is obtained a function that gives the reactivity depending on the separation of the core length. Both techniques are verified using a no multiplicative system. Reactivity values for different position of the core obtained by α-Rossi technique are: $[0 cm] = (-11+/-1) dollar, $[3 cm] = (-7+/-1) dollar, $[3.5 cm] (-5.5+/-0.8) dollar, $[4.2 cm] = (-3.8+/-0.3) dollar y $[4.5] = (-3.0+/-0.1) dollar (author)

  17. 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)

  18. Measurement of fast neutron spectra inside reactors with a Li{sup 6} semiconductor counter spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Ajdacic, V S; Lalovic, B I; Petrovic, B P [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1963-12-15

    The possibility of using the Li{sup 6} semiconductor counter spectrometer for measuring fast neutron spectra inside reactors has been investigated in details and some solutions of the difficulties associated with the high interference of thermal neutrons in well-moderated reactors are suggested and checked experimentally (author)

  19. Flux distribution by neutrons semi-conductors detectors during the startup of the EL4 reactor

    International Nuclear Information System (INIS)

    Fuster, S.; Tarabella, A.

    1967-01-01

    The Cea developed neutron semi-conductors detectors which allows a quasi-instantaneous monitoring of neutrons flux distribution, when placed in a reactor during the tests. These detectors have been experimented in the EL4 reactor. The experiment and the results are presented and compared with reference mappings. (A.L.B.)

  20. The neutron small-angle camera D11 at the high-flux reactor, Grenoble

    International Nuclear Information System (INIS)

    Ibel, K.

    1976-01-01

    The neutron small-angle scattering system at the high-flux reactor in Grenoble consists of three major parts: the supply of cold neutrons via bent neutron guides; the small-angle camera D11; and the data handling facilities. The camera D11 has an overall length of 80 m. The effective length of the camera is variable. The full length of the collimator before the fixed sample position can be reduced by movable neutron guides; the second flight path of 40 m full length contains detector sites in various positions. Thus a large range of momentum transfers can be used with the same relative resolution. Scattering angles between 5 x 10 -4 and 0.5 rad and neutron wavelengths from 0.2 to 2.0 nm are available. A large-area position-sensitive detector is used which allows simultaneous recording of intensities scattered at different angles; it is a multiwire proportional chamber. 3808 elements of 1 cm 2 are arranged in a two-dimensional matrix. (Auth.)

  1. Results of neutron dose measurements at the Rossendorf research reactors taking the actual neutron spectra into account

    International Nuclear Information System (INIS)

    Rimpler, A.; Kneschke, H.

    1985-01-01

    Based on a systematic evaluation of area dose studies at the beginning of the seventies, no individual routine neutron monitoring has been performed at the Rossendorf research reactors. To check this decision, a limited number of persons has been monitored with solid-state nuclear track detectors for several years. The dosemeters were calibrated on the basis of neutron spectra determined at the working places by means of the Bonner sphere method. Intermediate neutrons with a 1/E/sup α/ Fermi distribution were dominating. The fraction of fast neutrons was practically negligible. The obtained spectra, radiation, field quantities and results of individual dose measurements are presented. The dosemeter most appropriate for such neutron fields would be a 12-inch Bonner sphere rem counter. As the mean annual neutron exposure of research workers at the reactor amounted to only 2% of the maximum permissible dose, individual routine monitoring will, also in the future, not be neccessary. (author)

  2. Neutron radiography in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Pugliesi, R.; Moraes, A.P.V. de; Yamazaki, I.M.; Freitas Acosta, C. de.

    1988-08-01

    Neutronradiography of several materials have been obtained at the IEA-R1 Nuclear Research Reactor (IPEN-CNEN/SP), by means of two conversion techniques: a) (n, α) at the beam-hole n 0 3 where a collimated thermal neutron beam, exposure area 4 cm x 8cm and flux at the sample 10 5 n/s cm 2 is obtained. The film used was the CN-85 cellulose nitrate coated with lithium tetraborate (conversor). The time irradiation of the film was 15 minutes and in following was eteched during 30 minutes in a NaOH(10%) aqueous solution at a constant temperature of 60 0 C.; b) (n,γ) by using an experimental arrangement installed in the botton of the pool of the reactor. The flux of the collimated neutron beam is 10 5 n/s/cm 2 at the sample and the conversion is made by means of a dysprozium sheet. The film used was Kodak T-5. The irradiation and the transfering time was 2 hours and 20 hours respectively. (author) [pt

  3. Neutronic assessment of strontium-90 transmutation in fusion reactors

    International Nuclear Information System (INIS)

    Parish, T.A.

    1979-01-01

    Transmutation of 90 Sr using fusion neutrons has been suggested as a possible technique for disposing of this waste nuclide. For transmutation to be attractive, high transmutation rates relative to natural decay are required. Effective half-lives for 90 Sr were computed for fusion reactor blankets constructed of various materials. To obtain satisfactory transmutation rates, fusion reactors with high first wall neutron currents and with highly moderating blankets were found to be necessary. An effective half-life for 90 Sr of 90 Sr inventory and the number of burners required for various fission usage scenarios. Efficient and fast chemical separations were needed to reap the benefits of a short effective half-life. For the fusion burners considered, it was found that the 90 Sr inventory could not be reduced to less than one-fourth of the inventory without transmutation if fission usage continued at a constant rate. Such a reduction is not sufficient to justify the transmutation disposal of 90 Sr

  4. Reactor neutron activation analysis on reference materials from intercomparison runs

    International Nuclear Information System (INIS)

    Pantelica, A.; Salagean, M.

    2003-01-01

    A review of using the Instrumental Neutron Activation Analysis (INAA) technique in our laboratory to determine major, minor and trace elements in mineral and biological samples from international intercomparison runs organised by IAEA Vienna, IAEA-MEL Monaco, 'pb-anal' Kosice, INCT Warszawa and IPNT Krakow is presented. Neutron irradiation was carried out at WWR-S reactor in Bucharest (short and long irradiation) during 1982-1997 and at TRIGA reactor in Pitesti (long irradiation) during the later period. The following type of materials were analysed: soils, marine sediments, uranium phosphate ore, water sludge, copper flue dust, whey powder, yeast, cereal flour (rye and wheat), marine animal tissue (mussel, garfish and tuna fish), as well as vegetal tissue (seaweed, cabbage, spinach, alfalfa, algae, tea leaves and herbs). The following elements could be, in general, determined: Ag, As, Au, Ba, Br, Ca, Ce, Co, Cr, Cs, Eu, Fe, Hf, Hg, K, La, Lu, Mo, Na, Nd, Ni, Rb, Sb, Sc, Se, Sm, Sr, Ta, Tb, Th, U, W, Yb and Zn of long-lived radionuclides, as well as Al, Ca, Cl, Cu, Mg, Mn, and Ti of short-lived radionuclides. Data obtained in our laboratory for various matrix samples presented and compared with the intercomparison certified values. The intercomparison exercises offer to the participating laboratories the opportunity to test the accuracy of their analytical methods as well as to acquire valuable Reference Materials/ standards for future analytical applications. (authors)

  5. Characterization of a Neutron Beam Following Reconfiguration of the Neutron Radiography Reactor (NRAD Core and Addition of New Fuel Elements

    Directory of Open Access Journals (Sweden)

    Aaron E. Craft

    2016-02-01

    Full Text Available The neutron radiography reactor (NRAD is a 250 kW Mark-II Training, Research, Isotopes, General Atomics (TRIGA reactor at Idaho National Laboratory, Idaho Falls, ID, USA. The East Radiography Station (ERS is one of two neutron beams at the NRAD used for neutron radiography, which sits beneath a large hot cell and is primarily used for neutron radiography of highly radioactive objects. Additional fuel elements were added to the NRAD core in 2013 to increase the excess reactivity of the reactor, and may have changed some characteristics of the neutron beamline. This report discusses characterization of the neutron beamline following the addition of fuel to the NRAD. This work includes determination of the facility category according to the American Society for Testing and Materials (ASTM standards, and also uses an array of gold foils to determine the neutron beam flux and evaluate the neutron beam profile. The NRAD ERS neutron beam is a Category I neutron radiography facility, the highest possible quality level according to the ASTM. Gold foil activation experiments show that the average neutron flux with length-to-diameter ratio (L/D = 125 is 5.96 × 106 n/cm2/s with a 2σ standard error of 2.90 × 105 n/cm2/s. The neutron beam profile can be considered flat for qualitative neutron radiographic evaluation purposes. However, the neutron beam profile should be taken into account for quantitative evaluation.

  6. Safety analysis of RA reactor operation, I-II, Part I - RA reactor technical and operation characteristics; Analiza sigurnosti rada reaktora RA - I-III, I deo - Tehnicke i pogonske karakteristike reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    RA research reactor is a thermal, heavy water moderated system with graphite reflector having nominal power 6.5 MW. The 2% enriched metal uranium fuel in the reactor core produces mean thermal neutron flux of 2.9 10{sup 13} neutrons/cm{sup 2} s, and maximum neutron flux 5.5 10{sup 13} neutrons/cm{sup 2} s. main components of the reactor described in this report are: rector core, reflector, biological shield, heavy water cooling system, ordinary water cooling system, helium system, reactor control system, reactor safety system, dosimetry system, power supply system, and fuel transport system. Detailed reactor properties and engineering drawings of all the system are part of this volume.

  7. Direct measurement of the cross section of neutron-neutron scattering at the YAGUAR reactor. Substantiation of the experiment technique

    International Nuclear Information System (INIS)

    Chernukhin, Yu.G.; Kandiev, Ya.Z.; Lartsev, V.D.; Levakov, B.G.; Modestov, D.G.; Simonenko, V.A.; Streltsov, S.I.; Khmel'nitskij, D.V.

    2006-01-01

    The main stage of experiment for direct measurement of cross section of neutron-neutron scattering σ nn at low energies (E nn determination. It was shown, that for achieving the criterion ε ∼ 4% it will be necessary to have 40-50 pulses of a reactor [ru

  8. 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.

  9. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

    International Nuclear Information System (INIS)

    Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

    2010-01-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks. (authors)

  10. Characteristics of neutron irradiation facility and dose estimation method for neutron capture therapy at Kyoto University research reactor institute

    International Nuclear Information System (INIS)

    Kobayashi, T.; Sakurai, Y.; Kanda, K.

