WorldWideScience

Sample records for neutron nuclear reactors

  1. 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.

  2. 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

  3. 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.)

  4. 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...

  5. 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

  6. 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...

  7. 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

  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. 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

  10. 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)

  11. 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

  12. 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.)

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  1. 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.

  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. 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)

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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.)

  9. 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.)

  10. 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

  11. 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)

  12. 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.)

  13. 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

  14. 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

  15. 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

  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 analysis of the fuel of high temperature nuclear reactors; Analisis neutronico del combustible de reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    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)

  18. 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

  19. 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.

  20. Benchmark test of evaluated nuclear data files for fast reactor neutronics application

    International Nuclear Information System (INIS)

    Chiba, Go; Hazama, Taira; Iwai, Takehiko; Numata, Kazuyuki

    2007-07-01

    A benchmark test of the latest evaluated nuclear data files, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0, has been carried out for fast reactor neutronics application. For this benchmark test, experimental data obtained at fast critical assemblies and fast power reactors are utilized. In addition to comparing of numerical solutions with the experimental data, we have extracted several cross sections, in which differences between three nuclear data files affect significantly numerical solutions, by virtue of sensitivity analyses. This benchmark test concludes that ENDF/B-VII.0 predicts well the neutronics characteristics of fast neutron systems rather than the other nuclear data files. (author)

  1. 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)

  2. 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)

  3. Nuclear data requirements for fission reactor neutronics calculations

    International Nuclear Information System (INIS)

    Finck, P.

    1998-01-01

    The paper discusses current European nuclear data measurement and evaluation requirements for fission reactor technology applications and problems involved in meeting the requirements. Reference is made to the NEA High Priority Nuclear Data Request List and to the production of the new JEFF-3 library of evaluated nuclear data. There are requirements for both differential (or basic) nuclear data measurements and for different types of integral measurement critical facility measurements and isotopic sample irradiation measurements. Cross-section adjustment procedures are being used to take into account the simpler types of integral measurement, and to define accuracy needs for evaluated nuclear data

  4. 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.

  5. 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.)

  6. Nuclear reactor control device by vertical displacement of neutron absorber scram rods

    International Nuclear Information System (INIS)

    Defaucheux, Jacques; Pasqualini, Gilbert; Wiart, Albert; Martin, Jean.

    1981-01-01

    Nuclear reactor control system by vertical displacement of an assembly absorbing the neutrons inside a reactor core and drop of the absorbing assembly in maximum insertion position under the effect of its own weight for emergency shutdown. The absorbing assembly is secured to the bottom end of a vertical control rod, the displacement of which is actuated by an electro-magnetic device [fr

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

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1986-04-01

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

  8. 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

  9. Neutrons characterization of the nuclear reactor Ian-R1 of Colombia; Caracterizacion de los neutrones del reactor nuclear IAN-R1 de Colombia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez P, L. X.; Martinez O, S. A. [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Carretera Central del Norte Km. 1, Via Paipa, 150003 Tunja, Boyaca (Colombia); Vega C, H. R., E-mail: s.agustin.martinez@uptc.edu.co [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2014-08-15

    By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)

  10. 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)

  11. 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

  12. Neutron Resonance Theory for Nuclear Reactor Applications: Modern Theory and Practices.

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Richard N. [Argonne National Lab. (ANL), Argonne, IL (United States); Blomquist, Roger N. [Argonne National Lab. (ANL), Argonne, IL (United States); Leal, Luiz C. [Inst. de Radioprotection et de Sûrete Nucleaire (ISRN), Fontenay-aux-Roses (France); Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States)

    2016-09-24

    The neutron resonance phenomena constitute one of the most fundamental subjects in nuclear physics as well as in reactor physics. It is the area where the concepts of nuclear interaction and the treatment of the neutronic balance in reactor fuel lattices become intertwined. The latter requires the detailed knowledge of resonance structures of many nuclides of practical interest to the development of nuclear energy. The most essential element in reactor physics is to provide an accurate account of the intricate balance between the neutrons produced by the fission process and neutrons lost due to the absorption process as well as those leaking out of the reactor system. The presence of resonance structures in many major nuclides obviously plays an important role in such processes. There has been a great deal of theoretical and practical interest in resonance reactions since Fermi’s discovery of resonance absorption of neutrons as they were slowed down in water. The resonance absorption became the center of attention when the question was raised as to the feasibility of the self-sustaining chain reaction in a natural uranium-fueled system. The threshold of the nuclear era was crossed almost eighty years ago when Fermi and Szilard observed that a substantial reduction in resonance absorption is possible if the uranium was made into the form of lumps instead of a homogeneous mixture with water. In the West, the first practical method for estimating the resonance escape probability in a reactor cell was pioneered by Wigner et al in early forties.

  13. Neutronic design of a 22 MW MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.; Soleman, I.; Ghazi, N.

    2006-12-01

    The neutronic design calculations of a 22 MW MTR type nuclear research reactor are conducted in this project. This reactor type is selected by the Arab Atomic Energy Commission in a cooperated project. The design calculations are conducted in two methods: The deterministic method, solving the neutron transport and diffusion equations using the WIMSD4 and the CITATION codes, and the probabilistic method using the MCNP code. Good agreements are noticed between the results of the multiplication factor and the neutron flux distribution which prove the accuracy of our models using the two methods. (author)

  14. Neutronic design of a 22 MW MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.; Soleman, I.; Ghazi, N.

    2008-01-01

    The neutronic design calculations of a 22 MW MTR type nuclear research reactor are conducted in this project. This reactor type is selected by the Arab Atomic Energy Commission in a cooperated project. The design calculations are conducted in two methods: The deterministic method, solving the neutron transport and diffusion equations using the WIMSD4 and the CITATION codes, and the probabilistic method using the MCNP code. Good agreements are noticed between the results of the multiplication factor and the neutron flux distribution which prove the accuracy of our models using the two methods. (authors)

  15. Neutron dosimetry at nuclear power plants with light water reactors (LWR)

    International Nuclear Information System (INIS)

    Hofmann, B.; Schwarz, W.; Burgkhardt, B.; Piesch, E.

    1989-02-01

    During nuclear start-up of the Muelheim-Kaerlich nuclear power plant in 1986 the neutron radiation fields in the primary and auxiliary component rooms of the containment were investigated using the Single Sphere Albedo Technique and additional measurement techniques. For personnel monitoring albedo neutron dosemeters were used consisting of thermoluminescent detectors and track etch detectors combined with boron converters. Results: (1) The neutron radiation fields reach dose rate values up to 1000 mSv/h at the sleeves of the reactor coolant pipes, in the refuelling pool and the reactor cavity sump. The neutron component varies between 10% in the steam generator rooms up to 92% in the refuelling pool. (2) The mean value of the effective neutron energy at the different locations was found to be about 100 keV. Thermal neutrons contribute with about 10% to the area dose. (3) By direct intercomparisons and different evaluation methods of the Single Sphere Albedo Dosemeter it was shown, that rem-counters used within routine monitoring in the mixed radiation fields of the LWR overestimate the neutron dose rate only insignificantly (+20%) and are therefore usable for practical radiation protection work. (4) The sensitivity of albedo neutron dosemeters allows the detection of neutrons above 10 μSv. The contribution of neutrons to the total personnel dose was 25% in maximum. For the evaluation of albedo detectors a constant calibration factor can be applied. (orig./HP) [de

  16. Study on neutron diffusion and time dependence heat ina fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Vilhena, M.T. de.

    1988-01-01

    The purpose of this work is to model the neutron diffusion and heat transfer for a Fluidized Bed Nuclear Reactor and its solution by Laplace Transform Technique with numerical inversion using Fourier Series. Also Gaussian quadrature and residues techniques were applied for numerical inversion. The neutron transport, diffusion, and point Kinetic equation for this nuclear reactor concept are developed. A matricial and Taylor Series methods are proposed for the solution of the point Kinetic equation which is a time scale problem of Stiff type

  17. Neutron detection of the Triga Mark III reactor, using nuclear track methodology

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa, G., E-mail: espinosa@fisica.unam.mx; Golzarri, J. I. [Instituto de Física, Universidad Nacional Autónoma de México Circuito de la Investigación Científica, Ciudad Universitaria. México, DF (Mexico); Raya-Arredondo, R.; Cruz-Galindo, S. [Instituto Nacional de Investigaciones Nucleares (Mexico); Sajo-Bohus, L. [Universidad Simón Bolivar, Laboratorio de Física Nuclear, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 10{sup 12} n cm{sup −2} s{sup −1}, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer{sup ®} PADC as neutron detection material, covered by 3 mm Plexiglas{sup ®} as converter. After exposure, plastic detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.

  18. 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

  19. Validation of iron nuclear data for the neutron calculation of nuclear reactors

    International Nuclear Information System (INIS)

    Vaglio-Gaudard, C.

    2010-01-01

    The GEN-III and GEN-IV reactors will be equipped with heavy reflectors. However, the existing integral validation of the iron nuclear data in the latest JEFF3 European library in the frame of the neutron calculation of the heavy reflector is very partial: some results exist concerning fast reactors but there is no result corresponding to the LWR heavy reflector. No clear trend on the JEFF3 iron cross sections was brought into evidence up to now for fission reactor calculations. Iron nuclear data were completely re-evaluated in the JEFF3 library. Despite the fact that iron is widely used in the nuclear industry, large uncertainties are still associated with its nuclear data, particularly its inelastic cross section which is very important in the neutron slowing down. A validation of 56 Fe nuclear data was performed on the basis of the analysis of integral experiments. Two major critical experiments, the PERLE experiment and the Gas Benchmark, were interpreted with 3D reference Monte-Carlo calculations and the JEFF3.1.1 library. The PERLE experiment was recently performed in the EOLE zero-power facility (CEA Cadarache). This experiment is dedicated to heavy reflector physics in GEN-III light water reactors. It was especially conceived for the validation of iron nuclear data. The Gas Benchmark is representative of a Gas Fast Reactor with a stainless steel reflector (with no fertile blanket) in the MASURCA facility (CEA Cadarache). Radial traverses of reaction rates were measured to characterize flux attenuation at various energies in the reflector. The results of the analysis of both experiments show good agreement between the calculations and the measurements, which is confirmed by the analysis of complementary experiments (ZR-6M, MISTRAL4, CIRANO-ZONA2B). A process of re-estimating the 56 Fe nuclear data was implemented on the basis of feedback from these two experiments and the RDN code. This code relies on a non-linear regression method using an iterative

  20. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  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. Complex of two-dimensional multigroup programs for neutron-physical computations of nuclear reactor

    International Nuclear Information System (INIS)

    Karpov, V.A.; Protsenko, A.N.

    1975-01-01

    Briefly stated mathematical aspects of the two-dimensional multigroup method of neutron-physical computation of nuclear reactor. Problems of algorithmization and BESM-6 computer realisation of multigroup diffuse approximations in hexagonal and rectangular calculated lattices are analysed. The results of computation of fast critical assembly having complicated composition of the core are given. The estimation of computation accuracy of criticality, neutron fields distribution and efficiency of absorbing rods by means of computer programs developed is done. (author)

  3. 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)

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

    International Nuclear Information System (INIS)

    Friess, Friederike Renate

    2017-01-01

    should make them attractive for remote areas or electrical grids that are not large enough to support a standard-sized nuclear power plant. The last application of fast reactors this thesis investigates promises a solution to the problem of the radioactive waste from nuclear energy production. The separation of the spent fuel in different material streams (partitioning) and the irradiation of minor actinides in a fast neutron spectrum (transmutation) is claimed to solve this problem. Implementation of partitioning and transmutation (P and T) would require centuries of dedicated efforts, since several irradiation cycles and repeated reprocessing of the spent fuel elements between the irradiation cycles would be necessary. For all three applications, computer models of exemplary reactor systems were set up to perform criticality, depletion, and dose rate calculations. Based on the results, a specific critique on the viability of these fast reactor applications was conducted. Possible risks associated with their deployment were investigated.

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

    Energy Technology Data Exchange (ETDEWEB)

    Friess, Friederike Renate

    2017-07-12

    should make them attractive for remote areas or electrical grids that are not large enough to support a standard-sized nuclear power plant. The last application of fast reactors this thesis investigates promises a solution to the problem of the radioactive waste from nuclear energy production. The separation of the spent fuel in different material streams (partitioning) and the irradiation of minor actinides in a fast neutron spectrum (transmutation) is claimed to solve this problem. Implementation of partitioning and transmutation (P and T) would require centuries of dedicated efforts, since several irradiation cycles and repeated reprocessing of the spent fuel elements between the irradiation cycles would be necessary. For all three applications, computer models of exemplary reactor systems were set up to perform criticality, depletion, and dose rate calculations. Based on the results, a specific critique on the viability of these fast reactor applications was conducted. Possible risks associated with their deployment were investigated.

  6. Neutron physics and nuclear data measurements with accelerators and research reactors

    International Nuclear Information System (INIS)

    1988-08-01

    The report contains a collection of lectures devoted to the latest theoretical and experimental developments in the field of fast neutron physics and nuclear data measurements. The possibilities offered by particle accelerators and research reactors for research and technological applications in these fields are pointed out. Refs, figs and tabs

  7. Experimental researches of nuclear reactor neutron and gamma radiation scattering into the atmosphere

    International Nuclear Information System (INIS)

    Istomin, Yu.L.; Zelensky, D.I.; Cherepnin, Yu.S.; Orlov, Yu.V.; Netecha, M.E.; Avaev, V.N.; Vasel'ev, G.A.; Sakamoto, H.; Nomura, Y.; Naito, Y.

    1998-01-01

    In the report there are results of measuring radiation distribution on the caps of the RA and IWG.1M research reactors. Comparative analysis of the results is also in the report. There are neutron spectra in the interval of energies from 10 -9 to 13 MeV above RA and IWG.1M reactors. The spectra were measured with a set of activation detectors. Measurements were calculated to a nominal rate: for RA reactor - 300 kw, for IWG.1M - 7 MW. Thus, in the course of the experiment, vast experimental information relating to distribution of the RA and IWG.1M reactor gamma and neutron radiation scattered in the air for distances varying from 50 to 1000 m from the reactors has become available. The data obtained are to be used to verify the calculation codes and to validate the group nuclear constants

  8. Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

    International Nuclear Information System (INIS)

    Koenig, D.R.; Gido, R.G.; Brandon, D.I.

    1985-01-01

    The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations

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

    International Nuclear Information System (INIS)

    Pacio, Julio Cesar

    2008-01-01

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

  10. Results on the neutron energy distribution measurements at the RECH-1 Chilean nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aguilera, P., E-mail: paguilera87@gmail.com; Romero-Barrientos, J. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, La Reina, Santiago (Chile); Universidad de Chile, Dpto. de Física, Facultad de Ciencias, Las Palmeras 3425, Nuñoa, Santiago (Chile); Molina, F. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, La Reina, Santiago (Chile)

    2016-07-07

    Neutron activations experiments has been perform at the RECH-1 Chilean Nuclear Reactor to measure its neutron flux energy distribution. Samples of pure elements was activated to obtain the saturation activities for each reaction. Using - ray spectroscopy we identify and measure the activity of the reaction product nuclei, obtaining the saturation activities of 20 reactions. GEANT4 and MCNP was used to compute the self shielding factor to correct the cross section for each element. With the Expectation-Maximization algorithm (EM) we were able to unfold the neutron flux energy distribution at dry tube position, near the RECH-1 core. In this work, we present the unfolding results using the EM algorithm.

  11. 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.)

  12. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  13. Neutrons characterization of the nuclear reactor Ian-R1 of Colombia

    International Nuclear Information System (INIS)

    Gonzalez P, L. X.; Martinez O, S. A.; Vega C, H. R.

    2014-08-01

    By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)

  14. Control for nuclear reactor

    International Nuclear Information System (INIS)

    Ash, E.B.; Bernath, L.; Facha, J.V.

    1980-01-01

    A nuclear reactor is provided with several hydraulically-supported spherical bodies having a high neutron absorption cross section, which fall by gravity into the core region of the reactor when the flow of supporting fluid is shut off. (auth)

  15. GNES-R: Global nuclear energy simulator for reactors task 1: High-fidelity neutron transport

    International Nuclear Information System (INIS)

    Clarno, K.; De Almeida, V.; D'Azevedo, E.; De Oliveira, C.; Hamilton, S.

    2006-01-01

    A multi-laboratory, multi-university collaboration has formed to advance the state-of-the-art in high-fidelity, coupled-physics simulation of nuclear energy systems. We are embarking on the first-phase in the development of a new suite of simulation tools dedicated to the advancement of nuclear science and engineering technologies. We seek to develop and demonstrate a new generation of multi-physics simulation tools that will explore the scientific phenomena of tightly coupled physics parameters within nuclear systems, support the design and licensing of advanced nuclear reactors, and provide benchmark quality solutions for code validation. In this paper, we have presented the general scope of the collaborative project and discuss the specific challenges of high-fidelity neutronics for nuclear reactor simulation and the inroads we have made along this path. The high-performance computing neutronics code system utilizes the latest version of SCALE to generate accurate, problem-dependent cross sections, which are used in NEWTRNX - a new 3-D, general-geometry, discrete-ordinates solver based on the Slice-Balance Approach. The Global Nuclear Energy Simulator for Reactors (GNES-R) team is embarking on a long-term simulation development project that encompasses multiple laboratories and universities for the expansion of high-fidelity coupled-physics simulation of nuclear energy systems. (authors)

  16. Nitrogen determination in wheat by neutron activation analysis using fast neutron flux from a thermal nuclear reactor

    International Nuclear Information System (INIS)

    Ramirez G, T.

    1976-01-01

    This is a study of the technique for the determination of nitrogen and other elements in wheat flour through activation analysis with fast neutrons from a thermal nuclear reactor. The study begins with an introduction about the basis of the analytical methods, the equipment used in activation analysis and a brief description of the neutrons source. In the study are included the experiments carried out in order to determine the flux form in the site of irradiation, the N-13 half life and the interference due to the sample composition. (author)

  17. 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)

  18. Neutronics analysis of Dalat Nuclear Research Reactor by MVP/GMVP code

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Toru Obara

    2008-01-01

    The paper presents neutronics calculation for Dalat Nuclear Research Reactor (DNRR) to validate MVP/GMVP Code. Beside fresh core calculation, burnt core and burn up distribution were also carried out and compared with experimental data or result obtained from other codes. With complex geometry and operating history like DNRR, burn up calculation by Monte Carlo Method is the better choice owing to the use of exact geometry description and continuous neutron energy in calculation. The discrepancy between calculated data and experimental data is good to compare. By using Monte Carlo method, continuous neutron energy from JENDL3.3 library and combined with burn up calculation, MVP/GMVP Code is a very useful tool for reactor calculation. (author)

  19. Fast neutron nuclear reactor with lightened internal structure

    International Nuclear Information System (INIS)

    Artaud, R.; Aubert, M.; Renaux, C.

    1984-01-01

    The invention concerns an integrated type fast reactor. The inner vessel comprises two truncated shells, of which the large bases are connected either directly, or by a cylindrical shell of large diameter. The small base of the upper truncated shell is prolongated by a shell of small diameter and the small base of the lower truncated shell supports the reactor core. The invention allows the construction of simpler and less expansive fast reactors [fr

  20. 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)

  1. Characterization of the neutronic fields obtained by means of neutron traps inside the nuclear reactor core IPEN/MB-01

    International Nuclear Information System (INIS)

    Mura, Luiz Ernesto Credidio

    2011-01-01

    This paper presents the results of the neutron flux values obtained from the deployment of a Flux Trap of neutrons in the reactor core IPEN/MB-01. We analyzed several configurations of Flux Traps deployed in the reactor core IPEN/MB-01 in order to get elected to Flux Trap configuration more efficient. To characterize the neutron spectrum were irradiated in the center of the Flux Trap activation detectors of different materials (Au, Sc, In, Ti, Ni). The respective gamma spectroscopy of these elements after irradiation with and without cadmium cover, provided the experimental values of the nuclear reaction rates (saturation activity) by the target nuclei and their uncertainties used as input to the code SANDBP who calculated the energy spectrum of neutrons in the center of the 'Flux-Trap' in 50 energy groups, using the input spectra calculated at the irradiation position (center of the 'Flux Trap') by codes for Reactor Physics. The results found an increase in the thermal neutron flux in the center of the Flux Trap configuration 203 for the standard configuration (default) of about 350% without having the need to increase the reactor power. We also made comparisons between the spectra obtained by SANDBP deployed, compared to those calculated by MCNP-4C code and XSDRNPM. The spatial characterization of the thermal neutron flux is made with activation foils in the form of an infinitely dilute bulk alloy of 1% Au and 99% Al in some internal points of the configuration 203 (axially to Flux Trap a nd adjacent radial) and the results showed a significant increase in the magnitude of their values when compared to standard rectangular configuration. (author)

  2. Exxon nuclear neutronics design methods for pressurized water reactors. Supplement 2

    International Nuclear Information System (INIS)

    Skogen, F.B.; Stout, R.B.

    1977-01-01

    Modifications to the Exxon Nuclear PWR neutronic design calculational methods are presented as well as the results obtained when these improved methods are compared to reactor measurements. The basic PWR design tools remain unchanged; i.e., the XPOSE code is used for generating the basic nuclear parameters, the PDQ-7 code is used for calculating reactivity and x-y power distributions, and the XTG code is used for three-dimensional analysis. The recent start-up experiences at D. C. Cook Unit 1 and H. B. Robinson Unit 2 have provided a significant increase in the data base supporting the current ENC PWR neutronic methods. The verification comparisons contained in the supplement include reactor measurements from D. C. Cook Unit 1, Cycle 2; H. B. Robinson Unit 2, Cycles 4 and 5; Palisades Cycle 2, and R. E. Ginna, Cycle 7

  3. Some neutron absorbing elements and devices for fast nuclear reactors regulation systems

    International Nuclear Information System (INIS)

    Kervalishvili, P.J.

    2010-01-01

    It is shown that performed technological, physical-mechanical and radiation tests clearly indicate the prospects of using Neutron Absorbing Elements (NAE) based on B-10 and some rare-earth compounds during the creation of highly effective Control and Safety System (CSS) rods for fast neutron nuclear energetic reactors. Particular attention was paid to the development of new and upgrading of existing computing and real technologies for designing and preparing the optimizing NAE items characterized by all physical and strength properties for obtaining desirable operational parameters of CSS rods on their base

  4. Progress report on neutron activation analysis at Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Tuan, Nguyen Ngoc

    2003-01-01

    Neutron Activation Analysis (NAA) is one of most powerful techniques for the simultaneous multi-elements analysis. This technique has been studied and applied to analyze major, minor and trace elements in Geological, Biological and Environmental samples at Dalat Nuclear Research Reactor. At the sixth Workshop, February 8-11, 1999, Yojakarta, Indonesia we had a report on Current Status of Neutron Activation Analysis using Dalat Nuclear Research Reactor. Another report on Neutron Activation Analysis at the Dalat Nuclear Research Reactor also was presented at the seventh Workshop in Taejon, Korea from November 20-24, 2000. So in this report, we would like to present the results obtained of the application of NAA at NRI for one year as follows: (1) Determination of the concentrations of noble, rare earth, uranium, thorium and other elements in Geological samples according to requirement of clients particularly the geologists, who want to find out the mineral resources. (2) The analysis of concentration of radionuclides and nutrient elements in foodstuffs to attend the program on Asian Reference Man. (3) The evaluation of the contents of trace elements in crude oil and basement rock samples to determine original source of the oil. (4) Determination of the elemental composition of airborne particle in the Ho Chi Minh City for studying air pollution. The analytical data of standard reference material, toxic elements and natural radionuclides in seawater are also presented. (author)

  5. Progress report on neutron activation analysis at Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tuan, Nguyen Ngoc [Nuclear Research Institute, Dalat (Viet Nam)

    2003-03-01

    Neutron Activation Analysis (NAA) is one of most powerful techniques for the simultaneous multi-elements analysis. This technique has been studied and applied to analyze major, minor and trace elements in Geological, Biological and Environmental samples at Dalat Nuclear Research Reactor. At the sixth Workshop, February 8-11, 1999, Yojakarta, Indonesia we had a report on Current Status of Neutron Activation Analysis using Dalat Nuclear Research Reactor. Another report on Neutron Activation Analysis at the Dalat Nuclear Research Reactor also was presented at the seventh Workshop in Taejon, Korea from November 20-24, 2000. So in this report, we would like to present the results obtained of the application of NAA at NRI for one year as follows: (1) Determination of the concentrations of noble, rare earth, uranium, thorium and other elements in Geological samples according to requirement of clients particularly the geologists, who want to find out the mineral resources. (2) The analysis of concentration of radionuclides and nutrient elements in foodstuffs to attend the program on Asian Reference Man. (3) The evaluation of the contents of trace elements in crude oil and basement rock samples to determine original source of the oil. (4) Determination of the elemental composition of airborne particle in the Ho Chi Minh City for studying air pollution. The analytical data of standard reference material, toxic elements and natural radionuclides in seawater are also presented. (author)

  6. First records of thermal neutrons with the spectrometer for time of flight (TOF) in the RP-10 Nuclear Reactor

    International Nuclear Information System (INIS)

    Munive, M.; Baltuano, O; Soto, C; Ravello, Y

    2002-01-01

    To obtain the first spectrum of an emergent beam of neutrons of a nuclear reactor is the main parameter of the characterization in the use of this reactor; one of ways to get this spectrum is for the technique of time of flight, TOF, which registers the time that a neutron need to cover a certain distance, associating this time then to the kinetic energy of the neutron. The kinetic study of the beam of neutrons is carried out on neutron pulses that are generated by a revolving choke called Chopper; and the analysis in the time of the detected pulses is carried out for a system MCS. Using this technique it is achieved the record of the spectra in energy, or in wavelength , of the irradiation facilities No 2 and 4, and of the exit N o 5 of the thermal column of the Nuclear Reactor RP-10 of the Nuclear Center Oscar de la Guerra RACSO, Peru (au)

  7. Apparatus for controlling a nuclear reactor by vertical displacement of a unit absorbing neutrons

    International Nuclear Information System (INIS)

    Wiart, A.; Defaucheux, J.; Martin, J.; Pasqualini, G.

    1980-01-01

    Apparatus is described for controlling a nuclear reactor by vertical displacement of a unit absorbing neutrons, comprising, inside a sealed enclosure in communication with the interior of the reactor, a movable magnetic piece connected to a control shaft which is itself connected to the absorbent unit. This magnetic piece has at least two radial projections. The magnetic piece is displaced by an inductor with at least two pole shoes corresponding to the projections on the magnetic piece and allowing magnetic coupling between the inductor and the magnetic piece. The inductor and its displacement device are disposed outside the sealed enclosure. A control means allows the control shaft to be uncoupled from a member assuring its suspension so as to drop the absorbent unit in the event of emergency shutdown. The apparatus is particularly applicable to control rods of pressurized water nuclear reactors

  8. Electrical properties of InP irradiated by fast neutrons of a nuclear reactor

    International Nuclear Information System (INIS)

    Kolin, N.G.; Merkurisov, D.I.; Solov'ev, S.P.

    2000-01-01

    Electrophysical properties of InP single crystalline samples with different initial concentration of charge carriers have been studied in relation to irradiation conditions with fast neutrons of a nuclear reactor and subsequent heat treatments within the temperature range of 20-900 deg C. It has been shown that changes of the properties depend on the initial doping level. The annealing in the temperature range mentioned above results in the elimination of radiation defects. This makes possible to use the nuclear doping method for InP samples. In this respect the contribution of intermediate neutron reactions to the whole effect of the InP nuclear doping is estimated to be ∼ 10% [ru

  9. Non-traditional neutron activation analysis by use of a nuclear reactor

    International Nuclear Information System (INIS)

    Mukhammedov, S.

    2003-01-01

    Full text: Traditional reactor neutron activation analysis (NAA) based on (n, γ) - thermal neutron capture nuclear reaction has been developed into a reliable and powerful analytical method, for trace element analysis, allowing the determination of over 60 chemical elements, with good accuracy and low detection limits. Considering all possibilities of activation and a radiochemical separation of the indicator radionuclide, the majority of the elements of this group can be determined at the ppm concentration level and below. However, for solving a number of analytical problems NAA technique is not well suited or it cannot be used at all. An important limitation is that all light elements, some medium and heavy elements cannot be determined even at ppm concentration level by this method, for example, H, Be, Li, B, C, N, O, Ti, Nb, Pb, etc. Accurate determination of lithium, oxygen and other light elements in sub-microgram level is of importance in geochemical and material studies. Such examples are great many. On such instances, several non-traditional reactor activation analysis can be used which have increasingly been developed and applied to several fields of semiconductor industry, biology, geology in recent years. The purpose of this presentation is to review the modern status of non-traditional nuclear reactor activation analysis based on use of nuclear reactions excited by the flow of secondary charged particles which are produced by two methods. In first method the triton flow is produced by thermal neutrons flux which excites the nuclear reaction 6 Li(n, α)T on lithium. The neutron activation analysis associated with two consecutive reactions 6 Li(n, α)T + 16 O(T, n) 18 F is established to determine trace amounts either of lithium or of oxygen in different geological, ecological and technological samples. Besides, the triton flow can be used for the determination of other light elements, for instance, B, N, S, Mg. This nuclear reactor triton activation

  10. Characterization and adjustment of the neutron radiography facility of the RP-10 nuclear reactor

    International Nuclear Information System (INIS)

    Ravello R, Y.R.

