Utilization of the experimental reactor Osiris for the study and the development of fuels of the fast neutron reactor type

International Nuclear Information System (INIS)

Marcon, M.; Faugere, J.L.; Genthon, J.P.; Maillot, R.

1977-01-01

Nuclear fuel tests for the fast neutron reactor type have been carried out at the Osiris reactor: thermal study of (U,Pu)O 2 oxide by measurement with thermocouples in the core of the fuel pellet; study of the effects of power cycling on nuclear fuel; study of the mechanical interactions between oxide and cladding by measurement of the cladding deformation during irradiation [fr

Neutronics analysis of Dalat Research Reactor

International Nuclear Information System (INIS)

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

2006-01-01

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

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

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)

Vanadium Beta Emission Detectors for Reactor In-Core Neutron Monitoring

Energy Technology Data Exchange (ETDEWEB)

Andersson, I Oe; Soederlund, B

1969-06-15

In-core flux measurements are becoming increasingly important in both power reactors and test reactors. In particular power distribution measurements in large power reactors have to be performed with a great number of neutron detectors capable of withstanding high integrated flux values. This report presents a summary of the development and application of a new type of nuclear radiation sensor, a beta emission detector, for measurements at high neutron flux levels. The work has been carried out at the Section for Instrumentation and has been the basis for a type of neutron detector employed in the Marviken in-core system as well as for other types. The report describes the design and principle of operation, experiments and tests. Also included are the results and comments from a long-term irradiation of some detectors in the Halden reactor.

Neutronics analysis of the proposed 25-MW leu TRIGA Multipurpose Research Reactor

International Nuclear Information System (INIS)

Nurdin, M.; Bretscher, M.M.; Snelgrove, J.L.

1982-01-01

More than two years ago the government of Indonesia announced plans to purchase a research reactor for the Puspiptek Research Center in Serpong Indonesia to be used for isotope production, materials testing, neutron physics measurements, and reactor operator training. Reactors using low-enriched uranium (LEU) plate-type and rod-type fuel elements were considered. This paper deals with the neutronic evaluation of the rod-type 25-MW LEU TRIGA Multipurpose Research Reactor (MPRR) proposed by the General Atomic Company of the United States of America

Trace element analysis at the Livermore pool-type reactor using neutron activation techniques

International Nuclear Information System (INIS)

Ragaini, R.C.; Ralston, R.; Garvis, D.

1975-01-01

The capabilities of trace element analysis at the Livermore Pool-Type Reactor (LPTR) using instrumental neutron activation analysis (INAA) are discussed. A description is given of the technology and the methods employed, including sample preparation, irradiation, and analysis. Applications of the INAA technique in past and current projects are described. A computer program, GAMANAL, has been used for nuclide identification and quantification. (U.S.)

Calculation of the neutron parameters of fast thermal reactor

International Nuclear Information System (INIS)

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

1975-01-01

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

BWR type reactors

International Nuclear Information System (INIS)

Watanabe, Shoichi

1986-01-01

Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

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)

Utilization of cold neutron beams at intermediate flux reactors

International Nuclear Information System (INIS)

Clark, D.D.

1992-01-01

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

Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

Energy Technology Data Exchange (ETDEWEB)

Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

2017-05-15

This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

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

Cold neutron source conceptual designing for Tehran Research Reactor

International Nuclear Information System (INIS)

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

2016-01-01

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

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)

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

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.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

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

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

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

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.

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

Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

Directory of Open Access Journals (Sweden)

Kaiser Krista

2016-01-01

Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

Proposal of a neutron transmutation doping facility for n-type spherical silicon solar cell at high-temperature engineering test reactor.

Science.gov (United States)

Ho, Hai Quan; Honda, Yuki; Motoyama, Mizuki; Hamamoto, Shimpei; Ishii, Toshiaki; Ishitsuka, Etsuo

2018-05-01

The p-type spherical silicon solar cell is a candidate for future solar energy with low fabrication cost, however, its conversion efficiency is only about 10%. The conversion efficiency of a silicon solar cell can be increased by using n-type silicon semiconductor as a substrate. This study proposed a new method of neutron transmutation doping silicon (NTD-Si) for producing the n-type spherical solar cell, in which the Si-particles are irradiated directly instead of the cylinder Si-ingot as in the conventional NTD-Si. By using a 'screw', an identical resistivity could be achieved for the Si-particles without a complicated procedure as in the NTD with Si-ingot. Also, the reactivity and neutron flux swing could be kept to a minimum because of the continuous irradiation of the Si-particles. A high temperature engineering test reactor (HTTR), which is located in Japan, was used as a reference reactor in this study. Neutronic calculations showed that the HTTR has a capability to produce about 40t/EFPY of 10Ωcm resistivity Si-particles for fabrication of the n-type spherical solar cell. Copyright © 2018 Elsevier Ltd. All rights reserved.

Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor

International Nuclear Information System (INIS)

Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.

2006-10-01

Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)

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

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

Neutron cross-section libraries in the AMPX master interface format for thermal and fast reactors

International Nuclear Information System (INIS)

Bjerke, M.A.; Webster, C.C.

1981-12-01

Neutron cross-section libraries in the AMPX master interface format have been created for three reactor types. Included are an 84-group library for use with light-water reactors, a 27-group library for use with heavy-water CANDU reactors and a 126-group library for use with liquid metal fast breeder reactors. In general, ENDF/B data were used in the creation of these libraries, and the nuclides included in each library should be sufficient for most neutronic analyses of reactors of that type. Each library has been used successfully in fuel depletion calculations

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.)

FBR type reactor

International Nuclear Information System (INIS)

Hayase, Tamotsu.

1991-01-01

The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

BWR type reactors

International Nuclear Information System (INIS)

Nakajima, Yoshitaka

1983-01-01

Purpose: To decrease the control rod exchanging frequency by increasing the working life of control rods for ordinary operation with large neutron irradiation dose, to thereby decrease the exposure dose for operators performing exchanging work, as well as decrease the amount of radioactive wastes resulted upon exchange of the control rods. Constitution: Hafnium solid metal is employed as the neutron absorber of control rods for usual operation inserted into and withdrawn from fuel assemblies for the reactor power control over the entire cycle of the ordinary reactor operation and boron carbide powder is employed as the neutron absorber for emergency control rods to be inserted between the fuel assemblies only upon reactor scram or shutdown, whereby the working life of the control rods for ordinary reactor operation with greater neutron irradiation dose can be improved. Accordingly, the control rod exchanging frequency can be reduced to decrease the exposure dose to the operator for conducting the exchanging work. (Yoshihara, H.)

Calculations on neutron irradiation damage in reactor materials

International Nuclear Information System (INIS)

Sone, Kazuho; Shiraishi, Kensuke

1976-01-01

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

Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Aguilar H, F.

2015-09-01

The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO 3 ) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10 18 n/cm 2 , which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

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

Neutron streaming evaluation for the DREAM fusion power reactor

International Nuclear Information System (INIS)

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

2000-01-01

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

Intense neutron irradiation facility for fusion reactor materials

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

Neutron radiography (NRAD) reactor 64-element core upgrade

Energy Technology Data Exchange (ETDEWEB)

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

2014-03-01

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

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

International Nuclear Information System (INIS)

Less, T.J.

1987-01-01

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

Multicomponent activation detector measurements of reactor neutron spectra

International Nuclear Information System (INIS)

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

1984-01-01

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

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

A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

2007-11-01

The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

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

Actinide neutron induced cross section measurements using the oscillation technique in the Minerve reactor

Energy Technology Data Exchange (ETDEWEB)

Bernard, B.; Leconte, P.; Gruel, A.; Antony, M.; Di-Salvo, J.; Hudelot, J.P.; Pepino, A.; Lecluze, A. [CEA Cadarache, DEN/CAD/DER/SPRC/LEPh, 13 - Saint-Paul-lez-Durance (France)

2009-07-01

CEA is deeply involved research programs concerning nuclear fuel advanced studies (actinides, plutonium), waste management, the scientific and technical support of French PWR reactors and EPR reactor, and innovative systems. In this framework, specific neutron integral experiments have been carried out in the critical ZPR (zero power reactor) facilities of the CEA at Cadarache such as MINERVE, EOLE and MASURCA. This paper deals with MINERVE Pool Reactor experiments. MINERVE is mainly devoted to neutronics studies of different reactor core types. The aim is to improve the knowledge of the integral absorption cross sections of actinides (OSMOSE program), of new absorbers (OCEAN program) and also for fission Products (CBU program) in thermal, epithermal and fast neutron spectra. (authors)

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.)