    2001-01-01

    The neutron irradiation characteristics of the Heavy Water Neutron Irradiation Facility (HWNIF) at the Kyoto University Research Reactor Institute (KIJRRI) for boron neutron capture therapy (BNCT), is described. The present method of dose measurement and its evaluation at the KURRI, is explained. Especially, the special feature and noticeable matters were expounded for the BNCT with craniotomy, which has been applied at present only in Japan. (author)

  11. Surveillance of a nuclear reactor by pattern recognition analysis of the neutronic noise. Experience on Phenix LMFBR

    International Nuclear Information System (INIS)

    Le Guillou, G.; Bernardin, B.

    1980-05-01

    The purpose of this paper is to present some results of pattern recognition methods applied to the problem of supervising the behaviour of a nuclear reactor, especially the sodium cooled fast breeder Phenix. The problem to solve can be divided into two parts: first, from a set of data about the behaviour of the reactor, we have to find consistent classes of functionning. These classes must be interpreted by physical considerations; then, we have to elaborate a simple classification algorithm, which can be used in-line, to improve safety systems of fast breeder reactors, by giving helpful information for decision making. The neutronic noise has been chosen for our study

  12. On Line Neutron Flux Mapping in Fuel Coolant Channels of a Research Reactor

    International Nuclear Information System (INIS)

    Barbot, Loic; Domergue, Christophe; Villard, Jean-Francois; Destouches, Christophe; Braoudakis, George; Wassink, David; Sinclair, Bradley; Osborn, John-C.; Wu, Huayou; Blandin, C.; Thevenin, Mathieu; Corre, Gwenole; Normand, Stephane

    2013-06-01

    This work deals with the on-line neutron flux mapping of the OPAL research reactor. A specific irradiation device has been set up to investigate fuel coolant channels using subminiature fission chambers to get thermal neutron flux profiles. Experimental results are compared to first neutronic calculations and show good agreement (C/E ∼0.97). (authors)

  13. Consideration of LH2 and LD2 cold neutron sources in heavy water reactor reflector

    International Nuclear Information System (INIS)

    Potapov, I.A.; Serebrov, A.P.

    2001-01-01

    The reactor power, the required CNS dimensions and power of the cryogenic equipment define the CNS type with maximized cold neutron production. Cold neutron fluxes from liquid hydrogen (LH 2 ) and liquid deuterium (LD 2 ) cold neutron sources (CNS) are analyzed. Different CNS volumes, presents and absence of reentrant holes inside the CNS, different adjustment of beam tube and containment are considered. (orig.)

  14. Application of the fractional neutron point kinetic equation: Start-up of a nuclear reactor

    International Nuclear Information System (INIS)

    Polo-Labarrios, M.-A.; Espinosa-Paredes, G.

    2012-01-01

    Highlights: ► Neutron density behavior at reactor start up with fractional neutron point kinetics. ► There is a relaxation time associated with a rapid variation in the neutron flux. ► Physical interpretation of the fractional order is related with non-Fickian effects. ► Effect of the anomalous diffusion coefficient and the relaxation time is analyzed. ► Neutron density is related with speed and duration of the control rods lifting. - Abstract: In this paper we present the behavior of the variation of neutron density when the nuclear reactor power is increased using the fractional neutron point kinetic (FNPK) equation with a single-group of delayed neutron precursor. It is considered that there is a relaxation time associated with a rapid variation in the neutron flux and its physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. We analyzed the case of increase the nuclear reactor power when reactor is cold start-up which is a process of inserting reactivity by lifting control rods discontinuously. The results show that for short time scales of the start-up the neutronic density behavior with FNPK shows sub-diffusive effects whose absorption are government by control rods velocity. For large times scale, the results shows that the classical equation of the neutron point kinetics over predicted the neutron density regarding to FNPK.

  15. 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).

  16. Importance of the (n,gamma) Cm-247 Evaluation on Neutron Emission in Fast Reactor Fuel Cycle Analysis

    International Nuclear Information System (INIS)

    Benoit Forget; Mehdi Asgari; Rodolfo M. Ferrer

    2007-01-01

    As part of the GNEP program, it is envisioned to build a fast reactor for the transmutation of minor actinides. The spent nuclear fuel from the current fleet of light water reactors would be recycled, the current baseline is the UREX+1a process, and would act as a feed for the fast reactor. As the fuel is irradiated in a fast reactor a certain quantity of minor actinides would thus build up in the fuel stream creating possible concerns with the neutron emission of these minor actinides for fuel transportation, handling and fabrication. Past neutronic analyses had not tracked minor actinides above Cm-246 in the transmutation chain, because of the small influence on the overall reactor performance and cycle parameters. However, when trying to quantify the neutron emission from the recycled fuel with high minor actinide content, these higher isotopes play an essential role and should be included in the analysis. In this paper, the influence of tracking these minor actinides on the calculated neutron emission is presented. Also presented is the particular influence of choosing a different evaluated cross section data set to represent the minor actinides above Cm-246. The first representation uses the cross-sections provided by MC2-2 for all isotopes, while the second representation uses infinitely diluted ENDF/BVII.0 cross-sections for Cm-247 to Cf-252 and MC2-2 for all other isotopes

  17. Cold neutron PGAA facility developments at university research reactors in the USA

    International Nuclear Information System (INIS)

    Uenlue, K.; Rios-Martinez, C.

    2005-01-01

    The PGAA applications can be enhanced by using subthermal neutrons, cold neutrons at university research reactors. Only two cold neutron beam facilities were developed at the U.S. university research reactors, namely at Cornell University and the University of Texas at Austin. Both facilities used mesitylene moderator. The mesitylene moderator in the Cornell Cold Neutron Beam Facility (CNBF) was cooled by a helium cryorefrigerator via copper cold fingers to maintain the moderator below 30 K at full power reactor operation. Texas Cold Neutron Source (TCNS) also uses mesitylene moderator that is cooled by a cryorefrigerator via a neon thermosiphon. The operation of the TCNS is based on a helium cryorefrigerator, which liquefies neon gas in a 3-m long thermosiphon. The thermosiphon cools and maintains mesitylene moderator at about 30 K in a chamber. Neutrons streaming through the mesitylene chamber are moderated and thus reduce their energy to produce a cold neutron distribution. (author)

  18. A two-dimensional simulator of the neutronic behaviour of low power fast reactors

    International Nuclear Information System (INIS)

    Penha, M.A.V.R. da.

    1984-01-01

    A model to simulate the temporal neutronic behaviour of fast breeder reactors was developed. The effective cross-sections are corrected, whenever the reactor state change; by using linear correlations and interpolation schemes with data contained in a library previously compiled. This methodology was coupled with a simplified spatial neutronic calculation to investigate the temporal behaviour of neutronic parameters such as breeding gain, flux and power. (Author) [pt

  19. Diagnosis of mucoviscidosis by neutron activation analysis. Part 1

    International Nuclear Information System (INIS)

    Bellido, Luis F.; Bellido, Alfredo V.

    1997-02-01

    Symptoms pathology, incidence, and gravity of the inherent syndrome called mucoviscidosis, or cystic fibrosis are described in this Part I. The analytical methods used for its diagnosis, both the conventional chemical ones and by neutron activation analysis are also summarised. Finally, an analytical method to study the incidence of mucoviscidosis in Brazil is presented. This , essentially, consists in bromine determination, in fingernails, by resonance neutron activation analysis. (author)

  20. Development of SiC Neutron Detector Assembly to Measure the Neutron Flux of the Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Park, Se Hwan; Park, June Sic; Shin, Hee Sung; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yong Kyun [Hanyang University, Seoul (Korea, Republic of)

    2012-05-15

    At present, the conventional detector to measure the neutron at harsh environment is a Self Powered Neutron Detector (SPND). Rhodium(Rh)-103 is in the SPND. When neutron is incident on the Rhodium, the neutron capture reaction occurs, and the Rh-103 is converted to Rh-104. The Rh-104 is decayed to Pd-104 by {beta}-decay, and electrons are generated as the decay products. Because of the half life of Rh-104, approximately 5 minutes are required for the SPND output to reach the equilibrium condition. Therefore the on-line monitoring of the nuclear reactor state is limited if the neutron flux in the reactor core is monitored with the SPND. Silicon carbide (SiC) has the possibility to be developed as neutron detector at harsh environment, because the SiC can be operative at high temperature and high neutron flux conditions. Previously, the basic operation properties of the SiC detector were studied. Also, the radiation response of the SiC detector was studied at high neutron and gamma dose rate. The measurement results for an ex-core neutron flux monitor or a neutron flux monitor of the spent fuel were published. The SiC detector was also developed as neutron detector to measure the fissile material with active interrogation method. However, the studies about the development of SiC detector are still limited. In the present work, the radiation damage effect of the SiC detector was studied. The detector structure was determined based on the study, and a neutron detector assembly was made with the SiC detectors. The neutron and gamma-ray response of the detector assembly is presented in this paper. The detector assembly was positioned in the HANARO research reactor core, the performance test was done. The preliminary results are also included in this paper

  1. Effects of moderation level on core reactivity and. neutron fluxes in natural uranium fueled and heavy water moderated reactors

    International Nuclear Information System (INIS)

    Khan, M.J.; Aslam; Ahmad, N.; Ahmed, R.; Ahmad, S.I.

    2005-01-01

    The neutron moderation level in a nuclear reactor has a strong influence on core multiplication, reactivity control, fuel burnup, neutron fluxes etc. In the study presented in this article, the effects of neutron moderation level on core reactivity and neutron fluxes in a typical heavy water moderated nuclear research reactor is explored and the results are discussed. (author)

  2. Measurement of Relative Biological Effectiveness (RBE) for the Radiation Beam from Neutron Source Reactor YAYOI -Comparisons with Cyclotron Neutron and 60Co Gamma Ray-

    OpenAIRE

    HIROAKI, WAKABAYASHI; SHOZO, SUZUKI; AKIRA, ITO; Nuclear Engineering Research Laboratory, Faculty of Engineering, the University of Tokyo; Institute of Medical Science, the University of Tokyo; Institute of Medical Science, the University of Tokyo

    1983-01-01

    Radiation biology and/or therapy research and development for a research reactor beam need specific RBEs of neutrons as well as of specific reactions. RBEs for reactor beams measured in situ condition are interesting because actual radiation effects on each biological system are different depending on detailed conditions of irradiation. A small powered research reactor (Fast Neutron Source Reactor: YAYOI) was examined here as a neutron beam source for obtaining survival curves in a manner usu...