    2001-01-01

    The main aim of this work was to characterize and adjust the neutron radiography facility of the RP-10 nuclear reactor, and therefore be able to offer with this technique services to the industry and research centers in general. This technique will be complemented with others such as x-rays and gamma radiography. First, the shielding capacity of the facility was analyzed, proving that it complies with the radiological safety requirements established by the radiological safety code. Then gamma filtration tests were conducted in order to implement the direct method for image formation, optical density curves were built according to the thickness of the gamma filter, the type of film and the type of irradiation. Also, the indirect method for image formation was implemented for two types of converters: indium and dysprosium. Growth curves for optical density were also made according to contact time between converter-film, for different types of films. The resolution of the facility was also analyzed using two methods: Klasens (1946) and Harms (1986). Harms method came closer to the resolution of the human eye when compared to the Klasens method. Finally, the application fields of neutron radiography are presented, including those conducted at the neutron radiography facility of the RP-10 nuclear reactor. With this work, the RP-10 neutron radiography facility is ready to offer inspection and research services

  11. An improved slow neutron spectrometer at nuclear research reactor et-r r-1. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    Abu El-Ela, M A [Reactor and Neutron Physics, Nuclear Research Center, AEA, Cairo (Egypt)

    1996-03-01

    An improved slow neutron selector has been aligned at channel number 6 of the nuclear research reactor ET-R R-1 Inshas. The flight path is 4 meter. The collimator-rotor-collimator system has the dimensions 0.3 x 2.5 x 70 cm with the rotor diameter 16 cm and 3 slits of 0.3 x 2.5 cm cross section. The rotor rotation rate varies between 600 r.p.m. the counting system has one of the best modern high electronic advanced technology time analyzer with minimum dwell time 2 sec, 8192 channels and a double detector inputs of TTL and NEG NIM standard pulses. The analyzer external triggering signals are of TTL standard type. A special design {sup 3} He detector for time of flight spectrometry has been used in the SNS. The reactor bare thermal neutron spectrum has been successfully measured, to show good agreement with the previous data. 6 figs.

  12. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  13. Testing FLUKA on neutron activation of Si and Ge at nuclear research reactor using gamma spectroscopy

    Science.gov (United States)

    Bazo, J.; Rojas, J. M.; Best, S.; Bruna, R.; Endress, E.; Mendoza, P.; Poma, V.; Gago, A. M.

    2018-03-01

    Samples of two characteristic semiconductor sensor materials, silicon and germanium, have been irradiated with neutrons produced at the RP-10 Nuclear Research Reactor at 4.5 MW. Their radionuclides photon spectra have been measured with high resolution gamma spectroscopy, quantifying four radioisotopes (28Al, 29Al for Si and 75Ge and 77Ge for Ge). We have compared the radionuclides production and their emission spectrum data with Monte Carlo simulation results from FLUKA. Thus we have tested FLUKA's low energy neutron library (ENDF/B-VIIR) and decay photon scoring with respect to the activation of these semiconductors. We conclude that FLUKA is capable of predicting relative photon peak amplitudes, with gamma intensities greater than 1%, of produced radionuclides with an average uncertainty of 13%. This work allows us to estimate the corresponding systematic error on neutron activation simulation studies of these sensor materials.

  14. Distribution of the thermal neutron field around the graphite reflector of the Dalat Nuclear Research 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

    Thermal neutron flux distributions around the graphite reflector of the Dalat Nuclear Research Reactor are determined by the method for neutron activating Cu foils. The major results are as follows: a/The axial distributions at the inner and outer margins of the graphite reflector have unsymmetrical shapes, similar to axial distributions in the core. There is a dissimilarity between the distribution curves at the inner margin and those at the outer margin of the reflector. b/ The radial distribution on the upper surface of the graphite reflector is measured and is described by the two-group neutron diffusion theory. The maximal value of the curve lies at the position of R{sub m}ax = 22.5 cm. c/ The distribution in the twenty water irradiation holes around the rotary specimen rack is obtained. (author). 3 refs., 5 figs., 1 tab.

  15. Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel

    International Nuclear Information System (INIS)

    Unesaki, H.; Isaka, S.; Nakagome, Y.

    2006-01-01

    Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel is investigated through cell burnup calculations using SRAC code system. Comparison of k ∞ and nuclide composition was made between the results obtained by JENDL-3.3, ENDF/B-VI.8 and JEFF3.0 for (U, Th)O 2 fuels as well as UO 2 fuels, with special interest on the burnup dependence of the neutronic characteristics. The impact of nuclear data library difference on k ∞ of (U, Th)O 2 fuels was found to be significantly large compared to that of UO 2 fuels. Notable difference was also found in nuclide concentration of TRU nuclides. (authors)

  16. Displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor

    International Nuclear Information System (INIS)

    Wang, Zujun; Huang, Shaoyan; Liu, Minbo; Xiao, Zhigang; He, Baoping; Yao, Zhibin; Sheng, Jiangkun

    2014-01-01

    The experiments of displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor are presented. The CMOS APS image sensors are manufactured in the standard 0.35 μm CMOS technology. The flux of neutron beams was about 1.33 × 10 8 n/cm 2 s. The three samples were exposed by 1 MeV neutron equivalent-fluence of 1 × 10 11 , 5 × 10 11 , and 1 × 10 12 n/cm 2 , respectively. The mean dark signal (K D ), dark signal spike, dark signal non-uniformity (DSNU), noise (V N ), saturation output signal voltage (V S ), and dynamic range (DR) versus neutron fluence are investigated. The degradation mechanisms of CMOS APS image sensors are analyzed. The mean dark signal increase due to neutron displacement damage appears to be proportional to displacement damage dose. The dark images from CMOS APS image sensors irradiated by neutrons are presented to investigate the generation of dark signal spike

  17. Neutron noise in nuclear reactors; Le bruit neutronique des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Blaquiere, A. [Institut National des Sciences et Techniques Nucleaires (France); Pachowska, R. [Universite Technique de Varsovie (Poland)

    1961-06-15

    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) [French] La puissance d'un reacteur nucleaire, dans les conditions du regime, est affectee de fluctuations dont les causes sont tres diverses. Ce comportement aleatoire rentre dans le cadre general de l'etude des 'bruits'. Entre autres sources ce bruit, nous analysons ici les fluctuations dues: a) a l'emission discontinue des neutrons provenant d'une source autonome; b) a la multiplication des neutrons au sein du reacteur. La methode que nous introduisons exploite les analogies entre les lois qui regissent un reacteur nucleaire au regime et certains systemes radioelectriques, en particulier les circuits a boucle de reaction. Le reacteur est caracterise par sa 'bande passante' et est decrit comme un systeme soumis a une succession d'impulsions aleatoires. Dans les conditions de fonctionnement non lineaires, l'effet du bruit neutronique est precise en utilisant une fonctionnelle non lineaire, ce qui relie cette theorie a

  18. Neutronic study of a nuclear reactor of fused salts; Estudio neutronico de un reactor nuclear de sales fundidas

    Energy Technology Data Exchange (ETDEWEB)

    Garcia B, F. B.; Francois L, J. L., E-mail: faviolabelen@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    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)

  19. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  20. Method and apparatus for controlling the neutron flux in nuclear reactors

    International Nuclear Information System (INIS)

    Minnick, L.E.

    1979-01-01

    A control rod assembly in a nuclear reactor that automatically scrams the reactor when a loss of coolant flow occurs and that can also control the level of neutron flux in the reactor is described. The control rod assembly includes a separator plate having an orifice through which the reactor coolant flows and a sealing surface around the orifice. The control rod in the assembly has a complementary sealing surface. When the control rod and separator plate are brought into contact, the differential pressure across the separator plate caused by the flow of the primary coolant through the reactor core retains the two sealing surfaces together. If the flow of coolant stops or the differential pressure across the separator plate decreases for any reason, the control rod drops by gravity and the reactor is scrammed. The control rod is also automatically dropped as a result of the lateral vibration of an earthquake or by the downward motion of the rod drive shaft, either of which will open the sealing surfaces and reduce the sealing pressure

  1. Improvement to surface lagging systems in a nuclear reactor, particularly of the fast neutron type

    International Nuclear Information System (INIS)

    Lemercier, Guy.

    1979-01-01

    Improvements to surface lagging systems in a nuclear reactor, particularly of the fast neutron kind. This system is composed of an assembly of panels each formed of a stack of metal fabric or trellis held against the surface to be protected, by a double fixing system comprising (a) a tubular component passing through a hole in the panel and applying it against the surface through a bearing plate, and (b) a bolt fitted in the centre of the tubular component, also secured to the surface and holding a washer capable of preventing the fall of the tubular component and the panel should the tubular component fracture [fr

  2. Preliminary neutronic design of spock reactor: A nuclear system for space power generation

    International Nuclear Information System (INIS)

    Burgio, N.; Santagata, A.; Cumo, M.; Fasano, A.; Frullini, M.

    2007-01-01

    Aim of this paper is to preliminary investigates the neutronic features of an upgrade of the MAUS [1] nuclear reactor whose core will be able to supply a thermoelectric converter in order to generate 30 kW of electricity for space applications. The neutronic layout of SPOCK (Space Power Core Ka) is a compact, MOX fuelled, liquid metal cooled and totally reflected fast reactor with a control system based on neutron absorption. Spock, that during the heart and launch operation must be maintained in sub-critical state, has to start up in the outer space at 40 K temperatures with the coolant in a solid state and it will reach the operating steady condition at the maximum temperature of 1300 K with the coolant in the liquid state. The main design goal is to maintains, in the operating conditions of a typical space mission, the control of the appropriate criticality margin versus temperature and coolant physical state. For this purpose, a neutronic/thermal-hydraulic calculation chain able to assists the entire design process must be set up. As preliminary recognition, MCNPX 2.5.0 and FLUENT calculations were carried out. The emerging key features of SPOCK are: an equilateral triangular mesh of 91 cylindrical UO 2 fuel rods with a Molybdenum clad ensured by two grids of the same material, cooled by liquid Sodium and contained in an AISI 316 L vessel. The core is totally wrapped by a Beryllium reflector that hosts six absorber (B 4 C) rotating control rods. The reactor shape is cylindrical (radius = 30 cm and height = 60 cm) with a total mass of 275 kg. The excess reactivity was of 5000 PCM at 1300 K. A preliminary evaluation of the control rods worth and a power spatial distribution were also discussed. Through the definition of an ideal reference K e ff value at 300 K for the actual SPOCK configuration, a sensitivity analysis on various cross sections data and material physical properties was performed for the given mission temperature range, allowing consideration on

  3. Numerical Solution of Fractional Neutron Point Kinetics Model in Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Nowak Tomasz Karol

    2014-06-01

    Full Text Available This paper presents results concerning solutions of the fractional neutron point kinetics model for a nuclear reactor. Proposed model consists of a bilinear system of fractional and ordinary differential equations. Three methods to solve the model are presented and compared. The first one entails application of discrete Grünwald-Letnikov definition of the fractional derivative in the model. Second involves building an analog scheme in the FOMCON Toolbox in MATLAB environment. Third is the method proposed by Edwards. The impact of selected parameters on the model’s response was examined. The results for typical input were discussed and compared.

  4. Improvement of Sodium Neutronic Nuclear Data for the Computation of Generation IV Reactors

    International Nuclear Information System (INIS)

    Archier, P.

    2011-01-01

    The safety criteria to be met for Generation IV sodium fast reactors (SFR) require reduced and mastered uncertainties on neutronic quantities of interest. Part of these uncertainties come from nuclear data and, in the particular case of SFR, from sodium nuclear data, which show significant differences between available international libraries (JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0). The objective of this work is to improve the knowledge on sodium nuclear data for a better calculation of SFR neutronic parameters and reliable associated uncertainties. After an overview of existing 23 Na data, the impact of the differences is quantified, particularly on sodium void reactivity effects, with both deterministic and stochastic neutronic codes. Results show that it is necessary to completely re-evaluate sodium nuclear data. Several developments have been made in the evaluation code Conrad, to integrate new nuclear reactions models and their associated parameters and to perform adjustments with integral measurements. Following these developments, the analysis of differential data and the experimental uncertainties propagation have been performed with Conrad. The resolved resonances range has been extended up to 2 MeV and the continuum range begins directly beyond this energy. A new 23 Na evaluation and the associated multigroup covariances matrices were generated for future uncertainties calculations. The last part of this work focuses on the sodium void integral data feedback, using methods of integral data assimilation to reduce the uncertainties on sodium cross sections. This work ends with uncertainty calculations for industrial-like SFR, which show an improved prediction of their neutronic parameters with the new evaluation. (author) [fr

  5. Role and use of nuclear theories and models in practical evaluation of neutron nuclear data needed for fission and fusion reactor design and other nuclear applications

    International Nuclear Information System (INIS)

    Prince, A.

    1975-01-01

    A review of the various nuclear models used in the evaluation of neutron nuclear data for fission and fusion reactors is presented. Computer codes embodying the principles of the relevant nuclear models are compared with each other and with experimental data. The regions of validity and limitations of the conceptual formalisms are also included, along with the effects of the numerical procedures used in the codes themselves. Conclusions and recommendations for future demands are outlined.15 tables, 15 figures, 90 references

  6. The role and use of nuclear theories and models in practical evaluation of neutron nuclear data needed for fission and fusion reactor design and other nuclear applications

    International Nuclear Information System (INIS)

    Prince, A.

    1976-01-01

    A review of the various nuclear models used in the evaluation of neutron nuclear data for fission and fusion reactors is presented. Computer codes embodying the principles of the relevant nuclear models are compared with each other and with experimental data. The regions of validity and limitations of the conceptual formalisms are also included, along with the effects of the numerical procedures used in the codes themselves. Conclusions and recommendations for future demands are outlined. (author)

  7. Neutron activation analysis in an industrial laboratory using an off-site nuclear reactor

    International Nuclear Information System (INIS)

    Osborn, T.W.; Broering, W.B.

    1977-01-01

    A multifunctional research laboratory, such as Procter and Gamble's Miami Valley Laboratories, requires elemental analyses on many materials. A general survey technique is important even if the information it provides is incomplete or is less precise than single element analyses. Procter and Gamble has developed neutron activation analysis (NAA) capabilities using a nuclear reactor several hundred miles away. The concentration of 40 to 50 elements can be determined in a variety of matrices. We have found NAA to be a powerful supplement to some of the more classical analytical techniques even without having an on-site neutron source. We have also found an automated data acquisition system to be essential for the successful application of NAA in an industrial laboratory

  8. Determination of nitrogen in wheat flour through Activation analysis using Fast neutron flux of a Thermal nuclear reactor; Determinacion de nitrogeno en harina de trigo mediante analisis por activacion empleando el flujo de neutrones rapidos de un reactor nuclear termico

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, T

    1976-07-01

    In this work is done a technical study for determining Nitrogen (protein) and other elements in wheat flour Activation analysis, with Fast neutrons from a Thermal nuclear reactor. Initially it is given an introduction about the basic principles of the methods of analysis. Equipment used in Activation analysis and a brief description of the neutron source (Thermal nuclear reactor). The realized experiments for determining the flux form in the irradiation site, the half life of N-13 and the interferences due to the sample composition are included too. Finally, the obtained results by Activation and the Kjeldahl method are tabulated. (Author)

  9. Measurements and simulations of electromagnetic perturbations on neutronic and thermodynamic control systems of nuclear reactors

    International Nuclear Information System (INIS)

    Lecompte, J.-C.; Buisson, Jacques.

    1981-01-01

    Neutronics and thermodynamics measuring devices realize the control and nuclear safety of the reactor and the core. They cause the emergency stop of the reactor if the informations given by the sensors are out of the normal operation tolerances. Defining and assuring a protection level sufficiently high against electromagnetic disturbances is essential for avoiding those measuring devices of giving erroneous indications in the environment where they are located. To obtain this result, one procceds according to the following three steps: a) In situ measurements of the ambient level of disturbances (oscilloscope analysis, statistical measurements). b) Measurement in laboratory of the immunity against the disturbances caused by every constituent of the device (electronic cables, connectors) and then the complete device. c) In situ simulation on the devices of the perturbations caused by the measured ambient level [fr

  10. Campbell's MSV method the neutron-gamma discrimination in mixed field of nuclear reactor

    International Nuclear Information System (INIS)

    Stankovic, S. J.; Loncar, B.; Avramovic, I.; Osmokrovic, P.

    2003-10-01

    In this paper it is carried out the analysis some capabilities of Campbell's MSV (Mean Square Value) measuring chain on base the principles derived by Campbell's theorem. Nevertheless, measurements have performed with digitized MSV method and results have compared related to they attained with classic measuring chain, when the mean value of signal from detector output has measured. In our case, detector element was uncompensated ionization chamber for mixed n-gamma fields. Thermal neutron flux, absorbed dose rate, equivalent dose rate and exposure rate in surrounding the reactor vessel of system HERBE, at nuclear reactor RB in 'VINCA' Institute, are determined. The examination of discrimination for gamma relate to neutron component in signal of detector output is performed whereby experimental work and the calculation according to linear theoretical model. The dependencies of changes for variance and mean value output detector signal versus four-decade change of fission reactor power, in range from 10 mW to 22W, are obtained. The advantage of MSV method is confirmed and concluded that the order n-gamma discrimination in MSV signal processing is around fifty times larger than classical measuring method. (author)

  11. 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.

  12. 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

  13. Real-time neutron radiography at the Iea-R1 m nuclear research reactor

    International Nuclear Information System (INIS)

    Menezes, M.O. de; Pugliesi, R.; Pereira, M.A.S.; Andrade, M.L.G.

    2003-01-01

    A LIXI (Light Intensifier X-ray Image) device has been employed in a real-time neutron radiography system. The LIXI is coupled to a video camera and the real-time images can be observed in a TV monitor, and processed in a computer. In order to get the real-time system operational, the neutron radiography facility installed at the IEA-R1 m nuclear research reactor of the IPEN-CNEN/S P has been optimized. The most important improvements were the neutron/gamma ratio, the effective energy of the neutron beam, decrease of the scattered radiation at the irradiation position, and the additional shielding of the video camera. Several one-frame as well as computer processed images are presented. The overall Modulation Transfer Function for the real-time system was obtained from the resolution parameter p = 0:44 +- 0:04 mm; the system sensitivity, evaluated for a Perspex step wedge, was determined and the average value is 0:70 +- 0:09 mm. (author)

  14. Karlsruhe Nuclear Research Center, Institute of Neutron Physics and Reactor Engineering. Progress report on research and development work in 1993

    International Nuclear Information System (INIS)

    1994-03-01

    The Institute of Neutron Physics and Reactor Engineering is concerned with research work in the field of nuclear engineering related to the safety of thermal reactors as well as with specific problems of fusion reactor technology. Under the project of nuclear safety research, the Institute works on concepts designed to drastically improve reactor safety. Apart from that, methods to estimate and minimize the radiological consequences of reactor accidents are developed. Under the fusion technology project, the Institute deals with neutron physics and technological questions of the breeding blanket. Basic research covers technico-physical questions of the interaction between light ion radiation of a high energy density and matter. In addition and to a small extent, questions of employing hydrogen in the transport area are studied. (orig.) [de

  15. Determination of nitrogen in wheat flour through Activation analysis using Fast neutron flux of a Thermal nuclear reactor

    International Nuclear Information System (INIS)

    Ramirez G, T.

    1976-01-01

    In this work is done a technical study for determining Nitrogen (protein) and other elements in wheat flour Activation analysis, with Fast neutrons from a Thermal nuclear reactor. Initially it is given an introduction about the basic principles of the methods of analysis. Equipment used in Activation analysis and a brief description of the neutron source (Thermal nuclear reactor). The realized experiments for determining the flux form in the irradiation site, the half life of N-13 and the interferences due to the sample composition are included too. Finally, the obtained results by Activation and the Kjeldahl method are tabulated. (Author)

  16. 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

  17. 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

  18. Neutronic design of a plutonium-thorium burner small nuclear reactor

    International Nuclear Information System (INIS)

    Hartanto, Donny

    2010-02-01

    A small nuclear reactor using thorium and plutonium fuel has been designed from the neutronic point of view. The thermal power of the reactor is 150 MWth and it is proposed to be used to supply electricity in an island in Indonesia. Thorium and plutonium fuel was chosen because in recent years the thorium fuel cycle is one of the promising ways to deal with the increasing number of plutonium stockpiles, either from the utilization of uranium fuel cycle or from nuclear weapon dismantling. A mixed fuel of thorium and plutonium will not generate the second generation of plutonium which will be a better way to incinerate the excess plutonium compared with the MOX fuel. Three kinds of plutonium grades which are the reactor grade (RG), weapon grade (WG), and spent fuel grade (SFG) plutonium, were evaluated as the thorium fuel mixture in the 17x17 Westinghouse PWR Fuel assembly. The evaluated parameters were the multiplication factor, plutonium depletion, fissile buildup, neutron spectrum, and temperature reactivity feedback. An optimization was also done to increase the plutonium depletion by changing the Moderator to Fuel Ratio (MFR). The computer codes TRITON (coupled NEWT and ORIGEN-S) in SCALE version 6 were used as the calculation tool for this assembly level. From the evaluation and optimization of the fuel assembly, the whole core was designed. The core was consisted of 2 types of thorium fuel with different plutonium grade and it followed the checkerboard loading pattern. A new concept of enriched burnable poison was also introduced to the core. The core life is 6.4 EFPY or 75 GWd/MTHM. It can burn up to 58% of its total mass of initial plutonium. VENTURE was used as the calculation tool for the core level

  19. A new detector for the measurement of neutron flux in nuclear reactors

    International Nuclear Information System (INIS)

    Koch, L.; Labeyrie, J.; Tarassenko, S.

    1958-01-01

    The detector described is designed for the instantaneous measurement of thermal neutron fluxes, in the presence of high γ ray activity; this detector can withstand temperatures as high as 500 deg. C. It is based on the following principle: radioactive atoms resulting from heavy-nucleus fission are carried by a gas flow to a detector recording their β and γ disintegration. Thermal neutron fluxes as low as few neutrons per cm 2 per second can be measured. This detector may be used to control a nuclear reactor, to plot the thermal flux distribution with an excellent definition (1 mm 2 ) for fluxes higher than 10 8 n/cm 2 /s. The time response of the system to a sharp variation of flux is limited, in case of large fluxes, to the transit time of the gas flow between the fission product emitter and the detector; of the order of one tenth of a sec per meter of piping. The detector may also be applied for spectroscopy of fission products eider than 0,1 s. (author) [fr

  20. Neutron flux parameters for k{sub 0}-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

    Energy Technology Data Exchange (ETDEWEB)

    Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Sarmani, S. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Masood, Z. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia)

    2011-02-15

    The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (f), epithermal neutron flux shape factor ({alpha}), thermal neutron flux ({phi}{sub th}) and epithermal neutron flux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

  1. Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor

    Science.gov (United States)

    Ródenas, José

    2017-11-01

    All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.

  2. Nuclear reactor monitoring system

    International Nuclear Information System (INIS)

    Drummond, C.N.; Bybee, R.T.; Mason, F.L.; Worsham, H.J.

    1976-01-01

    The invention pertains to an improved monitoring system for the neutron flux in a nuclear reactor. It is proposed to combine neutron flux detectors, a thermoelement, and a background radiation detector in one measuring unit. The spatial arrangement of these elements is fixed with great exactness; they are enclosed by an elastic cover and are brought into position in the reactor with the aid of a bent tube. The arrangement has a low failure rate and is easy to maintain. (HP) [de

  3. Simultaneous estimation of neutron density and reactivity in a nuclear reactor using a bank of Kalman filters

    International Nuclear Information System (INIS)

    Cortina, E.; D'Atellis, C.E.

    1990-01-01

    This paper reports on the problem of simultaneously estimating neutron density and reactivity while operating a nuclear reactor. It is solved by using a bank of Kalman filters as an estimator and applying a probabilistic test to determine which filter of the bank has the best performance

  4. 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.

  5. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  6. Nuclear physics and neutronics

    International Nuclear Information System (INIS)

    Paya, D.

    1997-01-01

    After a brief review of the beginnings of the nuclear reaction physics in France in the 40's and 50's, the experimentation neutronics and nuclear physics studies are related and their uses presented, which aims were to provide data for the study of the various reactor concepts and to study fundamental physics. Progressively, pure nuclear physics lost its links with neutronics, and its influence decreases more or less. Long life radioactive waste reprocessing is an important domain where it could regain its contribution

  7. Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shmelev, A. N.; Kulikov, G. G.; Kryuchkov, E. F.; Apse, V. A.; Kulikov, E. G. [National Research Nuclear Univ. MEPhI, Kashirskoe shosse, 31, 115409, Moscow (Russian Federation)

    2012-07-01

    The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

  8. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and Mars. These reactors require robust automatic control systems using low mass, rapid...

  9. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and/or Mars. These reactors require robust automatic control systems using low mass, rapid...

  10. Neutron physics and nuclear data measurements with accelerators and research reactors

    International Nuclear Information System (INIS)

    1985-08-01

    The report contains a collection of lectures devoted to the latest theoretical and experimental developments in the field of fast neutron measurements and in the studies of neutron interactions with nuclei. The possibilities offered by particle accelerators and research reactors for research and technological applications in these fields are pointed out

  11. Virtual nuclear reactor for education of nuclear reactor physics

    International Nuclear Information System (INIS)

    Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

    2008-01-01

    As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  13. 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

  14. Neutron spectra calculation and doses in a subcritical nuclear reactor based on thorium

    International Nuclear Information System (INIS)

    Medina C, D.; Hernandez A, P. L.; Hernandez D, V. M.; Vega C, H. R.; Sajo B, L.

    2015-10-01

    This paper describes a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a source of 252 Cf, whose dose levels in the periphery allows its use in teaching and research activities. The design was done by the Monte Carlo method with the code MCNP5 where the geometry, dimensions and fuel was varied in order to obtain the best design. The result is a cubic reactor of 110 cm side with graphite moderator and reflector. In the central part they have 9 ducts that were placed in the direction of axis Y. The central duct contains the source of 252 Cf, of 8 other ducts, are two irradiation ducts and the other six contain a molten salt ( 7 LiF - BeF 2 - ThF 4 - UF 4 ) as fuel. For design the k eff , neutron spectra and ambient dose equivalent was calculated. In the first instance the above calculation for a virgin fuel was called case 1, then a percentage of 233 U was used and the percentage of Th was decreased and was called case 2. This with the purpose to compare two different fuels working inside the reactor. In the case 1 a value was obtained for the k eff of 0.13 and case 2 of 0.28, maintaining the subcriticality in both cases. In the dose levels the higher value is in case 2 in the axis Y with a value of 3.31 e-3 ±1.6% p Sv/Q this value is reported in for one. With this we can calculate the exposure time of personnel working in the reactor. (Author)

  15. Experimental investigation of the neutron physics characteristics of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang; Thong, Ha Van [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The investigation of the neutron physics characteristics of the Dalat Reactor has obtained the results as follows: 1/ The effective fraction of delayed photoneutrons and the extraneous neutron source left after reactor shut down are measured. 2/ The lowest power levels of critical states of the reactor are determined. 3/The perturbation effect is investigated when a water column or a plexiglass rod is substituted for a fuel element. 4/ The relative axial and radial distributions of the thermal neutrons measured and the geometrical parameters of the core such as the inhomogeneous coefficients, the buckling, the effective height and radius, the extrapolated distances are obtained. 4/ The thermal neutron distributions are measured around the old graphite reflector. (author). 10 refs., 10 figs., 2 tabs.

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  18. 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

  19. Theoretical and experimental study of the dark signal in CMOS image sensors affected by neutron radiation from a nuclear reactor

    Science.gov (United States)

    Xue, Yuanyuan; Wang, Zujun; He, Baoping; Yao, Zhibin; Liu, Minbo; Ma, Wuying; Sheng, Jiangkun; Dong, Guantao; Jin, Junshan

    2017-12-01

    The CMOS image sensors (CISs) are irradiated with neutron from a nuclear reactor. The dark signal in CISs affected by neutron radiation is studied theoretically and experimentally. The Primary knock-on atoms (PKA) energy spectra for 1 MeV incident neutrons are simulated by Geant4. And the theoretical models for the mean dark signal, dark signal non-uniformity (DSNU) and dark signal distribution versus neutron fluence are established. The results are found to be in good agreement with the experimental outputs. Finally, the dark signal in the CISs under the different neutron fluence conditions is estimated. This study provides the theoretical and experimental evidence for the displacement damage effects on the dark signal CISs.

  20. Neutrons and Nuclear Engineering

    International Nuclear Information System (INIS)

    Ekkebus, Allen E.

    2007-01-01

    Oak Ridge National Laboratory hosted two workshops in April 2007 relevant to nuclear engineering education. In the Neutron Stress, Texture, and Phase Transformation for Industry workshop (http://neutrons.ornl.gov/workshops/nst2/), several invited speakers gave examples of neutron stress mapping for nuclear engineering applications. These included John Root of National Research Council of Canada, Mike Fitzpatrick of the UK's Open University, and Yan Gao of GE Global Research on their experiences with industrial and academic uses of neutron diffraction. Xun-Li Wang and Camden Hubbard described the new instruments at ORNL that can be used for such studies. This was preceded by the Neutrons for Materials Science and Engineering educational symposium (http://neutrons.ornl.gov/workshops/edsym2007). It was directed to the broad materials science and engineering community based in universities, industry and laboratories who wish to learn what the neutron sources in the US can provide for enhancing the understanding of materials behavior, processing and joining. Of particular interest was the presentation of Donald Brown of Los Alamos about using 'Neutron diffraction measurements of strain and texture to study mechanical behavior of structural materials.' At both workshops, the ORNL neutron scattering instruments relevant to nuclear engineering studies were described. The Neutron Residual Stress Mapping Facility (NRSF2) is currently in operation at the High Flux Isotope Reactor; the VULCAN Engineering Materials Diffractometer will begin commissioning in 2008 at the Spallation Neutron Source. For characteristics of these instruments, as well as details of other workshops, meetings, capabilities, and research proposal submissions, please visit http://neutrons.ornl.gov. To submit user proposals for time on NRSF2 contact Hubbard at hubbardcratornl.gov

  1. Determination of the neutron flux for the Yankee Rowe experiment in the Ford Nuclear Reactor

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Petrusha, L.