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

International Nuclear Information System (INIS)

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

1974-01-01

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

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.)

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)

Strategies of development of reactor types

International Nuclear Information System (INIS)

Bacher, P.

2004-01-01

The development of nuclear energy in the coming decades will depend on the goals followed, on the available technologies and on the strategies implemented in the world in agreement with public acceptation. This article is limited to the technical aspects of the strategies of development of reactor types: 1 - objectives; 2 - common constraints to all reactor types: safety and terrorism risks, wastes, non-proliferation, economics; 3 - different reactor types: general considerations, proven technologies (PWR, BWR, Candu), non-proven technologies but having an important experience, technologies at the design stage; 4 - energy systems and 'Generation IV forum': systems based on thermal neutron reactors and low enrichment, systems for the valorization of 238 U, systems for Pu burning, systems allowing the destruction of minor actinides, thorium-based systems, the Gen IV international forum; 5 - conclusion. (J.S.)

INR TRIGA Research Reactors: A Neutron Source for Radioisotopes and Materials Investigation

International Nuclear Information System (INIS)

Barbos, D.; Ciocanescu, M.; Paunoiu, C.; Bucsa, A.F.

2013-01-01

At the INR there are 2 high intensity neutron sources. These sources are in fact the two nuclear TRIGA reactors: TRIGA SSR 14 MW and TRIGA ACPR. TRIGA stationary reactor is provided with several in-core irradiation channels. Other several out-of-core irradiation channels are located in the vertical channels in the beryllium reflector blocks. The maximum value of the thermal neutron flux (E 14 cm -2 s -1 and of fast neutron flux (E>1 MeV) is 6.89×10 13 cm -2 s -1 . For neutron activation analysis both reactors are used and k0-NAA method has been implemented. At INR Pitesti a prompt gamma ray neutron activation analysis devices has been designed, manufactured ant put into operation. For nuclear materials properties investigation neutron radiography methods was developed in INR. For these purposes two neutron radiography devices were manufacture, one of them underwater and other one dry. The neutron beams are used for investigation of materials properties and components produced or under development for applications in the energy sector (fission and fusion). At TRIGA 14 MW reactor a neutron difractormeter and a SANS devices are available for material residual stress and texture measurements. TRIGA 14 MW reactor is used for medical and industrial radioisotopes production ( 131 I, 125 I, 192 Ir, etc) and a method for 99 Mo- 99 Tc production from fission is under developing. At INR Pitesti several special programmes for new types of nuclear fuel behavior characterization are under development. (author)

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

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)

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

International Nuclear Information System (INIS)

Duran, I.; Bolshakova, I.; Holyaka, R.; Viererbl, L.; Lahodova, Z.; Sentkerestiova, J.; Bem, P.

2010-01-01

We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10 16 cm -2 was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

Neutron beam facilities at the replacement research reactor

International Nuclear Information System (INIS)

Kennedy, S.

1999-01-01

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

F-type centers in neutron-irradiated AIN

International Nuclear Information System (INIS)

Atobe, Kozo; Honda, Makoto; Fukuoka, Noboru; Okada, Moritami; Nakagawa, Masuo.

1990-01-01

The production of point defects by neutron irradiation and thermal decay in sintered AIN polycrystal are investigated. The absorption band at 370 nm is observed after reactor neutron irradiation to a dose of 10 16 n/cm 2 (E > 0.1 MeV). The defect corresponding to the band is tentatively assigned as an F-type center from the optical absorption and electron spin resonance. (author)

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.)

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Neutron noise measurement technique in a coupled reactor

International Nuclear Information System (INIS)

Genoud, J.P.

1976-01-01

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

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)

Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

Brauer, G.; Matz, W.; Liszkay, L.; Molnar, B.

1990-11-01

Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

Time changes of vertical profile of neutron fluence rate in LVR-15 reactor

International Nuclear Information System (INIS)

Viererbl, L.; Stehno, J.; Erben, O.; Lahodova, Z.; Marek, M.

2003-01-01

The LVR-15 reactor is a light water research type reactor, which is situated, in Nuclear Research Institute, Rez near Prague. The reactor is used as a multipurpose facility. For some experiments and material productions, e.g. for homogeneity of silicon resistance in production of radiation doped silicon, the time changes of vertical profile of neutron fluence rate are particularly important. The assembly used for silicon irradiation has two self-powered neutron detectors installed in a vertical irradiation channel in LVR-15 reactor. Vertical profile of thermal neutron fluence rate was automatically scanned during reactor operation. The results of measurements made in 2002 and 2003 with these detectors are presented. A set of vertical profile measurements was made during two 21-days reactor cycles. During the cycle the vertical profile slightly changes both in the position of its maximum and in the shape. The time dependences of the position of profile maximum and the profile width at half maximum during the cycle are given. (author)

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.)

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

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

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)

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

International Nuclear Information System (INIS)

Jammes, Ch.

2010-07-01

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

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)

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.)

Graphs of neutron cross section data for fusion reactor development

International Nuclear Information System (INIS)

Asami, Tetsuo; Tanaka, Shigeya

1979-03-01

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

Neutronic study of the two french heavy water reactors

International Nuclear Information System (INIS)

Horowitz, J.

1955-01-01

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

The neutron beam facility at the Australian replacement research reactor

International Nuclear Information System (INIS)

Hunter, B.; Kennedy, S.

1999-01-01

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

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

Studies on the molten salt reactor. Code development and neutronics analysis of MSRE-type design

International Nuclear Information System (INIS)

Zhuang Kun; Cao Liangzhi; Zheng Youqi; Wu Hongchun

2015-01-01

The molten salt reactor is characterized by its use of the fluid-fuel, which serves both as a fuel and as a coolant simultaneously. The position of delayed neutron precursors continuously changes both in the core and in the external loop due to the fuel circulation, and the fission products are extracted by an online fuel reprocessing unit, which all lead to the modeling methods for the conventional reactors using solid fuel not applicable. This study establishes suitable calculation models for the neutronics analysis of the molten salt reactor and develops a new code named MOREL based on the three-dimensional diffusion steady and transient calculations. Some numerical tests are chosen to verify the code and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactor. After verification, it is applied to analyze the characteristics of a typical molten salt reactor, including the steady characteristics, the influence of fuel circulation on the kinetic behaviors. Besides, the influence of online fuel reprocessing simulation is also examined. The results show that inherent safety is the character of the molten salt reactor from the aspect of reactivity feedback and the fuel circulation has great influence on the kinetic characteristics of molten salt reactor. (author)

WWER-440 type reactor core

International Nuclear Information System (INIS)

Mizov, J.; Svec, P.; Rajci, T.

1987-01-01

Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

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.