  3. Fabrication development for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Pace, B.W.; Copeland, G.L.

    1995-08-01

    This report presents the fuel fabrication development for the Advanced Neutron Source (ANS) reactor. The fuel element is similar to that successfully fabricated and used in the High Flux Isotope Reactor (HFIR) for many years, but there are two significant differences that require some development. The fuel compound is U 3 Si 2 rather than U 3 O 8 , and the fuel is graded in the axial as well as the radial direction. Both of these changes can be accomplished with a straightforward extension of the HFIR technology. The ANS also requires some improvements in inspection technology and somewhat more stringent acceptance criteria. Early indications were that the fuel fabrication and inspection technology would produce a reactor core meeting the requirements of the ANS for the low volume fraction loadings needed for the highly enriched uranium design (up to 1.7 Mg U/m 3 ). Near the end of the development work, higher volume fractions were fabricated that would be required for a lower- enrichment uranium core. Again, results look encouraging for loadings up to ∼3.5 Mg U/m 3 ; however, much less evaluation was done for the higher loadings

  4. Neutron activation analysis of essential elements in Multani mitti clay using miniature neutron source reactor

    International Nuclear Information System (INIS)

    Waheed, S.; Rahman, S.; Faiz, Y.; Siddique, N.

    2012-01-01

    Multani mitti clay was studied for 19 essential and other elements. Four different radio-assay schemes were adopted for instrumental neutron activation analysis (INAA) using miniature neutron source reactor. The estimated weekly intakes of Cr and Fe are high for men, women, pregnant and lactating women and children while intake of Co is higher in adult categories and Mn by pregnant women. Comparison of MM clay with other type of clays shows that it is a good source of essential elements. - Highlights: ► Multani mitti clay has been studied for 19 essential elements for human adequacy and safety using INAA and AAS. ► Weekly intakes for different consumer categories have been calculated and compared with DRIs. ► Comparison of MM with other type of clays depict that MM clay is a good source of essential elements.

  5. Neutron and gamma ray streaming experiments at the fast neutron source reactor 'YAYOI'

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Yanagisawa, Ichiro; Akiyama, Masatsugu; An, Shigehiro

    1979-07-01

    Neutron and gamma ray streaming experiments were performed in the ducts and cavities that were located in the heavy concrete shields of the fast neutron source reactor YAYOI of University of Tokyo. The configurations have the feature that the streaming through the ducts are occurred following the scattering in the cavity. The axes of the ducts are perpendicular to the source radiation from the core. The spectrum of the source was modified by putting a plug in the beam hole of the core. An aluminum plug and the plug which contains paraffin were used. The decay in the ducts, however, hardly depends on the source spectrum. The decay in the ducts is nearly exponential. (author)

  6. Formulary for neutron propagation in sodium-steel media for the fast reactor shields

    International Nuclear Information System (INIS)

    Bouteau, F.; Caumette, P.; Khairallah, A.; Oceraies, Y.; Devillers, C.

    1975-01-01

    The simplified calculational tool (''formulary'') for neutron propagation in the shields of fast reactors, being developed at CEA, has two objectives: to reduce the cost of the major part of design calculations, without a significant loss of accuracy; to facilitate the adjustment of the calculational tool with the results of the program of integral propagation experiments, which is conducted in parallel with the development of the calculational method. The version 0 (i.e. before any adjustment) of the formulary and a first test of its validity as compared to the results of integral measurements are presented [fr

  7. Amino acids analysis by total neutron cross-sections determinations: part V

    International Nuclear Information System (INIS)

    Voi, Dante L.; Ferreira, Francisco de O.; Rocha, Helio F. da

    2013-01-01

    Total neutron cross-sections of twenty essential and non-essential amino acids to human were determined using crystal spectrometer installed on the Argonauta reactor of IEN (Instituto de Engenharia Nuclear (CNEN-RJ) and compared with data generated by parceling and grouping methodologies developed at this institution. For each amino acid was calculated the respective neutron cross-section by molecular structure, conformation and chemistry analysis. The results obtained for eighteen of twenty amino acids confirm the specifications and product formulations indicated by manufactures. These initial results allow to build a neutron cross-sections database as part of quality control of the amino supplied to hospitals for production of nutriments for parenteral or enteral formulations used in critical patients dependent on artificial feed, and for application in future studies of structure and dynamics for more complex molecules, including proteins, enzymes, fatty acids, membranes, organelles and other cell components. (author)

  8. Epithermal neutron flux characterization of the TRIGA Mark III reactor, Salazar, Mexico, for use in Internal Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Herrera Peraza, E.

    1996-01-01

    The non ideality of the epithermal neutron flux distribution at a reactor site parameter (made, using Chloramine-T method. Radiochemical purity and stability of the labelled product were determined by radiochromatography. The labelled Melagenine-II showed two radioactive fractions thermal-to-epithermal neutron ratio (f) were determined in the 3 typical irradiations positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using the Cd-ratio for multi monitor and bare bi-isotopic monitor methods respectively. This characterization is of use in the K o - method of neutron activation analysis, recently introduced at the Institute

  9. Highlighted requests of neutron nuclear data measurements for fission reactors

    International Nuclear Information System (INIS)

    1977-01-01

    Highlighted requests were selected from lists of requests for neutron nuclear Data measurements. This work was made by WRENDA Working Group of Japanese Nuclear Data Committee in response to an action of the NEANDC. This will be submitted to a NEA meeting on request lists. Compiled requests in this note correspond finally to the parts of the WRENDA 76/77 with some modifications and additional comments. (auth.)

  10. Characterization of reactor neutron environments at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Kelly, J.G.; Luera, T.F.; Griffin, P.J.; Vehar, D.W.

    1994-01-01

    To assure quality in the testing of electronic parts in neutron radiation environments, Sandia National Laboratories (SNL) has incorporated modern techniques and procedures, developed in the last two decades by the radiation effects community, into all of its experimental programs. Attention to the application of all of these methodologies, experiment designs, nuclear data, procedures and controls to the SNL radiation services has led to the much more accurate and reliable environment characterizations required to correlate the effects observed with the radiation delivered

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

    International Nuclear Information System (INIS)

    1966-01-01

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

  12. Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields

    International Nuclear Information System (INIS)

    Fabry, A.; McElroy, W.N.; Kellogg, L.S.; Lippincott, E.P.; Grundl, J.A.; Gilliam, D.M.; Hansen, G.E.

    1976-01-01

    This paper is intended to review and critically discuss microscopic integral cross section measurement and calculation data for fundamental reactor dosimetry benchmark neutron fields. Specifically the review covers the following fundamental benchmarks: the spontaneous californium-252 fission neutron spectrum standard field; the thermal-neutron induced uranium-235 fission neutron spectrum standard field; the (secondary) intermediate-energy standard neutron field at the center of the Mol-ΣΣ, NISUS, and ITN-ΣΣ facilities; the reference neutron field at the center of the Coupled Fast Reactor Measurement Facility; the reference neutron field at the center of the 10% enriched uranium metal, cylindrical, fast critical; the (primary) Intermediate-Energy Standard Neutron Field

  13. Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields

    International Nuclear Information System (INIS)

    Fabry, A.; McElroy, W.N.; Kellogg, L.S.; Lippincott, E.P.; Grundl, J.A.; Gilliam, D.M.; Hansen, G.E.

    1976-10-01

    The paper is intended to review and critically discuss microscopic integral cross section measurement and calculation data for fundamental reactor dosimetry benchmark neutron fields. Specifically the review covers the following fundamental benchmarks: (1) the spontaneous californium-252 fission neutron spectrum standard field; (2) the thermal-neutron induced uranium-235 fission neutron spectrum standard field; (3) the (secondary) intermediate-energy standard neutron field at the center of the Mol-ΣΣ, NISUS, and ITN--ΣΣ facilities; (4) the reference neutron field at the center of the Coupled Fast Reactor Measurement Facility (CFRMF); (5) the reference neutron field at the center of the 10 percent enriched uranium metal, cylindrical, fast critical; and (6) the (primary) Intermediate-Energy Standard Neutron Field

  14. Neutron Tomography Application for Aircraft-parts and Root of Ginseng

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yi Kyung; Lee, Seung Wook; Sim, Chul Mu; Jeon, Jin Su; Kim, Tae Ju [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    The neutron computerized tomography is considered to be a complementary tool to X-ray tomography in the sense that neutron interacts with atomic nucleus, otherwise X-ray interacts with an orbital electron of atom. The neutron tomography compared with X-ray tomography has a relatively short history. It is employed in stationary, as neutron sources are produced by the nuclear reactor. The full potential of the neutron tomography has yet to be investigated. Since, HANARO NRF was installed in 1995, the neutron radiography research group of KAERI has been developed the non-destructive testing methods by the computer tomography served for aerospace industry and agricultural industry. Concerning to NDT for aerospace, research was cooperated with Korean Air force. At the first stage, research was focused to find the micro-cracks based on internal passages inside aircraft parts and residual core of turbine blade. Concerning to NDT for agriculture, research was cooperated with the Agricultural Development and Technology Center. Research was focused to find the alive roots of Korean ginseng.

  15. 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

  16. Research on neutron radiography in Research Reactor Institute, Kyoto University and activities related to it

    International Nuclear Information System (INIS)

    Fujine, Shigenori; Yoneda, Kenji

    1994-01-01

    The research on neutron radiography in Research Reactor Institute, Kyoto University was begun in 1974 using the E-2 experimental hole which was designed for neutron irradiation. It was reconstructed for the excellent performance as neutron radiography facility by fixing aluminum plugs, a collimator and so on. The research activities thereafter are briefly described. In 1989, the cold neutron facility was installed in the graphite thermal neutron facility, and the experiment on cold neutron radiography became feasible. The reactor in Kyoto University is of the thermal output of 5 MW, and is put to the joint utilization by universities and research institutes in whole Japan. The experimental items carried out so far are enumerated. At present, the main subjects of research are the development of the standard for establishing image evaluation method, the analysis of gas-liquid two-phase flow, the construction of the data base for the literatures and images of neutron radiography, the application of cold neutron radiography, the development of the imaging method using fast neutrons and so on. The thermal neutron radiography and the cold neutron radiography facilities of Kyoto University research reactor are described. The research and activities at Kyoto University research reactor and the investigation of problems are reported. (K.I.) 56 refs

  17. Neutron absorption profile in a reactor moderated by different mixtures of light and heavy waters

    International Nuclear Information System (INIS)

    Nagy, Mohamed E.; Aly, Mohamed N.; Gaber, Fatma A.; Dorrah, Mahmoud E.