    1994-01-01

    Yankee Atomic Electric Company undertook a Test Irradiation Program at the Ford Nuclear Reactor of the University of Michigan. The program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials. The program was also intended to remove uncertainties in the existing reactor vessel fluence and damage predictions on the Yankee Rowe reactor vessel steel. Since this is the first in-core experiment of this type for the Ford Nuclear Reactor, the measurement of the reaction rate and the estimate of the fluence are presented

  2. In-core program for on line measurements of neutron, photon and nuclear heating parameters inside Jules Horowitz MTR reactor

    International Nuclear Information System (INIS)

    Lyoussi, A.; Reynard-Carette, C.

    2014-01-01

    Accurate on-line measurements of key parameters inside experimental channels of Material Testing Reactor are necessary to dimension the irradiation devices and consequently to conduct smart experiments on fuels and materials under suitable conditions. In particular the quantification of nuclear heating, a relevant parameter to reach adapted thermal conditions, has to be improved. These works focus on an important collaborative program between CEA and Aix-Marseille University called INCORE (Instrumentation for Nuclear radiations and Calorimetry On-line in Reactor) dedicated to the development of a new measurement methodology to quantify both nuclear heating and accurate radiation flux levels (neutrons and photons). The methodology, which is based on experiments carried out under irradiation conditions with a multi-sensor device (ionization chamber, fission chamber, gamma thermometer, calorimeter, SPND, SPGD) as well as works performed out-of nuclear/radiative environment on a reference sensor used to measure nuclear heating (calorimeter), is presented (authors)

  3. ENEA TRIGA RC-1 reactor activities in the fields of nuclear medicine and neutron radiography

    International Nuclear Information System (INIS)

    Chiesa, Gianni; Festinesi, Armando; Palomba, Mario; Rosa, Roberto; Rossi, Gabriela; Sangiovanni, Gino; Santoro, Emilio; Sedda, Antioco Franco; Storelli, Lucio

    2008-01-01

    In the last three years, TRIGA RC-1 plant staff is involved in collaborations with some roman hospitals for the production of particular radioisotopes for the diagnosis and therapy in the field of human cancer. Further, the thermal column of TRIGA reactor has been prepared for neutron radiography and tomography. For another channel, instruments and equipment above neutron radiography and tomography are in preparation phase. This paper includes an overview of the experimental equipment properly developed by TRIGA staff. (authors)

  4. Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia); Sarmani, S.B. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, 43000 Kajang, Selangor (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia); Radir, M.H. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, 43000 Kajang, Selangor (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia)

    2011-05-15

    Determination of thermal to fast neutron flux ratio (f{sub fast}) and fast neutron flux ({phi}{sub fast}) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f{sub fast} and subsequently {phi}{sub fast} were determined using the absolute method. The f{sub fast} ranged from 48 to 155, and the {phi}{sub fast} was found in the range 1.03x10{sup 10}-4.89x10{sup 10} n cm{sup -2} s{sup -1}. These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  7. Calculation of energetic characteristics of C-14 emitted from Beloyarsk nuclear power plant plume with fast neutron reactor

    Science.gov (United States)

    Kolotkov, Gennady A.; Penin, Sergei

    2017-11-01

    The paper examines an update of comparative analysis of radionuclides released into the atmosphere from Beloyarsk nuclear power plant with fast-neutron reactor for nine years in a row, from 2008 to 2016. It has been shown that the main radionuclides throw out into the atmosphere from Beloyarsk nuclear power plant are beta-active radionuclides. Based on data releases of the RPA "Typhoon", it has been conclude that radiation situation become worse insignificantly; beside on the new reactor BN-800 was put in operation in 2016. Using Spencer-Fano's equation, it was carried out the summary spectrum of emitted radionuclides. On example of Beloyarsk nuclear power plant, it was considered a question about ability of remote detection of raised radioactivity in the atmospheric radioactive plume. It has been shown that it possible to detect raised radioactivity in the emission plume from Beloyarsk nuclear power plant.

  8. Moderator for nuclear reactor

    International Nuclear Information System (INIS)

    Milgram, M.S.; Dunn, J.T.; Hart, R.S.

    1995-01-01

    This invention relates to a moderator for a nuclear reactor and more specifically, to a composite moderator. A moderator is designed to slow down, or thermalize, neutrons which are released during nuclear reactions in the reactor fuel. Pure or almost pure materials like light water, heavy water, beryllium or graphite are used singly as moderators at present. All these materials, are used widely. Graphite has a good mechanical strength at high temperatures encountered in the nuclear core and therefore is used as both the moderator and core structural material. It also exhibits a low neutron-capture cross section and high neutron scattering cross section. However, graphite is susceptible to attach by carbon dioxide and/or oxygen where applicable, and releases stress energy under certain circumstances, although under normal operating conditions these reactions can be controlled. (author). 1 tab

  9. Implementation into a CFD code of neutron kinetics and fuel pin models for nuclear reactor transient analyses

    International Nuclear Information System (INIS)

    Chen Zhao; Chen, Xue-Nong; Rineiski, Andrei; Zhao Pengcheng; Chen Hongli

    2014-01-01

    Safety analysis is an important tool for justifying the safety of nuclear reactors. The traditional method for nuclear reactor safety analysis is performed by means of system codes, which use one-dimensional lumped-parameter method to model real reactor systems. However, there are many multi-dimensional thermal-hydraulic phenomena cannot be predicated using traditional one-dimensional system codes. This problem is extremely important for pool-type nuclear systems. Computational fluid dynamics (CFD) codes are powerful numerical simulation tools to solve multi-dimensional thermal-hydraulics problems, which are widely used in industrial applications for single phase flows. In order to use general CFD codes to solve nuclear reactor transient problems, some additional models beyond general ones are required. Neutron kinetics model for power calculation and fuel pin model for fuel pin temperature calculation are two important models of these additional models. The motivation of this work is to develop an advance numerical simulation method for nuclear reactor safety analysis by implementing neutron kinetics model and fuel pin model into general CFD codes. In this paper, the Point Kinetics Model (PKM) and Fuel Pin Model (FPM) are implemented into a general CFD code FLUENT. The improved FLUENT was called as FLUENT/PK. The mathematical models and implementary method of FLUENT/PK are descripted and two demonstration application cases, e.g. the unprotected transient overpower (UTOP) accident of a Liquid Metal cooled Fast Reactor (LMFR) and the unprotected beam overpower (UBOP) accident of an Accelerator Driven System (ADS), are presented. (author)

  10. Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of neutron flux distribution and its reactivity ratio

    Energy Technology Data Exchange (ETDEWEB)

    Bitelli, Ulysses d' Utra; Aredes, Vitor O.G.; Mura, Luiz E.C.; Santos, Diogo F. dos; Silva, Alexandre P. da, E-mail: ubitelli@ipen.br, E-mail: vitoraredes@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    When compared to a rectangular parallelepiped configuration the cylindrical configuration of a nuclear reactor core has a better neutron economy because in this configuration the probability of the neutron leakage is smaller, causing an increase in overall reactivity in the system to the same amount of fuel used. In this work we obtained a critical cylindrical configuration with the control rods 89.50% withdraw from the active region of the IPEN/MB-01 core. This is the cylindrical configuration minimum possible excess of reactivity. Thus we obtained a cylindrical configuration with a diameter of only 28 fuel rods with lowest possible excess of reactivity. For this purpose, 112 peripheral fuel rods are removed from standard reactor core (rectangular parallelepiped of 28x28 fuel rods). In this configuration the excesses of reactivity is approximated 279 pcm. From there, we characterize the neutron field by measuring the spatial distribution of the thermal and epithermal neutron flux for the reactor operating power of 83 watts measured by neutron noise analysis technique and 92.08± 0.07 watts measured by activation technique [10]. The values of thermal and epithermal neutron flux in different directions, axial, radial north-south and radial east-west, are obtained in the asymptotic region of the reactor core, away from the disturbances caused by the reflector and control bar, by irradiating thin gold foils infinitely diluted (1% Au - 99% Al) with and without (bare) cadmium cover. In addition to the distribution of neutron flux, the moderator temperature coefficient, the void coefficient, calibration of the control rods were measured. (author)

  11. 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

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  14. Algorithmic developments and qualification of the ERANOS nuclear code and data system for the characterization of fast neutron reactors

    International Nuclear Information System (INIS)

    Rimpault, G.

    2003-09-01

    In this report, the author discusses the algorithmic and methodological developments in the field of nuclear reactor physics, and more particularly the developments of the ERALIB1/ERANOS nuclear code and data system for the calculation of core critical mass and power of sodium-cooled fast neutron reactors (Phenix and Super Phenix), and of the CAPRA 4/94 core. After a brief recall of nuclear data and methods used to determine critical masses and powers, the author discusses the interpretation of start-up experiments performed on Super-Phenix. The methodology used to characterize the uncertainties of these parameters is then applied to the calculation of the Super-Phenix critical mass and power distribution. He presents the approach chosen to define the validity domain of the ERANOS form

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-10-15

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

  16. A safe private nuclear tool-the miniature neutron source reactor

    International Nuclear Information System (INIS)

    Zhou Yongmao

    1987-01-01

    The prototype miniature neutron source reactor (MNSR) designed by China Institute of Atomic Energy has been operated successfully for more than 3 years and the practical experience enriches the original design idea. The commercial MNSR is under study design and develop in following aspects: 1. Prolonging the control rod cycle duration and core burn-up life; 2. Increasing the neutron flux per unit power. Obviously, the MNSR will show more advantages in extending application area and in providing the core using low envichment fuel. (Liu)

  17. 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)

  18. Modeling and Simulation Monte Carlo by the MCNP code for determining neutron parameters of the nuclear reactor-subcritical assembly in CNSTN

    International Nuclear Information System (INIS)

    Romdhani, Ibtissem

    2014-01-01

    As part of developing its nuclear infrastructure base, the National Science and Technology Center Nuclear (CNSTN) examines the technical feasibility of setting up a new installation of subcritical assembly. Our study focuses on determining the neutron parameters of a nuclear zero power reactor based on Monte Carlo simulation MCNP. The objective of the simulation is to model the installation, determine the effective multiplication factor, and spatial distribution of neutron flux.

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  20. 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

  1. Reactivity determination in accelerator driven nuclear reactors by statistics from neutron detectors (Feynman-Alpha Method)

    International Nuclear Information System (INIS)

    Ceder, M.

    2002-03-01

    The Feynman-alpha method is used in traditional nuclear reactors to determine the subcritical reactivity of a system. The method is based on the measurement of the mean number and the variance of detector counts for different measurement times. The measurement is performed while a steady-state neutron flux is maintained in the reactor by an external neutron source, as a rule a radioactive source. From a plot of the variance-to-mean ratio as a function of measurement time ('gate length'), the reactivity can be determined by fitting the measured curve to the analytical solution. A new situation arises in the planned accelerator driven systems (ADS). An ADS will be run in a subcritical mode, and the steady flux will be maintained by an accelerator based source. Such a source has statistical properties that are different from those of a steady radioactive source. As one example, in a currently running European Community project for ADS research, the MUSE project, the source will be a periodically pulsed neutron generator. The theory of Feynman-alpha method needs to be extended to such nonstationary sources. There are two ways of performing and evaluating such pulsed source experiments. One is to synchronise the detector time gate start with the beginning of an incoming pulse. The Feynman-alpha method has been elaborated for such a case recently. The other method can be called stochastic pulsing. It means that there is no synchronisation between the detector time gate start and the source pulsing, i.e. the start of each measurement is chosen at a random time. The analytical solution to the Feynman-alpha formula from this latter method is the subject of this report. We have obtained an analytical Feynman-alpha formula for the case of stochastic pulsing by two different methods. One is completely based on the use of the symbolic algebra code Mathematica, whereas the other is based on complex function techniques. Closed form solutions could be obtained by both methods

  2. Reactivity determination in accelerator driven nuclear reactors by statistics from neutron detectors (Feynman-Alpha Method)

    Energy Technology Data Exchange (ETDEWEB)

    Ceder, M

    2002-03-01

    The Feynman-alpha method is used in traditional nuclear reactors to determine the subcritical reactivity of a system. The method is based on the measurement of the mean number and the variance of detector counts for different measurement times. The measurement is performed while a steady-state neutron flux is maintained in the reactor by an external neutron source, as a rule a radioactive source. From a plot of the variance-to-mean ratio as a function of measurement time ('gate length'), the reactivity can be determined by fitting the measured curve to the analytical solution. A new situation arises in the planned accelerator driven systems (ADS). An ADS will be run in a subcritical mode, and the steady flux will be maintained by an accelerator based source. Such a source has statistical properties that are different from those of a steady radioactive source. As one example, in a currently running European Community project for ADS research, the MUSE project, the source will be a periodically pulsed neutron generator. The theory of Feynman-alpha method needs to be extended to such nonstationary sources. There are two ways of performing and evaluating such pulsed source experiments. One is to synchronise the detector time gate start with the beginning of an incoming pulse. The Feynman-alpha method has been elaborated for such a case recently. The other method can be called stochastic pulsing. It means that there is no synchronisation between the detector time gate start and the source pulsing, i.e. the start of each measurement is chosen at a random time. The analytical solution to the Feynman-alpha formula from this latter method is the subject of this report. We have obtained an analytical Feynman-alpha formula for the case of stochastic pulsing by two different methods. One is completely based on the use of the symbolic algebra code Mathematica, whereas the other is based on complex function techniques. Closed form solutions could be obtained by both methods

  3. Neutronic study of heavy nucleus produced in nuclear reactor fuel cycle

    International Nuclear Information System (INIS)

    Giacometti, A.

    1978-01-01

    Importance of minor actinides (U, Np, Pu, Am and Cm isotopes) PWR and fast neutron reactors and their associated fuel cycle is examined in this thesis. The amount of actinides formed in the various types of fuels or reactors are given. The different ways of formation and their importance are described. Modifications of the core reactivity due to actinides are shown. After a review of the fuel cycle (enrichment, fabrication, reprocessing, transport) actinide evolution outside the core is described and main problems concerning radioactivity in the different steps of the cycle or long term storage are underlined [fr

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

    International Nuclear Information System (INIS)

    Stark, K.

    1974-01-01

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

  5. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  6. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  7. A simple setup for neutron tomography at the Portuguese nuclear research reactor

    International Nuclear Information System (INIS)

    Pereira, M.A. Stanojev; Marques, J.G.; Pugliesi, R.

    2012-01-01

    A simple setup for neutron radiography and tomography was recently installed at the Portuguese Research Reactor. The objective of this work was to determine the operational characteristics of the installed setup, namely the irradiation time to obtain the best dynamic range for individual images and the spatial resolution. The performance of the equipment was demonstrated by imaging a fragment of a seventeenth-century decorative tile. (author)

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  9. 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.

  10. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  11. Nuclear fission sustainability with subcritical reactors driven by external neutron sources

    International Nuclear Information System (INIS)

    Lafuente, A.; Piera, M.

    2011-01-01

    Although nuclear breeder reactors are a promising way to enhance the potential energy currently retrievable from the Uranium reserves, they still have disadvantages because of their safety features (i.e. poor stabilizing mechanisms) and the security of their fuel cycle (diversion of Pu for non-civilian purposes). Loading natural nuclear fuels to a reactor and completely burning them without reprocessing would be ideal, however, this is not possible in critical reactors due to the limitations imposed by the maximum achievable burn-up. An alternative option to attain very high percentages of nuclear natural materials exploitation, while meeting other objectives of Nuclear Sustainability, could consist of using externally-driven subcritical reactors to reach the desired high burn-ups (of the order of 30% and more) without reprocessing. Such scheme would lead to an efficient exploitation of the available raw material, without any risk of proliferation. Exploring this type of reactor concept, this paper analyzes the different ways to accomplish this goal while identifying potential setbacks.

  12. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    Barjon, Robert

    1975-01-01

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude [fr

  13. Investigations of the influence of feedback and coupling effects on neutron noise in a nuclear reactor

    International Nuclear Information System (INIS)

    Vath, W.

    1975-01-01

    Investigations are described of the influence of a known feedback, i.e., a control loop with known transfer function, on the spectra of neutron chamber signals. Theoretical formulas for the spectra are derived using the point reactor model. These formulas were verified by noise measurements in a zero power reactor. Special attention is given to the noise generated by the control loop, the influence of this feedback noise on the spectra being verified experimentally. In large reactors space dependent transfer functions must be taken into account. As a first approximation to handle the spatial dependence, the afore-mentioned investigations were extended to the two-point reactor model. Corresponding experimental work was done for the Argonaut Reactor Karlsruhe (ARK) with a symmetrical two-slab core loading. Coolant boiling in a BWR was investigated. The boiling must be considered as a feedback mechanism as well as an extrnal reactivity perturbation. In order to simulate the steam bubble content, nitrogen gas was injected into the water moderator of the ARK. By modulating the total gas flow according to the instantaneous reactor power the feedback effect was simulated. The gas flow produced a band-limited, white reactivity noise. The upper break frequency could be used to determine the travelling time of the bubbles through the core. (author)

  14. First-principles investigation of neutron-irradiation-induced point defects in B4C, a neutron absorber for sodium-cooled fast nuclear reactors

    Science.gov (United States)

    You, Yan; Yoshida, Katsumi; Yano, Toyohiko

    2018-05-01

    Boron carbide (B4C) is a leading candidate neutron absorber material for sodium-cooled fast nuclear reactors owing to its excellent neutron-capture capability. The formation and migration energies of the neutron-irradiation-induced defects, including vacancies, neutron-capture reaction products, and knocked-out atoms were studied by density functional theory calculations. The vacancy-type defects tend to migrate to the C–B–C chains of B4C, which indicates that the icosahedral cage structures of B4C have strong resistance to neutron irradiation. We found that lithium and helium atoms had significantly lower migration barriers along the rhombohedral (111) plane of B4C than perpendicular to this plane. This implies that the helium and lithium interstitials tended to follow a two-dimensional diffusion regime in B4C at low temperatures which explains the formation of flat disk like helium bubbles experimentally observed in B4C pellets after neutron irradiation. The knocked-out atoms are considered to be annihilated by the recombination of the close pairs of self-interstitials and vacancies.

  15. Core Power Control of the fast nuclear reactors with estimation of the delayed neutron precursor density using Sliding Mode method

    International Nuclear Information System (INIS)

    Ansarifar, G.R.; Nasrabadi, M.N.; Hassanvand, R.

    2016-01-01

    Highlights: • We present a S.M.C. system based on the S.M.O for control of a fast reactor power. • A S.M.O has been developed to estimate the density of delayed neutron precursor. • The stability analysis has been given by means Lyapunov approach. • The control system is guaranteed to be stable within a large range. • The comparison between S.M.C. and the conventional PID controller has been done. - Abstract: In this paper, a nonlinear controller using sliding mode method which is a robust nonlinear controller is designed to control a fast nuclear reactor. The reactor core is simulated based on the point kinetics equations and one delayed neutron group. Considering the limitations of the delayed neutron precursor density measurement, a sliding mode observer is designed to estimate it and finally a sliding mode control based on the sliding mode observer is presented. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. Sliding Mode Control (SMC) is one of the robust and nonlinear methods which have several advantages such as robustness against matched external disturbances and parameter uncertainties. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability.

  16. Utilization of fusion neutrons in the tokamak fusion test reactor for blanket performance testing and other nuclear engineering experiments

    International Nuclear Information System (INIS)

    Caldwell, C.S.; Pettus, W.G.; Schmotzer, J.K.; Welfare, F.; Womack, R.

    1979-01-01

    In addition to developing a set of reacting-plasma/blanket-neutronics benchmark data, the TFTR fusion application experiments would provide operational experience with fast-neutron dosimetry and the remote handling of blanket modules in a tokamak reactor environment; neutron streaming and hot-spot information invaluable for the optimal design of penetrations in future fusion reactors; and the identification of the most damage-resistant insulators for a variety of fusion-reactor components

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  19. Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5; Calculo de activacion neutronica de barras de control de un reactor nuclear, utilizando MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Pena V, J.D.

    2016-07-01

    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)

  20. Technology of nuclear reactors

    International Nuclear Information System (INIS)

    Ravelet, F.

    2016-01-01

    This academic report for graduation in engineering first presents operation principles of a nuclear reactor core. It presents core components, atomic nuclei, the notions of transmutation and radioactivity, quantities used to characterize ionizing radiations, the nuclear fission, statistical aspects of fission and differences between fast and slow neutrons, a comparison between various heat transfer fluids, the uranium enrichment process, and different types of reactor (boiling water, natural uranium and heavy water, pressurized water, and fourth generation). Then, after having recalled the French installed power, the author proposes an analysis of a typical 900 MWe nuclear power plant: primary circuit, reactor, fuel, spent fuel, pressurizer and primary pump, secondary circuit, aspects related to control-command, regulation, safety and exploitation. The last part proposes a modelling of the thermodynamic cycle of a pressurized water plant by using an equivalent Carnot cycle, a Rankine cycle, and a two-phase expansion cycle with drying-overheating

  1. k0-measurements and related nuclear data compilation for (n, γ) reactor neutron activation analysis Pt. 3b

    International Nuclear Information System (INIS)

    Corte, F. de; Simonits, A.

    1989-01-01

    k 0 -factors and related nuclear data are tabulated for 112 radionuclides of interest in (n, γ) reactor neutron activation analysis. Whenever relevant, critical comments are made with respect to the accuracy of literature data for e. g. isotopic abundances, half-lives, absolute gamma-intensities and 2200 m · s -1 (n, γ) cross sections. As to the latter, a comparison is made with the values calculated from the experimentally determined k 0 -factors, by introduction of selected literature data for the input parameters. References to the table (79 pages) include 156 items. (author) 7 refs.; 1 tab

  2. Fertile assembly for a fast neutron nuclear reactor cooled by liquid sodium, with regulation of the cooling rate

    International Nuclear Information System (INIS)

    Pradal, L.; Berte, M.; Chiarelli, C.

    1985-01-01

    The assembly has a casing in which are arranged the fertile elements, the liquid sodium flowing through the casing along these elements. It includes several apertured diaphragms transverse to the rods to regulate the liquid sodium flow rate. At least one diaphragm, in its central part around its aperture, of a material soluble in liquid sodium, such as copper. The invention applies, more particularly, to fast neutron nuclear reactor having a heterogeneous core. The coolant flow can increase with time to match the increased power generated by the fertile assembly along its life [fr

  3. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  5. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Movafeghi, Amir

    2016-11-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  6. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    International Nuclear Information System (INIS)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser; Movafeghi, Amir

    2016-01-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  7. One procedure for determination of the neutron flux in the nuclear reactor fuel

    International Nuclear Information System (INIS)

    Bulovic, V.; Krtil, J.; Maksimovic, Z.; Martinc, R.

    1979-09-01

    Possibility of determination of the neutron flux in the fuel of a heavy water reactor has been examined. In determination of the flux an iterative procedure was used to compare calculated and measured contents of several fission products. The former contents were determined by calculation of the burning process balance and the latter by non-destructive gamma-spectrometric analysis of fuel. The obtained results prove the possibility of such determination of not only the average value of the flux but also of the change of its intensity during utilization of fuel (author) [sr

  8. Maximum neutron flux at thermal nuclear reactors; Maksimum neutronskog fluksa kod termalnih reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1968-10-15

    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. [Serbo-Croat] Savremeni reaktori za fizicka i tehnoloska istrazivanja predstavljaju tehnicki komplikovanu i skupu masinu. Iz tog razloga su opravdana nastojanja da se podesnim rasporedom goriva u jezgru reaktora dodje do sto ekonomicnijeg rjesenja. U literaturi postoji vise radova, cak i konkretnih realizacija u vidu reaktora sa reflektorom u centru, koji se bave odredjivanjem takve prostorne zavisnosti koncentracije goriva koja pod odredjenim uslovima daje najveci neutronski fluks. Zajednicki nedostatak svih pomenutih rjesenja je u tome sto se polazi od pretpostavljenih prostornih distribucija

  9. The radial distribution of the neutron field in the core of the Dalat Nuclear Research 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

    Determination of the radial distribution of the thermal neutron field in the core of the Dalat reactor is done by the Cu foil activation method. The measured data are fitted by the least square method to determine several physical parameters of the reactor, as follows: 1. Buckling B{sub r}{sup 2}=(84.6{+-}5.5)10{sup -4}/cm{sup 2}. 2. The effective radius R{sub eff}=(27.6{+-}1.0)cm. 3. The extrapolation distance {lambda}=(8.7{+-}1.0)cm. 4. The unequal coefficient of the effective multiplication K{sub r}=1.77{+-}0.11. (author). 2 refs., 4 figs., 1 tab.

  10. Measuring device for the spatial neutron density distribution within a nuclear reactor

    International Nuclear Information System (INIS)

    Fracke, A.; Wachtler, H.

    1974-01-01

    A solid probe in a pneumatic tube is lead from the core to a measuring device outside the pressure vessel and reversely, in order to measure the local neutron density distribution inside a reactor core. The activiable solid probe is in the form of a steel spiral spring with densely open coils and semi-spherical end pieces. A good curve negotiating characteristic of the measuring probe and defined duration times are secured in the reactor core. Furthermore, the interior of the spiral can be filled with a lubricating medium, e.g. molybdenum sulphite, so that a better sliding of the measuring probe into the tubes of the pneumatic tube is ensured. (DG) [de

  11. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  12. 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)

  13. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  14. Determination of contaminants in nuclear materials by measuring the capture gamma rays of thermal neutrons in a reactor internal geometry

    International Nuclear Information System (INIS)

    Suarez, A.A.

    1980-01-01

    A new method for analysis of impurities in nuclear fuel material was developed. Prompt gamma rays following thermal neutron capture, from a sample placed inside the research reactor were analyzed with a solid state high resolution detector. A number of improvements were introduced to improve the background-to-signal ratio, and the sensitivity of the method: use of collimeters for gamma rays and 6 Li 2 CO 3 filters to eliminate thermal neutrons from the beam were supplemented with the application of a pair spectrometer. Using a 42.5 cm 3 true coaxial Ge(Li) detector, and two optically separated NaI (Tl) scintillation detector, the sensitivity of the method for quantitative determination of impurities reached 30 p.p.m. The reproducibility of the results was better than 2%

  15. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

    International Nuclear Information System (INIS)

    March-Leuba, J.; Rey, J.M.

    1992-01-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of 'unexpected' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a 'new and improved' state of the art has emerged recently. (authors). 6 figs., 57 refs., 1 appendix

  16. Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries

    International Nuclear Information System (INIS)

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

    2010-01-01

    The aim of this study is to analyze the neutronic parameters of TRIGA Mark-II research reactor using the chain of NJOY-WIMS-CITATION computer codes based on evaluated nuclear data libraries CENDL-2.2 and JEFF-3.1.1. The nuclear data processing code NJOY99.0 has been employed to generate the 69 group WIMS library for the isotopes of TRIGA core. The cell code WIMSD-5B was used to generate the cross sections in CITATION format and then 3-dimensional diffusion code CITTATION was used to calculate the neutronic parameters of the TRIGA Mark-II research reactor. All the analyses were performed using the 7-group macroscopic cross section library. The CITATION test-runs using different cross section sets based on different models applied in WIMS calculations have shown a strong influence of those models on the final integral parameters. Some of the cells were specially treated with PRIZE options available in WIMSD-5B to take into account the fine structure of the flux gradient in the fuel-reflector interface region. It was observed that two basic parameters, the effective multiplication factor, k eff and the thermal neutron flux, were in good agreement among the calculated results with each other as well as the measured values. The maximum power densities at the hot spot were 1.0446E02 W/cc and 1.0426E02 W/cc for the libraries CENDL-2.2 and JEFF-3.1.1 respectively. The calculated total peaking factors 5.793 and 5.745 were compared to the original SAR value of 5.6325 as well as MCNP result. Consequently, this analysis will be helpful to enhance the neutronic calculations and also be used for the further thermal-hydraulics study of the TRIGA core.

  17. The final power calibration of the IPEN/MB-01 nuclear reactor for various configurations obtained from the measurements of the absolute average neutron flux

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandre Fonseca Povoa da, E-mail: alexandre.povoa@mar.mil.br [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto Credidio; Lima, Ana Cecilia de Souza; Betti, Flavio; Santos, Diogo Feliciano dos, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The use of neutron activation foils is a widely spread technique applied to obtain nuclear parameters then comparing the results with those calculated using specific methodologies and available nuclear data. By irradiation of activation foils and subsequent measurement of its induced activity, it is possible to determine the neutron flux at the position of irradiation. The power level during operation of the reactor is a parameter which is directly proportional to the average neutron flux throughout the core. The objective of this work is to gather data from irradiation of gold foils symmetrically placed along a cylindrically configured core which presents only a small excess reactivity in order to derive the power generated throughout the spatial thermal and epithermal neutron flux distribution over the core of the IPEN/MB-01 Nuclear Reactor, eventually lending to a proper calibration of its nuclear channels. The foils are fixed in a Lucite plate then irradiated with and without cadmium sheaths so as to obtain the absolute thermal and epithermal neutron flux. The correlation between the average power neutron flux resulting from the gold foils irradiation, and the average power digitally indicated by the nuclear channel number 6, allows for the calibration of the nuclear channels of the reactor. The reactor power level obtained by thermal neutron flux mapping was (74.65 ± 2.45) watts to a mean counting per seconds of 37881 cps to nuclear channel number 10 a pulse detector, and 0.719.10{sup -5} ampere to nuclear linear channel number 6 (a non-compensated ionization chamber). (author)

  18. Australia's new nuclear reactor

    International Nuclear Information System (INIS)

    Kemeny, L.