Measurement of the energy spectrum of the neutrons inside the neutron flux trap assembled in the center of the reactor core IPEN/MB-01

Energy Technology Data Exchange (ETDEWEB)

Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto Credidio; Santos, Diogo Feliciano dos; Jerez, Rogerio; Mura, Luis Felipe Liamos, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

This paper presents the neutron energy spectrum in the central position of a neutron flux trap assembled in the core center of the research nuclear reactor IPEN/MB-01 obtained by an unfolding method. To this end, have been used several different types of activation foils (Au, Sc, Ti, Ni, and plates) which have been irradiated in the central position of the reactor core (setting number 203) at a reactor power level of 64.57 ±2.91 watts . The activation foils were counted by solid-state detector HPGe (gamma spectrometry). The experimental data of nuclear reaction rates (saturated activity per target nucleus) and a neutron spectrum estimated by a reactor physics computer code are the main input data to get the most suitable neutron spectrum in the irradiation position obtained through SANDBP code: a neutron spectra unfolding code that use an iterative adjustment method. The adjustment resulted in 3.85 ± 0.14 10{sup 9} n cm{sup -2} s{sup -1} for the integral neutron flux, 2.41 ± 0.01 10{sup 9} n cm{sup -2} s{sup -1} for the thermal neutron flux, 1.09 ± 0.02 10{sup 9} n cm{sup -2} s{sup -1} for intermediate neutron flux and 3.41± 0.02 10{sup 8} n cm{sup -2} s{sup -1} for the fast neutrons flux. These results can be used to verify and validate the nuclear reactor codes and its associated nuclear data libraries, besides show how much is effective the use of a neutron flux trap in the nuclear reactor core to increase the thermal neutron flux without increase the operation reactor power level. The thermal neutral flux increased 4.04 ± 0.21 times compared with the standard configuration of the reactor core. (author)

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.

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.)

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

International Nuclear Information System (INIS)

1989-03-01

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

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

Neutron beam facilities at the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Kim, S.

2003-01-01

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

A multi-group neutron noise simulator for fast reactors

International Nuclear Information System (INIS)

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

2013-01-01

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

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

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

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki; An, Shigehiro.

1983-01-01

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

The IAEA collaborating centre for neutron activation based methodologies of research reactors

International Nuclear Information System (INIS)

Bode, P.

2010-01-01

The Reactor Institute Delft of the Delft University of Technology houses the Netherlands' only academic nuclear research reactor, with associated instrumentation and laboratories, for scientific education and research with ionizing radiation. The Institute's swimming pool type research reactor reached first criticality in 1963 and is currently operated at 2MW thermal powers on a 100 h/week basis. The reactor is equipped with neutron mirror guides serving ultra modern neutron beam physics instruments and with a very bright positron facility. Fully automated gamma-ray spectrometry systems are used by the laboratory for neutron activation analysis, providing large scale services under an ISO/IEC 17025:2005 compliant management system, being (since 1993) the first accredited laboratory of its kind in the world. Already for several years, this laboratory is sustainable by rendering these services to both the public and the private sector. The prime user of the Institute's fac ilities is the scientific Research Department of Radiation, Radionuclide and Reactors of the Faculty of Applied Sciences, housed inside the building. All reactor facilities are also made available for use by or for services to, external clients (industry, government, private sector, other (international research institutes and universities). The Reactor Institute Delft was inaugurated in May 2009 as a new lAEA Collaborating Centre for Neutron Activation Based Methodologies of Research Reactors. The collaboration involves education, research and development in (I) Production of reactor-produced, no-carrier added radioisotopes of high specific activity via neutron activation; (II) Neutron activation analysis with emphasis on automation as well as analysis of large samples, and radiotracer techniques and as a cross-cutting activity, (IIl) Quality assurance and management in research and application of research reactor based techniques and in research reactor operations. This c ollaboration will

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

Commissioning of the Opal reactor cold neutron source

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Neutronic characteristics of linear-assembly breed-and-burn reactors

International Nuclear Information System (INIS)

Petroski, Robert; Forget, Benoit; Forsberg, Charles

2012-01-01

Highlights: ► Simple models used to characterize general behavior of linear-assembly B and B reactors. ► Diffusion theory model developed to explain axial distributions, height vs. reactivity. ► Neutron excess concept reformulated to include linear-assembly B and B reactors. ► Designed model of B and B reactor started using melt-refined B and B reactor used fuel. ► Computed doubling time of fuel cycle requiring no chemical separations. - Abstract: Linear-assembly breed-and-burn (B and B) reactors are B and B reactors that use axially connected assemblies similar to conventional LWR or fast reactor fuel assemblies. Methods for analyzing linear-assembly B and B reactors and their fuel cycles are developed and applied. General neutronic characteristics of linear-assembly B and B reactors are analyzed, including the effects that burnup, shuffling sequence, and radial and axial size have on equilibrium-cycle k-effective. The mechanisms that give rise to a highly peaked axial burnup distribution are explained, and a method for predicting peak burnup vs. k-effective based on infinite-medium depletion calculations is developed. Next, the neutron excess concept from previous studies of B and B reactors is extended to apply to linear-assembly B and B reactors, which allows the amount of starter fuel needed to establish a given equilibrium cycle to be calculated. Several example applications of the neutron excess formulation are given. First, an example model of a linear-assembly B and B reactor is analyzed to find the neutron excess cost of an equilibrium cycle. Second, simple one-dimensional models are used to predict the neutron excess value obtainable from different starter fuel configurations. Finally, these ideas are applied to design a fuel cycle consisting of linear-assembly B and B reactors and fuel recycling via a melt refining process. The neutron excess concept is used to design an appropriate starter fuel configuration made from melt refined fuel, which

Photon and fast neutron dosimetry using aluminium oxide thermoluminescence dosemeters in a pool-type research reactor

International Nuclear Information System (INIS)

Santos, J.P.; Marques, J.G.; Fernandes, A.C.; Osvay, M.

2007-01-01

Al 2 O 3 :Mg,Y thermoluminescence (TL) dosemeters were used to measure photon and fast neutron doses in the mixed radiation field of the Portuguese Research Reactor. The dosemeters were irradiated in core positions under a photon dose rate of the order of 10 4 Gy/h and a fast neutron flux in the range of 10 9 -10 11 n/cm 2 /s. In order to evaluate the ability of the TL dosemeters for mixed field dosimetry at the research reactor, the measurements were compared with results obtained via conventional methods. The agreement between the different methods is better than 13% for the determination of photon doses and within 5% for the determination of neutron fluxes in mixed fields

In-pile modelling of nuclear fuel element for the MTR type reactors. Pt. 2

Energy Technology Data Exchange (ETDEWEB)

Farhadi, Kazem [AEOI, Tehran (Iran, Islamic Republic of). Radiations Application Research School

2014-06-15

In part two of the present paper, neutronic properties of the pool-type research reactor core are used to assess the similitude laws derived for out-of-pile modelling of the fuel element. The benchmark reactor used for this purpose is an IAEA 5 MW thermal pool-type research reactor currently in operation. The neutronic properties analysis are based on typical 2 200 m/sec and neutrons having 0.025 eV energy. The non-leakage capability of the system is estimated in terms of diffusion length. Also the slowing down power and the moderating ratio of the modelled methanol coolant are calculated in terms of lethargy of the diffusing medium. It is shown that the Iron which is substituted for Aluminium cladding is a relatively low absorber of neutrons but has a high neutron leakage. Methanol which replaced ordinary water as coolant is not a suitable coolant due to high neutrons absorbing substance. It is concluded that although Iron as a cladding material and methanol as a coolant meet the modelling out-of-pile criteria but are not satisfying neutronic properties. Therefore, use of them as a model clad and coolant are not suggested for research reactors. (orig.)

Neutron radiography

International Nuclear Information System (INIS)

Hrdlicka, Z.

1977-01-01

Neutron radiography is a radiographic method using a neutron beam of a defined geometry. The neutron source usually consists of a research reactor, a specialized neutron radiography reactor or the 252 Cf radioisotope source. There are two types of the neutron radiography display system, viz., a system producing neutron radiography images by a photographic process or a system allowing a visual display, eg., using a television monitor. The method can be used wherever X-ray radiography is used except applications in the radiography of humans. The neutron radiography unit at UJV uses the WWR-S reactor as the neutron source and both types of the above mentioned display system. (J.P.)

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)

Three-dimensional reactor dynamics code for VVER type nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

Three-dimensional reactor dynamics code for VVER type nuclear reactors

International Nuclear Information System (INIS)

Kyrki-Rajamaeki, R.

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

Neutron beam utilization at the TRIGA Mark II reactor Vienna

International Nuclear Information System (INIS)

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

1999-01-01

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

Research Reactors Types and Utilization

International Nuclear Information System (INIS)

Abdelrazek, I.D.