    2014-01-01

    Highlights: • We studied neutron absorption spectra in a mixed water moderated reactor. • Changing D 2 O% in moderator induced neutron energy spectral shift. • Most of the neutrons absorbed in control rods were epithermal. • Control rods worth changes were not proportional to changes of D 2 O% in moderator. • Control rod arrangement influenced the neutronic behavior of the reactor. - Abstract: A Monte-Carlo parametric study was carried out to investigate the neutron absorption profile in a model of LR-0 reactor when it is moderated by different mixtures of heavy/light waters at molecular ratios ranging from 0% up to 100% D 2 O at increments of 10% in D 2 O. The tallies included; neutron absorption profiles in control rods and moderator, and neutron capture profile in 238 U. The work focused on neutron absorption in control rods entailing; total mass of control rods needed to attain criticality, neutron absorption density and total neutron absorption in control rods at each of the studied mixed water moderators. The aim was to explore whether thermal neutron poisons are the most suitable poisons to be used in control rods of nuclear reactors moderated by mixed heavy/light water moderators

  18. 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.

  19. Diagnosis of mucoviscidosis by neutron activation analysis. Part 1; Diagnostico da mucoviscidose utilizando analise por ativacao com neutrons. Parte 1

    Energy Technology Data Exchange (ETDEWEB)

    Bellido, Luis F.; Bellido, Alfredo V

    1997-02-01

    Symptoms pathology, incidence, and gravity of the inherent syndrome called mucoviscidosis, or cystic fibrosis are described in this Part I. The analytical methods used for its diagnosis, both the conventional chemical ones and by neutron activation analysis are also summarised. Finally, an analytical method to study the incidence of mucoviscidosis in Brazil is presented. This , essentially, consists in bromine determination, in fingernails, by resonance neutron activation analysis. (author) 33 refs., 13 figs.

  20. Neutron cross-section libraries in the AMPX master interface format for thermal and fast reactors

    International Nuclear Information System (INIS)

    Bjerke, M.A.; Webster, C.C.

    1981-12-01

    Neutron cross-section libraries in the AMPX master interface format have been created for three reactor types. Included are an 84-group library for use with light-water reactors, a 27-group library for use with heavy-water CANDU reactors and a 126-group library for use with liquid metal fast breeder reactors. In general, ENDF/B data were used in the creation of these libraries, and the nuclides included in each library should be sufficient for most neutronic analyses of reactors of that type. Each library has been used successfully in fuel depletion calculations

  1. Utilization of the Dalat Research Reactor for Radioisotope Production, Neutron Activation Analysis, Research and Training

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Duong Van Dong; Cao Dong Vu; Nguyen Xuan Hai

    2013-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool type reactor loaded with a mixed core of HEU (36% enrichment) and LEU (19.75% enrichment) fuel assemblies. The reactor is used as a neutron source for the purposes of radioisotopes production, neutron activation analysis, basic and applied research and training. The reactor is operated mainly in continuous runs of 108 hours for cycles of 3–4 weeks for the above mentioned purposes. The current status of safety, operation and utilization of the reactor is given and some aspects for improvement of commercial products and services of the DNRR are also discussed in this paper. (author)

  2. Effects of loading reactivity at dynamic state on wave of neutrons in burst reactor

    International Nuclear Information System (INIS)

    Gao Hui; Liu Xiaobo; Fan Xiaoqiang

    2013-01-01

    Based on the point reactor model, the program for simulating the burst of reactors, including delay neutron, thermal feedback and reactivity of rod, was developed. The program proves to be suitable to burst reactor by experimental data. The program can describe the process of neutron-intensity change in burst reactors. With the program, the parameters of burst (wave of burst, power of peak and reactivity of reactor) under the condition of dynamic reactivity can be calculated. The calculated result demonstrates that the later the burst is initiated, the greater its power of peak and yield are and that the maximum yield coordinates with the yield under static state. (authors)

  3. Neutron irradiation embrittlement of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Steele, L.E.

    1975-01-01

    The reliability of nuclear power plants depends on the proper functioning of complex components over the whole life on the plant. Particular concern for reliability is directed to the primary pressure boundary. This report focuses on the portion of the primary system exposed to and significantly affected by neutron radiation. Experimental evidence from research programmes and from reactor surveillance programmes has indicated radiation embrittlement of a magnitude sufficient to raise doubts about reactor pressure vessel integrity. The crucial nature of the primary vessel function heightens the need to be alert to this problem, to which, fortunately, there are positive aspects: for example, steels have been developed which are relatively immune to radiation embrittlement. Further, awareness of such embrittlement has led to designs which can accomodate this factor. The nature of nuclear reactors, of the steels used in their construction, and of the procedures for interpreting embrittlement and minimizing the effects are reviewed with reference to the reactors that are expected to play a major role in electric power production from now to about the turn of the century. The report is intended as a manual or guidebook; the aim has been to make each chapter or major sub-division sufficiently comprehensive and self-contained for it to be understood and read independently of the rest of the book. At the same time, it is hoped that the whole is unified enough to make a complete reading useful and interesting to the several classes of reader that are involved with only specific aspects of the topic

  4. Determination of the effective delayed neutron fraction in the Coral-I Reactor; Determinacion de la fraccion efectiva de neutrones retardados en el Reactor Coral-1

    Energy Technology Data Exchange (ETDEWEB)

    Francisco, J L. de; Perez-Navarro, A; Rodriguez-Mayquez, E

    1973-07-01

    The effective delayed neutron fraction, {beta} eff, has been determined from the measurement of E / {beta}{sup 2}, by means of reactor noise analysis in the time domain, and the neutron detector efficiency, {epsilon}. For the {epsilon} measurement it is necessary to determine the fission rate in the reactor. This value can be obtained from the absolute measurement of the fission rate per cm{sup 3}, at a certain point of the reactor, and the determination of these two values ratio, which has been calculated by the Monte Cario method and also measured with results in good agreement. (Author)

  5. 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)

  6. Neutronic study using oxide and nitride fuels for the Super Phenix 2 reactor

    International Nuclear Information System (INIS)

    Batista, J.L.; Renke, C.A.C.

    1991-11-01

    This report presents a neutronic analysis and a description of the Super Phenix 2 reactor, taken as reference. We present the methodology and results for cell and global reactor calculations for oxide (U O 2 - Pu O 2 ) and nitride (U N - Pu N) fuels. To conclude we compare the performance of oxide and nitride fuels for the reference reactor. (author)

  7. The Effect Of Beryllium Interaction With Fast Neutrons On the Reactivity Of ETRR-2 Research Reactor

    International Nuclear Information System (INIS)

    Aziz, M.; El Messiry, A.M.

    2000-01-01

    The effect of beryllium interactions with fast neutrons is studied for Etrr 2 research reactors. Isotope build up inside beryllium blocks is calculated under different irradiation times. a new model for the Etrr 2 research reactor is designed using MCNP code to calculate the reactivity and flux change of the reactor due to beryllium poison

  8. Instrumentation for PSD based neutron diffractometers at Dhruva reactor

    International Nuclear Information System (INIS)

    Pande, S.S.; Borkar, S.P.; Prafulla, S.; Srivastava, V.D.; Behare, A.; Mukhopadhyay, P.K.; Ghodgaonkar, M.D.; Kataria, S.K.

    2004-01-01

    Linear position sensitive detectors (PSDs) are widely used to configure neutron diffractometers and other instruments. Necessary front-end electronics and a data acquisition system is developed to cater to such instruments built around the Dhruva research reactor in BARC. These include three diffractometers with multiple PSDs and four with single PSD. The front-end electronics consists of high voltage units, preamplifiers, shaping amplifiers, ratio ADCs (RDC). The data acquisition system consists of an interface card and software. Commercially available hardware like temperature controller or stepper motor controller connected over GPIB or RS232 are also integrated in the data acquisition system. The data acquisition is automated so that it can continue unattended for control parameter like temperature, thus enabling optimum utilization of available beam time. The instrumentation is scalable and can be easily configured for various instrumental requirements. The front-end electronics and the data acquisition system are described here. (author)

  9. Instrumentation for PSD-based neutron diffractometers at Dhruva reactor

    Science.gov (United States)

    Pande, S. S.; Borkar, S. P.; Prafulla, S.; Srivastava, V. D.; Behare, A.; Mukhopadhyay, P. K.; Ghodgaonkar, M. D.; Kataria, S. K.

    2004-08-01

    Linear position sensitive detectors (PSDs) are widely used to configure neutron diffractometers and other instruments. Necessary front-end electronics and a data acquisition system [1] is developed to cater to such instruments built around the Dhruva research reactor in BARC. These include three diffractometers with multiple PSDs and four with single PSD. The front-end electronics consists of high voltage units, preamplifiers [2], shaping amplifiers, ratio ADCs (RDC) [3]. The data acquisition system consists of an interface card and software. Commercially available hardware like temperature controller or stepper motor controller connected over GPIB or RS232 are also integrated in the data acquisition system. The data acquisition is automated so that it can continue unattended for control parameter like temperature, thus enabling optimum utilization of available beam time. The instrumentation is scalable and can be easily configured for various instrumental requirements. The front-end electronics and the data acquisition system are described here.

  10. Joint Neutron Noise Measurements on Metallic Reactor Caliban

    International Nuclear Information System (INIS)

    Chapelle, Amaury; Authier, Nicolas; Pierre, Casoli; Richard, Benoit; Myers, Will; Hutchinson, Jesson; Sood, Avneet; Rooney, Brian

    2013-06-01

    The aim of the experiments concerning neutron noise measurements presented in this article is to compare the measured parameters to the simulated ones. The results of these measurements must therefore be very accurate, with controlled uncertainties. To determine the relative contribution of uncertainties to the final result, a table presents the prompt multiplication obtained by a French Team and a U.S. team. The different sources of uncertainties are then explored, distinguishing them between three categories, those linked to the experimental configuration, to the detection process and finally to the analysis process. These experiments improve the safety task of reactivity control far from criticality, with static methods, and the knowledge of the behaviour of a subcritical reactor. (authors)

  11. Design of the Mechanical Parts for the Neutron Guide System at HANARO

    International Nuclear Information System (INIS)

    Shin, J. W.; Cho, Y. G.; Cho, S. J.; Ryu, J. S.