    2007-01-01

    On 19 and 20 April 2007, the Australian Nuclear Science and Technology Organisation (ANSTO) celebrated the recent commissioning of its new, world-class, OPAL (Open Pool Australian Lightwater) research reactor at the Lucas Heights. On the 19th, scientists, business leaders and academics were introduced to the reactor and its technical capacity for the manufacture of radiopharmaceuticals, its material science applications, its environmental services and its neutron scattering facilities for business applications. The formal OPAL opening function took place that evening and, on the 20th, Prime Minister John Howard visited ANSTO to be briefed about OPAL and to be shown the work being carried out at Lucas Heights

  19. The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. S.; Ju, H. G.; Jeon, T. H. and others

    2005-03-15

    A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well.

  20. The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

    International Nuclear Information System (INIS)

    Kim, K. S.; Ju, H. G.; Jeon, T. H. and others

    2005-03-01

    A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well

  1. Last Improvements of the CALMOS Calorimeter Dedicated to Thermal Neutron Flux and Nuclear Heating Measurements inside the OSIRIS Reactor

    Directory of Open Access Journals (Sweden)

    Carcreff H.

    2018-01-01

    Full Text Available Nuclear heating inside an MTR reactor needs to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. To improve the nuclear heating knowledge, an innovative calorimetric system CALMOS has been studied, manufactured and tested for the 70MWth OSIRIS reactor operated by CEA. This device is based on a mobile calorimetric probe which can be inserted in any in-core experimental location and can be moved axially from the bottom of the core to 1000 mm above the core mid-plane. Obtained results and advantages brought by the first CALMOS-1 equipment have been already presented. However, some difficulties appeared with this first version. A thermal limitation in cells did not allow to monitor nuclear heating up to the 70 MW nominal power, and some significant discrepancies were observed at high heating rates between results deduced from the calibration and those obtained by the “zero method”. Taking this feedback into account, the new CALMOS-2 calorimeter has been designed both for extending the heating range up to 13W.g-1 and for improving the “zero method” measurement thanks to the implementation of a 4-wires technique. In addition, the new calorimeter has been designed as a real operational measurement system, well suited to characterize and to follow the radiation field evolution throughout the reactor cycle. To meet this requirement, a programmable system associated with a specific software allows automatic complete cell mobility in the core, the data acquisition and the measurements processing. This paper presents the analysis of results collected during the 2015 comprehensive measurement campaign. The 4-wires technique was tested up to around a 4 W.g-1 heating level and allowed to quantify discrepancies between “zero” and calibration methods. Thermal neutron flux and nuclear heating measurements from CALMOS-1 and CALMOS-2 are compared. Thermal neutron flux distributions

  2. Last Improvements of the CALMOS Calorimeter Dedicated to Thermal Neutron Flux and Nuclear Heating Measurements inside the OSIRIS Reactor

    Science.gov (United States)

    Carcreff, H.; Salmon, L.; Lepeltier, V.; Guyot, J. M.; Bouard, E.

    2018-01-01

    Nuclear heating inside an MTR reactor needs to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. To improve the nuclear heating knowledge, an innovative calorimetric system CALMOS has been studied, manufactured and tested for the 70MWth OSIRIS reactor operated by CEA. This device is based on a mobile calorimetric probe which can be inserted in any in-core experimental location and can be moved axially from the bottom of the core to 1000 mm above the core mid-plane. Obtained results and advantages brought by the first CALMOS-1 equipment have been already presented. However, some difficulties appeared with this first version. A thermal limitation in cells did not allow to monitor nuclear heating up to the 70 MW nominal power, and some significant discrepancies were observed at high heating rates between results deduced from the calibration and those obtained by the "zero method". Taking this feedback into account, the new CALMOS-2 calorimeter has been designed both for extending the heating range up to 13W.g-1 and for improving the "zero method" measurement thanks to the implementation of a 4-wires technique. In addition, the new calorimeter has been designed as a real operational measurement system, well suited to characterize and to follow the radiation field evolution throughout the reactor cycle. To meet this requirement, a programmable system associated with a specific software allows automatic complete cell mobility in the core, the data acquisition and the measurements processing. This paper presents the analysis of results collected during the 2015 comprehensive measurement campaign. The 4-wires technique was tested up to around a 4 W.g-1 heating level and allowed to quantify discrepancies between "zero" and calibration methods. Thermal neutron flux and nuclear heating measurements from CALMOS-1 and CALMOS-2 are compared. Thermal neutron flux distributions, obtained with the Self-Power Neutron

  3. About neutron capture therapy method development at WWR-SM reactor in institute of Nuclear Physics of Uzbekistan Academy of Sciences

    International Nuclear Information System (INIS)

    Abdullaeva, G.A.; Baytelesov, S.A.; Dosimbaev, A.A.; Koblik, Yu.N.; Gritsay, O.O.

    2006-01-01

    Full text: Neutron capture therapy (NCT) is developing method of swellings treatment, on which specialists set one's serious hopes, as at its realization the practical possibilities of the effect on any swellings open. The essence of method is simple and lies in the fact that to the swelling enter preparation containing boron or gadolinium, which one have a large capture cross-section of the thermal and slow neutrons. Then the swelling is irradiated once with the slow (epithermal) neutron beam with fluency about 10 9 neutrons /sm 2 s for a short time and single. As a result of thermal neutrons capture by the boron (or gadolinium) nuclei secondary radiation which affecting swelling cells is emitted. NCT of oncologic diseases makes the specific demands to physical parameters of neutron beams. Now research reactors are often used for NCT. However, research reactor WWR-SM (INP, Uzbekistan AS, Tashkent) doesn't provide with the epithermal neutron beams and to develop this technique the reactor, first of all, needs for obtaining the epithermal neutron beams with energy spectrum in range from 1 eV up to 10 keV and with intensity ∼ 10 9 neutron /sm 2 s. Practically it is connected with upgrade of at least one of existed reactor channels, namely with equipping with the special equipment (filters), forming from the reactor spectrum the beam of necessary energy neutrons. It requires realization of preliminary model calculations, including calculations of capture cross-sections, of filters types and their geometrical parameters on the basis of optimal selected materials. Such calculations, as a rule, are carried out on the basis of Monte-Carlo method and designed software for calculation of nuclear reactor physical and technical characteristics [1]. In this work the calculation results of devices variants and problems discussion, related with possibility of WWR-SM reactor using for NCT are presented. (author)

  4. Nuclear reactor shield including magnesium oxide

    International Nuclear Information System (INIS)

    Rouse, C.A.; Simnad, M.T.

    1981-01-01

    An improvement is described for nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux. The reactor shielding includes means providing structural support, neutron moderator material, neutron absorber material and other components, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron

  5. Simulating the dynamics of the neutron flux in a nuclear reactor by locally recurrent neural networks

    International Nuclear Information System (INIS)

    Cadini, F.; Zio, E.; Pedroni, N.

    2007-01-01

    In this paper, a locally recurrent neural network (LRNN) is employed for approximating the temporal evolution of a nonlinear dynamic system model of a simplified nuclear reactor. To this aim, an infinite impulse response multi-layer perceptron (IIR-MLP) is trained according to a recursive back-propagation (RBP) algorithm. The network nodes contain internal feedback paths and their connections are realized by means of IIR synaptic filters, which provide the LRNN with the necessary system state memory

  6. Modelling of heterogenous neutron leakages in a nuclear reactor; Modelisation des fuites heterogenes de neutrons dans un reacteur nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Wohleber, X

    1997-11-17

    The TIBERE Model is a neutron leakage method based on B{sub 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{sub 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{sub 1} homogeneous leakage method, and with a shorter computational time. (author)

  7. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

    2004-01-01

    Requirements of neutron, thermohydraulic and safety analysis calculation are very important because of issuing new version of SAR for DNRR, research on construction of new research reactor and nuclear power plant. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor has been done in the frame work of research theme in the year 2002-2003. The purposes of the research are maintaining safety operation of the DNRR and enhancement of man power and calculation and safety analysis tool potential. (author)

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  9. Effects of neutron source ratio on nuclear characteristics of D-D fusion reactor blankets and shields

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Nakao, Yasuyuki; Ohta, Masao

    1978-01-01

    An examination is made of the dependence shown by the nuclear characteristics of the blanket and shield of D-D fusion reactors on S sub( d d)/S sub( d t), the ratio between the 2.45 MeV neutrons resulting from the D-D reaction and those of 14.06 MeV from the D-T reaction. Also, an estimate is presented of this neutron source ratio S sub( d d)/S sub( d t) for the case of D-D reactors, taken as an example. It is shown that an increase of S sub( d d)/S sub( d t) reduces the amount of nuclear heating per unit source neutron, while at the same time improving the shielding characteristics. This is accountable to lowering of the energy and penetrability of incident neutrons into the blanket brought about by the increase of S sub( d d)/S sub( d t). The value of S sub( d d)/S sub( d t) in a steady state D-D fusioning plasma core is estimated to be 1.46 -- 1.72 for an ion temperature ranging from 60 -- 180 keV. The reductions obtained on H sub( t)sup( b) (total heating in the blanket), H sub( t)sup( m g)/H sub( t)sup( b) (shielding indicator = ratio between total heating in superconducting magnet and that in the blanket) and phi sup( m g)/phi sup( w) (ratio of fast neutron fluxes between that at the magnet inner surface and that at the first wall inner surface) brought about by increasing S sub( d d)/S sub( d t) from unity to the value cited above do not differ to any appreciable extent, whichever is adopted among the design models considered here, the differences being at most about 10, 15 and 25%, respectively, for these three parameters. These results would broaden the validity of the conclusion derived in the previous paper for the case of S sub( d d)/S sub( d t) = 1.0, that the blanket-shield concept would appear to be the most suitable for D-D fusion reactors. (author)

  10. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  11. Proposal for a radiation shielding study aiming the implantation of neutrons beam shutter in the J-9 radiation channel of the Argonauta reactor of the Nuclear Engineering Institute

    Energy Technology Data Exchange (ETDEWEB)

    Xavier, Larissa R.P.; Cardoso, Domingos D’Oliveira, E-mail: larissa.xavier@cnen.gov.br, E-mail: domingosoliveiralvr71@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Ferreira, Francisco José de Oliveira; Voi, Dante Luiz, E-mail: fferreira@ien.gov.br, E-mail: dante@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Argonauta, the only nuclear research reactor situated in Rio de Janeiro, located at the Institute of Nuclear Engineering (IEN), regularly serves a network of users focused on research and development, and also provides its infrastructure for experimental classes and completion work course. Due to increasing demand for non-destructive thermal neutron assays and production of radioisotopes, there is a search for new procedures and/or devices that optimize users' exposure to neutrons. The implementation of mechanisms that allow access to the irradiation channels without the reactor being turned off and with a shielding configuration that limits the occupational doses at this location is very useful for the operation of the reactor. In order to achieve this, the present work proposes the establishment of a neutron beam shutter of the J-9 irradiation channel of the IEN's Argonauta reactor. In a first step, experimental measurements were made in the irradiation channel of the reactor using a BF3 detector, which is coupled to a spectrometer. In this phase, the neutron beam was aligned to the spectrometer, and different materials were used as shields, aiming the attenuation of the beam. To validate and/or change the configuration of the barrier that best meets the material irradiation needs, a second planned phase is involving the neutron flux simulation of the reactor and the various shields with different boundary conditions using the particle transport code, Monte Carlo N-Particle Extended (MCNP- X). (author)

  12. Proposal for a radiation shielding study aiming the implantation of neutrons beam shutter in the J-9 radiation channel of the Argonauta reactor of the Nuclear Engineering Institute

    International Nuclear Information System (INIS)

    Xavier, Larissa R.P.; Cardoso, Domingos D’Oliveira; Ferreira, Francisco José de Oliveira; Voi, Dante Luiz

    2017-01-01

    Argonauta, the only nuclear research reactor situated in Rio de Janeiro, located at the Institute of Nuclear Engineering (IEN), regularly serves a network of users focused on research and development, and also provides its infrastructure for experimental classes and completion work course. Due to increasing demand for non-destructive thermal neutron assays and production of radioisotopes, there is a search for new procedures and/or devices that optimize users' exposure to neutrons. The implementation of mechanisms that allow access to the irradiation channels without the reactor being turned off and with a shielding configuration that limits the occupational doses at this location is very useful for the operation of the reactor. In order to achieve this, the present work proposes the establishment of a neutron beam shutter of the J-9 irradiation channel of the IEN's Argonauta reactor. In a first step, experimental measurements were made in the irradiation channel of the reactor using a BF3 detector, which is coupled to a spectrometer. In this phase, the neutron beam was aligned to the spectrometer, and different materials were used as shields, aiming the attenuation of the beam. To validate and/or change the configuration of the barrier that best meets the material irradiation needs, a second planned phase is involving the neutron flux simulation of the reactor and the various shields with different boundary conditions using the particle transport code, Monte Carlo N-Particle Extended (MCNP- X). (author)

  13. Development of a nuclear data uncertainties propagation code on the residual power in fast neutron reactors

    International Nuclear Information System (INIS)

    Benoit, J.-C.

    2012-01-01

    This PhD study is in the field of nuclear energy, the back end of nuclear fuel cycle and uncertainty calculations. The CEA must design the prototype ASTRID, a sodium cooled fast reactor (SFR) and one of the selected concepts of the Generation IV forum, for which the calculation of the value and the uncertainty of the decay heat have a significant impact. In this study is developed a code of propagation of uncertainties of nuclear data on the decay heat in SFR. The process took place in three stages. The first step has limited the number of parameters involved in the calculation of the decay heat. For this, an experiment on decay heat on the reactor PHENIX (PUIREX 2008) was studied to validate experimentally the DARWIN package for SFR and quantify the source terms of the decay heat. The second step was aimed to develop a code of propagation of uncertainties: CyRUS (Cycle Reactor Uncertainty and Sensitivity). A deterministic propagation method was chosen because calculations are fast and reliable. Assumptions of linearity and normality have been validated theoretically. The code has also been successfully compared with a stochastic code on the example of the thermal burst fission curve of 235 U. The last part was an application of the code on several experiments: decay heat of a reactor, isotopic composition of a fuel pin and the burst fission curve of 235 U. The code has demonstrated the possibility of feedback on nuclear data impacting the uncertainty of this problem. Two main results were highlighted. Firstly, the simplifying assumptions of deterministic codes are compatible with a precise calculation of the uncertainty of the decay heat. Secondly, the developed method is intrusive and allows feedback on nuclear data from experiments on the back end of nuclear fuel cycle. In particular, this study showed how important it is to measure precisely independent fission yields along with their covariance matrices in order to improve the accuracy of the calculation of

  14. Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    After the reactor has reached physical criticality with the core configuration exempt from central neutron trap on 1 November 1983, the core configuration with a central neutron trap has been arranged in the reactor and the reactor has reached physical criticality with this core configuration at 17h48 on 18 December 1983. The integral worths of different control rods are determined with accuracy. 2 refs., 24 figs., 18 tabs

  15. Nuclear reactors

    International Nuclear Information System (INIS)

    Pearson, K.G.

    1977-01-01

    Reference is made to auxiliary means of cooling the nuclear fuel clusters used in light or heavy water cooled nuclear reactors. One method is to provide one or more spray cooling tubes. From holes in the side walls of those tubes coolant water may be sprayed laterally into the cluster against the rods. The flow of main coolant may thus be supplemented or even replaced by the auxiliary coolant. A difficulty, however, is that only those fuel rods close to a spray cooling tube can readily be reached by the auxiliary coolant. In the arrangement described, where the fuel rods are spaced apart by transverse grids, at least one of the interspaces between the grids is provided with an axially extending auxiliary coolant conduit having lateral holes through which an auxiliary coolant is sprayed into the cluster. A deflector is provided that extends from a transverse grid into a position in front of the holes and deflects auxiliary coolant on to parts of the fuel rods otherwise inaccessible to the auxiliary coolant. The construction of the deflector is described. (U.K.)

  16. Application of COMSOL in the solution of the neutron diffusion equations for fast nuclear reactors in stationary state; Aplicacion de COMSOL en la solucion de las ecuaciones de difusion de neutrones para reactores nucleares rapidos en estado estacionario

    Energy Technology Data Exchange (ETDEWEB)

    Silva A, L.; Del Valle G, E., E-mail: evalle@ipn.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico)

    2012-10-15

    This work shows an application of the program COMSOL Multi physics Ver. 4.2a in the solution of the neutron diffusion equations for several energy groups in nuclear reactors whose core is formed by assemblies of hexagonal transversal cut as is the cas of fast reactors. A reference problem of 4 energy groups is described of which takes the cross sections which are processed by means of a program that prepares the definition of the constants utilized in COMSOL for the generic partial differential equations that this uses. The considered solution domain is the sixth part of the core which is applied frontier conditions of reflection and incoming flux zero. The discretization mesh is elaborated in automatic way by COMSOL and the solution method is one of finite elements of Lagrange grade two. The reference problem is known as the Knk with and without control rod which led to propose the calculation of the effective multiplication factor in function of the control rod fraction from a value 0 (completely inserted control rod) until the value 1 (completely extracted control rod). Besides this the reactivity was determined as well as the change of this in function of control rod fraction. The neutrons scalar flux for each energy group with and without control rod is proportioned. The reported results show a behavior similar to the one reported in other works but using the discreet ordinates S{sub 2} approximation. (Author)

  17. An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

    Science.gov (United States)

    Merk, Bruno; Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J

    2017-01-01

    A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

  18. Fundamentals of Nuclear Reactor Physics

    CERN Document Server

    Lewis, E E

    2008-01-01

    This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

  19. The Oklo natural nuclear reactors: neutron parameters, age and duration of the reactions, uranium and fission products migrations

    International Nuclear Information System (INIS)

    Ruffenach, J.-C.

    1979-09-01

    Mass spectrometry and isotopic dilution technique are used in order to carry out, on various samples from the fossil nuclear reactors at Oklo, Gabon, isotopic and chemical analyses of some particular elements involved in the nuclear reactions: uranium, lead, bismuth, thorium, rare gases (krypton, xenon), rare earths (neodymium, samarium, europium, gadolinium, dysprosium), ruthenium and palladium. Interpretations of these analyses lead to the determination of many neutron parameters such as the neutron fluence received by the samples, the spectrum index, the conversion coefficient, and also the percentages of fissions due to uranium-238 and plutonium-239 and the total number of fissions relative to uranium. All these results make it possible to determine the age of the nuclear reactions by measuring the amounts of fission rare earths formed, i.e. 1.97 billion years. This study brings some informations to the general problem of radioactive wastes storage in deep geological formations, the storage of uranium, plutonium and many fission products having been carried out naturally, and for about two billion years [fr

  20. Calibration of the nuclear power channels for the cylindrical configuration of the IPEN/MB-01 reactor obtained from the measurements of the spatial neutron flux distribution in the reactor core through the irradiation of gold foils

    Energy Technology Data Exchange (ETDEWEB)

    Bitelli, Ulysses d' Utra; Silva, Alexandre F. Povoa da; Mura, Luiz Ernesto Credidio; Aredes, Vitor Ottoni Garcia; Santos, Diogo Feliciano dos, E-mail: ubitelli@ipen.br, E-mail: alexpovoa@yahoo.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The activation foil is one of the most used techniques to obtain and compare nuclear parameters from the nuclear data libraries, given by a gamma spectrometry system. Through the measurements of activity induced in the foils, it is possible to determine the neutron flux profile exactly where it has been irradiated. The power level operation of the reactor is a parameter directly proportional to the average neutron flux in the core. The objective of this work is to obtain, for a cylindrical configuration, the power generation through a spatial thermal neutron flux distribution in the core of IPEN/MB-01 Reactor, by irradiating gold foils positioned symmetrically into the core. They are put in a Lucite plate which will not interfere in the analysis of the neutron flux, because of its low microscopic absorption cross section for the analyzed neutrons. The foils are irradiated with and without cadmium covered small plates, to obtain the thermal and epithermal neutron flux, through specific equations. The correlation between the average power neutron flux, as a result of the foil's irradiation, and the average power digital neutron flux of the nuclear power channels, allows the calibration of the nuclear channels of the reactor. This same correlation was done in 2008 with the reactor in a rectangular configuration, which resulted in a specific calibration of the power level operation. This calibration cannot be used in the cylindrical configuration, because the nuclear parameters could change, which may lead to a different neutron profile. Furthermore, the precise knowledge of the power neutron flux in the core also validates the mathematics used to calculate the power neutron flux. (author)

  1. Application of COMSOL in the solution of the neutron diffusion equations for fast nuclear reactors in stationary state

    International Nuclear Information System (INIS)

    Silva A, L.; Del Valle G, E.

    2012-10-01

    This work shows an application of the program COMSOL Multi physics Ver. 4.2a in the solution of the neutron diffusion equations for several energy groups in nuclear reactors whose core is formed by assemblies of hexagonal transversal cut as is the cas of fast reactors. A reference problem of 4 energy groups is described of which takes the cross sections which are processed by means of a program that prepares the definition of the constants utilized in COMSOL for the generic partial differential equations that this uses. The considered solution domain is the sixth part of the core which is applied frontier conditions of reflection and incoming flux zero. The discretization mesh is elaborated in automatic way by COMSOL and the solution method is one of finite elements of Lagrange grade two. The reference problem is known as the Knk with and without control rod which led to propose the calculation of the effective multiplication factor in function of the control rod fraction from a value 0 (completely inserted control rod) until the value 1 (completely extracted control rod). Besides this the reactivity was determined as well as the change of this in function of control rod fraction. The neutrons scalar flux for each energy group with and without control rod is proportioned. The reported results show a behavior similar to the one reported in other works but using the discreet ordinates S 2 approximation. (Author)

  2. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  3. Aspects of the physics and chemistry of water radiolysis by fast neutrons and fast electrons in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCracken, D.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Tsang, K.T. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Laughton, P.J

    1998-09-01

    Detailed radiation physics calculations of energy deposition have been done for the coolant of CANDU reactors and Pressurized Water Reactors (PWRs). The geometry of the CANDU fuel channel was modelled in detail. Fluxes and energy-deposition rates for neutrons, recoil ions, photons, and fast electrons have been calculated using MCNP4B, WIMS-AECL, and specifically derived energy-transfer factors. These factors generate the energy/flux spectra of recoil ions from fast-neutron energy/flux spectra. The energy spectrum was divided into 89 discrete ranges (energy bins).The production of oxidizing species and net coolant radiolysis can be suppressed by the addition of hydrogen to the coolant of nuclear reactors. It is argued that the net dissociation of coolant by gamma rays is suppressed by lower levels of excess hydrogen than when dissociation is by ion recoils. This has consequences for the modelling of coolant radiolysis by homogeneous kinetics. More added hydrogen is required to stop water radiolysis by recoil ions acting alone than if recoil ions and gamma rays acted concurrently in space and time. Homogeneous kinetic models and experimental data suggest that track overlap is very inefficient in providing radicals from gamma-ray tracks to recombine molecular products in ion-recoil tracks. An inhomogeneous chemical model is needed that incorporates ionizing-particle track structure and track overlap. Such a model does not yet exist, but a number of limiting cases using homogeneous kinetics are discussed. There are sufficient uncertainties and contradictions in the data relevant to the radiolysis of reactor coolant that the relatively high CHC's (critical hydrogen concentration) observed in NRU reactor experiments (compared to model predictions) may be explainable by errors in fundamental data and understanding of water radiolysis under reactor conditions. The radiation chemistry program at CRL has been focused to generate quantitative water-radiolysis data in a

  4. 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)

  5. Nuclear rocket engine reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lanin, Anatoly

    2013-07-01

    Covers a new technology of nuclear reactors and the related materials aspects. Integrates physics, materials science and engineering Serves as a basic book for nuclear engineers and nuclear physicists. The development of a nuclear rocket engine reactor (NRER) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  7. 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

  8. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  10. Technique of nuclear reactors controls

    International Nuclear Information System (INIS)

    Weill, J.

    1953-12-01

    This report deal about 'Techniques of control of the nuclear reactors' in the goal to achieve the control of natural uranium reactors and especially the one of Saclay. This work is mainly about the measurement into nuclear parameters and go further in the measurement of thermodynamic variables,etc... putting in relief the new features required on behalf of the detectors because of their use in the thermal neutrons flux. In the domain of nuclear measurement, we indicate the realizations and the results obtained with thermal neutron detectors and for the measurement of ionizations currents. We also treat the technical problem of the start-up of a reactor and of the reactivity measurement. We give the necessary details for the comprehension of all essential diagrams and plans put on, in particular, for the reactor of Saclay. (author) [fr

  11. Neutron spectra calculation and doses in a subcritical nuclear reactor based on thorium; Calculo de espectros de neutrones y dosis en un reactor nuclear subcritico a base de Torio

    Energy Technology Data Exchange (ETDEWEB)

    Medina C, D.; Hernandez A, P. L.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: dmedina_c@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080A (Venezuela, Bolivarian Republic of)

    2015-10-15

    This paper describes a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a source of {sup 252}Cf, whose dose levels in the periphery allows its use in teaching and research activities. The design was done by the Monte Carlo method with the code MCNP5 where the geometry, dimensions and fuel was varied in order to obtain the best design. The result is a cubic reactor of 110 cm side with graphite moderator and reflector. In the central part they have 9 ducts that were placed in the direction of axis Y. The central duct contains the source of {sup 252}Cf, of 8 other ducts, are two irradiation ducts and the other six contain a molten salt ({sup 7}LiF - BeF{sub 2} - ThF{sub 4} - UF{sub 4}) as fuel. For design the k{sub eff}, neutron spectra and ambient dose equivalent was calculated. In the first instance the above calculation for a virgin fuel was called case 1, then a percentage of {sup 233}U was used and the percentage of Th was decreased and was called case 2. This with the purpose to compare two different fuels working inside the reactor. In the case 1 a value was obtained for the k{sub eff} of 0.13 and case 2 of 0.28, maintaining the subcriticality in both cases. In the dose levels the higher value is in case 2 in the axis Y with a value of 3.31 e-3 ±1.6% p Sv/Q this value is reported in for one. With this we can calculate the exposure time of personnel working in the reactor. (Author)

  12. Method of fueling for a nuclear reactor

    International Nuclear Information System (INIS)

    Igarashi, Takao.

    1983-01-01

    Purpose: To enable the monitoring of reactor power with sufficient accuracy, upon starting even without existence of neutron source in case of a low average burnup degree in the reactor core. Constitution: Each of fuel assemblies is charged such that neutron source region monitors for the start-up system in a reactor core neutron instrumentation system having nuclear fuel assemblies and a neutron instrumentation system are surrounded with 4 or 16 fuel assemblies of a low burnup degree. Then, the average burnup degree of the fuel assemblies surrounding the neutron source region monitors are increased than the reactor core burnup degree, whereby neutrons released from the peripheral fuels are increased, sufficient number of neutron counts can be obtained even with no neutron sources upon start-up and the reactor power can be monitored at a sufficient accuracy. (Sekiya, K.)

  13. One procedure for determination of the neutron flux in the nuclear reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bulovic, V; Krtil, J; Maksimovic, Z; Martinc, R [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1979-09-15

    Possibility of determination of the neutron flux in the fuel of a heavy water reactor has been examined. In determination of the flux an iterative procedure was used to compare calculated and measured contents of several fission products. The former contents were determined by calculation of the burning process balance and the latter by non-destructive gamma-spectrometric analysis of fuel. The obtained results prove the possibility of such determination of not only the average value of the flux but also of the change of its intensity during utilization of fuel (author) U radu je ispitivana mogucnost odredjivanja fluksa neutrona u gorivu teskovodnog nuklearnog reaktora. Pri odredjivanju fluksa koriscen je iterativni postupak, u kome se porede izracunati i izmereni sadrzaji nekoliko gama-radioaktivnih fisionih produkata. Prvi sadrzaji su odredjivani preko proracuna bilansa procesa sagorevanja, a drugi - nedestruktivnom gama-spektrometrijskom analizom goriva. Dobijeni rezultati potvrdjuju mogucnost ovakvog odredjivanja ne samo srednje vrednosti fluksa, vec i promene njegovog intenziteta u toku koriscenja goriva (author)

  14. Nuclear data for nuclear reactor analyses

    International Nuclear Information System (INIS)

    Pearlstein, S.

    1984-01-01

    A discussion of nuclear data is presented emphasizing to what extent data are known and to what accuracy. The principal data of interest is that for neutron cross-sections. The changing status of data, evaluated nuclear data files and data validation and improvement are described. Although the discussion relates to nuclear data for reactor analysis may of the results also apply to fusion, accelerator, shielding, biomedical, space and defense studies. (U.K.)