2008-01-01

A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235 U or 239 Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

Maximum neutron flux in thermal reactors

International Nuclear Information System (INIS)

Strugar, P.V.

1968-12-01

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

Development of long-life neutron detectors for the prototype heavy water reactor 'Fugen'

International Nuclear Information System (INIS)

Ohteru, Shigeru; Shirayama, Shimpey.

1981-01-01

The development of long-life neutron detectors as the flux monitors for the prototype heavy water reactor has been made. Three kinds of neutron monitors, namely start-up monitor (SUM), power up monitor (PUM) and local power monitor (LPM), are provided. The LPM consists of 4 ion chamber type neutron detectors and a guide tube of power calibration monitor (PCM). This is useful for reactor control and fuel soundness monitor. The improvement of the neutron detectors was made for the operation under high neutron flux and gamma-ray heating. For the long-life operation, U-234 was mixed into U-235 for the conversion in the detectors. The ratio of U-234 to U-235 is 3 to 1. The PCM is also an ion chamber type detector with U-235. The mixing ratio of U-234 to U-235 was determined by a test with the JMTR. The characteristic performance was also investigated by the JMTR. After the completion of Fugen, various tests on the long-life detectors were performed with Fugen. It was hard to test the output linearity of the detectors with a large scale reactor. Therefore, it was tested that the operation range of the detectors is within the linear region of detector output. The voltage-current characteristics and the correlation of output current and saturation current were measured. The variation of the neutron sensitivity of the detectors with the cumulative dose was also studied. (Kato, T.)

Utilization of the Dalat Research Reactor for Radioisotope Production, Neutron Activation Analysis, Research and Training

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Duong Van Dong; Cao Dong Vu; Nguyen Xuan Hai

2013-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool type reactor loaded with a mixed core of HEU (36% enrichment) and LEU (19.75% enrichment) fuel assemblies. The reactor is used as a neutron source for the purposes of radioisotopes production, neutron activation analysis, basic and applied research and training. The reactor is operated mainly in continuous runs of 108 hours for cycles of 3–4 weeks for the above mentioned purposes. The current status of safety, operation and utilization of the reactor is given and some aspects for improvement of commercial products and services of the DNRR are also discussed in this paper. (author)

Residual stress analysis by neutron time-of-flight at a reactor source

International Nuclear Information System (INIS)

Priesmeyer, H.G.; Schroder, J.

1990-01-01

Non-destructive neutron diffractometry for stress analysis will be a powerful experimental tool in material science research performed at the GKSS 5 MW reactor FRG-1. Arguments which show the advantages of the time-of-flight method are given and a suitable high-resolution neutron-efficient type of spectrometer is introduced. First results derived from this method are presented

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

International Nuclear Information System (INIS)

Kennedy, Shane J.

2006-01-01

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

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-06-01

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

Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

International Nuclear Information System (INIS)

Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

2012-01-01

The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

International Nuclear Information System (INIS)

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

2011-01-01

Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

Experimental measurements and theoretical simulations for neutron flux in self-serve facility of Dhruva reactor

International Nuclear Information System (INIS)

Rana, Y.S.; Mishra, Abhishek; Singh, Tej

2016-06-01

Dhruva is a 100 MW th tank type research reactor with natural metallic uranium as fuel and heavy water as coolant, moderator and reflector. The reactor is utilized for production of a large variety of radioisotopes for fulfilling growing demands of various applications in industrial, agricultural and medicinal sectors, and neutron beam research in condensed matter physics. The core consists of two on-power tray rods for radioisotope production and fifteen experimental beam holes for neutron beam research. Recently, a self-serve facility has also been commissioned in one of the through tubes in the reactor for carrying out short term irradiations. To get accurate information about neutron flux spectrum, measurements have been carried out in self-serve facility of Dhruva reactor. The present report describes measurement method, analysis technique and results. Theoretical estimations for neutron flux were also carried out and a comparison between theoretical and experimental results is made. (author)

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)

The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

International Nuclear Information System (INIS)

Zheng Youqi; Zu Tiejun; Wu Hongchun; Cao Liangzhi; Yang Chao

2012-01-01

In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.

Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

Energy Technology Data Exchange (ETDEWEB)

Lescop, B.; Badeau, G.; Ivanovic, S.; Foulon, F. [National Institute for Nuclear science and Technology French Atomic Energy and Alternative Energies Commission (CEA), Saclay Research Center, 91191 Gif-sur-Yvette (France)

2015-07-01

Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. The facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also

Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

International Nuclear Information System (INIS)

Lescop, B.; Badeau, G.; Ivanovic, S.; Foulon, F.

2015-01-01

Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. The facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10 12 n/cm 2 .s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also modify the

Evaluation of uncertainties of key neutron parameters of PWR-type reactors with slab fuel, application to neutronic conformity

International Nuclear Information System (INIS)

Bernard, D.

2001-12-01

The aim of this thesis was to evaluate uncertainties of key neutron parameters of slab reactors. Uncertainties sources have many origins, technologic origin for parameters of fabrication and physical origin for nuclear data. First, each contribution of uncertainties is calculated and finally, a factor of uncertainties is associated to key slab parameter like reactivity, isotherm reactivity coefficient, control rod efficiency, power form factor before irradiation and life-time. This factors of uncertainties were computed by Generalized Perturbations Theory in case of step 0 and by directs calculations in case of irradiation problems. One of neutronic conformity applications was about fabrication and nuclear data targets precision adjustments. Statistic (uncertainties) and deterministic (deviations) approaches were studied. Then, neutronics key slab parameters uncertainties were reduced and so nuclear performances were optimized. (author)

Self operation type reactor control device

International Nuclear Information System (INIS)

Saito, Makoto; Gunji, Minoru.

1990-01-01

A boiling-requefication chamber containing transporting materials having somewhat higher boiling point that the usual reactor operation temperature and liquid neutron absorbers having a boiling point sufficiently higher than that of the transporting materials is disposed near the coolant exit of a fuel assembly and connected with a tubular chamber in the reactor core with a moving pipe at the bottom. Since the transporting materials in the boiling-requefication chamber is boiled and expanded by heating, the liquid neutron absorbers are introduced passing through the moving pipe into the cylindrical chamber to control the nuclear reactions. When the temperature is lowered by the control, the transporting materials are liquefied to contract the volume and the liquid neutron absorbers in the cylindrical chamber are returned passing through the moving tube into the boiling-liquefication chamber to make the nuclear reaction vigorous. Thus, self-operation type power conditioning and power stopping are enabled not by way of control rods and not requiring external control, to prevent scram failure or misoperation. (N.H.)

Neutron flux measurements in PUSPATI Triga Reactor

International Nuclear Information System (INIS)

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

1983-01-01

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

Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

International Nuclear Information System (INIS)

Tariq Siddique, M.; Kim, Myung Hyun

2014-01-01

Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM

Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

Energy Technology Data Exchange (ETDEWEB)

Tariq Siddique, M.; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of)

2014-05-15

Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM.

BWR type reactor

International Nuclear Information System (INIS)

Watanabe, Shoichi

1988-01-01

Purpose: To inhibit the lowering of the neutron moderation effect due to voids in the upper portion of the reactor core, thereby flatten the axial power distribution. Constitution: Although it has been proposed to enlarge the diameter at the upper portion of a water rod thereby increasing the moderator in the upper portion, since the water rod situates within the channel box, the increase in the capacity thereof is has certain limit. In the present invention, it is designed such that the volume of the region at the outside of the channel box for the fuel assembly to which non-boiling water in the non-boiling water region can enter is made greater in the upper portion than in the lower portion of the reactor core. Thus, if the moderator density in the upper portion of the reactor core should be decreased due to the generation of the voids, the neutron moderating effect in the upper portion of the reactor core is not lowered as compared with the lower portion of the reactor core and, accordingly, the axial power distribution can be flattening more as compared with that in the conventional nuclear reactors. (Takahashi, M.)