    2008-01-01

    The research reactor HANARO (High-flux Advanced Neutron Application ReactOr) in Korea will be equipped with a neutron guide system, in order to transport cold neutrons from the neutron source to the neutron scattering instruments in the neutron guide hall near the reactor building. The neutron guide system of HANARO consists of the in-pile plug assembly with in-pile guides, the primary shutter with in-shutter guides, the neutron guides in the guide shielding room with dedicated secondary shutters, and the neutron guides connected to the instruments in the neutron guide hall. Functions of the in-pile plug assembly are to shield the reactor environment from nuclear radiation and to support the neutron guides and maintain them precisely oriented. The primary shutter is a mechanical structure to be installed just after the in-pile plug assembly, which stops neutron flux on demand. This paper describes the design of the in-pile assembly and the primary shutter for the neutron guide system at HANARO. The design of the guide shielding assembly for the primary shutter and the neutron guides is also presented

  12. Conversion of reactor neutrons by lithium deuteride; Konverzija reaktorskih neutrona pomocu litijumdeuterida

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P; Altiparmakov, D [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1979-07-01

    Nuclear reactors are powerful neutron sources. But for many purposes, neutrons of higher than fission energy are needed. Such neutrons may be produced by the conversion of reactor neutrons into 14 MeV neutrons using lithium deuteride converter. For that converter is generally employed {sup 6}LiD, the usual material for the thermonuclear weapons, and therefore hardly accessible in needed quantity. That was the reason to analyse such converter made of LiD with the lithium of natural isotopic content. This analysis starts with the basic conversion relations, takes into account neutron absorption and tritium generation, and finally, estimates the 14 MeV neutron flux and the heat generated in proposed converter, when the converter was coupled to each of three Yugoslav nuclear reactors. Results show that the converter made of LiD with the natural lithium is 50% less efficient than the converter of {sup 6}LiD. Intensity of 14 MeV neutrons is within limits 5. 10{sup 5} - 10{sup 10} (n/cm{sup 2}.s) for the converter used either as external converter with reactor RB, within the thermal column in the reactor TRIGA or as a 'fuel' segment at the reactor RA. (author)

  13. Conceptual design study of quasi-steady state fusion experimental reactor (FER-Q), part 2

    International Nuclear Information System (INIS)

    1985-12-01

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. Starting from 1984 FER design is being reviewed and redesigned. This report is a part of the interim report which describes the results obtained in the review and redesign activities in FY 1984. The results of the following design items are included: heating/current drive system, plasma position control, power supply, diagnostics, neutronics, blanket test module, repair and maintenance and safety. (author)

  14. Comparison of pressure vessel neutron fluences for the Balakovo-3 reactor with measurements and investigation of the influence of neutron cross sections and number of groups on the results

    Energy Technology Data Exchange (ETDEWEB)

    Barz, H U; Boehmer, B; Konheiser, J; Stephan, I

    1998-10-01

    The general methodical questions of experimental and theoretical determination of neutron fluences have been described in connection with the measurements and 3-D Monte Carlo calculation for the Rovno-3 reactor. The same calculation and measurement methods were applied for the Balakovo-3 reactor. In the first part, the results of the comparison for Balakovo will be given and discussed. However, for this reactor the main attention was focussed on investigations of the accuracy of the calculation. In this connection an important question is the influence of neutron data on the results. With this respect not only the source of the data but also the number of energy groups is important. (orig.)

  15. Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (P fus ∼ 10-100 MW, β N ∼ 2-3, Q p ∼ 2-5, R ∼ 3-5 m, I ∼ 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-a-vis a neutron source is reviewed. (author)

  16. INR TRIGA Research Reactors: A Neutron Source for Radioisotopes and Materials Investigation

    International Nuclear Information System (INIS)

    Barbos, D.; Ciocanescu, M.; Paunoiu, C.; Bucsa, A.F.

    2013-01-01

    At the INR there are 2 high intensity neutron sources. These sources are in fact the two nuclear TRIGA reactors: TRIGA SSR 14 MW and TRIGA ACPR. TRIGA stationary reactor is provided with several in-core irradiation channels. Other several out-of-core irradiation channels are located in the vertical channels in the beryllium reflector blocks. The maximum value of the thermal neutron flux (E 14 cm -2 s -1 and of fast neutron flux (E>1 MeV) is 6.89×10 13 cm -2 s -1 . For neutron activation analysis both reactors are used and k0-NAA method has been implemented. At INR Pitesti a prompt gamma ray neutron activation analysis devices has been designed, manufactured ant put into operation. For nuclear materials properties investigation neutron radiography methods was developed in INR. For these purposes two neutron radiography devices were manufacture, one of them underwater and other one dry. The neutron beams are used for investigation of materials properties and components produced or under development for applications in the energy sector (fission and fusion). At TRIGA 14 MW reactor a neutron difractormeter and a SANS devices are available for material residual stress and texture measurements. TRIGA 14 MW reactor is used for medical and industrial radioisotopes production ( 131 I, 125 I, 192 Ir, etc) and a method for 99 Mo- 99 Tc production from fission is under developing. At INR Pitesti several special programmes for new types of nuclear fuel behavior characterization are under development. (author)

  17. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  18. An ultracold neutron source at the NC State University PULSTAR reactor

    Science.gov (United States)

    Korobkina, E.; Wehring, B. W.; Hawari, A. I.; Young, A. R.; Huffman, P. R.; Golub, R.; Xu, Y.; Palmquist, G.

    2007-08-01

    Research and development is being completed for an ultracold neutron (UCN) source to be installed at the PULSTAR reactor on the campus of North Carolina State University (NCSU). The objective is to establish a university-based UCN facility with sufficient UCN intensity to allow world-class fundamental and applied research with UCN. To maximize the UCN yield, a solid ortho-D 2 converter will be implemented coupled to two moderators, D 2O at room temperature, to thermalize reactor neutrons, and solid CH 4, to moderate the thermal neutrons to cold-neutron energies. The source assembly will be located in a tank of D 2O in the space previously occupied by the thermal column of the PULSTAR reactor. Neutrons leaving a bare face of the reactor core enter the D 2O tank through a 45×45 cm cross-sectional area void between the reactor core and the D 2O tank. Liquid He will cool the disk-shaped UCN converter to below 5 K. Independently, He gas will cool the cup-shaped CH 4 cold-neutron moderator to an optimum temperature between 20 and 40 K. The UCN will be transported from the converter to experiments by a guide with an inside diameter of 16 cm. Research areas being considered for the PULSTAR UCN source include time-reversal violation in neutron beta decay, neutron lifetime determination, support measurements for a neutron electric-dipole-moment search, and nanoscience applications.

  19. RA reactor operation and maintenance in 1996, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

    1996-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. Since the RA reactor is shutdown since 1984, it is high time for decision making of its future status. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown. Control and maintenance of the reactor instrumentation and devices was done regularly but dependent on the availability of the spare parts and financial means. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  20. Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, David William, E-mail: hummeld@mcmaster.ca; Novog, David Raymond

    2016-03-15

    Highlights: • A coupled spatial kinetics and thermalhydraulics model of the PT-SCWR was created. • Positive power excursions were demonstrated during accident-like transients. • The reactor will inherently self-shutdown in such transients with some delay. • A fast-acting shutdown system would limit the consequences of the power pulse. - Abstract: The Canadian Supercritical Water-cooled Reactor concept, as an evolution of the CANada Deuterium Uranium (CANDU) reactor, includes both pressure tubes and a low temperature heavy water moderator. The current Pressure Tube type SCWR (PT-SCWR) concept features 64-element fuel assemblies placed within High Efficiency Re-entrant Channels (HERCs) that connect to core inlet and outlet plena. Among current SCWR concepts the PT-SCWR is unique in that the HERC separates multiple coolant and moderator regions, giving rise to coupled neutronic-thermalhydraulic feedbacks beyond those present in CANDU or contemporary Light Water Reactors. The objective of this work was thus to model the coupled neutronic-thermal hydraulic properties of the PT-SCWR to establish the impact of these multiple regions on the core's transient behavior. To that end, the features of the PT-SCWR were first modeled with the neutron transport code DRAGON to create a database of homogenized and condensed cross-sections and thermalhydraulic feedback coefficients. These were used as input to a core-level neutron diffusion model created with the code DONJON. The behavior of the primary heat transport system was modeled with the thermalhydraulic system code CATHENA. A procedure was developed to couple the outputs of DONJON and CATHENA, facilitating three-dimensional spatial neutron kinetics and coupled thermalhydraulic analysis of the PT-SCWR core. Several postulated transients were initiated within the coupled model by changing the core inlet and outlet boundary conditions. Decreasing coolant density around the fuel was demonstrated to produce positive

  1. Modelling of heterogenous neutron leakages in a nuclear reactor

    International Nuclear Information System (INIS)

    Wohleber, X.

    1997-01-01

    The TIBERE Model is a neutron leakage method based on B 1 heterogeneous transport equation resolution. In this work, we have studied the influence of the reflection mode at the boundary of the assembly. In particular the White boundary condition has been implemented in the APOLLO2 neutron transport code. We have compared the two TIBERE kinds of boundary conditions (specular and white) with the classical B 1 homogeneous leakage method in the modelling of some reactors. We have remarked the better capability of the TIBERE Model to compute voided assemblies. The white boundary condition is also able to compute a completely voided assembly and, besides, wins a factor 10 in CPU time in comparison with the specular boundary condition. These two heterogenous leakage formalisms have been tested on a partially voided experiment and have shown that the TIBERE Model can compute this kind of situation with a greater precision than the classical B 1 homogeneous leakage method, and with a shorter computational time. (author)

  2. Design, construction and characterization of a new neutron beam for neutron radiography at the Tehran Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choopan Dastjerdi, M.H., E-mail: mdastjerdi@aeoi.org.ir [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Department of Energy Engineering and Physics, Amirkabir University of Technology, Tehran (Iran, Islamic Republic of); Khalafi, H.; Kasesaz, Y.; Mirvakili, S.M.; Emami, J.; Ghods, H.; Ezzati, A. [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of)

    2016-05-11

    To obtain a thermal neutron beam for neutron radiography applications, a neutron collimator has been designed and implemented at the Tehran Research Reactor (TRR). TRR is a 5 MW open pool light water moderated reactor with seven beam tubes. The neutron collimator is implemented in the E beam tube of the TRR. The design of the neutron collimator was performed using MCNPX Monte Carlo code. In this work, polycrystalline bismuth and graphite have been used as a gamma filter and an illuminator, respectively. The L/D parameter of the facility was chosen in the range of 150–250. The thermal neutron flux at the image plane can be varied from 2.26×10{sup 6} to 6.5×10{sup 6} n cm{sup −2} s{sup −1}. Characterization of the beam was performed by ASTM standard IQI and foil activation technique to determine the quality of neutron beam. The results show that the obtained neutron beam has a good quality for neutron radiography applications.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. Nuclear reactor with a fixed system of neutron poison, which can be burnt up, introduced into the reactor core

    International Nuclear Information System (INIS)

    Mueller, E.; Roegler, H.J.; Wickert, M.