  15. Graphite materials for nuclear reactors

    International Nuclear Information System (INIS)

    Oku, Tatsuo

    1991-01-01

    Graphite materials have been used in the nuclear fission reactors from the beginning of the reactor development for the speed reduction and reflection of neutron. Graphite materials are used both as a moderator and as a reflector in the core of high temperature gas-cooled reactors, and both as a radiation shielding material and as a reflector in the surrounding of the core for the fast breeder reactor. On the other hand, graphite materials are being positively used as a first wall of plasma as it is known that low Z materials are useful for holding high temperature plasma in the nuclear fusion devices. In this paper the present status of the application of graphite materials to the nuclear fission reactors and fusion devices (reactors) is presented. In addition, a part of results on the related properties to the structural design and safety evaluation and results examined on the subjects that should be done in the future are also described. (author)

  16. A new detector for the measurement of neutron flux in nuclear reactors; Nouvelle methode de mesure des flux de neutrons dans les reacteurs atomiques

    Energy Technology Data Exchange (ETDEWEB)

    Koch, L; Labeyrie, J; Tarassenko, S [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The detector described is designed for the instantaneous measurement of thermal neutron fluxes, in the presence of high {gamma} ray activity; this detector can withstand temperatures as high as 500 deg. C. It is based on the following principle: radioactive atoms resulting from heavy-nucleus fission are carried by a gas flow to a detector recording their {beta} and {gamma} disintegration. Thermal neutron fluxes as low as few neutrons per cm{sup 2} per second can be measured. This detector may be used to control a nuclear reactor, to plot the thermal flux distribution with an excellent definition (1 mm{sup 2}) for fluxes higher than 10{sup 8} n/cm{sup 2}/s. The time response of the system to a sharp variation of flux is limited, in case of large fluxes, to the transit time of the gas flow between the fission product emitter and the detector; of the order of one tenth of a sec per meter of piping. The detector may also be applied for spectroscopy of fission products eider than 0,1 s. (author)Fren. [French] On decrit un appareil permettant la mesure instantanee des flux de neutrons thermiques accompagnes de flux intenses de rayons {gamma} et situes dans des enceintes pouvant etre portees a des temperatures superieures a 500 deg. C. On utilise la radioactivite des atomes resultant de la fission des noyaux lourds; ces atomes sont entraines par un courant gazeux vers un detecteur de radioactivite qui enregistre leurs desintegrations {beta} et {gamma}. On peut mesurer des flux partir de quelques neutrons thermiques par cm{sup 2} et par seconde. L'appareil permet de suivre la puissance d'un reacteur atomique, de tracer des cartes de densite de neutrons avec une tres bonne definition (1 mm{sup 2}) dans le cas de flux superieurs a 10{sup 8} cm{sup 2}/s. Le temps de reponse du systeme a une variation du flux de neutrons est limite, poes flux importants, par le temps de transit du gaz entre l'emetteur de produits de fission et le detecteur: soit quelques dizaines de

  17. 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.)

  18. Installation and development of neutron radiography in the nuclear reactor (IEAR-1) of the Instituto de Energia Atomica, Brazil

    International Nuclear Information System (INIS)

    Fuga, R.

    1979-01-01

    Investigations on the field of Neutron Radiography have been performed at the IEAR-1, swimming pool reactor utilizing a collimated neutron beam and the so-called photographic transfer method as a mean of detection. The test object (sample) is placed between the neutron source (reactor core) and the gold foil. The acitivity of its different points is the inverse measure of the neutrons absorbed in the test sample at the corresponding points. The activity distribution on the gold foil is determined again by exposing it to an X-ray film. A multichannel type collimator consisting of an assemblage of stainless steel tubes inside an aluminium mantle (tube) was used as a direction beam selector. Improvements have been introduced in respect to the reduction of angular divergence and neutron scattering. To improve further the quality of the radiographs another collimator type has been developed using boric acid as a neutron absorber and moderator. Flux measurements by means of gold foil activation at reactor positions of interest were necessary to eliminate errors originating of different neutron flux values. The dependence of film darkening upon the neutron flux and other factors have been discussed. Finally neutron-and gama-radiographs of the same objects were evaluated in comparison. (author) [pt

  19. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  20. Applications in nuclear data and reactor physics

    International Nuclear Information System (INIS)

    Cullen, D.E.; Muranaka, R.; Schmidt, J.

    1986-01-01

    This book presents the papers given at a conference on reactor kinetics and nuclear data collections. Topics considered at the conference included nuclear data processing, PWR core design calculations, reactor neutron dosimetry, in-core fuel management, reactor safety analysis, transients, two-phase flow, fuel cycles of research reactors, slightly enriched uranium, highly enriched uranium, reactor start-up, computer codes, and the transport of spent fuel elements

  1. Spatial and energy distributions of skyshine neutron and gamma radiation from nuclear reactors on the ground-air boundary

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Y.; Netecha, M.E.; Vasiliev, A.P.; Avaev, V.N.; Vasiliev, G.A. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Zelensky, D.I.; Istomin, Y.L.; Cherepnin, Y.S. [Institute of Atomic Energy of the National Nuclear Center of the Republic of Kazakhstan, Semipalatinsk-21 (Kazakhstan); Nomura, Y. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-03-01

    A set of measurements on skyshine radiation was conducted at two special research reactors. A broad range of detectors was used in the measurements to record neutron and gamma radiations. Dosimetric and radiometric field measurements of the neutrons and gamma quanta of the radiation scattered in the air were performed at distances of 50 to 1000 m from the reactor during different weather conditions. The neutron spectra in the energy range of 1 eV to 10 MeV and the gamma quanta spectra in the range of 0.1-10 MeV were measured. (author)

  2. Calibration of the nuclear power channels of the IPEN/MB-01 reactor obtained from the measurements of the spatial thermal neutron flux distribution in the reactor core through the irradiation of infinitely diluted gold foils

    International Nuclear Information System (INIS)

    Goncalves, Lucas Batista

    2008-01-01

    Several nuclear parameters are obtained through the gamma spectrometry of targets irradiated in a research reactor core and this is the case of the activation foils which make possible, through the measurements of the activity induced, to determine the neutron flux in the place where they had been irradiated. The power level operation of the reactor is a parameter directly proportional to the average neutron flux in the core. This work aims to get the power operation of the reactor through of spatial neutron flux distribution in the core of IPEN/MB-01 reactor by the irradiation of infinitely diluted gold foils and prudently located in its interior. These foils were made in the form of metallic alloy in concentration levels such that the phenomena of flux disturbance, as the self-shielding factors to neutrons become worthless. These activation foils has only 1% of dispersed gold atoms in an aluminium matrix content of 99% of this element. The irradiations of foils have been carried through with and without cadmium plate. The total correlation between the average thermal neutron flux obtained by irradiation of infinitely diluted activation foils and the average digital value of current of the nuclear power channels 5 and 6 (non-compensated ionization chambers - CINC), allow the calibration of the nuclear channels of the IPEN/MB-01 reactor. (author)

  3. An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

    Directory of Open Access Journals (Sweden)

    Bruno Merk

    Full Text Available A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

  4. An innovative way of thinking nuclear waste management – Neutron physics of a reactor directly operating on SNF

    Science.gov (United States)

    Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J.

    2017-01-01

    A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60’s for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient. PMID:28749952

  5. 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

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  7. Physical start up of the Dalat nuclear research reactor with the core configuration exempt from neutron trap

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    The nominal power of the reconstructed Dalat reactor is of 500 KW. After a meticulous preparation the Russian and Vietnamese teams have proceeded to the physical reactor start-up in November 1983 with the core configuration exempt from the neutron trap. The reactor has reached the physical criticality at 19h50 on 1 November 1983. The report delineates different steps of the start-up procedure. 2 refs., 3 figs., 7 tabs

  8. Compaction in optical fibres and fibre Bragg gratings under nuclear reactor high neutron and gamma fluence

    Energy Technology Data Exchange (ETDEWEB)

    Remy, L.; Cheymol, G. [CEA, French Nuclear Energy Commission, Nuclear Energy Division, DPC/SEARS/LISL Bat 467 CEA Saclay 91191 Gif/Yvette Cedex (France); Gusarov, A. [SCK.CEN - Belgian Nuclear Research center, Boeretang 200 2400 Mol (Belgium); Morana, A.; Marin, E.; Girard, S. [Universite de Saint-Etienne, Laboratoire Hubert Curien, UMR CNRS5516, 18, rue du Pr. Lauras, F-42000 Saint-Etienne (France)

    2015-07-01

    In the framework of the development by CEA and SCK.CEN of a Fabry Perot Sensor (FPS) able to measure dimensional changes in Material Testing Reactor (MTR), the first goal of the SAKE 1 (Smirnof extention - Additional Key-tests on Elongation of glass fibres) irradiation was to measure the linear compaction of single mode fibres under high fast neutron fluence. Indeed, the compaction of the fibre which forms one side of the Fabry Perot cavity, may in particular cause a noticeable measurement error. An accurate quantification of this effect is then required to predict the radiation-induced drift and optimize the sensor design. To achieve this, an innovative approach was used. Approximately seventy uncoated fibre tips (length: 30 to 50 mm) have been prepared from several different fibre samples and were installed in the SCK.CEN BR2 reactor (Mol Belgium). After 22 days of irradiation a total fast (E > 1 MeV) fluence of 3 to 5x10{sup 19} n{sub fast}/cm{sup 2}, depending on the sample location, was accumulated. The temperature during irradiation was 291 deg. C, which is not far from the condition of the intended FPS use. A precise measurement of each fibre tip length was made before the irradiation and compared to the post irradiation measurement highlighting a decrease of the fibres' length corresponding to about 0.25% of linear compaction. The amplitude of the changes is independent of the capsule, which could mean that the compaction effect saturates even at the lowest considered fluence. In the prospect of performing distributed temperature measurement in MTR, several fibre Bragg gratings written using a femtosecond laser have been also irradiated. All the gratings were written in radiation hardened fibres, and underwent an additional treatment with a procedure enhancing their resistance to ionizing radiations. A special mounting made it possible to test the reflection and the transmission of the gratings on fibre samples cut down to 30 to 50 mm. The comparison

  9. 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

  10. Random processes in nuclear reactors

    CERN Document Server

    Williams, M M R

    1974-01-01

    Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

  11. Neutron use in nuclear medicine

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; May, R.; Moss, R. [HFR-Unit, European Commission, IAM, Petten (Netherlands); Askienazy, S. [Departement Central de Medicine Nucleaire et Biophysique, Saint Antoine Hospital, Paris (France); Hildebrand, J. [Neurology Department, Erasmus Hospital, Brussels (Belgium)

    1999-07-01

    Neutrons produced by research reactors are being used in nuclear medicine and other medical applications in several ways. The High Flux Reactor (HFR) based in Petten (The Netherlands), owned by the European Commission, has been working increasingly in this field of health care for the European citizen. On the basis of this experience, a survey has been carried out on the main possibilities of neutrons used in nuclear medicine. The most important and most well known is the production of radioisotopes for diagnosis and therapy. Ten million patients receive nuclear medicine in Europe each year, with more than 8 million made with the products issued from research reactors. The survey of the market and the techniques (cyclotron, PET) shows that this market will continue to increase in the future. The direct use of reactors in medicine is actually made by the Boron Neutron capture Therapy (BNCT) for the treatment of glioblastoma, which kills about 15.000 people in Europe each year. For this promising technique, HFR is the most advanced for experimental possibilities and treatment studies. Medical research is also made in other promising fields: the use beam tubes for characterizing of prostheses and bio-medical materials, alpha-immuno therapy products, new types of radioisotopes, new types of illness to be treated by BNCT, etc. (author)

  12. Neutron use in nuclear medicine

    International Nuclear Information System (INIS)

    Guidez, J.; May, R.; Moss, R.; Askienazy, S.; Hildebrand, J.

    1999-01-01

    Neutrons produced by research reactors are being used in nuclear medicine and other medical applications in several ways. The High Flux Reactor (HFR) based in Petten (The Netherlands), owned by the European Commission, has been working increasingly in this field of health care for the European citizen. On the basis of this experience, a survey has been carried out on the main possibilities of neutrons used in nuclear medicine. The most important and most well known is the production of radioisotopes for diagnosis and therapy. Ten million patients receive nuclear medicine in Europe each year, with more than 8 million made with the products issued from research reactors. The survey of the market and the techniques (cyclotron, PET) shows that this market will continue to increase in the future. The direct use of reactors in medicine is actually made by the Boron Neutron capture Therapy (BNCT) for the treatment of glioblastoma, which kills about 15.000 people in Europe each year. For this promising technique, HFR is the most advanced for experimental possibilities and treatment studies. Medical research is also made in other promising fields: the use beam tubes for characterizing of prostheses and bio-medical materials, alpha-immuno therapy products, new types of radioisotopes, new types of illness to be treated by BNCT, etc. (author)

  13. Characterization and adjustment of the neutron radiography facility of the RP-10 nuclear reactor; Caracterizacion y puesta a punto de la facilidad de neutrografia del reactor nuclear RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Ravello R, Y R

    2001-07-01

    The main aim of this work was to characterize and adjust the neutron radiography facility of the RP-10 nuclear reactor, and therefore be able to offer with this technique services to the industry and research centers in general. This technique will be complemented with others such as x-rays and gamma radiography. First, the shielding capacity of the facility was analyzed, proving that it complies with the radiological safety requirements established by the radiological safety code. Then gamma filtration tests were conducted in order to implement the direct method for image formation, optical density curves were built according to the thickness of the gamma filter, the type of film and the type of irradiation. Also, the indirect method for image formation was implemented for two types of converters: indium and dysprosium. Growth curves for optical density were also made according to contact time between converter-film, for different types of films. The resolution of the facility was also analyzed using two methods: Klasens (1946) and Harms (1986). Harms method came closer to the resolution of the human eye when compared to the Klasens method. Finally, the application fields of neutron radiography are presented, including those conducted at the neutron radiography facility of the RP-10 nuclear reactor. With this work, the RP-10 neutron radiography facility is ready to offer inspection and research services.

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  16. Neutronic calculations for the reactor pressure vessel of Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Lerner, Ana M.; Madariaga, Marcelo R.

    1999-01-01

    In 1974 a surveillance program for the Atucha I nuclear power plant pressure vessel was initiated which included the construction of different types of specimens, distributed in 30 irradiation capsules located under the core at the lower part of some of the fuel channels. The capsules containing the irradiated specimens were withdrawn in two stages; the first set (SET 1) of 15 specimens in 1980 and the second one (SET 2) of the remaining 15, in 1987. Both fracture mechanic tests and dosimetry analysis were carried out by the designer (KWU) for SET1 and by the owner National Atomic Energy Commission (CNEA) for SET2. The calculations performed in the case of SET1 showed that there was a significant spectrum difference between the position where the specimens had been and the reactor pressure vessel (RPV) - inner surface (IS). It was established that the ratio of thermal flux (E 1 MeV) varied, approximately, from 1000 to 10 from the irradiation position to the RPV- IS. The purpose of this report is to show the calculations recently performed at the Nuclear Regulatory Authority, with particular emphasis on the difference in the results generated by the modification to sightly enriched fuel. A simplified 1-D calculations show that there is a slight increase (4% approximately) in the flux along the whole energy range. As it has already been mentioned, this is due, more than to the isotopic composition of the new fuel, to the difference in power density spatial distribution, which is a consequence of a different fuel management, necessary to preserve operational limits below their maximum allowed values with the same total thermal power generated. More detailed calculations are nevertheless foreseen in order to verify these first results. (author)

  17. Nuclear reactor kinetics and control

    International Nuclear Information System (INIS)

    Lewins, J.

    1978-01-01

    A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

  18. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

  19. Design and neutronic investigation of the Nano fluids application to VVER-1000 nuclear reactor with dual cooled annular fuel

    International Nuclear Information System (INIS)

    Ansarifar, G.R.; Ebrahimian, M.

    2016-01-01

    Highlights: • The change in neutronic parameters to the use of nanofluid as coolant is presented. • Nanoparticle deposition on fuel clad is investigated. • Radial and axial local power peaking factors are presented. • ZrO 2 and Al 2 O 3 have the lowest rate of K eff drop off. - Abstract: Nowadays, many efforts have been made to improve the efficiency of nuclear power plants. One of which is use of the dual cooled annular fuel which is an internally and externally cooled annular fuel with many advantages in heat transfer characteristics. Another is the use of nanoparticle/water (nanofluid) as coolant. In this paper, by combining these two methods, the change in neutronic parameters of the VVER-1000 nuclear reactor core with dual cooled annular fuel attributable to the use of nanoparticle/water (nanofluid) as coolant is presented. Optimization of type and volume fraction of nanoparticles in water that affect the safety enhancement of core primary parameters is intended in this study. Reactivity change, radial and axial local power peaking factors (LPPF), and the consequence of nanoparticle deposition on fuel clad are investigated. As a result of changing the effective multiplication factor and PPF calculations for six types of nanoparticles which have been studied extensively for their heat transfer properties including Alumina, Aluminum, Copper oxide, Copper, Titania, and Zirconia with different volume fractions, it can be concluded that at low concentration (0.03 volume fraction), Zirconia and Alumina are the optimum nanoparticles for normal operation. The maximum radial and axial PPF are found to be invariant to the type of nanofluid at low volume fractions. With an increase in nanoparticle deposition thickness on the outer and inner clad, a flux and K eff depression occurred and ZrO 2 and Al 2 O 3 have the lowest rate of drop off.

  20. Neutron fluence measurement in nuclear facilities

    International Nuclear Information System (INIS)

    Camacho L, M.E.

    1997-01-01

    The objective of present work is to determine the fluence of neutrons in nuclear facilities using two neutron detectors designed and built at Instituto Nacional de Investigaciones Nucleares (ININ), Mexico. The two neutron detectors are of the passive type, based on solid state nuclear tracks detectors (SSNTD). One of the two neutron detectors was used to determine the fluence distribution of the ports at the nuclear research reactor TRIGA Mark III, which belongs to ININ. In these facilities is important to know the neutron fluence distribution characteristic to carried out diverse kind of research activities. The second neutron detector was employed in order to carry out environmental neutron surveillance. The detector has the property to separate the thermal, intermediate and fast components of the neutron fluence. This detector was used to measure the neutron fluence at hundred points around the primary container of the first Mexican Nuclear Power plant 'Laguna Verde'. This last detector was also used to determine the neutron fluence in some points of interest, around and inside a low scattering neutron room at the 'Centro de Metrologia de Radiaciones Ionizantes' of the ININ, to know the background neutron field produced by the neutron sources used there. The design of the two neutron detector and the results obtained for each of the surveying facilities, are described in this work. (Author)

  1. Nuclear reactors; graphical symbols

    International Nuclear Information System (INIS)

    1987-11-01

    This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

  2. Guidebook to nuclear reactors

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

  3. Neutron tomography McClellan Nuclear Radiation Center (MNRC) TRIGA reactor

    International Nuclear Information System (INIS)

    Crump, J.C. III; Shields, K.; Newell, D.; Kiger, K.; Friday, C.

    1994-01-01

    At the MNRC Reactor Facility we are currently in the process of converting one of our real-time radiography bays into a neutron tomography facility. The facility's initial use will be for the determination of the amount of hydrogen in titanium compressor blades. An increase in the amount of hydrogen within the blade can result in embriddlement and later structural failure of the fan blade. The blades are 'baked' (heated to high temperatures to.force the hydrogen from the blade) after a number of flying hours. The goal of the system is to enable the end user to know when blades have reached potentially damaging levels of hydrogen (greater than 125 ppm). The system also needs to be able to process a large number of blades in a short time (i.e. a production mode). To achieve the goal of detecting the hydrogen at the appropriate levels; maxmize through put, and give the system flexibility to be used in other applications several unique features were incorporated. These design features included; a high resolution CCD camera for imaging, multiple high capacity turn tables to minimize system fixturing time, and automatic image analysis for hydrogen content. The high resolution CCD camera was selected based on previous work at the facility done by Dr. Eric McFarlan of University of California at Santa Barbara. He originally used the ST-4 and ST-6 which had small rectangular arrays of 375x242. For the system we chose the Princeton Instruments TE-1024K which has a segmentable (can be partitioned in smaller areas) array of 1024x1024. The camera also has the ability to digitize image information at 16 bit resolution or as low as 8 bit for high speed. Additionally, the CCD camera sensitivity (quantum efficiency) was matched closely to the wavelength of light emitted by the scintillation screen. The turn tables used have a weight capacity of 2000 pounds. This is well above the weight of the fan blade (less than one pound). To increase the through put we placed 3 blades on each

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  7. 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

  8. Reactor physics aspects of burning actinides in a nuclear reactor

    International Nuclear Information System (INIS)

    Hage, W.; Schmidt, E.

    1978-01-01

    A short review of the different recycling strategies of actinides other than fuel treated in the literature, is given along with nuclear data requirements for actinide build-up and transmutation studies. The effects of recycling actinides in a nuclear reactor on the flux distribution, the infinite neutron multiplication factor, the reactivity control system, the reactivity coefficients and the delayed neutron fraction are discussed considering a notional LWR or LMFBR as an Actinide Trasmutaton Reactor. Some operational problems of Actinide Transmutation reactors are mentioned, which are caused by the α-decay heat and the neutron sources of Actinide Target Elements

  9. Nuclear characteristic simulation device for reactor core

    International Nuclear Information System (INIS)

    Arakawa, Akio; Kobayashi, Yuji.

    1994-01-01

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

  10. 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)

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  12. Nuclear Reactor Sharing Program

    International Nuclear Information System (INIS)

    1994-01-01

    The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system

  13. 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.)

  14. Summary report of the IAEA advisory group meeting on nuclear data for neutron multiplication in fusion-reactor first-wall and blanket materials

    International Nuclear Information System (INIS)

    Muir, D.W.; Pashchenko, A.B.

    1992-09-01

    The present Report contains the Summary of the IAEA Advisory Group Meeting on Nuclear Data for Neutron Multiplication in Fusion-Reactor First-Wall and Blanket Materials, which was hosted by the Southwest Institute of Nuclear physics and Chemistry (SWINPC) at Chengdu, China and held from 19-21 November 1990. This AGM was organized by the IAEA Nuclear Data Section (NDS), with the cooperation and assistance of local organizers at the SWINPC. The papers which the participants prepared for and presented at the meeting will be published as an INDC report. (author)

  15. Intensive neutron source based on powerful electron linear accelerator LIA-30 and pulsed nuclear reactor FR-1

    Energy Technology Data Exchange (ETDEWEB)

    Bossamykin, V S; Koshelev, A S; Gerasimov, A I; Gordeev, V S; Grishin, A V; Averchenkov, V Ya; Lazarev, S A; Maslov, G N; Odintsov, Yu M [All-Russian Scientific Research Institute of Experimental Physics, Sarov (Russian Federation)

    1997-12-31

    Some results are given of investigations on joint operation modes of the linear induction electron accelerator LIA-30 ({approx} 40 MeV, {approx} 100 kA, {approx} 20 ns) and the pulsed reactor FR-1 with a compact metal core, aimed at achieving high intensity neutron fluxes. The multiplication factor Q for prompt neutrons in the FR-1 booster mode operation increased from 100 to 4500. The total output of prompt neutrons from FR-1 at Q = 2570 was 1.4 x 10{sup 16} 1/pulse with a pulse half width of {approx} 25 {mu}s. (author). 4 figs., 4 refs.

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    In the system described the fuel elements are arranged vertically in groups and are supported in such a manner as to tend to tilt them towards the center of the respective group, the fuel elements being urged laterally into abutment with one another. The elements have interlocking bearing pads, whereby lateral movement of adjacent elements is resisted; this improves the stability of the reactor core during refuelling operations. Fuel elements may comprise clusters of parallel fuel pins enclosed in a wrapper of hexagonal cross section, with bearing pads in the form of spline-like ribs located on each side of the wrapper and extending parallel to the longitudinal axis of the fuel element, being interlockable with ribs on pads of adjacent fuel elements. The arrangement is applicable to a reactor core in which fuel elements and control rod guide tubes are arranged in modules each of which comprises a cluster of at least three fuel elements, one of which is rigidly supported whilst the others are resiliently tilted towards the center of the cluster so as to lean on the rigidly supported element. It is also applicable to modules comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements located outside the clusters and also resiliently tilted towards the central voids, the latter being used to accommodate control rod guide tubes. The need for separate structural members to act as leaning posts is thus avoided. Such structural members are liable to irradiation embrittlement, that could lead to core failure. (U.K.)

  18. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  19. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR

    International Nuclear Information System (INIS)

    Moranchel y R, M.

    2012-01-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  20. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  2. 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)

  3. Process and device for identifying nuclear reactor neutron absorber rod etancheity defect

    International Nuclear Information System (INIS)

    Pelletier, J.; Parrat, D.

    1990-01-01

    For identifying defects in the sealing of neutron absorbing rods. The rod is placed in a pressure tight enclosure filled with a chemically agressive solution. After a time the pressure is released to allow the solution come out of the rod. An analysis of the solution allows the detection of radioactive isotopes of metals which are in the rod [fr

  4. Experimental study of angular neutron flux spectra on a slab surface to assess nuclear data and calculational methods for a fusion reactor design

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1988-06-01

    This paper presents an experimental approach to interpret the results of integral experiments for fusion neutronics research. The measurement is described of the angular neutron flux on a restricted area of slab assemblies with D-T neutron bombardment by using the time-of-flight (TOF) method with an NE213 neutron detector over an energy range from 0.05 to 15 MeV. A two bias scheme was developed to obtain an accurate detection efficiency over a wide energy range. The detector-collimator response function was introduced to define the restricted surface area and to determine the effective measured area. A series of measurements of the angular neutron flux on slabs of fusion blanket materials, i.e., Be, C, and Li 2 O, as functions of neutron leaking angle and slab thickness have been performed to examine neutron transport characteristics in bulk materials. The calculational analyses of the experimental results have been also carried out by using Monte Carlo neutron transport codes, i.e., MORSE-DD and MCNP. The existing nuclear data files, i.e., JENDL-3PR1, -3PR2, ENDF/B-IV and -V were tested by comparing with the experimental results. From the comparisons, the data on C and 7 Li in the present files are fairly sufficient. Those on beryllium, however, is insufficient for the estimation of high threshold reactions such as tritium production in a fusion reactor blanket design. It is also found that the total and elastic cross sections are more important for accurate predictions of neutronic parameters at deep position. The comparisons between the measured and calculated results provide information to understand the results of the previous integral experiments for confirmation of accuracy of fusion reactor designs. (author)

  5. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

    This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

  6. Comparison between different flux traps assembled in the core of the nuclear reactor IPEN/MB-01 by measuring of the thermal and epithermal neutron fluxes using activation foils

    International Nuclear Information System (INIS)

    Mura, Luiz Ernesto Credidio; Bitelli, Ulysses d'Utra; Mura, Luis Felipe Liambos; Carluccio, Thiago; Andrade, Graciete Simoes de

    2011-01-01

    The production of radioisotopes is one of the most important applications of nuclear research reactors. This study investigated a method called Flux Trap, which is used to increase the yield of production of radioisotopes in nuclear reactors. The method consists in the rearrangement of the fuel rods to allow the increase of the thermal neutron flux in the irradiation region inside the reactor core, without changing the standard reactor power level. Various configurations were assembled with the objective of finding the configuration with the highest thermal neutron flux in the region of irradiation. The method of activation analysis was used to measure the thermal neutron flux and determine the most efficient reactor core configuration . It was found that there was an increase in the thermal neutron flux of 337% in the most efficient configuration, which demonstrates the effectiveness of the method. (author)

  7. 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.)

  8. Transmutation of nuclear waste in nuclear reactors

    International Nuclear Information System (INIS)

    Abrahams, K.; Kloosterman, J.L.; Pilate, S.; Wehmann, U.K.

    1996-03-01

    The objective of this joint study of ECN, Belgonucleaire, and Siemens is to investigate possibilities for transmutation of nuclear waste in regular nuclear reactors or in special transmutation devices. Studies of possibilities included the limits and technological development steps which would be needed. Burning plutonium in fast reactors, gas-cooled high-temperature reactors and light water reactors (LWR) have been considered. For minor actinides the transmutation rate mainly depends on the content of the minor actinides in the reactor and to a much less degree on the fact whether one uses a homogeneous system (with the actinides mixed into the fuel) or a heterogeneous system. If one wishes to stabilise the amount of actinides from the present LWRs, about 20% of all nuclear power would have to be generated in special burner reactors. It turned out that reactor transmutation of fission products would require considerable recycling efforts and that the time needed for a substantial transmutation would be rather long for the presently available levels of the neutron flux. If one would like to design burner systems which can serve more light water reactors, a large effort would be needed and other burners (possibly driven by accelerators) should be considered. (orig.)

  9. 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)

  10. 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)

  11. 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

  12. Nuclear reactor power control device

    International Nuclear Information System (INIS)

    Koshi, Yuji; Sakata, Akira; Karatsu, Hiroyuki.

    1987-01-01

    Purpose: To control abrupt changes in neutron fluxes by feeding back a correction signal obtained from a deviation between neutron fluxes and heat fluxes for changing the reactor core flow rate to a recycling flow rate control system upon abrupt power change of a nuclear reactor. Constitution: In addition to important systems, that is, a reactor pressure control system and a recycling control system in the power control device of a BWR type power plant, a control circuit for feeding back a deviation between neutron fluxes and heat fluxes to a recycling flow rate control system is disposed. In the suppression circuit, a deviation signal is prepared in an adder from neutron flux and heat flux signals obtained through a primary delay filter. The deviation signal is passed through a dead band and an advance/delay filter into a correction signal, which is adapted to be fed back to the recycling flow rate control system. As a result, the reactor power control can be conducted smoothly and it is possible to effectively suppress the abrupt change or over shoot of the neutron fluxes and abrupt power change. (Kamimura, M.)

  13. Program MCU for Monte-Carlo calculations of neutron-physical characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    Abagyan, L.P.; Alekseev, N.I.; Bryzgalov, V.I.; Glushkov, A.E.; Gomin, E.A.; Gurevich, M.I.; Kalugin, M.A.; Majorov, L.V.; Marin, S.V.; Yhdkevich, M.S.

    1994-01-01

    A description of the MCU data modification is presented. The calculation results by the MCU-2 and MCU-3 codes are compared for the critical assemblies of a different reactor types. The full list of the critical assemblies calculation results obtained by all MCU code versions is given. 32 refs.; 32 tabs

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

    International Nuclear Information System (INIS)

    Benmansour, L.