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

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

International Nuclear Information System (INIS)

Bisanti, Paola; Lovesey, S.W.

1987-05-01

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

Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor; Estimacion de la fluencia de neutrones rapidos en probetas de acero tipo Laguna Verde en el reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO{sub 3}) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10{sup 18} n/cm{sup 2}, which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

Report on neutron reflectometry for the Australian Replacement Reactor

International Nuclear Information System (INIS)

James, M.

2001-01-01

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

Behavior of Type 316 stainless steel under simulated fusion reactor irradiation

International Nuclear Information System (INIS)

Wiffen, F.W.; Maziasz, P.J.; Bloom, E.E.; Stiegler, J.O.; Grossbeck, M.L.

1978-05-01

Fusion reactor irradiation response in alloys containing nickel can be simulated in thermal-spectrum fission reactors, where displacement damage is produced by the high-energy neutrons and helium is produced by the capture of two thermal neutrons in the reactions: 58 Ni + n → 59 Ni + γ; 59 Ni + n → 56 Fe + α. Examination of type 316 stainless steel specimens irradiated in HFIR has shown that swelling due to cavity formation and degradation of mechanical properties are more severe than can be predicted from fast reactor irradiations, where the helium contents produced are far too low to simulate fusion reactor service. Swelling values are greater and the temperature dependence of swelling is different than in the fast reactor case

Pulsed TRIGA reactor as substitute for long pulse spallation neutron source

International Nuclear Information System (INIS)

Whittemore, W.L.

1999-01-01

TRIGA reactor cores have been used to demonstrate various pulsing applications. The TRIGA reactor fuel (U-ZrH x ) is very robust especially in pulsing applications. The features required to produce 50 pulses per second have been successfully demonstrated individually, including pulse tests with small diameter fuel rods. A partially optimized core has been evaluated for pulses at 50 Hz with peak pulsed power up to 100 MW and an average power up to 10 MW. Depending on the design, the full width at half power of the individual pulses can range between 2000 μsec to 3000 μsec. Until recently, the relatively long pulses (2000 μsec to 3000 μsec) from a pulsed thermal reactor or a long pulse spallation source (LPSS) have been considered unsuitable for time-of-flight measurements of neutron scattering. More recently considerable attention has been devoted to evaluating the performance of long pulse (1000 to 4000 μs) spallation sources for the same type of neutron measurements originally performed only with short pulses from spallation sources (SPSS). Adequate information is available to permit meaningful comparisons between CW, SPSS, and LPSS neutron sources. Except where extremely high resolution is required (fraction of a percent), which does require short pulses, it is demonstrated that the LPSS source with a 1000 msec or longer pulse length and a repetition rate of 50 to 60 Hz gives results comparable to those from the 60 MW ILL (CW) source. For many of these applications the shorter pulse is not necessarily a disadvantage, but it is not an advantage over the long pulse system. In one study, the conclusion is that a 5 MW 2000 μsec LPSS source improves the capability for structural biology studies of macromolecules by at least a factor of 5 over that achievable with a high flux reactor. Recent studies have identified the advantages and usefulness of long pulse neutron sources. It is evident that the multiple pulse TRIGA reactor can produce pulses comparable to

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

International Nuclear Information System (INIS)

Kennedy, J.S.

2005-01-01

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

Updated neutron spectrum characterization of SNL baseline reactor environments

International Nuclear Information System (INIS)

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

1994-04-01

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

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

International Nuclear Information System (INIS)

Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

2015-01-01

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF 2 -ThF 4 - 233 UF 4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

Experimental determination of neutron temperature distribution in reactor cell

International Nuclear Information System (INIS)

Bosevski, T.

1965-12-01

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

Neutronics comparative analysis between MNSR and slowpoke-II reactors

International Nuclear Information System (INIS)

Khamis, I.; Khattab, K.

1999-01-01

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

System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

International Nuclear Information System (INIS)

Moiseyev, A.V.

2008-01-01

There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k eff , control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

2008-07-01

There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Thermal neutron flux distribution in ET-RR-2 reactor thermal column

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

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)

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

International Nuclear Information System (INIS)

Hawari, Ayman

2014-01-01

This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

Experimental study of neutron streaming through steel-walled annular ducts in reactor shields

International Nuclear Information System (INIS)

Toshimas, M.; Nobuo, S.

1983-01-01

For the purpose of providing experimental data to assess neutron streaming calculations, neutron flux measurements were performed along the axes of the steel-walled annular ducts set up in a water shield of the pool-type reactor JRR-4. An annular duct simulated the air gap around the main coolant pipe. Another duct simulated the streaming path around the primary circulating pump of the integrated-type marine reactor. A 90-deg bend annular duct was also studied. In a set of measurements, the distance Z between the core center and the duct axis and the annular gap width delta were taken as parameters, that is, Z = 0, 80, and 160 cm and delta = 2.2, 4.7, and 10.1 cm. The reaction rates and the fluxes measured by the activation method are given in terms of absolute magnitude within an accuracy of + or - 30%. An empirical formula is derived based on those measured data, which describes the axial distribution of the neutron flux in the steel-walled annular duct in reactor shields. It is expressed by a simple function of the axial distance in units of the square root of the line-of-sight area, S /SUB l/ . The accuracy of the formula is examined by taking into account the duct location with respect to the reactor core, the neutron energy, the steel wall thickness, and the media outside of the steel wall. The accuracy of the formula is, in general, <30% in the axial distance between 3√S /SUB l/ and 30√S /SUB l/

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

Science.gov (United States)

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

2018-02-01

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

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

Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

2008-10-29

The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

A new neutron noise technique for fast reactors

International Nuclear Information System (INIS)

Zhuo Fengguan; Jin Manyi; Yao Shigui; Su Zhuting

1987-12-01

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

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

International Nuclear Information System (INIS)

Menlove, H.O.

1981-08-01

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

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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)

Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

International Nuclear Information System (INIS)

Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

2016-01-01

One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

Personal neutron dosimetry at a research reactor facility

International Nuclear Information System (INIS)

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

2001-01-01

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

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

High conversion burner type reactor

International Nuclear Information System (INIS)

Higuchi, Shin-ichi; Kawashima, Masatoshi

1987-01-01

Purpose: To simply and easily dismantle and reassemble densified fuel assemblies taken out of a high conversion ratio area thereby improve the neutron and fuel economy. Constitution: The burner portion for the purpose of fuel combustion is divided into a first burner region in adjacent with the high conversion ratio area at the center of the reactor core, and a second burner region formed to the outer circumference thereof and two types of fuels are charged therein. Densified fuel assemblies charged in the high conversion ratio area are separatably formed as fuel assemblies for use in the two types of burners. In this way, dense fuel assembly is separated into two types of fuel assemblies for use in burner of different number and arranging density of fuel elements which can be directly charged to the burner portion and facilitate the dismantling and reassembling of the fuel assemblies. Further, since the two types of fuel assemblies are charged in the burner portion, utilization factor for the neutron fuels can be improved. (Kamimura, M.)

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.)

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

CERN Document Server

Ogawa, Y; Sagawa, H; Tsujimoto, T

2002-01-01

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

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

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

International Nuclear Information System (INIS)

Koch, R.; Findeisen, A.

1986-04-01

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

Evaluation of neutron streaming through injection ports in a tokamak-type fusion reactor

International Nuclear Information System (INIS)

Ide, Takahiro; Seki, Yasushi; Iida, Hiromasa

1976-03-01

The effects of neutron streaming through injection ports in the fusion reactor designed in JAERI have been studied, especially those on tritium breeding ratio and the shielding of the superconducting magnet. In placement of the injection ports in the blanket, the tritium breeding ratio decreases by up to 1.3%, and shielding problem of the superconducting magnet is very important. (auth.)

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

International Nuclear Information System (INIS)

Gold, R.