    1985-01-01

    The fixed system consists of neutron poison which can be burnt up, in an uneven distribution, and with adjustable absorber rods for output control, which are driven into the reactor core from the side along the fuel elements. There is an excess of neutron poison which can be burnt up, overall, on the side of the reactor core away from the absorber rods. The reactor core is free of neutron poison which can be burnt up on the side where the absorber rods are driven in, so that the ratio of maximum to mean power density with reference to a possible absorber rod positions is less than for homogeneous distribution of the neutron poison which can be burnt up. (orig./HP) [de

  5. Estimation of damage by inmates of a PWR Reactor neutron irradiation. Project ZIRP; Estimacion del Dano por Irradiacion Neutronica en los Internos de un Reactor PWR. Proyecto ZIRP

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.

    2013-07-01

    The study presented here focuses on the analysis of neutron and gamma irradiation damage suffered by the inmates of the JC NPP reactor metallic materials throughout its operational life. Such analysis of radiation are part of a project of great international impact, led by EPRI (Electric Power Research Institute) from the MRP (Materials Reliability Program), which aims to relate the degradation of the properties of metallic materials of the inmates of the reactor, with the conditions of operation and irradiation to which have been subjected during the operational life of the plant.

  6. Preliminary validation of computational model for neutron flux prediction of Thai Research Reactor (TRR-1/M1)

    Science.gov (United States)

    Sabaibang, S.; Lekchaum, S.; Tipayakul, C.

    2015-05-01

    This study is a part of an on-going work to develop a computational model of Thai Research Reactor (TRR-1/M1) which is capable of accurately predicting the neutron flux level and spectrum. The computational model was created by MCNPX program and the CT (Central Thimble) in-core irradiation facility was selected as the location for validation. The comparison was performed with the typical flux measurement method routinely practiced at TRR-1/M1, that is, the foil activation technique. In this technique, gold foil is irradiated for a certain period of time and the activity of the irradiated target is measured to derive the thermal neutron flux. Additionally, the flux measurement with SPND (self-powered neutron detector) was also performed for comparison. The thermal neutron flux from the MCNPX simulation was found to be 1.79×1013 neutron/cm2s while that from the foil activation measurement was 4.68×1013 neutron/cm2s. On the other hand, the thermal neutron flux from the measurement using SPND was 2.47×1013 neutron/cm2s. An assessment of the differences among the three methods was done. The difference of the MCNPX with the foil activation technique was found to be 67.8% and the difference of the MCNPX with the SPND was found to be 27.8%.

  7. 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

  8. Physical model study of neutron noise induced by vibration of reactor internals

    International Nuclear Information System (INIS)

    Liu Jinhui; Gu Fangyu

    1999-01-01

    The author presents a physical model of neutron noise induced by reactor internals vibration in frequency domain. Based on system control theory, the reactor dynamic equations are coupled with random vibration equation, and non-linear terms are also taken into accounted while treating the random vibration. Experiments carried out on a zero-power reactor show that the model can be used to describe dynamic character of neutron noise induced by internals' vibration. The model establishes a method to help to determine internals'vibration features, and to diagnosis anomalies through neutron noise

  9. Measurement of neutrons in the RA reactor cell; Merenje neutrona u elementarnoj celiji reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Bosevski, T [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    A special experimental device was constructed for measuring the neutron flux distribution in the RA reactor cell. This device simulated the reactor cell in order to avoid disturbance in the reactor core. It was made of an aluminium cylindrical vessel having outer diameter same as the vertical experimental channel and contained three fuel slugs. Hole was made in through the center of the fuel slugs and a copper wire was placed in the hole for measuring the thermal neutron flux distribution. It was placed in the experimental channel VK-5 in the location of highest neutron flux. Handling of samples for irradiation was quite simple.

  10. The advanced neutron source - A world-class research reactor facility

    International Nuclear Information System (INIS)

    Thompson, P.B.; Meek, W.E.

    1993-01-01

    The advanced neutron source (ANS) is a new facility being designed at the Oak Ridge National Laboratory that is based on a heavy-water-moderated reactor and extensive experiment and user-support facilities. The primary purpose of the ANS is to provide world-class facilities for neutron scattering research, isotope production, and materials irradiation in the United States. The neutrons provided by the reactor will be thermalized to produce sources of hot, thermal, cold, very cold, and ultracold neutrons usable at the experiment stations. Beams of cold neutrons will be directed into a large guide hall using neutron guide technology, greatly enhancing the number of research stations possible in the project. Fundamental and nuclear physics, materials analysis, and other research pro- grams will share the neutron beam facilities. Sufficient laboratory and office space will be provided to create an effective user-oriented environment

  11. Optimization study of ultracold neutron sources at TRIGA reactors using MCNP

    International Nuclear Information System (INIS)

    Pokotilovskij, Yu.N.; Rogov, A.D.

    1997-01-01

    Monte Carlo simulation for the optimization of ultracold and very cold neutron sources for TRIGA reactors is performed. The calculations of thermal and cold neutron fluxes from the TRIGA reactor for different positions and configurations of a very cold solid methane moderator were performed with using the MCNP program. The production of neutrons in the ultracold and very cold energy range was calculated for the most promising final moderators (converters): very cold solid deuterium and heavy methane. The radiation energy deposition was calculated for the optimized solid methane-heavy methane cold neutron moderator

  12. Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

    2008-01-01

    A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

  13. Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

    2009-01-01

    A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

  14. 239Pu prompt fission neutron spectra impact on a set of criticality and experimental reactor benchmarks

    International Nuclear Information System (INIS)

    Peneliau, Y.; Litaize, O.; Archier, P.; De Saint Jean, C.

    2014-01-01

    A large set of nuclear data are investigated to improve the calculation predictions of the new neutron transport simulation codes. With the next generation of nuclear power plants (GEN IV projects), one expects to reduce the calculated uncertainties which are mainly coming from nuclear data and are still very important, before taking into account integral information in the adjustment process. In France, future nuclear power plant concepts will probably use MOX fuel, either in Sodium Fast Reactors or in Gas Cooled Fast Reactors. Consequently, the knowledge of 239 Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239 Pu are part of these relevant data (an IAEA working group is even dedicated to PFNS) and the work presented here deals with this particular topic. The main international data files (i.e. JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0, BRC-2009) have been considered and compared with two different spectra, coming from the works of Maslov and Kornilov respectively. The spectra are first compared by calculating their mathematical moments in order to characterize them. Then, a reference calculation using the whole JEFF-3.1.1 evaluation file is performed and compared with another calculation performed with a new evaluation file, in which the data block containing the fission spectra (MF=5, MT=18) is replaced by the investigated spectra (one for each evaluation). A set of benchmarks is used to analyze the effects of PFNS, covering criticality cases and mock-up cases in various neutron flux spectra (thermal, intermediate, and fast flux spectra). Data coming from many ICSBEP experiments are used (PU-SOL-THERM, PU-MET-FAST, PU-MET-INTER and PU-MET-MIXED) and French mock-up experiments are also investigated (EOLE for thermal neutron flux spectrum and MASURCA for fast neutron flux spectrum). This study shows that many experiments and neutron parameters are very sensitive to

  15. Utilization of boron irradiation filters in reactor neutron activation via epithermal (n,γ) and fast neutron reactions

    International Nuclear Information System (INIS)

    Chisela, F.

    1986-01-01

    The technique of instrumental neutron activation analysis based on irradiation with reactor epithermal and fast neutrons has been described and evaluated. Important characteristics of boron neutron absorbers used to remove thermal neutrons from the reactor neutron spectrum have been examined and compared with those of cadmium. Three boron compound shields, have been designed and constructed at the BER II 5MW reactor for use in epithermal neutron activation analysis of biological materials. The major advantages offered by these filters in this application include the flexibility of varying the filter thickness, the low radioactivity induced in the filters during irradiation, ease of fabrication and the relatively low cost of the filter materials. The radiation heating due to the 10 B(n,α) 7 Li-reaction has been experimentally investigated for the filters used and the results obtained confirm the necessity for efficient cooling of these filters during irradiation. Three irradiation facilities have been characterized with respect to the neutron flux density and the flux spatial distribution. An experiment has been designed and carried out to compensate the flux inhomogeneity in two irradiation positions of the DBV facility caused by the reactor geometry. Several biological samples including well characterized reference materials have been analysed after epithermal activation and the results compared with those obtained with the classical thermal neutron activation method. Improved sensitivity of determination has been found for elements with high resonance integral to thermal neutron cross section ratios (RI/σ 0 ). The range of elements that can be determined instrumentally is extended and the time scale of analysis is considerably reduced. (orig.) [de

  16. A CONCEPTUAL DESIGN OF NEUTRON COLLIMATOR IN THE THERMAL COLUMN OF KARTINI RESEARCH REACTOR FOR IN VITRO AND IN VIVO TEST OF BORON NEUTRON CAPTURE THERAPY

    Directory of Open Access Journals (Sweden)