    1992-01-01

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

  15. 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

  16. Dosimetric evaluation of semiconductor detectors for application in neutron dosimetry and microdosimetry in nuclear reactor and radiosurgical facilities

    International Nuclear Information System (INIS)

    Cardenas, Jose Patricio Nahuel

    2010-01-01

    The main objective of this research is the dosimetric evaluation of semiconductor components (surface barrier detectors and PIN photodiodes) for applications in dose equivalent measurements on low dose fields (fast and thermal fluxes) using an AmBe neutron source, the IEA-R1 reactor neutrongraphy facility (epithermal and thermal fluxes) and the Critical Unit facility IPEN/MB-01 (fast fluxes). As moderator compound to fast neutrons flux from the AmBe source was used paraffin and boron and polyethylene as converter for thermal and fast neutrons measurements. The resulting fluxes were used to the irradiation of semiconductor components (SSB - Surface Barrier Detector and PIN photodiodes). A mixed converter made of a borated polyethylene foil (Kodak) was also used. Monte Carlo simulation methodology was employed to evaluate analytically the optimal paraffin thickness. The obtained results were similar to the experimental data and allowed the evaluation of emerging neutron flux from moderator, as well as the fast neutron flux reaching the polyethylene covering the semiconductor sensitive surface. Gamma radiation levels were evaluated covering the whole detector with cadmium foil 1 mm thick, allowing thermal neutrons blockage and gamma radiation measurements. The IPEN/MB-01 facility was employed to evaluate the detector response for high neutron flux. The results were in good agreement with other studies published. Using the obtained spectra an approach to dose equivalent calculation was established. (author)

  17. Neutron nuclear data evaluation for actinide nucleic

    International Nuclear Information System (INIS)

    Chen Guochang; Yu Baosheng; Duan Junfeng; Ge Zhigang; Cao Wentian; Tang Guoyou; Shi Zhaomin; Zou Yubin

    2010-01-01

    The nuclear data with high accuracy for minor actinides are playing an important role in nuclear technology applications, including reactor design and operation, fuel cycle concepts, estimation of the amount of minor actinides in high burn-up reactors and the minor actinides transmutation. Through describe the class of nuclear data and nuclear date library, and introduce the procedure of neutron nuclear data evaluation. 234 U(n, f) and 237 Np(n, 2n) reaction experimental data evaluation was evaluated. The fission nuclear data are updated and improved. (authors)

  18. Assessment of the impact of neutronic/thermal-hydraulic coupling on the design and performance of nuclear reactors for space propulsion

    International Nuclear Information System (INIS)

    Aithal, S.M.; Aldemir, T.; Vafai, K.

    1994-01-01

    A series of studies has been performed to investigate the potential impact of the coupling between neutronics and thermal hydraulics on the design and performance assessment of solid core reactors for nuclear thermal space propulsion, using the particle bed reactor (PBR) concept as an example system. For a given temperature distribution in the reactor, the k eff and steady-state core power distribution are obtained from three-dimensional, continuous energy Monte Carlo simulations using the MCNP code. For a given core power distribution, determination of the temperature distribution in the core and hydrogen-filled annulus between the reflector and pressure vessel is based on a nonthermal equilibrium analysis. The results show that a realistic estimation of fuel, core size, and control requirements for PBRs using hydrogenous moderators, as well as optimization of the overall engine design, may require coupled neutronic/thermal-hydraulic studies. However, it may be possible to estimate the thermal safety margins and propellant exit temperatures based on power distributions obtained from neutronic calculations at room temperature. The results also show that, while variation of the hydrogen flow rate in the annulus has been proposed as a partial control mechanism for PBRs, such control mechanism may not be feasible for PBRs with high moderator-to-fuel ratios and hence soft core neutron spectra

  19. 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

  20. Assessment of the relative biological effectiveness of LVR-15 nuclear reactor neutron beam by a simple animal model

    Czech Academy of Sciences Publication Activity Database

    Mareš, Vladislav; Burian, J.; Prokeš, K.

    2002-01-01

    Roč. 78, - (2002), s. 5-19 ISSN 1212-3137 R&D Projects: GA MZd NC6473; GA MPO FD-K/048 Institutional research plan: CEZ:AV0Z5011922 Keywords : effectiveness of LVR-15 nuclear reactor * body irradiation * animal model Subject RIV: BG - Nuclear , Atomic and Molecular Physics, Colliders

  1. 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)

  2. Neutronic calculation of safety parameters for the RP-0 and RP-10 nuclear reactors

    OpenAIRE

    Lázaro, Gerardo; Deen, James R.; Woodruff, William L.

    2002-01-01

    Theoretical safety calculations were done with proved codes utilized by the staff of the RERTR program in the HEU to LEU core conversions. The studies were designed to evaluate the reactivity coefficients and kinetics parameters of the reactor involved in the evolution of peak power transients by reactivity insertion accidents. It was done to show the trend of these reactivity coefficients as a function of the core size and fuel depletion for RP10 cores. It was useful to get a better underst...

  3. Separated type nuclear superheating reactor

    International Nuclear Information System (INIS)

    Hida, Kazuki.

    1993-01-01

    In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

  4. Nuclear fusion and neutron processes

    International Nuclear Information System (INIS)

    Orlov, V.V.; Shatalov, G.E.; Sherstnev, K.E.

    1984-01-01

    Problems of providing development of the design of an experimental fusion reactor with necessary neutron-physical data are discussed. Isotope composition of spent fuel in the blanket of a hybride fusion reactor (HFR) is given. Neutron balance of the reactor with Li-blanket and neutron balance of the reactor with Pb-multiplier are disclosed. A simplified scheme of neutron and energy balance in the HFR blanket is given. Development and construction of the experimental power reactor is shown to become the nearest problem of the UTS program. Alongside with other complex physical and technical problems solution of this problem requires realization of a wide program of neutron-physical investigations including measurements with required accuracy of neutron cross sections, development of methodical, program and constant basis of neutron calculations and macroscopic experiments on neutron sources

  5. Validation of The Deterministic Diffusion Method For The Neutronic Calculations of Thermal Research Reactors of TRIGA-Type Using The Wisdom-IAEA-69 Nuclear Data Library

    International Nuclear Information System (INIS)

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

    2011-01-01

    The objective of this paper is to assess the suitability and the accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA type research reactors in proposed condensed energy spectra of five and seven groups with one and three thermal groups respectively, using the calculational line: WIMSD-IAEA-69 nuclear data library/ WIMSD-5B lattice and cell calculations code/ CITVAP v3.1 core calculations code. Firstly, The assessment goes through analyzing the integral parameters - k e ff, ρ 238 , σ 235 , σ 238 , and C * - of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra, which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark- III Thai research reactor, using the CITVAP v3.1 code and macroscopic cross-section libraries generated using the WIMSD-5B code at the proposed energy spectra separately. The results include the excess reactivities and the worth of control rods, which were compared with previous Monte Carlo results and experimental values, that show good agreement with the references at both energy spectra, albeit better accuracies are shown with the five groups spectrum. The results also includes neutron flux distributions which are settled for future comparisons with other calculational techniques, even, they are comparable to reactors and fuels of the same type. The study reflects the adequacy of using the pre-stated calculational line at the condensed energy spectra for evaluation of the neutronic parameters of the TRIGA type reactors, and future comparisons of the un-benchmarked results could assure this result for wider range of neutronics or safety-related parameters

  6. A device for heat-proofing the closing slab of a fast neutron nuclear reactor

    International Nuclear Information System (INIS)

    Lemercier, Guy.

    1973-01-01

    A device proofing the closing-slab comprising a metal-shoe against which is applied said device, the latter being constituted by a plurality of panels, each of which is formed by a padding of superimposed metal-cloths arranged in parallel relationship to the slab to be heat-proofed. Said device is characterized in that each panel is contained in a casing applied against the slab, constituted by two metal half-boxes of parallelepipedic shape embedded into each other. This can be applied to high-power reactors for protecting the slab against the aerosols of the coolant liquid-metal [fr

  7. NF-6 program complex for BESM-6 computation of the basic neutron-physical characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    Zizin, M.N.; Savochkina, O.A.; Chukhlova, O.P.

    1978-01-01

    A structure of standard designations is described and semantics of a number of standard values used in a NF-6 program complex is given. Main source data and results of neutron-physical reactor calculation are standard values, the peculiarities of FORTRAN and ALGOL-GDR algorithm languages in the DUBNA monitoring system were taken account of. As a base of standard values list the FIHAR system list, supplemented with new standard designations for integral reactor characteristics, is used. Developed is also a list of standard values to organize the exchange with external memory in the process of task solution and long-range storage

  8. Neutron fluence measurement in nuclear facilities.; Medicion de flujos de neutrones en instalaciones nucleares.

    Energy Technology Data Exchange (ETDEWEB)

    Camacho L, M E

    1997-12-01

    The objective of present work is to determine the fluence of neutrons in nuclear facilities using two neutron detectors designed and built at Instituto Nacional de Investigaciones Nucleares (ININ), Mexico. The two neutron detectors are of the passive type, based on solid state nuclear tracks detectors (SSNTD). One of the two neutron detectors was used to determine the fluence distribution of the ports at the nuclear research reactor TRIGA Mark III, which belongs to ININ. In these facilities is important to know the neutron fluence distribution characteristic to carried out diverse kind of research activities. The second neutron detector was employed in order to carry out environmental neutron surveillance. The detector has the property to separate the thermal, intermediate and fast components of the neutron fluence. This detector was used to measure the neutron fluence at hundred points around the primary container of the first Mexican Nuclear Power plant `Laguna Verde`. This last detector was also used to determine the neutron fluence in some points of interest, around and inside a low scattering neutron room at the `Centro de Metrologia de Radiaciones Ionizantes` of the ININ, to know the background neutron field produced by the neutron sources used there. The design of the two neutron detector and the results obtained for each of the surveying facilities, are described in this work. (Author).

  9. Calculation models for a nuclear reactor

    International Nuclear Information System (INIS)

    Tashanii, Ahmed Ali

    2010-01-01

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

  10. TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

    1984-06-01

    The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

  11. A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushuev, A. V.; Kozhin, A. F., E-mail: alexfkozhin@yandex.ru; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E. [National Research Nuclear University MEPhI (Russian Federation)

    2016-12-15

    An active neutron method for measuring the residual mass of {sup 235}U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual {sup 235}U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of {sup 238}U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

  12. Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of spatial and energetic distribution of neutron flux distribution

    International Nuclear Information System (INIS)

    Aredes, Vitor Ottoni Garcia

    2014-01-01

    In this work was conducted the mapping of the thermal and epithermal neutrons flux and the energy spectrum of the neutrons in the reactor core IPEN/MB-01 for a cylindrical core configuration with minor excess reactivity, which is 28 x 28 fuel rods arranged in north-south and east-west directions. The calibration of control rods for this configuration determined their excess reactivity. The lower excess reactivity in the core decreased neutron flux disturbance caused by the neutron absorbing rods , given that the nuclear reactor was operated with the rods almost completely removed . Was used the 'Activation Analysis Technique' with the thin foil activation detectors ( infinitely diluted and hyper-pure), of different materials that work in different energy ranges, to calculate the saturation activity, used for determining the neutron flux and in the SANDBP code as input for the calculation of the neutrons energy spectrum. To discriminate thermal and epithermal flux , was used the 'Cadmium RatioTechnique' . The activation detectors were distributed in a total of 140 radial and axial positions in the reactor core and 16 irradiation, with bare and covered with cadmium activation foils. A model of this configuration was simulated by MCNP-5 code to determine the cadmium correction factor and comparison of the results obtained experimentally. The cylindrical configuration desired, with 17% less fuel than the standard rectangular configuration (28 x 26 fuel rods), reached criticality with the control rods approximately 90% removed, which decreased considerably the disturbance in neutron flux. Given the highest power density of the 28 x 28 cylindrical core, the neutron flux increased by over 50% in the central regions of the core compared to the values of the 28 x 26 standard rectangular core. (author)

  13. Some reactor properties of the new designed nuclear fuels after neutron irradiation

    International Nuclear Information System (INIS)

    Bajan, M.; Necas, V.

    2013-01-01

    The main goal of this paper was perform the optimisation of the fuel assemblies from the profiling point of view as well as the enrichment of individual rods in such a way that the power peaking factor is steady as possible and also the stock of reactivity for six year fuel cycle. For this reason the limit for maximum fuel rod enrichment was increased to 5.95%. The power in the individual rods is the factor, which can limit the total reactor's power, it is very important to minimise the power peaking factor as possible. At the first the power peaking factor of selected fuel assemblies used in VVER-440 reactor were investigated and from results was based perspective designs which was divided into four parts according to the position of pins with gadolinium burnable absorber and according to the shroudless design. From every part the most perspective fuel assembly was chosen. The results are shown in the Fig. 7. The best result is using the shroudless design. As the second best design is fuel assembly with three gadolinium rods in the middle of the assembly. The power peaking factor unsteadiness is much lower as the reference fuel assembly Gd-2. Also it was demonstrate that the increase of enrichment to 5.95% is perspective, because in several designs the difference in enrichment in individual pins was 1% "2'3"5U. Considering only the present allowed value (max 5%) it would not be possible to reach such good power peaking factor and the reactivity sufficient for 6-years fuel cycle. Profiling optimisation together with modernization of structural changes of assembly was achieved the low power peaking factor unsteadiness in individual pins and higher average enrichment of "2"3"5U. So the optimisation can be summarized as very prosperous and perspective. (authors)

  14. Nuclear reactor theory

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi

    2007-09-01

    This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

  15. Nuclear reactor core flow baffling

    International Nuclear Information System (INIS)

    Berringer, R.T.

    1979-01-01

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

  16. Nuclear reactors in remote earth

    International Nuclear Information System (INIS)

    Garzon, L.; Cavero, A.

    1999-01-01

    Same basic geological principles along with other facts, have allowed us to establish the existence in the remote past (Between 2.5 and 4 x 10''9 years ago) of the uranium deposits and/or uranium mineralized volumes, which be-have as nuclear reactors. A simplified neutronic diffusion model have allowed us to describe the main characteristics of such systems. The obtained results indicate that this phenomenon was a rather frequent fact. (Author) 7 refs

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

    Science.gov (United States)

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

    2016-10-01

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

  18. 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.)

  19. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  20. Refuelling nuclear reactors

    International Nuclear Information System (INIS)

    Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

    1981-01-01

    An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

  1. A three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition in graphite components of advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, D.O.; Robinson, A.T.; Allen, D.A.; Picton, D.J.; Thornton, D.A. [TCS, Serco, Rutherford House, Olympus Park, Quedgeley, Gloucester, Gloucestershire GL2 4NF (United Kingdom); Shaw, S.E. [EDF Energy, Barnet Way, Barnwood, Gloucester GL4 3RS (United Kingdom)

    2011-07-01

    This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK's Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition. (authors)

  2. Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

    International Nuclear Information System (INIS)

    Trellue, Holly R.

    2004-01-01

    The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

  3. 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

  4. Experimental determination of spectral ratios and of neutrons energy spectrum in the fuel of the IPEN/MB-01 nuclear reactor

    International Nuclear Information System (INIS)

    Nunes, Beatriz Guimaraes

    2012-01-01

    This study aims to determine the spectral ratios and the neutron energy spectrum inside the fuel of IPEN/MB-01 Nuclear Reactor. These parameters are of great importance to accurately determine spectral physical parameters of nuclear reactors like reaction rates, fuel lifetime and also security parameters such as reactivity. For the experiment, activation detectors in the form of thin metal foils were introduced in a collapsible fuel rod. Then the rod was placed in the central position of the core which has a standard rectangular configuration of 26 x 28 fuel rods. There were used activation detectors from different elements such Au-197, U-238, Sc-45, Ni-58, Mg-24, Ti-47 and In-115 to cover a large range of the neutron energy spectrum. After the irradiation, the activation detectors were submitted to gamma spectrometry using a counting system with high purity Germanium, to obtain the reaction rates (saturation activity) per target nucleus. The spectral ratios were compared with calculated values obtained by the Monte Carlo method using the MCNP-4C code. The neutron energy spectrum was obtained inside the fuel rod using the SANDBP code with an input spectrum obtained by the MCNP-4C code, based on the saturation activity per target nucleus values of the activation detectors irradiated. (author)

  5. Utilization of nuclear research reactors

    International Nuclear Information System (INIS)

    1980-01-01

    training. With these considerations in mind, and with the object of providing a proper perspective to scientists and engineers from developing Member States on the potentials for optimum utilization of research reactors as neutron sources in physics, chemistry, biology, and industrial applications, and to familiarize them with up-to-date developments in research reactor technology, the IAEA, through its technical assistance programme, included this training course in its activities for 1979. Since the utilization and operation of research reactors covers many diverse subjects, the programme included a wide variety of topics of interest. Professor S.H. Levine from Pennsylvania State University (USA) delivered a series of lectures on fundamental reactor physics which served as an excellent starting point for the rest of the lectures. Fundamental neutron physics, research reactor techniques and development, modern nuclear electronics and instrumentation, principles of radiation protection at research reactors and the use of microcomputers and microprocessors in reactor operation, were among the basic subjects of the theoretical lectures. Regarding applications, quite a few lectures were devoted to neutron activation analysis, semiconductor gamma ray spectrometry and isotope production in low- and medium-flux reactors. The morning lectures were complemented by some 18 laboratory exercises which dealt with many relevant aspects of research reactor utilization. Some of the topics covered in these experiments were: shielding measurements in mixed neutron and gamma fields, thermoluminescent dosimetry, determination of neutron dose intensity, reactor simulator measurements, control rod calibration, critical and sub-critical experiments, thermal neutron spectra and flux measurements, neutron radiography, semiconductor spectrometry and instrumental neutron activation analysis in several matrices. The laboratory manual prepared by the staff of the host institution and distributed

  6. Problems of nuclear reactor safety. Vol. 2

    International Nuclear Information System (INIS)

    Goncharov, L.A.

    1995-01-01

    Theses of proceedings of the 9 Topical Meeting on problems of nuclear power plant safety are presented. Reports include results of neutron-physical experiments carried out for reactor safety justification. Concepts of advanced reactors with improved safety are considered. Results of researches on fuel cycles are given too

  7. Reactor physics for non-nuclear engineers

    International Nuclear Information System (INIS)

    Lewis, E.E.

    2011-01-01

    A one-term undergraduate course in reactor physics is described. The instructional format is strongly influenced by its intended audience of non-nuclear engineering students. In contrast to legacy treatments of the subject, the course focuses on the physics of nuclear power reactors with no attempt to include instruction in numerical methods. The multi-physics of power reactors is emphasized highlighting the close interactions between neutronic and thermal phenomena in design and analysis. Consequently, the material's sequencing also differs from traditional treatments, for example treating kinetics before the neutron diffusion is introduced. (author)

  8. Reactors physics. Bases of nuclear physics

    International Nuclear Information System (INIS)

    Diop, Ch.M.

    2006-01-01

    The aim of nuclear reactor physics is to quantify the relevant macroscopic data for the characterization of the neutronic state of a reactor core and to evaluate the effects of radiations (neutrons and gamma radiations) on organic matter and on inorganic materials. This first article presents the bases of nuclear physics in the context of nuclear reactors: 1 - reactor physics and nuclear physics; 2 - atomic nucleus - basic definitions: nucleus constituents, dimensions and mass of the atomic nucleus, mass defect, binding energy and stability of the nucleus, strong interaction, nuclear momentums of nucleons and nucleus; 3 - nucleus stability and radioactivity: equation of evolution with time - radioactive decay law; alpha decay, stability limit of spontaneous fission, beta decay, electronic capture, gamma emission, internal conversion, radioactivity, two-body problem and notion of radioactive equilibrium. (J.S.)

  9. Shield structure for a nuclear reactor

    International Nuclear Information System (INIS)

    Rouse, C.A.; Simnad, M.T.

    1979-01-01

    An improved nuclear reactor shield structure is described for use where there are significant amounts of fast neutron flux above an energy level of approximately 70 keV. The shield includes structural supports and neutron moderator and absorber systems. A portion at least of the neutron moderator material is magnesium oxide either alone or in combination with other moderator materials such as graphite and iron. (U.K.)

  10. Nuclear fast neutron reactor cooled by a liquid metal and of which internal structures are equipped with a thermal protection device

    International Nuclear Information System (INIS)

    Lemercier, G.; Lions, N.

    1986-01-01

    The internal structures of a nuclear fast neutron reactor are covered at least partially, on the most hot side, by a thermal protection device. This device comprises modular plates arranged end to end with a certain play between themselves and taking approximately the shape of the internal structures. Each plate is fixed in its center on the internal structures by a stud. A small plate fixed at one of the corners of each plate and covering partially the adjacent plates ensures the safety fixing of these ones [fr

  11. Neutron skyshine from nuclear facilities

    International Nuclear Information System (INIS)

    Nakamura, Takashi; Hayashi, Katsumi.

    1984-01-01

    The advance in neutron skyshine research and the significance are first described. Then, skyshine calculation methods in 1980s particularly and the skyshine experiment in Japan with various nuclear facilities (reactors, D-T neutron sources, accelerators) are reviewed. In comparison with such experiment usable as bench mark, the skyshine calculation methods (Monte Carlo method, transport calculation method) are evaluated for their accuracy and merits and demerits. The values by Monte Carlo calculation were in agreement within about 30 % with the experimental values. Those by DOT 3.5 calculation were twice as large as the experimental values. Those by PALLAS calculation were in good agreement in dose with the experimental values, but the spectra were considerably different. The values by SKYSHINE-2 were in good agreement with the experimental values, but since the ground effect was ignored, the values may deviate from the experimental ones if it is taken into account. (Mori, K.)

  12. 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)

  13. Solution of the neutron diffusion equation to study the 3D distribution of power, applied to nuclear reactors

    International Nuclear Information System (INIS)

    Costa, Danilo Leite

    2013-01-01

    This work aims to present a study about the power distribution behavior in a PWR type reactor, considering both intensity and migration of power peaks due to insertion of control rods into the core. Employing the multidimensional steady-state neutron diffusion equation in order to simulate the neutron flux, and using the Finite Difference Method. Furthermore, based on the axial power distribution on the largest heat flux rod, is carried out thermal analysis of this rod and associated coolant channel. For this purpose is employed the FueLRod 3 D code, it uses the Finite Element Method to model the fuel rod and the associated coolant channel, allowing the thermohydraulics simulation of a single rod discretized in three dimensions, considering the heat flux from the pellet, crossing the gap and the cladding until it reaches the coolant. (author)

  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. Analysis of the micro-structural damages by neutronic irradiation of the steel of reactor vessels of the nuclear power plant of Laguna Verde. Characterization of the design steel; Analisis de los danos micro-estructurales por irradiacion neutronica del acero de la vasija de los reactores de la Central Nuclear de Laguna Verde. Caracterizacion del acero de diseno

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y Rodriguez, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Av. Luis Enrique Erro s/n, Unidad Profesional Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.m [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2010-09-15

    The vessel of a nuclear reactor is one of the safety barriers more important in the design, construction and operation of the reactor. If the vessel results affected to the grade of to have fracture and/or cracks it is very probable the conclusion of their useful life in order to guarantee the nuclear safety and the radiological protection of the exposure occupational personnel, of the public and the environment avoiding the exposition to radioactive sources. The materials of the vessel of a nuclear reactor are exposed continually to the neutronic irradiation that generates the same nuclear reactor. The neutrons that impact to the vessel have the sufficient energy to penetrate certain depth in function of the energy of the incident neutron until reaching the repose or to be absorbed by some nucleus. In the course of their penetration, the neutrons interact with the nuclei, atoms, molecules and with the same crystalline nets of the vessel material producing vacuums, interstitial, precipitate and segregations among other defects that can modify the mechanical properties of the steel. The steel A533-B is the material with which is manufactured the vessel of the nuclear reactors of nuclear power plant of Laguna Verde, is an alloy that, among other components, it contains atoms of Ni that if they are segregated by the neutrons impact this would favor to the cracking of the same vessel. This work is part of an investigation to analyze the micro-structural damages of the reactor vessels of the nuclear power plant of Laguna Verde due to the neutronic irradiation which is exposed in a continuous way. We will show the characterization of the design steel of the vessel, what offers a comprehension about their chemical composition, the superficial topography and the crystalline nets of the steel A533-B. It will also allow analyze the existence of precipitates, segregates, the type of crystalline net and the distances inter-plains of the design steel of the vessel. (Author)

  16. 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)

  17. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  18. Indian advanced nuclear reactors

    International Nuclear Information System (INIS)

    Saha, D.; Sinha, R.K.

    2005-01-01

    For sustainable development of nuclear energy, a number of important issues like safety, waste management, economics etc. are to be addressed. To do this, a number of advanced reactor designs as well as fuel cycle technologies are being pursued worldwide. The advanced reactors being developed in India are the AHWR and the CHTR. Both the reactors use thorium based fuel and have many passive features. This paper describes the Indian advanced reactors and gives a brief account of the international initiatives for the sustainable development of nuclear energy. (author)

  19. 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

  20. 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

  1. 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)

  2. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2001-01-01

    This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

  3. Analysis of the micro-structural damages by neutronic irradiation of the steel of reactor vessels of the nuclear power plant of Laguna Verde. Characterization of the design steel

    International Nuclear Information System (INIS)

    Moranchel y Rodriguez, M.; Garcia B, A.; Longoria G, L. C.

    2010-09-01

    The vessel of a nuclear reactor is one of the safety barriers more important in the design, construction and operation of the reactor. If the vessel results affected to the grade of to have fracture and/or cracks it is very probable the conclusion of their useful life in order to guarantee the nuclear safety and the radiological protection of the exposure occupational personnel, of the public and the environment avoiding the exposition to radioactive sources. The materials of the vessel of a nuclear reactor are exposed continually to the neutronic irradiation that generates the same nuclear reactor. The neutrons that impact to the vessel have the sufficient energy to penetrate certain depth in function of the energy of the incident neutron until reaching the repose or to be absorbed by some nucleus. In the course of their penetration, the neutrons interact with the nuclei, atoms, molecules and with the same crystalline nets of the vessel material producing vacuums, interstitial, precipitate and segregations among other defects that can modify the mechanical properties of the steel. The steel A533-B is the material with which is manufactured the vessel of the nuclear reactors of nuclear power plant of Laguna Verde, is an alloy that, among other components, it contains atoms of Ni that if they are segregated by the neutrons impact this would favor to the cracking of the same vessel. This work is part of an investigation to analyze the micro-structural damages of the reactor vessels of the nuclear power plant of Laguna Verde due to the neutronic irradiation which is exposed in a continuous way. We will show the characterization of the design steel of the vessel, what offers a comprehension about their chemical composition, the superficial topography and the crystalline nets of the steel A533-B. It will also allow analyze the existence of precipitates, segregates, the type of crystalline net and the distances inter-plains of the design steel of the vessel. (Author)

  4. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  5. Neutrons and fusion nuclear technology

    International Nuclear Information System (INIS)

    Hirayama, Shoichi

    1991-01-01

    The strategy of the devolopment of the fusion reactor has been compared with the history of the development of the fission reactor. More than 50 neutron reactors (neutron sources for research and development of reactor components and materials, and for Pu production) have been constructed and operated before the introduction of demonstration power reactors. This fact suggests us to introduce a new path of neutron reactor in the strategy of the development of fusion power reactor in addition to the orthodox approach which goes through the break-even, self-ignition, ETR, and DEMO. One of the benefits of the introduction of such neutron reactor or into the strategy of the fusion reactor development has been studied numerically. The results demonstrate that the introduction of fission-fusion hybrid reactor in 2030, can save ∝20% of natural uranium by 2100 in Japan, in comparison with the case when the fast breeder reactor is introduced in 2030. This saving is recognized large enough to justify earlier construction of the fusion neutron reactor. (orig.)

  6. 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)

  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. Nuclear reactor control rod

    International Nuclear Information System (INIS)

    Cearley, J.E.; Izzo, K.R.

    1987-01-01

    This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured

  9. Dynamic simulation of a two-phase control absorber for neutron flux regulation in a nuclear reactor

    International Nuclear Information System (INIS)

    Plourde, J.A.; Lepp, R.M.

    1979-08-01

    A dynamic simulation of the two-phase control absorber being proposed for future Canadian nuclear power reactors has been developed at Chalk River Nuclear Laboratories. The model, implemented on a hybrid computer, was developed to study absorber dynamics at different circuit operating conditions and with different circuit configurations. The simulation is modular, with as much correspondence as possible between individual modules and the physical entities. The dynamics of several of the modules are described by partial differential equations, with space and time as independent variables. These are solved via the Continuous Space/Discrete Time technique. The simulation has been validated with data from the Two-Phase Absorber Experimental (TOPAX) Rig installed at the ZED-2 test reactor. (author)

  10. 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

  11. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

    An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  12. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

    The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  13. Nuclear reactor core catcher

    International Nuclear Information System (INIS)

    1977-01-01

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

  14. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

  15. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    Baptista, Vinicius Damas

    1996-01-01

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

  16. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  17. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

    Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

  18. Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core

    International Nuclear Information System (INIS)

    Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G.

    2006-01-01

    Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of 62 149 Sm and its dependence on the shift of a resonance position E r (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73≤ΔE r ≤62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant α. We obtain new, more accurate limits of -4x10 -17 ≤α·/α≤3x10 -17 yr -1 . Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress

  19. Nuclear data for structural materials of fission and fusion reactors

    International Nuclear Information System (INIS)

    Goulo, V.