1975-08-01

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

Conversion of reactor neutrons by lithium deuteride; Konverzija reaktorskih neutrona pomocu litijumdeuterida

Energy Technology Data Exchange (ETDEWEB)

Strugar, P; Altiparmakov, D [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1979-07-01

Nuclear reactors are powerful neutron sources. But for many purposes, neutrons of higher than fission energy are needed. Such neutrons may be produced by the conversion of reactor neutrons into 14 MeV neutrons using lithium deuteride converter. For that converter is generally employed {sup 6}LiD, the usual material for the thermonuclear weapons, and therefore hardly accessible in needed quantity. That was the reason to analyse such converter made of LiD with the lithium of natural isotopic content. This analysis starts with the basic conversion relations, takes into account neutron absorption and tritium generation, and finally, estimates the 14 MeV neutron flux and the heat generated in proposed converter, when the converter was coupled to each of three Yugoslav nuclear reactors. Results show that the converter made of LiD with the natural lithium is 50% less efficient than the converter of {sup 6}LiD. Intensity of 14 MeV neutrons is within limits 5. 10{sup 5} - 10{sup 10} (n/cm{sup 2}.s) for the converter used either as external converter with reactor RB, within the thermal column in the reactor TRIGA or as a 'fuel' segment at the reactor RA. (author)

Impact of neutron resonance treatments on reactor calculation

International Nuclear Information System (INIS)

Leszczynski, F.

1988-01-01

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

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

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

Solving the uncommon reactor core neutronics problems

International Nuclear Information System (INIS)

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

1980-01-01

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

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.

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)

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

Energy Technology Data Exchange (ETDEWEB)

Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)

2015-09-30

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

Analysis of fuel management pattern of research reactor core of the MTR type design

International Nuclear Information System (INIS)

Lily Suparlina; Tukiran Surbakti

2014-01-01

Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x10 15 n/cm 2 s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 10 15 n/cm 2 s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

Fueling method in LMFBR type reactors

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Inoue, Kotaro.

1985-01-01

Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

The development of ex-core neutron flux monitoring system for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S

2004-12-01

Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying

The development of ex-core neutron flux monitoring system for integral reactor

International Nuclear Information System (INIS)

Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S.

2004-12-01

Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying the ex

A neutron radiography facility on the IRT-2000 reactor

International Nuclear Information System (INIS)

Khadduri, I.Y.

1976-01-01

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

An ultracold neutron source at the NC State University PULSTAR reactor

Science.gov (United States)

Korobkina, E.; Wehring, B. W.; Hawari, A. I.; Young, A. R.; Huffman, P. R.; Golub, R.; Xu, Y.; Palmquist, G.

2007-08-01

Research and development is being completed for an ultracold neutron (UCN) source to be installed at the PULSTAR reactor on the campus of North Carolina State University (NCSU). The objective is to establish a university-based UCN facility with sufficient UCN intensity to allow world-class fundamental and applied research with UCN. To maximize the UCN yield, a solid ortho-D 2 converter will be implemented coupled to two moderators, D 2O at room temperature, to thermalize reactor neutrons, and solid CH 4, to moderate the thermal neutrons to cold-neutron energies. The source assembly will be located in a tank of D 2O in the space previously occupied by the thermal column of the PULSTAR reactor. Neutrons leaving a bare face of the reactor core enter the D 2O tank through a 45×45 cm cross-sectional area void between the reactor core and the D 2O tank. Liquid He will cool the disk-shaped UCN converter to below 5 K. Independently, He gas will cool the cup-shaped CH 4 cold-neutron moderator to an optimum temperature between 20 and 40 K. The UCN will be transported from the converter to experiments by a guide with an inside diameter of 16 cm. Research areas being considered for the PULSTAR UCN source include time-reversal violation in neutron beta decay, neutron lifetime determination, support measurements for a neutron electric-dipole-moment search, and nanoscience applications.

Materials research with neutron beams from a research reactor

Energy Technology Data Exchange (ETDEWEB)

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

2015-03-15

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Design study of a medical reactor for Boron Neutron Capture Therapy

International Nuclear Information System (INIS)

Sasaki, M.; Hirota, J.; Tamao, S.; Kanda, K.; Mishima, Y.

1992-01-01

A new design study of a medical reactor for Boron Neutron Capture Therapy (BNCT) has been carried out. The reactor is to be used exclusively for the treatment of malignant melanoma and other cancers as well as for the further biomedical research. Main specifications of the reactor are as follows; thermal power of 2 MW, water cooling by natural convection, semitight core of triangular lattice, UO 2 fuel rod of 9.5 mm diameter and no refueling in the reactor-life. Three horizontal and one vertical neutron beam hole are to be provided to deliver thermal and epithermal neutrons. N-γ coupling Sn transport calculations indicate that the patient treatment period will be about 30 minutes with minimal fast neutron and gamma contaminants. (author)

Burnup influence on the VVER-1000 reactor vessel neutron fluence evaluation

International Nuclear Information System (INIS)

Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

2009-01-01

The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of the Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in 1/4 depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (authors)

Burnup influence on the WWER-1000 reactor vessel neutron fluence evaluation

International Nuclear Information System (INIS)

Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

2009-01-01

The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of WWER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in ? depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (Authors)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Determination of neutron energy spectrum at a pneumatic rabbit station of a typical swimming pool type material test research reactor

International Nuclear Information System (INIS)

Malkawi, S.R.; Ahmad, N.

2002-01-01

The method of multiple foil activation was used to measure the neutron energy spectrum, experimentally, at a rabbit station of Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type material test research reactor. The computer codes MSITER and SANDBP were used to adjust the spectrum. The pre-information required by the adjustment codes was obtained by modelling the core and its surroundings in three-dimensions by using the one dimensional transport theory code WIMS-D/4 and the multidimensional finite difference diffusion theory code CITATION. The input spectrum covariance information required by MSITER code was also calculated from the CITATION output. A comparison between calculated and adjusted spectra shows a good agreement

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

International Nuclear Information System (INIS)

Taryo, Taswanda

2002-01-01

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

The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

Science.gov (United States)

Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

2016-02-01

The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

Directory of Open Access Journals (Sweden)

Wagemans Jan

2016-01-01

Full Text Available The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

Characteristics of Fabricated SiC Neutron Detectors for Neutron Flux Monitoring

Energy Technology Data Exchange (ETDEWEB)

Kim, Han Soo; Ha, Jang Ho; Park, Se Hwan; Lee, Kyu Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Ho [Hanyang University, Seoul (Korea, Republic of)

2011-05-15

An SPND (Self-powered Neutron Detector) is commonly used for neutron detection in NPP (Nuclear Power Plant) by virtue of un-reactivity for gamma-rays. But it has a drawback, which is that it cannot detect neutrons in real time due to beta emissions (about > 48 s) after reactions between neutrons and {sup 103}Rh in an SPND. And Generation IV reactors such as MSR (Molten-salt reactor), SFR (Sodium-cooled fast reactor), and GFR (Gas-cooled fast reactor) are designed to compact size and integration type. For GEN IV reactor, neutron monitor also must be compact-sized to apply such reactor easily and much more reliable. The wide band-gap semiconductors such as SiC, AlN, and diamond make them an attractive alternative in applications in harsh environments by virtue of the lower operating voltage, faster charge-collection times compared with gas-filled detectors, and compact size.1) In this study, two PIN-type SiC semiconductor neutron detectors, which are for fast neutron detection by elastic and inelastic scattering SiC atoms and for thermal neutron detection by charged particle emissions of 6LiF reaction, were designed and fabricated for NPP-related applications. Preliminary tests such as I-V and alpha response were performed and neutron responses at ENF in HANARO research reactor were also addressed. The application feasibility of the fabricated SiC neutron detector as an in-core neutron monitor was discussed

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Neutronics design of the next tokamak. (Swimming pool type)

International Nuclear Information System (INIS)

Seki, Y.; Iida, H.; Kitamura, K.; Minato, A.; Sako, K.; Mori, S.; Nishida, H.