    Nina Fauziah

    2015-03-01

    Full Text Available Studies were carried out to design a collimator which results in epithermal neutron beam for IN VITRO and IN VIVO of Boron Neutron Capture Therapy (BNCT at the Kartini research reactor by means of Monte Carlo N-Particle (MCNP codes. Reactor within 100 kW of thermal power was used as the neutron source. The design criteria were based on recommendation from the International Atomic Energy Agency (IAEA. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 5 cm thick of Ni as collimator wall, 60 cm thick of Al as moderator, 15 cm thick of 60Ni as filter, 2 cm thick of Bi as γ-ray shielding, 3 cm thick of 6Li2CO3-polyethylene as beam delimiter, with 1 to 5 cm varied aperture size, epithermal neutron beam with maximum flux of 7.65 x 108 n.cm-2.s-1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 1.76 x 10-13 Gy.cm2.n-1 and 1.32 x 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.008, and maximum directionality of 0.73. It did not fully pass the IAEA’s criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 108 n.cm-2.s-1. When it was assumed that the graphite inside the thermal column was not discharged but only the part which was going to be replaced by the collimator, the performance of the collimator became better within the positive effect from the surrounding graphite that the beam resulted passed all criteria with epithermal neutron flux up to 1.68 x 109 n.cm-2.s-1. Keywords: design, collimator, epithermal neutron beam, BNCT, MCNP, criteria   Telah dilakukan penelitian tentang desain kolimator yang menghasilkan radiasi netron epitermal untuk uji in vitro dan in vivo pada Boron Neutron Capture Therapy (BNCT di Reaktor Riset Kartini dengan menggunakan program Monte

  17. Analysis of the neutron flux in an annular pulsed reactor by using finite volume method

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mário A.B. da; Narain, Rajendra; Bezerra, Jair de L., E-mail: mabs500@gmail.com, E-mail: narain@ufpe.br, E-mail: jairbezerra@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociências. Departamento de Energia Nuclear

    2017-07-01

    Production of very intense neutron sources is important for basic nuclear physics and for material testing and isotope production. Nuclear reactors have been used as sources of intense neutron fluxes, although the achievement of such levels is limited by the inability to remove fission heat. Periodic pulsed reactors provide very intense fluxes by a rotating modulator near a subcritical core. A concept for the production of very intense neutron fluxes that combines features of periodic pulsed reactors and steady state reactors was proposed by Narain (1997). Such a concept is known as Very Intense Continuous High Flux Pulsed Reactor (VICHFPR) and was analyzed by using diffusion equation with moving boundary conditions and Finite Difference Method with Crank-Nicolson formalism. This research aims to analyze the flux distribution in the Very Intense Continuous Flux High Pulsed Reactor (VICHFPR) by using the Finite Volume Method and compares its results with those obtained by the previous computational method. (author)

  18. Analysis of the neutron flux in an annular pulsed reactor by using finite volume method

    International Nuclear Information System (INIS)

    Silva, Mário A.B. da; Narain, Rajendra; Bezerra, Jair de L.

    2017-01-01

    Production of very intense neutron sources is important for basic nuclear physics and for material testing and isotope production. Nuclear reactors have been used as sources of intense neutron fluxes, although the achievement of such levels is limited by the inability to remove fission heat. Periodic pulsed reactors provide very intense fluxes by a rotating modulator near a subcritical core. A concept for the production of very intense neutron fluxes that combines features of periodic pulsed reactors and steady state reactors was proposed by Narain (1997). Such a concept is known as Very Intense Continuous High Flux Pulsed Reactor (VICHFPR) and was analyzed by using diffusion equation with moving boundary conditions and Finite Difference Method with Crank-Nicolson formalism. This research aims to analyze the flux distribution in the Very Intense Continuous Flux High Pulsed Reactor (VICHFPR) by using the Finite Volume Method and compares its results with those obtained by the previous computational method. (author)

  19. Pursuing nuclear energy with no nuclear contamination - from neutron flux reactor to deuteron flux reactor

    International Nuclear Information System (INIS)

    Li, X. Z.; Wei, Q. M.; Liu, B.; Zhu, X. G.; Ren, S. L.

    2007-01-01

    Pursuing nuclear energy with no nuclear contamination has been a long endeavor since the first fission reactor in 1942. Four major concepts have been the key issues: i.e. resonance, negative feed back, self-sustaining, nuclear radiation. When nuclear energy was just discovered in laboratory, the key issue was to enlarge it from the micro-scale to the macro-scale. Slowing-down the neutrons was the key issue to enhance the fission cross-section in order to build-up the neutron flux through the chain-reactions using resonance between neutron and fissile materials. Once the chain-reaction was realized, the negative feed-back was the key issue to keep the neutron flux at the allowable level. The negative reaction coefficient was introduced by the thermal expansion, and the resonant absorption in cadmium or boron was used to have a self-sustaining fission reactor with neutron flux. Then the strong neutron flux became the origin of all nuclear contamination, and a heavy shielding limits the application of the nuclear energy. The fusion approach to nuclear energy was much longer; nevertheless, it evolved with the similar issues. The resonance between deuteron and triton was resorted to enlarge the fusion cross section in order to keep a self-sustaining hot plasma. However, the 14 MeV neutron emission became the origin of all nuclear contamination again. Deuteron plus helium-3 fusion reaction was proposed to avoid neutron emission although there are two more difficulties: the helium-3 is supposed to be carried back from the moon; and much more higher temperature plasma has to be confined while 50 years needed to realized the deuteron-triton plasma already. Even if deuteron plus helium-3 fusion plasma might be realized in a much higher temperature plasma, we still have the neutron emission from the deuteron-deuteron fusion reaction in the deuteron plus helium-3 fusion plasma. Polarized deuteron-deuteron fusion reaction was proposed early in 1980's to select the neutron

  20. The neutron utilization and promotion program of TRR-II research reactor project in Taiwan

    International Nuclear Information System (INIS)

    Gone, J.K.; Huang, Y.H.

    2001-01-01

    The objective of the Taiwan research reactor system improvement and utilization promotion project is to reconstruct the old Taiwan research reactor (TRR), which was operated by the Institute of Nuclear Energy Research (INER) between 1973 and 1988, into a multi-purpose medium flux research reactor (TRR-II). The project started in 1998, and the new reactor is scheduled to have its first critical in June of 2006. The estimated maximum unperturbed thermal neutron flux (E 14 n/cm 2 sec, and it is about one order of magnitude higher than other operating research reactors in Taiwan. The new reactor will equip with secondary neutron sources to provide neutrons with different energies, which will be an essential tool for advanced material researches in Taiwan. One of the major tasks of TRR-II project is to promote domestic utilization of neutrons generated at TRR-II. The traditional uses of neutrons in fuel/material research, trace element analysis, and isotope production has been carried out at INER for many years. On the other hand, it is obvious that promotions of neutron spectrometric technique will be a major challenge for the project team. The limited neutron flux from operating research reactors had discouraged domestic users in developing neutron spectrometric technique for many years, and only few researchers in Taiwan are experienced in using spectrometers. It is important for the project team to encourage domestic researchers to use neutron spectrometers provided by TRR-II as a tool for their future researches in various fields. This paper describes the current status of TRR-II neutron utilization and promotion program. The current status and future plans for important issues such as staff recruiting, personnel training, international collaboration, and promotion strategy will be described. (orig.)

  1. Burnup influence on the VVER-1000 reactor vessel neutron fluence evaluation

    International Nuclear Information System (INIS)

    Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

    2009-01-01

    The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of the Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in 1/4 depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (authors)

  2. Burnup influence on the WWER-1000 reactor vessel neutron fluence evaluation

    International Nuclear Information System (INIS)

    Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

    2009-01-01

    The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of WWER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in ? depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (Authors)

  3. Measurement and analysis of the neutron noise of the pool research reactor at IPEN

    International Nuclear Information System (INIS)

    Simoes, Graciete Pedro

    1979-01-01

    Variations in the neutron density or power of a nuclear reactor (the neutron noise) operating at nominally constant power are generally random and can only be described in terms of statistical parameters. Random variations in the power of a power reactor are produced by one or more driving functions. In this work the neutron noise of the pool reactor IEAR-1 (2 MW nominal power) has been studied using two compensated ionization chambers ( Westinghouse VJL6377) and related to three possible-driving functions, namely vibration of the control bar and reactor support bridge and the temperature of the water entering the core. The CIC detectors were located in rigid tubes in turn positively located in the reactor lattice plate. Conventional accelerometers were used. Temperature measurements were made with a NiCr/Ni thermocouple (wire diam ∼ 0.2mm) located 10 mm above the top of a fuel element. Although the correlation between the measured neutron signals was high ( > 0,4) for frequencies in the range 0 to 10 Hz no resonances were identified in the neutron noise. A significant correlation (> 0,4) between the control bar acceleration and the neutron flux was obtained in the frequency range 0 to 10 Hz. The measured correlation between the neutron noise and both the bridge vibration and the reactor water inlet temperature was insignificant. (author)

  4. Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

    Energy Technology Data Exchange (ETDEWEB)

    Lescop, B.; Badeau, G.; Ivanovic, S.; Foulon, F. [National Institute for Nuclear science and Technology French Atomic Energy and Alternative Energies Commission (CEA), Saclay Research Center, 91191 Gif-sur-Yvette (France)

    2015-07-01

    modify the settings of the electronic system, such as the high voltage and the discrimination level in order to obtain all the characteristic curves of the detectors. These curves are used to define the right setting of the electronic system and to discuss the expected degradation of the detector signal resulting from the detector damage under the integrated neutron and gamma fluxes. Moreover, in addition to the study of the neutron detection systems itself, the integration of the measurements made by these detection systems in the logic of the safety system of the nuclear reactor is also addressed. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. In addition to on-reactor training, Internet Reactor Laboratory capability has been implemented on the ISIS reactor in 2014. For the Internet Reactor Laboratory an extensive video conference system has been implemented on ISIS reactor. The system includes 4 cameras and the transmission of the video signal given by the supervision system of the reactor which records and processes the data of the reactor. According to the pedagogic needs during the training courses, the lecturer on the ISIS reactor chooses to broadcast the relevant information at each stage of the course. For example, graph showing the histogram of the counting and current as a function of the time, or the electrical signal observed on the oscilloscope, can be broadcasted trough internet. By interacting through the video conference, the remote classroom is able to ask for changes in the reactor power or settings of the detection systems. They can also ask for the

  5. Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

    International Nuclear Information System (INIS)

    Lescop, B.; Badeau, G.; Ivanovic, S.; Foulon, F.