    1989-06-01

    The document presents the status of nuclear reaction theory concerning optical model development, level density models and pre-equilibrium and direct processes used in calculation of neutron nuclear data for structural materials of fission and fusion reactors. 6 refs

  20. Nuclear reactors safety issues

    International Nuclear Information System (INIS)

    Barre, Francois; Seiler, Nathalie

    2008-01-01

    Full text of publication follows: Since the seventies, economic incentives have led the utilities to drive a permanent evolution of the light water reactor (LWR). The evolution deals with the reactor designs as well as the way to operate them in a more flexible manner. It is for instance related to the fuel technologies and management. On the one hand, the technologies are in continuous evolution, such as the fuel pellets (MOX, Gd fuel, or Cr doped fuels..) as well as advanced cladding materials (M5 TM , MDA or ZIRLO). On the other hand, the fuel management is also subject to continuous evolution in particular in terms of increasing the level of burn-up, the reactor (core) power, the enrichment, as well as the duration of reactor cycles. For instance, in a few years in France, the burn-up has raised beyond the value of 39 GWj/t, initially authorized up to 52 GWj/t for the UO 2 fuel. In the near future, utilities foreseen to reach fuel burn-up of 60 GWj/t for MOX fuel and 70 GWj/t for UO 2 fuel. Furthermore, the future reactor of fourth generation will use new fuels of advanced conception. Furthermore with the objective of improving the safety margins, methods and calculation tools used by the utilities in the elaboration of their safety demonstrations submitted to the Safety Authority, are in movement. The margin evaluation methodologies often consist of a calculation chain of best-estimate multi-field simulations (e.g. various codes being coupled to simulate in a realistic way the evolution of the thermohydraulic, neutronic and mechanic state of the reactor). The statistical methods are more and more sophisticated and the computer codes are integrating ever-complex physical models (e.g. three-dimensional at fine scale). Following this evolution, the Institute of Radioprotection and Nuclear Safety (IRSN), whose one of the roles is to examine the safety records and to rend a technical expertise, considers the necessity of reevaluating the safety issues for advanced

  1. 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)

  2. Overview moderator material for nuclear reactor components

    International Nuclear Information System (INIS)

    Mairing Manutu Pongtuluran; Hendra Prihatnadi

    2009-01-01

    In order for a reactor design is considered acceptable absolute technical requirement is fulfilled because the most important part of a reactor design. Safety considerations emphasis on the handling of radioactive substances emitted during the operation of a reactor and radioactive waste handling. Moderator material is a layer that interacts directly with neutrons split the nuclear fuel that will lead to changes in physical properties, nuclear properties, mechanical properties and chemical properties. Reviews moderator of this time is of the types of moderator is often used to meet the requirements as nuclear material. (author)

  3. 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

  4. Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3

    International Nuclear Information System (INIS)

    Altaf, M.H.; Badrun, N.H.; Chowdhury, M.T.

    2015-01-01

    Highlights: • SRAC-PIJ code and SRAC-CITATION have been utilized to model the core. • Most of the simulated results show no significant differences with references. • Thermal peak flux varies a bit due to up condition of TRIGA. • ENDF/B-VII.0 and JENDL-3.3 libraries perform well for neutronics analysis of TRIGA. - Abstract: Important kinetic parameters such as effective multiplication factor, k eff , excess reactivity, neutron flux and power distribution, and power peaking factors of TRIGA Mark II research reactor in Bangladesh have been calculated using the comprehensive neutronics calculation code system SRAC 2006 with the evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3. In the code system, PIJ code was employed to obtain cross section of the core cells, followed by the integral calculation of neutronic parameters of the reactor conducted by CITATION code. All the analyses were performed using the 7-group macroscopic cross section library. Results were compared to the experimental data, the safety analysis report (SAR) of the reactor provided by General Atomic as well as to the simulated values by numerically benchmarked MCNP4C, WIMS-CITATION and SRAC-CITATION codes. The maximum power densities at the hot spot were found to be 169.7 W/cc and 170.1 W/cc for data libraries ENDF/B-VII.0 and JENDL-3.3, respectively. Similarly, the total peaking factors based on ENDF/B-VII.0 and JENDL-3.3 were calculated as 5.68 and 5.70, respectively, which were compared to the original SAR value of 5.63, as well as to MCNP4C, WIMS-CITATION and SRAC-CITATION results. It was found in most cases that the calculated results demonstrate a good agreement with our experiments and published works. Therefore, this analysis benchmarks the code system and will be helpful to enhance further neutronics and thermal hydraulics study of the reactor

  5. The use of nuclear reactor in radiation biology

    International Nuclear Information System (INIS)

    Ujeno, Yowri

    1991-01-01

    The Kyoto University Reactor (KUR) is widely used not only in biology, but also in applied biology, today. These studies were surveyed in the present paper and the future possibility to use KUR in radiation biology was discussed. The researches on the effects of thermal neutrons on various normal tissues, the biological effects of neutrons except thermal neutrons, especially intermediate neutrons between thermal and high speed neutrons or cold neutrons, the adaptive response of cells to thermal neutron radiation, the application of nuclear reactor-produced radionuclides including 195m Pt to biology, and the mutation in botanical science and so on, should be continued using nuclear reactor. The necessity of nuclear reactor in biology and applied biology is emphasized. (author)

  6. Neutrons in basic and applied nuclear research - a review

    International Nuclear Information System (INIS)

    Bhattacharya, Sailajananda

    2013-01-01

    Energetic neutron sources, both white and mono-energetic, are widely used In basic nuclear physics as well as various multidisciplinary research. Precise measurement of various neutron induced reaction cross-sections are crucial for the design and development of new generation of reactors, like accelerator driven subcritical systems, nuclear incinerators, etc. A review of some recent trends in neutron induced basic and applied nuclear research will be presented in this talk. (author)

  7. HOMOGENEOUS NUCLEAR POWER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  8. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  9. Stochastic model of energetic nuclear reactor

    International Nuclear Information System (INIS)

    Bojko, R.V.; Ryazanov, V.V.

    2002-01-01

    Behaviour of nuclear reactor was treated using the theory of branching processes. As mathematical model descriptive the neutron number in time the Markov occasional process is proposed. Application of branching occasional processes with variable regime to the description of neutron behaviour in the reactor makes possible conducting strong description of critical operation regime and demonstrates the severity of the process. Three regimes of the critical behaviour depending on the sign of manipulated variables and feedbacks were discovered. Probability regularities peculiar to the behaviour of the reactor are embodied to the suggested stochastic model [ru

  10. 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)

  11. Review of Non-Neutron and Neutron Nuclear Data, 2004

    International Nuclear Information System (INIS)

    Holden, Norman E.

    2005-01-01

    Review articles are in preparation for the 2004 edition of the CRC Handbook of Chemistry and Physics dealing with the evaluation of both non-neutron and neutron nuclear data. Data on the discovery of element 110, Darmstadtium, and element 111 have been officially accepted, while data on element 118 have been withdrawn. Data to be presented include revised values for very short-lived nuclides, long-lived nuclides, and beta-beta decay measurements. There has been a reassessment of the spontaneous fission (sf) half-lives, which distinguishes between sf decay half-lives and cluster decay half-lives, and with cluster-fission decay. New measurements of isotopic abundance values for many elements will be discussed with an emphasis on the minor isotopes of interest for use in neutron activation analysis measurements. Neutron resonance integrals will be discussed emphasizing the differences between the calculated values obtained from the analytical integration over neutron resonances and the measured values in a neutron reactor-spectrum, which does not quite conform to the assumed 1/E neutron energy spectrum. The method used to determine the neutron resonance integral from measurement, using neutron activation analysis, will be discussed

  12. Nuclear data usage for research reactors

    International Nuclear Information System (INIS)

    Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

    1996-01-01

    In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)

  13. Energy from nuclear reactors

    International Nuclear Information System (INIS)

    Hospe, J.

    1977-01-01

    This VDI-Nachrichten series has the target to provide a technical-objective basis for the discussion of the pros and cons of nuclear power. The first part deals with LWR-type reactors which so far have prevailed in nuclear power generation. (orig.) [de

  14. 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

  15. 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

  16. 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

  17. 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

  18. Neutron spectroscopy, nuclear structure, related topics. Abstracts

    International Nuclear Information System (INIS)

    Sukhovoj, A.M.

    1996-01-01

    Neutron spectroscopy, nuclear structure and related topics are considered. P, T-breaking, neutron beta decay, neutron radiative capture and neutron polarizability are discussed. Reaction with fast neutrons, methodical aspect low-energy fission are considered too

  19. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Guidez, Joel; Jarriand, Paul.

    1975-01-01

    The invention concerns a fast neutron nuclear reactor cooled by a liquid metal driven through by a primary pump of the vertical drive shaft type fitted at its lower end with a blade wheel. To each pump is associated an exchanger, annular in shape, fitted with a central bore through which passes the vertical drive shaft of the pump, its wheel being mounted under the exchanger. A collector placed under the wheel comprises an open upward suction bell for the liquid metal. A hydrostatic bearing is located above the wheel to guide the drive shaft and a non detachable diffuser into which at least one delivery pipe gives, envelopes the wheel [fr

  20. Nuclear reactors to come

    International Nuclear Information System (INIS)

    Lung, M.

    2002-01-01

    The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

  1. Nuclear reactors theory

    International Nuclear Information System (INIS)

    Naudan, G.; Nigon, J.L.

    1993-01-01

    After principles of chain reaction and criticality notion, a descriptive model of neutrons behaviour is exposed from a local point of view (this model is called four factors model). One justifies the use of middle values for the calculation of the distribution in space of reactor, quantities representing heterogeneous middle from a local point of view (fuel, moderator, can or clad, and so on ...) by substitution of an equivalent homogeneous middle. Time dependence, dynamical behaviour of reactor are studied. Long term effects of evolution of constituents elements of heart under irradiation, and ways to balance this evolution are in the last paragraph. 18 refs., 26 figs

  2. Overview of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Nguyen Thai Sinh; Luong Ba Vien

    2016-01-01

    The present reactor called Dalat Nuclear Research Reactor (DNRR) has been reconstructed from the former TRIGA Mark II reactor which was designed by General Atomic (GA, San Diego, California, USA), started building in early 1960s, put into operation in 1963 and operated until 1968 at nominal power of 250 kW. In 1975, all fuel elements of the reactor were unloaded and shipped back to the USA. The DNRR is a 500-kW pool-type research reactor using light water as both moderator and coolant. The reactor is used as a neutron source for the purposes of: (1) radioactive isotope production; (2) neutron activation analysis; and (3) research and training

  3. 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

  4. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    International Nuclear Information System (INIS)

    Kim, M. C.; Lee, B. S.; Oh, Y. J.

    2003-01-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T sp ), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T 41J . T sp from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T 0 ) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen

  5. 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)

  6. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Butterfield, C.E.; Waite, E.

    1982-01-01

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  7. Nuclear reactor container

    International Nuclear Information System (INIS)

    Hosaka, Seiichi.

    1988-01-01

    Cables coverd with non-halogen covering material are used as electric wire cables wired for supplying electric power to a reactor recycling pump. Silicone rubber having specified molecular formula is used for the non-halogen covering material. As a result, formation of chlorine in a nuclear reactor container can be eliminated and increase in the deposited salts to SUS pipeways, etc. can be prevented, to avoid the occurrence of stress corrosion cracks. (H.T.)

  8. 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)

  9. 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.

  10. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  11. Nuclear reactor buildings

    International Nuclear Information System (INIS)

    Nagashima, Shoji; Kato, Ryoichi.

    1985-01-01

    Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

  12. 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)

  13. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1983-01-01

    The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

  14. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  15. Graphical method for determination of critical dimensions and neutron flux distribution in multi zone nuclear reactors; Graficka metoda za odredjivanje kriticnih dimenzija i raspodele fluksa kod multiregionalnih nuklearnih reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Radanovic, Lj; Bingulac, S; Lazarevic, B; Matausek, M [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1964-07-01

    This paper describes the graphical method for calculating the neutron flux distribution by using normalized Riccati equations. It was shown that the solutions of adequately normalized Riccati equations could be used as standard curves for determining the critical dimensions and radial flux distribution in multi zone nuclear reactors. the methos is applicable irrelevant of the number and position of the region in the core.

  16. Nuclear reactor control column

    International Nuclear Information System (INIS)

    Bachovchin, D.M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

  17. 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)

  18. Migros-3: a code for the generation of group constants for reactor calculations from neutron nuclear data in KEDAK format

    International Nuclear Information System (INIS)

    Broeders, I.; Krieg, B.

    1977-01-01

    The code MIGROS-3 was developed from MIGROS-2. The main advantage of MIGROS-3 is its compatibility with the new conventions of the latest version of the Karlsruhe nuclear data library, KEDAK-3. Moreover, to some extent refined physical models were used and numerical methods were improved. MIGROS-3 allows the calculation of microscopic group cross sections of the ABBN type from isotopic neutron data given in KEDAK-format. All group constants, necessary for diffusion-, consistent P 1 - and Ssub(N)-calculations can be generated. Anisotropy of elastic scattering can be taken into account up to P 5 . A description of the code and the underlying theory is given. The input and output description, a sample problem and the program lists are provided. (orig.) [de

  19. 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.

  20. The physics of nuclear reactors

    CERN Document Server

    Marguet, Serge

    2017-01-01

    This comprehensive volume offers readers a progressive and highly detailed introduction to the complex behavior of neutrons in general, and in the context of nuclear power generation. A compendium and handbook for nuclear engineers, a source of teaching material for academic lecturers as well as a graduate text for advanced students and other non-experts wishing to enter this field, it is based on the author’s teaching and research experience and his recognized expertise in nuclear safety. After recapping a number of points in nuclear physics, placing the theoretical notions in their historical context, the book successively reveals the latest quantitative theories concerning: •   The slowing-down of neutrons in matter •   The charged particles and electromagnetic rays •   The calculation scheme, especially the simplification hypothesis •   The concept of criticality based on chain reactions •   The theory of homogeneous and heterogeneous reactors •   The problem of self-shielding �...

  1. 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

  2. Analog computing for a new nuclear reactor dynamic model based on a time-dependent second order form of the neutron transport equation

    International Nuclear Information System (INIS)

    Pirouzmand, Ahmad; Hadad, Kamal; Suh, Kune Y.

    2011-01-01

    This paper considers the concept of analog computing based on a cellular neural network (CNN) paradigm to simulate nuclear reactor dynamics using a time-dependent second order form of the neutron transport equation. Instead of solving nuclear reactor dynamic equations numerically, which is time-consuming and suffers from such weaknesses as vulnerability to transient phenomena, accumulation of round-off errors and floating-point overflows, use is made of a new method based on a cellular neural network. The state-of-the-art shows the CNN as being an alternative solution to the conventional numerical computation method. Indeed CNN is an analog computing paradigm that performs ultra-fast calculations and provides accurate results. In this study use is made of the CNN model to simulate the space-time response of scalar flux distribution in steady state and transient conditions. The CNN model also is used to simulate step perturbation in the core. The accuracy and capability of the CNN model are examined in 2D Cartesian geometry for two fixed source problems, a mini-BWR assembly, and a TWIGL Seed/Blanket problem. We also use the CNN model concurrently for a typical small PWR assembly to simulate the effect of temperature feedback, poisons, and control rods on the scalar flux distribution

  3. 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.)

  4. Survey of nuclear parameters from the TRIGA Mark I IPR R1 Brazilian reactor with concentric configuration aiming the application of K0 neutron activation technique

    International Nuclear Information System (INIS)

    Franco, Milton Batista

    2006-01-01

    This research intended to determine the nuclear parameters a, f, spectral index and neutron temperature in several irradiations positions of the TRIGA Mark 1 IPR-R1 reactor, for use on the parametric method K 0 in the CDTN. K 0 is a monostandard method of neutron activation analysis. It is, on the whole, experimentally simple, flexible and an important tool for accurate and convenient standardization in instrumental multi-element analysis. At the time the parameters were determined at the rotatory rack, lower layer and in the central thimble: alpha was calculated applying the three bare monitor method using 197 Au, 94 Zr and 96 Zr; f determination was done according to the bare bi-isotopic method; neutron temperature was calculated through the direct method using 176 Lu, 94 Zr, 96 Zr and 197 Au and the Westcott's g(Tn) function for the 176 Lu was calculated and the result was interpolated in the Grintakis and Kim (1975) Table, determining the neutron temperature. The procedure to check the parameters consisted in using standard solutions of Au (metal foil, NBS), Lu (LuO 2 , Johnson Mattey Company - JMC) and Zr (ZrO 2 and metal foil, Johnson Mattey Company 99,99% and Zry - 4: 98,14% of Zr, National Bureau of Standard- NBS). Several certified reference materials and two samples of intercomparisons (samples of sediment of the IAEA/ARCAL XXVI project) have been analysed by means of k 0 - INAA in order to verify the efficiency of the method and the quality of the parameters. The certified reference materials were: GXR-2, GXR-5 and GXR-6 of the United States Geological Survey (USGS) and Soil-5, Soil-7 and SL-1 of the International Atomic Energy Agency (IAEA). (author)

  5. Analysis of space and energy homogenization techniques for the solving of the neutron transport equation in nuclear reactor

    International Nuclear Information System (INIS)

    Magat, Ph.

    1997-04-01

    Today neutron transport in PWR's core is routinely computed through the transport-diffusion(2 groups) scheme. This method gives satisfactory results for reactors operating in normal conditions but the 2 group diffusion approximation is unable to take into account interface effects or anisotropy. The improvement of this scheme is logically possible through the use of a simplified P N method (SP N ) for the modeling of the core. The comparison between S N calculations and SP N calculations shows an excellent agreement on eigenvalues as well as on power maps. We can notice that: -) it is no use extending the development beyond P 3 , there is no effect; -) the P 1 development is adequate; and -) the P 0 development is totally inappropriate. Calculations performed on the N4 core of the Chooz power plant have enabled us to compare diffusion operators with transport operators (SP 1 , SP 3 , SP 5 and SP 7 ). These calculations show that the implementation of the SP N method is feasible but the extra-costs in computation times and memory are important. We recommend: SP 5 P 1 calculations for heterogeneous 2-dimension geometry and SP 3 P 1 calculations for the homogeneous 3-dimension geometry. (A.C.)

  6. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sasaki, Y.; Tashima, J.

    1975-01-01

    A description is given of nuclear reactor fuel assemblies arranged in the form of a lattice wherein there is attached to the interface of one of two adjacent fuel assemblies a plate spring having a concave portion curved toward said interface and to the interface of the other fuel assembly a plate spring having a convex portion curved away from said interface

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  8. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    Banerjee, S.; Kamath, H.S.

    2005-01-01

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO 2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO 2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  9. Nuclear reactor assembly

    International Nuclear Information System (INIS)

    Dorner, H.; Scholz, M.; Jungmann, A.

    1975-01-01

    A nuclear reactor assembly includes a reactor pressure tank having a substantially cylindrical side wall surrounded by the wall of a cylindrical cavity formed by a biological shield. A rotative cylindrical wall is interposed between the walls and has means for rotating it from outside of the shield, and a probe is carried by the rotative wall for monitoring the pressure tank's wall. The probe is vertically movable relative to the rotative cylindrical wall, so that by the probe's vertical movement and rotation of the rotative cylinder, the reactor's wall can be very extensively monitored. If the reactor pressure tank's wall fails, it is contained by the rotative wall which is backed-up by the shield cavity wall. (Official Gazette)

  10. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

    Kakaria, B. K.

    1994-01-01

    AECL has over 40 years of experience in the nuclear field. Over the past 20 years, this unique Canadian nuclear technology has made a worldwide presence, In addition to 22 CANDU reactors in Canada, there are also two in India, one in Pakistan, one in Argentina, four in Korea and five in Romania. CANDU advancements are based on evolutionary plant improvements. They consist of system performance improvements, design technology improvements and research and development in support of advanced nuclear power. Given the good performance of CANOU plants, it is important that this CANDU operating experience be incorporated into new and repeat designs

  11. Materials for generation-IV nuclear reactors

    International Nuclear Information System (INIS)

    Alvarez, M. G.

    2009-01-01

    Materials science and materials development are key issues for the implementation of innovative reactor systems such as those defined in the framework of the Generation IV. Six systems have been selected for Generation IV consideration: gas-cooled fast reactor, lead-cooled fast reactor, molten salt-cooled reactor, sodium-cooled fast reactor, supercritical water-cooled reactor, and very high temperature reactor. The structural materials need to resist much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. For this reason, the first consideration in the development of Generation-IV concepts is selection and deployment of materials that operate successfully in the aggressive operating environments expected in the Gen-IV concepts. This paper summarizes the Gen-IV operating environments and describes the various candidate materials under consideration for use in different structural applications. (author)

  12. DOE fundamentals handbook: Nuclear physics and reactor theory

    International Nuclear Information System (INIS)

    1993-01-01

    The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance

  13. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  14. 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

  15. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Marmonier, Pierre; Mesnage, Bernard; Nervi, J.C.

    1975-01-01

    This invention refers to fuel assemblies for a liquid metal cooled fast neutron reactor. Each assembly is composed of a hollow vertical casing, of regular polygonal section, containing a bundle of clad pins filled with a fissile or fertile substance. The casing is open at its upper end and has a cylindrical foot at its lower end for positioning the assembly in a housing provided in the horizontal diagrid, on which the core assembly rests. A set of flat bars located on the external surface of the casing enables it to be correctly orientated in its housing among the other core assemblies [fr

  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. Improvements in streaking nuclear reactors

    International Nuclear Information System (INIS)

    Pedrick, A.P.

    1976-01-01

    In this type of reactor atomic nuclei are stripped of their electron shells by heating to form a very high temperature plasma which is passed at high speed through a chamber in which they are forced into contact with a 'wall' formed by a unidirectional stream of photons from continuous laser beams. In this way it should be possible to brush off from the surface of the nuclei protons and neutrons, with release of their binding energy. The energy thus produced can be subjected to much more gentle control than with a fission or fusion reactor. Furthermore, if this concept can be successfully applied to elements of high atomic number which are normally regarded as stable and unfissionable, a vast new source of nuclear energy release will have been made available. It also seems possible that an atomic nucleus might be spun sufficiently in such a reactor to disintegrate it completely into nucleons by simple centrifugal action, with great release of binding energy. The reactor described has a central body with radial ducts through which the nuclei are passed, and a number of lasers are provided whose beams are arranged so that the nuclei are discharged at the cross-over point of two or more laser beams which form a corner at the junction of two or more photon walls. (U.K.)

  18. Nuclear reactions excited by recoil protons on a nuclear reactor

    International Nuclear Information System (INIS)

    Mukhammedov, S.; Khaydarov, A.; Barsukova, E.G.

    2006-01-01

    The nuclear reactions excited by recoil protons and of the detection possibility of the various chemical elements with the use of these secondary nucleus reactions were investigated. The recoil protons are produced on a nuclear reactor in the result of (n, p) inelastic and elastic scattering interaction of fast neutrons with nuclei of hydrogen. It is well known that the share of fast neutrons in energetic spectrum of reactor's neutrons in comparison with the share of thermal neutrons is small. . Consequently, the share of recoil protons produced in the result of fast neutron interaction with nuclei of light elements, capable to cause the nuclear reactions, is also small, des, due to Coulomb barrier of nuclei the recoil protons can cause the nuclear reactions only on nuclei of light and some middle elements. Our studies show that observable yields have radio nuclides excited in the result of nuclear reactions on Li, B, O, V and Cu. Our experimental results have demonstrated that the proton activation analysis based on the application of secondary nuclear reactions is useful technique to determine large contents of various light and medium chemical elements. Detection limits for studied chemical elements are estimated better than 10 ppm

  19. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    Aspden, G.J.

    1980-01-01

    A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

  20. Nuclear reactor installation

    International Nuclear Information System (INIS)

    Keller, W.

    1976-01-01

    A nuclear reactor installation includes a pressurized-water coolant reactor vessel and a concrete biological shield surrounding this vessel. The shield forms a space between it and the vessel large enough to permit rapid escape of the pressurized-water coolant therefrom in the event the vessel ruptures. Struts extend radially between the vessel and shield for a distance permitting normal radial thermal movement of the vessel, while containing the vessel in the event it ruptures, the struts being interspaced from each other to permit rapid escape of the pressurized-water coolant from the space between the shield and the vessel

  1. Refueling of nuclear reactor

    International Nuclear Information System (INIS)

    Kaufmann, J.W.; Swidwa, K.J.; Hornak, L.P.

    1989-01-01

    This patent describes an apparatus for refueling a nuclear reactor, the reactor being disposed for refueling under water in a pit in a containment, the apparatus including a bridge to be mounted moveably over the pit on the containment, first means connected to the bridge, for moving the bridge forward and backward on the containment over the pit along a first path, a first pulse generator, connected to the moving means, responsive to the movement of the bridge, for producing pulses, means, connected to the generator,for counting the pulses, the count of the pulses being dependent on the distance of the movement of the bridge

  2. Nuclear data for fusion reactor technology

    International Nuclear Information System (INIS)

    1988-06-01

    The meeting was organized in four sessions and four working groups devoted to the following topics: Requirements of nuclear data for fusion reactor technology (6 papers); Status of experimental and theoretical investigations of microscopic nuclear data (10 papers); Status of existing libraries for fusion neutronic calculations (5 papers); and Status of integral experiments and benchmark tests (6 papers). A separate abstract was prepared for each of these papers

  3. 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

  4. 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...

  5. 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

  6. Report of scientific results 1976. Section nuclear chemistry and reactor

    International Nuclear Information System (INIS)

    1976-01-01

    The report of the section Nuclear Chemistry and Reactor presents the results of R and D in the fields of neutron scattering, radiation damage in solids, reactor chemistry, trace elements research in biomedicine, geochemistry, reactor operation, radioisotope production, and gives a survey of publications and lectures. (HK) [de

  7. Operational methods of the fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Borges, V.; Sefidvash, F.

    1993-01-01

    The operational curve of reactivity as a function of porosity of the Fluidized Bed Nuclear Reactor is presented. The strategies for start-up, shut-down and maintaining the reactor critical during operation are described. The inherent safety of the reactor from neutronic point of view under steady state condition is demonstrated. (author)

  8. Licensing of nuclear reactor operators

    International Nuclear Information System (INIS)

    1979-09-01

    Recommendations are presented for the licensing of nuclear reactor operators in units licensed according to the legislation in effect. They apply to all physical persons designated by the Operating Organization of the nuclear reactor or reactors to execute any of the following functional activities: a) to manipulate the controls of a definite reactor b) to direct the authorized activities of the reactor operators licesed according to the present recommendations. (F.E.) [pt

  9. Poly(L-lactic acid) microspheres containing neutron-activatable holmium-165: a study of the physical characteristics of microspheres before and after irradiation in a nuclear reactor.

    Science.gov (United States)

    Mumper, R J; Jay, M

    1992-01-01

    The solvent evaporation technique was employed to prepare poly(L-lactic acid) (PLA) microspheres with 165Ho acetylacetonate (Ho-AcAc). Particle size, percentage Ho-165, percent residual solvent, and retentive ability of the spheres were found to be strongly affected by preparatory conditions. Differential scanning calorimetry (DSC) thermograms suggested that the Ho-AcAc existed in the PLA matrix as a molecular dispersion. High neutron flux irradiations of the PLA spheres in a nuclear reactor produced Ho-166, a therapeutic radionuclide that emits high-energy negatrons (Emax = 1.84 MeV; half-life = 26.9 hr). The gamma radiation dose (53-75 Mrad) from the core of the reactor provided an overkill of all bioburdens in the PLA spheres. Gel permeation chromatography (GPC) analysis showed that these irradiations caused a reduction in PLA molecular weight. Infrared spectra, 13C NMR spectra, 1H NMR spectra, and DSC thermograms further confirmed the presence of lower molecular weight PLA but proved the overall maintenance of PLA structure.

  10. 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

  11. Nuclear reactor shutdown control rod assembly

    International Nuclear Information System (INIS)

    Bilibin, K.

    1988-01-01

    This patent describes a nuclear reactor having a reactor core and a reactor coolant flowing therethrough, a temperature responsive, self-actuated nuclear reactor shutdown control rod assembly, comprising: an upper drive line terminating at its lower end with a substantially cylindrical wall member having inner and outer surfaces; a lower drive line having a lower end adapted to be attached to a neutron absorber; a ring movable disposed about the outer surface of the wall member of the upper drive line; thermal actuation means adapted to be in heat exchange relationship with coolant in an associated reactor core and in contact with the ring, and balls located within the openings in the upper drive line. When reactor coolant approaches a predetermined design temperature the actuation means moves the ring sufficiently so that the balls move radially out from the recess and into the space formed by the second portion of the ring thereby removing the vertical support for the lower drive line such that the lower drive line moves downwardly and inserts an associated neutron absorber into an associated reactor core resulting in automatic reduction of reactor power

  12. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  13. Nuclear reactor container

    International Nuclear Information System (INIS)

    Ishiyama, Takenori.

    1989-01-01

    This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

  14. The nuclear reactor systems

    International Nuclear Information System (INIS)

    Bacher, P.

    2008-01-01

    This paper describes the various nuclear reactor systems, starting with the Generation II, then the present development of the Generation III and the stakes and challenges of the future Generation IV. Some have found appropriate to oppose reactor systems or generations one to another, especially by minimizing the enhancements of generation III compared to generation II or by expecting the earth from generation IV (meaning that generation III is already obsolete). In the first part of the document (chapter 2), some keys are given to the reader to develop its proper opinion. Chapter 3 describes more precisely the various reactor systems and generations. Chapter 4 discusses the large industrial manoeuvres around the generation III, and the last chapter gives some economical references, taking into account, for the various means of power generation, the impediments linked to climate protection

  15. Nuclear reactor refueling system

    International Nuclear Information System (INIS)

    Wade, E.E.