1983-01-01

A swimming pool type tokamak reactor (SPTR) has been proposed in the Japan Atomic Energy Research Institute as a candidate for the next generation tokamak reactor after the JT-60. The concept of the SPTR evolved from an incentive to relieve the difficulties of repair and maintenance procedures of a tokamak reactor. After about two years of the reactor design studies, several advantages of the SPTR over the conventional tokamak reactors such as the ease of penetration shielding, reduction in solid radwaste have been shown. On the other hand, some drawbacks and uncertainties of the SPTR have also been pointed out but so far no serious defect negating the concept has been found. This paper describes the neutronics aspect of the SPTR based mostly on the result of one dimensional calculations. At first, the radiation shielding capability of water is compared with those of other candidate materials used in the blanket and shield of fusion reactors. Based on the result of the comparison and other requirements such as tritium breeding, thermal mechanical design, repair and maintenance procedures, the material arrangements of the blanket and shield are determined. The result of the blanket neutronics calculations, the radiation shielding calculations for the superconducting magnets, shutdown dose calculations are given together with major penetration shielding considerations. (author)

Conceptual designs of power tokamak-type thermonuclear reactors

International Nuclear Information System (INIS)

Shejndlin, A.E.; Nedospasov, A.V.

1978-01-01

Physico-technical and ecological aspects of conceptual designing power tokamak-type reactors have been briefly considered. Only ''pure'' (''non-hybride'') reactors are discussed. Presented are main plasma-physical parameters, characteristics of blankets and magnetic systems of the following projects: PPPL; V-2; V-3; Culham-2, JAERI; TBEh-2500; TFTR. Two systems of the first wall protection have been considered: divertor one and by means of a layer of a cool turbulent plasma. Examined are the following problems: fuel loading, choice of the first wall material, blanket structure, magnetic system, environmental contamination. The comparison of relative hazards of fast neutron reactors and fusion reactors has shown that in respect of fusion reactors the biological hazard potential value is less by one-two orders

Delayed Neutron Fraction (beta-effective) Calculation for VVER 440 Reactor

International Nuclear Information System (INIS)

Hascik, J.; Michalek, S.; Farkas, G.; Slugen, V.

2008-01-01

Effective delayed neutron fraction (β eff ) is the main parameter in reactor dynamics. In the paper, its possible determination methods are summarized and a β eff calculation for a VVER 440 power reactor as well as for training reactor VR1 using stochastic transport Monte Carlo method based code MCNP5 is made. The uncertainties in determination of basic delayed neutron parameters lead to the unwished conservatism in the reactor control system design and operation. Therefore, the exact determination of the β eff value is the main requirement in the field of reactor dynamics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of experiments and measurements what do you mean differences between different calculation approaches and experimental results. In consequence of difficulties in β eff experimental measurement, this value in exact state is determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. An accurate estimate of β eff is essential for converting reactivity, as measured in dollars, to an absolute reactivity and/or to an absolute k eff . In the past, k eff has been traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum-weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. The summary of the possible β eff determination methods can be found in this work and also a calculation of β eff first for the training reactor VR1 in one operation state and then for VVER 440 power reactor in two different operation states are made using the prompt method, by MCNP5 code.(author)

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

Science.gov (United States)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has

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

International Nuclear Information System (INIS)

Robinson, G.S.

1984-03-01

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

Neutron activation analysis of essential elements in Multani mitti clay using miniature neutron source reactor

International Nuclear Information System (INIS)

Waheed, S.; Rahman, S.; Faiz, Y.; Siddique, N.

2012-01-01

Multani mitti clay was studied for 19 essential and other elements. Four different radio-assay schemes were adopted for instrumental neutron activation analysis (INAA) using miniature neutron source reactor. The estimated weekly intakes of Cr and Fe are high for men, women, pregnant and lactating women and children while intake of Co is higher in adult categories and Mn by pregnant women. Comparison of MM clay with other type of clays shows that it is a good source of essential elements. - Highlights: ► Multani mitti clay has been studied for 19 essential elements for human adequacy and safety using INAA and AAS. ► Weekly intakes for different consumer categories have been calculated and compared with DRIs. ► Comparison of MM with other type of clays depict that MM clay is a good source of essential elements.

Trends and Developments for Fast Neutron Reactors and Related Fuel Cycles

International Nuclear Information System (INIS)

Carré, Frank

2013-01-01

• FR13 – A unique and dedicated framework to share updates on national programs of Fast Reactor developments, projects of new builds and plans for the future: - Near term projects of sodium and lead-alloy Fast Reactors; - Gen-IV visions of sodium-cooled and alternative types of Fast Neutron Reactors (GFR, LFR…). • FR13 – A special emphasis put on Fast Reactor Safety, Sustainability of nuclear fuel cycle and Young Generation perspective. • FR13 – A catalyst for further collaborations and alliances: - To share visions of goals and advisable options for future Fast Reactors and Nuclear Fuel Cycle; - To share cost of R&D and large demonstrations (safety, security, recycling); - To progress towards harmonized international standards; - To integrate national projects into a consistent international roadmap

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

Conceptual design of neutron diagnostic systems for fusion experimental reactor

International Nuclear Information System (INIS)

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

1994-01-01

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

Sensitivity analysis of reflector types and impurities in 10 MW MTR type nuclear research reactor

International Nuclear Information System (INIS)

Khattab, K.; Khamis, I.

2007-01-01

The 2-D and 3-D neutronics models for 10 MW nuclear research reactor of MTR type have been developed and presented in this paper. Our results agree very well with the results of seven countries mentioned in the IAEA-TECDOC-233. To study the effect of reflector types on the reactor effective multiplication factor, five types of reflectors such as pure beryllium, beryllium, heavy water, carbon and water are selected for this study. The pure beryllium is found to be the most efficient reflector in this group. The effect of the most important impurities, which exist on the beryllium reflector such as iron, silicon and aluminium on the reactor multiplication factor, have been analyzed as well. It is found that the iron impurity affects the reactor multiplication factor the most compared to silicon and aluminium impurities. (author)

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...

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

International Nuclear Information System (INIS)

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

1981-01-01

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

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

Researches at the University of Tokyo fast neutron sources reactor, YAYOI

International Nuclear Information System (INIS)

Koshizuka, S.; Oka, Y.; Saito, I.

1992-01-01

The Fast neutron source reactor YAYOI was critical in 1971 at the Nuclear Engineering Research Laboratory, the Faculty of Engineering, the University of Tokyo (UTNL). The core is fueled with the enriched uranium surrounded by the depleted uranium. YAYOI is the first fast reactor in Japan. Many types of studies have been carried out by the researchers of the University of Tokyo in these 20 years. It also contributed to the Japan's national project of developing fast breeder reactors. The reactor is opened to the visiting researchers from universities and research institutes. YAYOI has also been utilized for education of undergraduate and graduate students of the Department of Nuclear Engineering of the University of Tokyo. The present paper briefly summerizes past and present researchers. (author)

Fast neutron flux in the RA reactor experimental channels; Fluks brzih neutrona u eksperimentalnim kanalima reaktora RA

Energy Technology Data Exchange (ETDEWEB)

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

1961-12-15

Fast neutron flux in the RA reactor experimental channels was determined by using threshold reaction detectors. The (n,p) type reactions S{sub 32} (n,p)P{sub 32}, and Al{sub 24} (n,p)Na{sub 24}. Prepared sulphur and phosphorous foils were placed in cadmium boxes and irradiated in experimental channels VK-5, VK-7 and VK-9. Gold foils were irradiated simultaneously for controlling the reactor power. Reactor power was 100 kW during irradiation of half an hour. Activity of P{sub 32} and S{sub 31} after reactor shutdown was measured by 4{pi} counter and three calibrated GM counters. Absolute neutron flux was determined by using thus obtained data.

Neutron noise analysis for malfunction diagnosis at sodium cooled reactors

International Nuclear Information System (INIS)

Hoppe, P.

1978-09-01

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

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

Neutronics of a mixed-flow gas-core reactor

International Nuclear Information System (INIS)

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

1977-11-01

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

Transport calculation of neutron flux distribution in reflector of PW reactor

International Nuclear Information System (INIS)

Remec, I.