    2015-01-01

    settings of the electronic system, such as the high voltage and the discrimination level in order to obtain all the characteristic curves of the detectors. These curves are used to define the right setting of the electronic system and to discuss the expected degradation of the detector signal resulting from the detector damage under the integrated neutron and gamma fluxes. Moreover, in addition to the study of the neutron detection systems itself, the integration of the measurements made by these detection systems in the logic of the safety system of the nuclear reactor is also addressed. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. In addition to on-reactor training, Internet Reactor Laboratory capability has been implemented on the ISIS reactor in 2014. For the Internet Reactor Laboratory an extensive video conference system has been implemented on ISIS reactor. The system includes 4 cameras and the transmission of the video signal given by the supervision system of the reactor which records and processes the data of the reactor. According to the pedagogic needs during the training courses, the lecturer on the ISIS reactor chooses to broadcast the relevant information at each stage of the course. For example, graph showing the histogram of the counting and current as a function of the time, or the electrical signal observed on the oscilloscope, can be broadcasted trough internet. By interacting through the video conference, the remote classroom is able to ask for changes in the reactor power or settings of the detection systems. They can also ask for the broadcast of some

  6. Neutronics analysis of the proposed 25-MW leu TRIGA Multipurpose Research Reactor

    International Nuclear Information System (INIS)

    Nurdin, M.; Bretscher, M.M.; Snelgrove, J.L.

    1982-01-01

    More than two years ago the government of Indonesia announced plans to purchase a research reactor for the Puspiptek Research Center in Serpong Indonesia to be used for isotope production, materials testing, neutron physics measurements, and reactor operator training. Reactors using low-enriched uranium (LEU) plate-type and rod-type fuel elements were considered. This paper deals with the neutronic evaluation of the rod-type 25-MW LEU TRIGA Multipurpose Research Reactor (MPRR) proposed by the General Atomic Company of the United States of America

  7. Investigating The Integral Control Rod Worth Of The Miniature Neutron Source Reactor MNSR

    International Nuclear Information System (INIS)

    Nguyen Hoang Hai; Do Quang Binh

    2011-01-01

    Determining control rod characteristics is an essential problem of nuclear reactor analysis. In this research, the integral control rod worth of the miniature neutron source reactor MNSR is investigated. Some other parameters of the nuclear reactor, such as core excess reactivity, shut down margin, are also calculated. Group constants for all reactor components are generated by the WIMSD code and then are used in the CITATION code to solve the neutron diffusion equations. The maximum relative error of the calculated results compared with the measurement data is about 3.5%. (author)

  8. AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup

  9. Fast neutron dosimetry in research reactors; Dosimetrie en neutrons rapides dans les reacteurs de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Eckert, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    This work chiefly concerns the measurement of fast neutron fluxes by means of threshold detectors. It is shown first that the cross sections to use for measurements by threshold detectors depend largely on the neutron spectrum, that is the position in which the measurement is performed. The spectrum is determined by calculation for several positions in the piles EL2 and EL3; from this can be deduced the cross-sections to be used for the measurements carried out in these positions. In the last part of the report, possible methods for the experimental determination of the spectrum are indicated. (author) [French] On etudie principalement la mesure des flux de neutrons rapides a l'aide de detecteurs a seuil. On montre d'abord que les sections efficaces a utiliser pour les mesures par detecteurs a seuil, dependent grandement du spectre des neutrons, c'est-a-dire de l'emplacement ou s'effectue la mesure. La determination du spectre est effectuee par le calcul pour plusieurs emplacements des piles EL2 et EL3; on en deduit les sections efficaces a utiliser pour les mesures effectuees a ces emplacements. Dans la derniere partie du rapport, on indique quelles methodes sont possibles pour la determination experimentale du spectre. (auteur)

  10. A hybrid source-driven method to compute fast neutron fluence in reactor pressure vessel - 017

    International Nuclear Information System (INIS)

    Ren-Tai, Chiang

    2010-01-01

    A hybrid source-driven method is developed to compute fast neutron fluence with neutron energy greater than 1 MeV in nuclear reactor pressure vessel (RPV). The method determines neutron flux by solving a steady-state neutron transport equation with hybrid neutron sources composed of peripheral fixed fission neutron sources and interior chain-reacted fission neutron sources. The relative rod-by-rod power distribution of the peripheral assemblies in a nuclear reactor obtained from reactor core depletion calculations and subsequent rod-by-rod power reconstruction is employed as the relative rod-by-rod fixed fission neutron source distribution. All fissionable nuclides other than U-238 (such as U-234, U-235, U-236, Pu-239 etc) are replaced with U-238 to avoid counting the fission contribution twice and to preserve fast neutron attenuation for heavy nuclides in the peripheral assemblies. An example is provided to show the feasibility of the method. Since the interior fuels only have a marginal impact on RPV fluence results due to rapid attenuation of interior fast fission neutrons, a generic set or one of several generic sets of interior fuels can be used as the driver and only the neutron sources in the peripheral assemblies will be changed in subsequent hybrid source-driven fluence calculations. Consequently, this hybrid source-driven method can simplify and reduce cost for fast neutron fluence computations. This newly developed hybrid source-driven method should be a useful and simplified tool for computing fast neutron fluence at selected locations of interest in RPV of contemporary nuclear power reactors. (authors)

  11. Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

    Directory of Open Access Journals (Sweden)

    Kaiser Krista

    2016-01-01

    Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  12. 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)

  13. The epithermal neutron-flux distribution in the reactor RA - Vinca

    International Nuclear Information System (INIS)

    Marinkov, V.; Bikit, I.; Martinc, R.; Veskovic, M.; Slivka, J.; Vaderna, S.

    1987-01-01

    The distribution of the epithermal neutron flux in the reactor RA - Vinca has been measured by means of Zr - activation detectors. In the channel VK-8 non-homogeneous flux distribution was observed (author) [sr

  14. Design study of a medical reactor for Boron Neutron Capture Therapy

    International Nuclear Information System (INIS)

    Sasaki, M.; Hirota, J.; Tamao, S.; Kanda, K.; Mishima, Y.

    1992-01-01

    A new design study of a medical reactor for Boron Neutron Capture Therapy (BNCT) has been carried out. The reactor is to be used exclusively for the treatment of malignant melanoma and other cancers as well as for the further biomedical research. Main specifications of the reactor are as follows; thermal power of 2 MW, water cooling by natural convection, semitight core of triangular lattice, UO 2 fuel rod of 9.5 mm diameter and no refueling in the reactor-life. Three horizontal and one vertical neutron beam hole are to be provided to deliver thermal and epithermal neutrons. N-γ coupling Sn transport calculations indicate that the patient treatment period will be about 30 minutes with minimal fast neutron and gamma contaminants. (author)

  15. Progress in study of a medical reactor for boron neutron capture therapy

    International Nuclear Information System (INIS)

    Sasaki, Makoto; Hirota, Jitsuya; Tamao, Shigeo; Kanda, Keiji; Mishima, Yutaka.

    1993-01-01

    A design study of a medical reactor for Boron Neutron Capture Therapy has made progress. Main specifications of the reactor are as follows; thermal power of 2 MW, water cooling by natural convection, semitight core of hexagonal lattice, UO 2 fuel rod of 9.5 mm diameter and no refueling in the reactor-life. Three horizontal and one vertical neutron beam holes are to be provided for simultaneous treatments by thermal and epithermal neutrons and for further biomedical research. The design objectives for the beam holes are to deliver the therapeutic doses in a modest time (30 to 60 min) with minimal fast neutron and gamma contaminants. The n-γ coupling Sn transport calculations have been carried out using n-21 and γ-9 group cross sections on 2-dim. practical models. The calculated results indicate that the design objectives will be achievable even if the thermal power of the reactor is reduced to 1 MW. (author)

  16. The use of neutron sources in nuclear reactors start-up after long shutdown periods

    International Nuclear Information System (INIS)

    Ponzoni Filho, P.; Borges, J.B.

    1990-01-01

    The reasons for the use of neutron sources in nuclear reactors, the different kinds of sources used and the alternatives to obtain the required minimum neutron counts in the external source range detectors after long maintenance and refueling periods are presented and discussed. The paper presents a formulation based in physics principles and experimental data, to calculate the power and time of reactor operation required to increase the effective fluence of secondary neutron sources. The option of using actinides produced during operation of the reactor as an additional source of neutrons is also discussed in depth to allow similar calculations in other kinds of reactors. The re-utilization of primary sources is considered as a last option. (author)

  17. Experimental measurements and theoretical simulations for neutron flux in self-serve facility of Dhruva reactor

    International Nuclear Information System (INIS)

    Rana, Y.S.; Mishra, Abhishek; Singh, Tej

    2016-06-01

    Dhruva is a 100 MW th tank type research reactor with natural metallic uranium as fuel and heavy water as coolant, moderator and reflector. The reactor is utilized for production of a large variety of radioisotopes for fulfilling growing demands of various applications in industrial, agricultural and medicinal sectors, and neutron beam research in condensed matter physics. The core consists of two on-power tray rods for radioisotope production and fifteen experimental beam holes for neutron beam research. Recently, a self-serve facility has also been commissioned in one of the through tubes in the reactor for carrying out short term irradiations. To get accurate information about neutron flux spectrum, measurements have been carried out in self-serve facility of Dhruva reactor. The present report describes measurement method, analysis technique and results. Theoretical estimations for neutron flux were also carried out and a comparison between theoretical and experimental results is made. (author)

  18. Surface erosion of fusion reactor components due to radiation blistering and neutron sputtering

    International Nuclear Information System (INIS)

    Das, S.K.; Kaminsky, M.

    1975-01-01

    Radiation blistering and neutron sputtering can lead to the surface erosion of fusion reactor components exposed to plasma radiations. Recent studies of methods to reduce the surface erosion caused by these processes are discussed

  19. Design of neutron radiography facility in pool for the reactor RA-10

    International Nuclear Information System (INIS)

    Peirone, M.; Coleff, A.; Sanchez, F.; Chiaraviglio, N.

    2013-01-01

    RA-10 project consists in the design and construction of a multipurpose reactor for multiple applications, including radioisotopes production, material testing and an in pool facility for neutron imaging. Neutron imaging is a powerful tool for studies of materials and offer several advantages among other attenuation-based techniques. In this study mechanical and neutronic requirements for the RA-10 in pool neutron imaging facility are described. The MCNP neutronic model and the mechanical design satisfying these requirements in a first engineering stage are described. (author)

  20. Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor

    International Nuclear Information System (INIS)

    Yavar, A.R.; Sarmani, S.B.; Wood, A.K.; Fadzil, S.M.; Radir, M.H.; Khoo, K.S.

    2011-01-01

    Determination of thermal to fast neutron flux ratio (f fast ) and fast neutron flux (φ fast ) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f fast and subsequently φ fast were determined using the absolute method. The f fast ranged from 48 to 155, and the φ fast was found in the range 1.03x10 10 -4.89x10 10 n cm -2 s -1 . These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.