    1978-01-01

    A system for transferring fuel assemblies between a nuclear reactor core and a fuel storage area while the fuel assembies remain completely submerged in a continuous body of coolant is described. The system comprises an in-vessel fuel transfer machine located inside the reactor vessel and an ex-vessel fuel transfer machine located in a fuel storage tank. The in-vessel fuel transfer machine comprises two independently rotatable frames with a pivotable fuel transfer apparatus disposed on the lower rotatable frame. The ex-vessel fuel transfer machine comprises one frame with a pivotable fuel transfer apparatus disposed thereon. The pivotable apparatuses are capable of being aligned with each other to transfer a fuel assembly between the reactor vessel and fuel storage tank while the fuel assembly remains completely submerged in a continuous body of coolant. 9 claims, 7 figures

  16. Heat dissipating nuclear reactor

    Science.gov (United States)

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  17. Nuclear waste disposal utilizing a gaseous core reactor

    Science.gov (United States)

    Paternoster, R. R.

    1975-01-01

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

  18. Decommissioning a nuclear reactor

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1991-01-01

    The process of decommissioning a facility such as a nuclear reactor or reprocessing plant presents many waste management options and concerns. Waste minimization is a primary consideration, along with protecting a personnel and the environment. Waste management is complicated in that both radioactive and chemical hazardous wastes must be dealt with. This paper presents the general decommissioning approach of a recent project at Los Alamos. Included are the following technical objectives: site characterization work that provided a thorough physical, chemical, and radiological assessment of the contamination at the site; demonstration of the safe and cost-effective dismantlement of a highly contaminated and activated nuclear-fuelded reactor; and techniques used in minimizing radioactive and hazardous waste. 12 figs

  19. 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

  20. Nuclear reactor operator licensing

    International Nuclear Information System (INIS)

    Bursey, R.J.

    1978-01-01

    The Atomic Energy Act of 1954, which was amended in 1974 by the Energy Reorganization Act, established the requirement that individuals who had the responsibility of operating the reactors in nuclear power plants must be licensed. Section 107 of the act states ''the Commission shall (1) prescribe uniform conditions for licensing individuals; (2) determine the qualifications of such individuals; and (3) issue licenses to such individuals in such form as the Commission may prescribe.'' The article discusses the types of licenses, the selection and training of individuals, and the administration of the Nuclear Regulatory Commission licensing examinations

  1. Laboratory neutrons - a breakthrough in non-nuclear disciplines

    International Nuclear Information System (INIS)

    Jervis, R.E.

    1983-01-01

    The availability of laboratory neutrons at SLOWPOKE Nuclear reactor facility, has greatly facilitated interdisciplinary applied research there. Examples of the uses of the laboratory neutrons include those involved with environmental dispersal of inorganic pollutants, and those associated with public health investigations. (UK)

  2. Shields for nuclear reactors

    International Nuclear Information System (INIS)

    Aspden, G.J.

    1984-01-01

    The patent concerns shields for nuclear reactors. The roof shield comprises a normally fixed radial outer portion, a radial inner portion rotatable about a vertical axis, and a connection between the inner and outer portions. In the event of hypothecal core disruption conditions, a cantilever system on the inner wall allows the upward movement of the inner wall, in order to prevent loss of containment. (UK)

  3. Compact nuclear reactor

    International Nuclear Information System (INIS)

    Juric, S.I.

    1975-01-01

    A compact nuclear reactor of the pressurized-water variety is described which has two separate parts separably engageable for ease of inspection, maintenance and repair. One of the parts is a pressure vessel having an active core and the other of the parts is a closure adapted on its lower surface with an integral steam generator. An integral pump, external pressurizer and control rods are provided which communicate with the active core when engaged to form a total unit. (U.S.)

  4. Nuclear reactor instrumentation

    International Nuclear Information System (INIS)

    Duncombe, E.; McGonigal, G.

    1975-01-01

    A liquid metal cooled nuclear reactor is described which has an equal number of fuel sub-assemblies and sensing instruments. Each instrument senses temperature and rate of coolant flow of a coolant derived from a group of three sub-assemblies so that an abnormal value for one sub-assembly will be indicated on three instruments thereby providing for redundancy of up to two of the three instruments. The abnormal value may be a precurser to unstable boiling of coolant

  5. A swivelling transfer device for nuclear reactors

    International Nuclear Information System (INIS)

    Allain, Albert; Mulot, Pierre; Filloleau, Etienne

    1974-01-01

    The invention relates to a swivelling transfer device for fuel-assemblies. According to the invention, the device comprises, within a protective enclosure, a swivelling system comprising two sets of rails rotatable about an axis and so arranged that the lower and thereof penetrates into the extensions of the extremities of ramps dipped into the reactor and into a storage enclosure. This can apply to the transfer of nuclear reactor fuel assemblies, in particular for reactors of the molten sodium fast neutron type [fr

  6. Nuclear reactor safety: physics and engineering aspects

    International Nuclear Information System (INIS)

    Kinchin, G.H.

    1982-01-01

    In order to carry out the sort of probabilistic analysis referred to by Farmer (Contemp. Phys.; 22:349(1981)), it is necessary to have a good understanding of the processes involved in both normal and accident conditions in a nuclear reactor. Some of these processes, for a variety of different reactor systems, are considered in sections dealing with the neutron chain reaction, the removal of heat from the reactor, material problems, reliability of protective systems and a number of specific topics of particular interest from the point of view of physics or engineering. (author)

  7. Nuclear power reactor technology

    International Nuclear Information System (INIS)

    1978-09-01

    Risoe National Laboratory was established more than twenty years ago with research and development of nuclear reactor technology as its main objective. The Laboratory has by now accumulated many years of experience in a number of areas vital to nuclear reactor technology. The work and experience of, and services offered by the Laboratory within the following fields are described: Health physics site supervision; Treatment of low and medium level radioactive waste; Core performance evaluation; Transient analysis; Accident analysis; Fuel management; Fuel element design, fabrication and performance evaluation; Non-destructive testing of nuclear fuel; Theoretical and experimental structural analysis; Reliability analysis; Site evaluation. Environmental risk and hazard calculation; Review and analysis of safety documentation. Risoe has already given much assistance to the authorities, utilities and industries in such fields, carrying out work on both light and heavy water reactors. The Laboratory now offers its services to others as a consultant, in education and training of staff, in planning, in qualitative and quantitative analysis, and for the development and specification of fabrication techniques. (author)

  8. Nuclear data needs for fusion reactors

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

    The nuclear design of fusion components (e.g., first wall, blanket, shield, magnet, limiter, divertor, etc.) requires an accurate prediction of the radiation field, the radiation damage parameters, and the activation analysis. The fusion nucleonics for these tasks are reviewed with special attention to point out nuclear data needs and deficiencies which effect the design process. The main areas included in this review are tritium breeding analyses, nuclear heating calculations, radiation damage in reactor components, shield designs, and results of uncertainty analyses as applied to fusion reactor studies. Design choices and reactor parameters that impact the neutronics performance of the blanket are discussed with emphasis on the tritium breeding ratio. Nuclear data required for kerma factors, shielding analysis, and radiation damage are discussed. Improvements in the evaluated data libraries are described to overcome the existing problems. 84 refs., 11 figs., 9 tabs

  9. Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of spatial and energetic distribution of neutron flux distribution; Caracterizacao do nucleo cilindrico de menor excesso de reatividade do reator IPEN/MB-01, pela medida da distribuicao espacial e energetica do fluxo de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Aredes, Vitor Ottoni Garcia

    2014-07-01

    In this work was conducted the mapping of the thermal and epithermal neutrons flux and the energy spectrum of the neutrons in the reactor core IPEN/MB-01 for a cylindrical core configuration with minor excess reactivity, which is 28 x 28 fuel rods arranged in north-south and east-west directions. The calibration of control rods for this configuration determined their excess reactivity. The lower excess reactivity in the core decreased neutron flux disturbance caused by the neutron absorbing rods , given that the nuclear reactor was operated with the rods almost completely removed . Was used the 'Activation Analysis Technique' with the thin foil activation detectors ( infinitely diluted and hyper-pure), of different materials that work in different energy ranges, to calculate the saturation activity, used for determining the neutron flux and in the SANDBP code as input for the calculation of the neutrons energy spectrum. To discriminate thermal and epithermal flux , was used the 'Cadmium RatioTechnique' . The activation detectors were distributed in a total of 140 radial and axial positions in the reactor core and 16 irradiation, with bare and covered with cadmium activation foils. A model of this configuration was simulated by MCNP-5 code to determine the cadmium correction factor and comparison of the results obtained experimentally. The cylindrical configuration desired, with 17% less fuel than the standard rectangular configuration (28 x 26 fuel rods), reached criticality with the control rods approximately 90% removed, which decreased considerably the disturbance in neutron flux. Given the highest power density of the 28 x 28 cylindrical core, the neutron flux increased by over 50% in the central regions of the core compared to the values of the 28 x 26 standard rectangular core. (author)

  10. A neutron irradiator to perform nuclear activation

    International Nuclear Information System (INIS)

    Zamboni, C. B.; Zahn, G.S.; Figueredo, A. M. G.; Madi, T. F.; Yoriyaz, H.; Lima, R. B.; Shtejer, K.; Dalaqua Jr, L.

    2001-01-01

    The development of appropriate nuclear instrumentation to perform neutron activation analyze (NAA), using thermal and fast neutrons, can be useful to investigate materials outside the reactor premises. Considering this fact, a small size neutron irradiator prototype was developed at IPEN facilities (Instituto de Pesquisas Energeticas e Nucleares - Brazil). Basically, this prototype consists of a cylinder of 1200 mm long and 985 mm diameter (filled with paraffin) with two Am-Be sources (600GBq each) arranged in the longitudinal direction of its geometric center. The material to be irradiated is positioned at a radial direction of the cylinder between the two Am-Be sources. The main advantage of this irradiator is a very stable neutron flux eliminating the use of standard material (measure of the induced activity in the sample by comparative method). This way the process became agile, practical and economic, but quantities at mg levels of samples are necessary to achieve good sensitivity, when the material has a low microscopy neutron cross section. As fast and thermal neutron can be used, the flux distribution, for both, were calculated and the prototype performance is discussed

  11. Calculation of the neutron flux and fluence in the covering of the nucleus and the vessel of a BWR; Calculo del flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor nuclear BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: evalle@esfm.ipn.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    One of the main objectives related with the safety in any nuclear power plant, including the nuclear power plant of Laguna Verde, is to guarantee the structural integrity of the pressure vessel of the reactor. To identify and quantifying the damage caused be neutron irradiation in the vessel of any nuclear reactor, is necessary to know as much the neutron flux as the fluence that it has been receiving during their time of operation life, since the observables damages by means of tests mechanics are products of micro-structural effects, induced by neutron irradiation, therefore, is important the study and prediction of the neutron flux to have a better knowledge of the damage that are receiving these materials. In our calculation the code DORT was used, which solves the transport equation in discreet coordinates and in two dimensions (x-y, r-{theta} and r-z), in accord to the regulator guide, it requires to make and approach of the neutron flux in three dimensions by means of the Synthesis Method. Whit this method is possible to achieve a representation of the flux in 3D combining or synthesizing the calculated fluxes by DORT code in r-{theta}, r-z and r. In this work the application of the Synthesis Method is presented, according to the Regulator Guide 1.190, to determine the fluxes 3D in the interns of a BWR using three different space meshes. (Author)

  12. Training and research on the nuclear reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    1998-01-01

    The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

  13. Nuclear reactor cavity streaming shield

    International Nuclear Information System (INIS)

    Klotz, R.J.; Stephen, D.W.

    1978-01-01

    The upper portion of a nuclear reactor vessel supported in a concrete reactor cavity has a structure mounted below the top of the vessel between the outer vessel wall and the reactor cavity wall which contains hydrogenous material which will attenuate radiation streaming upward between vessel and the reactor cavity wall while preventing pressure buildup during a loss of coolant accident

  14. 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

  15. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  16. Nuclear research reactors in Brazil

    International Nuclear Information System (INIS)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias

    2011-01-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  17. Seminar Neutronika-2012. Neutron-physical problems of nuclear-power engineering. Program and abstracts

    International Nuclear Information System (INIS)

    2012-01-01

    On October, 30 - November, 2 in State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky a seminar Neutron-physical problems of nuclear power engineering - Neutronika-2012 took place. On the seminar the following problems were discussed: justification of neutron-physical characteristics of reactor facilities and innovation projects; constant support of neutron-physical calculations of nuclear power installations; numerical simulation during solving reactor physics problems; simulation of neutron-physical processes in reactor facilities by Monte Carlo method; development and verification of programs for reactor facilities neutron-physical calculations; algorithms and programs for solving nonstationary problems of neutron-physical calculation of nuclear reactors; analysis of integral and reactor experiments, experimental database; justification of nuclear and radiation safety of fuel cycle [ru

  18. Neutron spectrometry by diamond detector for nuclear radiation

    International Nuclear Information System (INIS)

    Kozlov, S.F.; Konorova, E.A.; Barinov, A.L.; Jarkov, V.P.

    1975-01-01

    Experiments on fast neutron spectrometry using the nuclear radiation diamond detector inside a horizontal channel of a water-cooled and water-moderated reactor are described. It is shown that the diamond detector enables neutron spectra to be measured within the energy range of 0.3 to 10 MeV against reactor gamma-radiation background and has radiation resistance higher than that of conventional semiconductor detectors. (U.S.)

  19. 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)

  20. 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)

  1. 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)

  2. 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

  3. 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

  4. Nuclear reactor sealing system

    International Nuclear Information System (INIS)

    McEdwards, J.A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system is disclosed. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel

  5. 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

  6. Nuclear reactor trip system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    Each parameter of the processes of a nuclear reactor and components operatively associated with it is monitored by a set of four like sensors. A trip system normally operates on a ''two out four'' configuration; i.e., to trip the reactor it is necessary that at least two sensors of a set sense an off-normal parameter. This assumes that all sensors are in normal operating condition. However, when a sensor is in test or is subject to maintenance or is defective or disabled, the ''two out of four''configuration would be reduced to a ''one out of three'' configuration because the affected sensor is taken out of service. This would expose the system to the possibility that a single sensor failure, which may be spurious, will cause a trip of the reactor. To prevent this, it is necessary that the affected sensor be bypassed. If only one sensor is bypassed, the system operates on a ''two out of three'' configuration. With two sensors bypassed, the sensing of an off-normal parameter by a third sensor trips the reactor. The by-pass circuit also disables the circuit coupling the by-passed sensor to the trip circuit. (author)

  7. Nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F

    1974-07-11

    The core of the fast neutron reactor consisting, among other components, of fuel elements enriched in plutonium is divided into modules. Each module contains a bundle of four or six elongated components (fuel elements and control rods). In the arrangement with four components, one is kept rigid while the other three are elastically yielding inclined towards the center and lean against the rigid component. In the modules with six pieces, each component is elastically yielding inclined towards a central cavity. In this way, they form a circular arc. A control rod may be placed in the cavity. In order to counteract a relative lateral movement, the outer surfaces of the components which have hexagonal cross-sections have interlocking bearing cushions. The bearing cushions consist of keyway-type ribs or grooves with the wedges or ribs gripping in the grooves of the neighbouring components. In addition, the ribs have oblique entering surfaces.

  8. The Change of the Seebeck Coefficient Due to Neutron Irradiation and Thermal Fatigue of Nuclear Reactor Pressure Vessel Steel and its Application to the Monitoring of Material Degradation

    International Nuclear Information System (INIS)

    Niffenegger, M.; Reichlin, K.; Kalkhof, D.

    2002-05-01

    The monitoring of material degradation, that might be caused by neutron irradiation and thermal fatigue, is an important topic in lifetime extension of nuclear power plants. We therefore investigated the application of the Seebeck effect for determining material degradation of common reactor pressure vessel steel. The Seebeck coefficient (SC) of several irradiated Charpy specimens made from Japanese JRQ-steel were measured. The specimens suffered a fluence from 0 up to 4.5 x 10 19 neutrons per cm 2 with energies higher than 1 MeV. The measured changes of the SC within this range were about 500 nV, increasing continuously in the range under investigation. Some indications of saturation appeared at fluencies larger than 4.55 x 10 19 neutrons per cm 2 . We obtained a linear dependency between the SC and the temperature shift ΔT 41 of the Charpy-Energy- Temperature curve which is widely used to characterize material embrittlement. Similar measurements were performed on specimens made from the widely used austenitic steel X6CrNiTi18-10 (according to DIN 1.4541) that were fatigued by applying a cyclic strain amplitude of 0.28%. For this kind of fatigue the observed change of SC was somewhat smaller than for the irradiated specimens. Further investigations were made to quantify the size of the gage volume in which the thermoelectric power is generated. It appeared that the information gathered from a Thermo Electric Power (TEP) measurement is very local. To overcome this problem we propose a novel TEP-method using a Thermoelectric Scanning Microscope (TSM). We finally conclude that the change of the SC has a potential for monitoring of material degradation due to neutron irradiation and thermal fatigue, but it has to be taken into account that several influencing parameters could contribute to the TEP in either an additional or extinguishing manner. A disadvantage of the method is the requirement of a clean surface without any oxide layer. A part of this disadvantage can

  9. Nuclear data for neutron emission in the fission process

    International Nuclear Information System (INIS)

    Ganesan, S.

    1991-11-01

    This document contains the proceedings of the IAEA Consultants' Meeting on Nuclear Data for Neutron Emission in the Fission Process, Vienna, 22 - 24 October 1990. Included are the conclusions and recommendations reached at the meeting and the papers presented by the meeting participants. These papers provide a review of the status of experimental and theoretical data on neutron emission in spontaneous and neutron induced fission with reference to the data needs for reactor applications oriented towards actinide burner studies. The specific topics covered are the following: experimental measurements and theoretical predictions and evaluations of fission neutron energy spectra, average prompt fission neutron multiplicity, correlation in neutron emission from complementary fragments, neutron emission during acceleration of fission fragments, statistical properties of neutron rich nuclei by study of emission spectra of neutrons from the excited fission fragments, integral qualification of nu-bar for the major fissile isotopes, nu-bar total of 239 Pu and 235 U, and related problems. Refs figs and tabs

  10. 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)

  11. Nuclear reactor fuelling machine

    International Nuclear Information System (INIS)

    Peberdy, J.M.

    1976-01-01

    The refuelling machine described comprises a rotatable support structure having a guide tube attached to it by a parellel linkage mechanism, whereby the guide tube can be displaced sideways from the support structure. A gripper unit is housed within the guide tube for gripping the end of a fuel assembly or other reactor component and has means for maintenance in the engaging condition during travel of the unit along the guide tube, except for a small portion of the travel at one end of the guide tube, where the inner surface of the guide tube is shaped so as to maintain the gripper unit in a disengaging condition. The gripper unit has a rotatable head, means for moving it linearly within the guide tube so that a component carried by the unit can be housed in the guide tube, and means for rotating the head of the unit through 180 0 relative to its body, to effect rotation of a component carried by the unit. The means for rotating the head of the gripper unit comprises ring and pinion gearing, operable through a series of rotatable shafts interconnected by universal couplings. The reason for provision for 180 0 rotation is that due to the variation in the neutron flux across the reactor core the side of a fuel assembly towards the outside of the core will be subjected to a lower neutron flux and therefore will grow less than the side of the fuel assembly towards the inside of the core. This can lead to bowing and possible jamming of the fuel assemblies. Full constructional details are given. See also BP 1112384. (U.K.)

  12. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y R, M.

    2012-07-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  13. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  14. The measurement of subcritical reactivity in nuclear reactors by use of a high frequency sine-wave modulated neutron source

    International Nuclear Information System (INIS)

    Guppy, C.B.

    1964-11-01

    In this report the frequency response characteristics for phase and gain of the fundamental reactor mode of the zero power kinetics are given for various subcritical reactivities in a fast reactor and in a thermal reactor. Results, of a study on harmonic effects based on a small zero energy thermal reactor are presented which demonstrate the importance of spatial harmonic effects. A harmonic theory for thermal reactors is developed. A new method of measuring, subcritical reactivity at moderately high frequencies is suggested which circumvents the harmonic problem. It is shown that at high frequencies there is more sensitivity than at low frequencies and that this could lead to an increased range over which subcritical reactivity can be measured. (author)

  15. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

  16. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The seals described are for use in a nuclear reactor where there are fuel assemblies in a vessel, an inlet and an outlet for circulating a coolant in heat transfer relationship with the fuel assemblies and a closure head on the vessel in a tight fluid relationship. The closure head comprises rotatable plugs which have mechanical seals disposed in the annulus around each plug while allowing free rotation of the plug when the seal is not actuated. The seal is usually an elastomer or copper. A means of actuating the seal is attached for drawing it vertically into the annulus for sealing. When the reactor coolant is liquid sodium, contact with oxygen must be avoided and argon cover gas fills the space between the bottom of the closure head and the coolant liquid level and the annuli in the closure head. (U.K.)

  17. Nuclear reactor container

    International Nuclear Information System (INIS)

    Yamaki, Rika; Kawabe, Ryuhei.

    1989-01-01

    A venturi scrubber is connected to a nuclear reactor container. Gases containing radioactive aerosols in the container are introduced into the venturi scrubber in the form of a high speed stream under the pressure of the container. The radioactive aerosols are captured by inertia collision due to the velocity difference between the high speed gas stream and water droplets. In the case of the present invention, since the high pressure of the reactor container generated upon accident is utilized, compressor, etc. is no more required, thereby enabling to reduce the size of the aerosol removing device. Further, since no external power is used, the radioactive aerosols can be removed with no starting failure upon accidents. (T.M.)

  18. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a nuclear reactor fuel assembly comprising a cluster of fuel elements supported by transversal grids so that their axes are parallel to and at a distance from each other, in order to establish interstices for the axial flow of a coolant. At least one of the interstices is occupied by an axial duct reserved for an auxiliary cooling fluid and is fitted with side holes through which the auxiliary cooling fluid is sprayed into the cluster. Deflectors extend as from a transversal grid in a position opposite the holes to deflect the cooling fluid jet towards those parts of the fuel elements that are not accessible to the auxiliary coolant. This assembly is intended for reactors cooled by light or heavy water [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-10-01

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

  20. 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)

  1. 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

  2. Fast reactors in nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkovskii, O

    1981-02-01

    The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

  3. Diagnosis of electric equipment at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Truong Sinh

    1999-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type of its kind in the world: Soviet-designed core and control system harmoniously integrated into the left-over infrastructure of the former American-made TRIGA MARK II reactor, which includes the reactor tank and shielding, graphite reflector, beam tubes and thermal column. The reactor is mainly used for radioisotope and radiopharmaceutical production, elemental analysis using neutron activation techniques, neutron beam exploitation, silicon doping, and reactor physics experimentation. For safe operation of the reactor maintenance work has been carried out for the reactor control and instrumentation, reactor cooling, ventilation, radiomonitoring, mechanical, normal electric supply systems as well as emergency electric diesel generators and the water treatment station. Technical management of the reactor includes periodical maintenance as required by technical specifications, training, re-training and control of knowledge for reactor staff. During recent years, periodic preventive maintenance (PPM) has been carried out for the electric machines of the technological systems. (author)

  4. Advanced nuclear reactor and nuclear fusion power generation

    International Nuclear Information System (INIS)

    2000-04-01

    This book comprised of two issues. The first one is a advanced nuclear reactor which describes nuclear fuel cycle and advanced nuclear reactor like liquid-metal reactor, advanced converter, HTR and extra advanced nuclear reactors. The second one is nuclear fusion for generation energy, which explains practical conditions for nuclear fusion, principle of multiple magnetic field, current situation of research on nuclear fusion, conception for nuclear fusion reactor and economics on nuclear fusion reactor.

  5. 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

  6. 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.)

  7. 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.)

  8. Nuclear reactor plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1977-01-01

    The invention is concerned with a quick-closing valve on the main-steam pipe of a nuclear reactor plant. The quick-closing valve serves as isolating valve and as safety valve permitting depressurization in case of an accident. For normal operation a tube-shaped gate valve is provided as valve disc, enclosing an auxiliary valve disc to be used in case of accidents and which is opened at increased pressure to provide a smaller flow cross-section. The design features are described in detail. (RW) [de

  9. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E. D.

    1984-01-01

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

  10. Nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E. D.

    1984-10-16

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

  11. Nuclear reactor control assembly

    International Nuclear Information System (INIS)

    Negron, S.B.

    1991-01-01

    This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other

  12. Nuclear reactor facility

    International Nuclear Information System (INIS)

    Wampole, N.C.

    1978-01-01

    In order to improve the performance of manitenance and inspections it is proposed for a nuclear reactor facility with a primary circuit containing liquid metal to provide a thermally insulated chamber, within which are placed a number of components of the primary circuit, as e.g. valves, recirculation pump, heat exchangers. The isolated placement permit controlled preheating on one hand, but prevents undesirable heating of adjacent load-bearing elements on the other. The chamber is provided with heating devices and, on the outside, with cooling devices; it is of advantage to fill it with an inert gas. (UWI) 891 HP [de

  13. Solution of the neutron diffusion equation to study the 3D distribution of power, applied to nuclear reactors; Solucao da equacao de difusao de neutrons para o estudo da distribuicao de potencia em 3D, aplicado a reatores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Danilo Leite

    2013-07-01

    This work aims to present a study about the power distribution behavior in a PWR type reactor, considering both intensity and migration of power peaks due to insertion of control rods into the core. Employing the multidimensional steady-state neutron diffusion equation in order to simulate the neutron flux, and using the Finite Difference Method. Furthermore, based on the axial power distribution on the largest heat flux rod, is carried out thermal analysis of this rod and associated coolant channel. For this purpose is employed the FueLRod{sub 3}D code, it uses the Finite Element Method to model the fuel rod and the associated coolant channel, allowing the thermohydraulics simulation of a single rod discretized in three dimensions, considering the heat flux from the pellet, crossing the gap and the cladding until it reaches the coolant. (author)

  14. Nuclear reactor container

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1989-01-01

    Aerosol filters considered so far for nuclear reactor containers in conventional BWR type nuclear power plants make the facility larger and involve a risk of clogging. In view of the above, in the present invention, the diameter of a flow channel of gases entering from a bent pipe to a suppression pool is made smaller thereby decreasing the diameter of gas bubbles in the supperssional pool. Since this reduces the force of surface tension, the diameter of resulted gas bubbles is made remarkably smaller as compared with the case where the gases are released from the lower end of the bent pipe. Since the absorption velocity of bubble-entrained aerosols into water is in proportion to the square of the bubble diameter, the absorption efficiency can be increased remarkably by reducing the diameter of the gas bubbles. Accordingly, it is possible to improve the efficiency of eliminating radioactivity of released gases. (K.M.)

  15. Polycrystalline semiconductor probes for monitoring the density distribution of an intense thermal neutron flux in nuclear reactors

    International Nuclear Information System (INIS)

    Graul, J.; Mueller, R.G.; Wagner, E.

    1975-05-01

    The applicability of semiconductor detectors for high thermal neutron flux densities is theoretically estimated and experimentally examined. For good thermal stability and low radiation capture rate silicon carbide is used as semiconductor material, produced in polycristalline layers to achieve high radiation resistance. The relations between crystallinity, photoelectric sensitivity and radiation resistance are shown. The radiation resistance of polycrystalline SiC-probes is approximately 100 times greater than that of conventional single crystal radiation detectors. For thermal neutron measurement they can be used in the flux range of approx. 10 10 13 (cm -2 sec -1 ) with operation times of 1.6 a >= tsub(b,max) >= 30 d, resp. (orig.) [de

  16. Surface harmonics method for two-dimensional time-dependent neutron transport problems of square-lattice nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V. F.; Kondrushin, A. E.; Fomichenko, P. A. [National Research Centre Kurchatov Institute, Kurchatov Sq. 1, Moscow (Russian Federation)

    2013-07-01

    Time-dependent equations of the Surface Harmonics Method (SHM) have been derived from the time-dependent neutron transport equation with explicit representation of delayed neutrons for solving the two-dimensional time-dependent problems. These equations have been realized in the SUHAM-TD code. The TWIGL benchmark problem has been used for verification of the SUHAM-TD code. The results of the study showed that computational costs required to achieve necessary accuracy of the solution can be an order of magnitude less than with the use of the conventional finite difference method (FDM). (authors)

  17. Nuclear polarization and neutrons

    International Nuclear Information System (INIS)

    Glaettli, H.

    1985-01-01

    Different possibilities for the use of polarized nuclei in thermal neutron scattering on condensed matter are reviewed. Highly polarized nuclei are the starting point for studying dipolar magnetic order. Systematic measurement of spin-dependent scattering lengths is possible on samples with polarized nuclei. Highly polarized hydrogen should help to unravel complicated structures in chemistry and biology. The use of polarized proton targets as an energy-independent neutron polarizer in the thermal and epithermal region should be considered afresh. (author)

  18. 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

  19. Thermionic nuclear reactor systems

    International Nuclear Information System (INIS)

    Kennel, E.B.

    1986-01-01

    Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

  20. Nuclear reactor installation

    International Nuclear Information System (INIS)

    Jungmann, A.

    1976-01-01

    A nuclear reactor metal pressure vessel is surrounded by a concrete wall forming an annular space around the vessel. Thermal insulation is in this space and surrounds the vessel, and a coolant-conductive layer is also in this space surrounding the thermal insulation, coolant forced through this layer reducing the thermal stress on the concrete wall. The coolant-conductive layer is formed by concrete blocks laid together and having coolant passages, these blocks being small enough individually to permit them to be cast from concrete at the reactor installation, the thermal insulation being formed by much larger sheet-metal clad concrete segments. Mortar is injected between the interfaces of the coolant-conductive layer and concrete wall and the interfaces between the fluid-conductive layer and the insulation, a layer of slippery sheet material being interposed between the insulation and the mortar. When the pressure vessel is thermally expanded by reactor operation, the annular space between it and the concrete wall is completely filled by these components so that zero-excursion rupture safeguard is provided for the vessel. 4 claims, 1 figure