1982-01-01

Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

Utilization of boron irradiation filters in reactor neutron activation via epithermal (n,γ) and fast neutron reactions

International Nuclear Information System (INIS)

Chisela, F.

1986-01-01

The technique of instrumental neutron activation analysis based on irradiation with reactor epithermal and fast neutrons has been described and evaluated. Important characteristics of boron neutron absorbers used to remove thermal neutrons from the reactor neutron spectrum have been examined and compared with those of cadmium. Three boron compound shields, have been designed and constructed at the BER II 5MW reactor for use in epithermal neutron activation analysis of biological materials. The major advantages offered by these filters in this application include the flexibility of varying the filter thickness, the low radioactivity induced in the filters during irradiation, ease of fabrication and the relatively low cost of the filter materials. The radiation heating due to the 10 B(n,α) 7 Li-reaction has been experimentally investigated for the filters used and the results obtained confirm the necessity for efficient cooling of these filters during irradiation. Three irradiation facilities have been characterized with respect to the neutron flux density and the flux spatial distribution. An experiment has been designed and carried out to compensate the flux inhomogeneity in two irradiation positions of the DBV facility caused by the reactor geometry. Several biological samples including well characterized reference materials have been analysed after epithermal activation and the results compared with those obtained with the classical thermal neutron activation method. Improved sensitivity of determination has been found for elements with high resonance integral to thermal neutron cross section ratios (RI/σ 0 ). The range of elements that can be determined instrumentally is extended and the time scale of analysis is considerably reduced. (orig.) [de

Small-angle neutron scattering at pulsed sources compared to reactor sources

International Nuclear Information System (INIS)

Hjelm, R.P. Jr.; Seeger, P.A.; Thiyagarajan, P.

1990-01-01

Detailed comparisons of measurements made on small-angle neutron scattering instruments at pulsed spallation and reactor sources show that the results from the two types of instruments are comparable. It is further demonstrated that spallation instruments are preferable for measurements in the mid-momentum transfer domain or when a large domain is needed. 8 refs., 2 figs

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)

Determination of neutron flux densities in WWR-S reactor core

International Nuclear Information System (INIS)

Tomasek, F.

1989-04-01

The method is described of determining neutron flux densities and neutron fluences using activation detectors. The basic definitions and relations for determining reaction rates, fluence and neutron flux as well as the characteristics of some reactions and of sitable activation detectors are reported. The flux densities were determined of thermal and fast neutrons and of gamma quanta in the WWR-S reactor core. The data measured in the period 1984-1987 are tabulated. Cross sections for the individual reactions were determined from spectra measurements processed using program SAND-II and cross section library ENDF-B IV. Neutron flux densities were also measured for the WWR-S reactor vertical channels. (E.J.). 10 figs., 8 tabs., 111 refs

Neutron polarizing set-up of the Sofia IRT research reactor

International Nuclear Information System (INIS)

Krezhov, K.; Mikhajlova, V.; Okorokov, A.

1990-01-01

Neutron polarizing set-up of one of the horizontal beam tubes of the IRT-200 research reactor of the Bulgarian Institute of Nuclear Research and Nuclear Energy is presented. Neutron mirrors are extensively used in an effort to compensate the moderate reactor beam intensity by the high reflected intensity and wide-band transmittance of the mirror neutron guides. Time-to-flight technique using a slotted neutron absorbing chopper with a horizontal rotation axis has been applied to obtain the exit neutron spectra. Beam polarization and flipping ratios have been determined. Cadmium ratio in the polarized beam has been found almost 10 4 and the average polarization has been measured to be higher than 96%. 3 figs, 3 refs

Criticality analysis of thermal reactors for two energy groups applying Monte Carlo and neutron Albedo method

International Nuclear Information System (INIS)

Terra, Andre Miguel Barge Pontes Torres

2005-01-01

The Albedo method applied to criticality calculations to nuclear reactors is characterized by following the neutron currents, allowing to make detailed analyses of the physics phenomena about interactions of the neutrons with the core-reflector set, by the determination of the probabilities of reflection, absorption, and transmission. Then, allowing to make detailed appreciations of the variation of the effective neutron multiplication factor, keff. In the present work, motivated for excellent results presented in dissertations applied to thermal reactors and shieldings, was described the methodology to Albedo method for the analysis criticality of thermal reactors by using two energy groups admitting variable core coefficients to each re-entrant current. By using the Monte Carlo KENO IV code was analyzed relation between the total fraction of neutrons absorbed in the core reactor and the fraction of neutrons that never have stayed into the reflector but were absorbed into the core. As parameters of comparison and analysis of the results obtained by the Albedo method were used one dimensional deterministic code ANISN (ANIsotropic SN transport code) and Diffusion method. The keff results determined by the Albedo method, to the type of analyzed reactor, showed excellent agreement. Thus were obtained relative errors of keff values smaller than 0,78% between the Albedo method and code ANISN. In relation to the Diffusion method were obtained errors smaller than 0,35%, showing the effectiveness of the Albedo method applied to criticality analysis. The easiness of application, simplicity and clarity of the Albedo method constitute a valuable instrument to neutronic calculations applied to nonmultiplying and multiplying media. (author)

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)

Safety classification of systems, structures, and components for pool-type research reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Ryong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-08-15

Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

The advanced neutron source - A world-class research reactor facility

International Nuclear Information System (INIS)

Thompson, P.B.; Meek, W.E.

1993-01-01

The advanced neutron source (ANS) is a new facility being designed at the Oak Ridge National Laboratory that is based on a heavy-water-moderated reactor and extensive experiment and user-support facilities. The primary purpose of the ANS is to provide world-class facilities for neutron scattering research, isotope production, and materials irradiation in the United States. The neutrons provided by the reactor will be thermalized to produce sources of hot, thermal, cold, very cold, and ultracold neutrons usable at the experiment stations. Beams of cold neutrons will be directed into a large guide hall using neutron guide technology, greatly enhancing the number of research stations possible in the project. Fundamental and nuclear physics, materials analysis, and other research pro- grams will share the neutron beam facilities. Sufficient laboratory and office space will be provided to create an effective user-oriented environment

A neutron amplifier: prospects for reactor-based waste transmutation

International Nuclear Information System (INIS)

Blanovsky, A.

2004-01-01

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

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

The effect of temperature and the control rod position on the spatial neutron flux distribution in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

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

2007-01-01

The effect of water and fuel temperature increase and changes in the control rod positions on the spatial neutron flux distribution in the Syrian Miniature Neutron Source Reactor (MNSR) is discussed. The cross sections of all the reactor components at different temperatures are generated using the WIMSD4 code. These group constants are used then in the CITATION code to calculate the special neutron flux distribution using four energy groups. This work shows that water and fuel temperature increase in the reactor during the reactor daily operating time does not affect the spatial neutron flux distribution in the reactor. Changing the control rod position does not affect as well the spatial neutron flux distribution except in the region around the control rod position. This stability in the spatial neutron flux distribution, especially in the inner and outer irradiation sites, makes MNSR as a good tool for the neutron activation analysis (NAA) technique and production of radioisotopes with medium or short half lives during the reactor daily operating time. (author)

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

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

International Nuclear Information System (INIS)

Khalafi, H.

1999-01-01

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

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

Early detection of coolant boiling in research reactors with MTR-type fuel

International Nuclear Information System (INIS)

Kozma, R.; Turkcan, E.; Verhoef, J.P.

1992-10-01

A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-08-18

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

Applications of neutron activation analysis technique in the IPR-R1 research reactor

International Nuclear Information System (INIS)

Sabino, C.V.S.; Mansur, N.

1986-01-01

A review is made of the neutron activation analysis technique used in the IPR-R1 reactor of the Centro de Desenvolvimento da Tecnologia Nuclear - NUCLEBRAS. Some characteristics of the method are described, types of samples and elements analyzed are also mentioned. (Author) [pt

Code of practice for in-core instrumentation for neutron fluence rate (flux) measurements in power reactors

International Nuclear Information System (INIS)

Anon.

1982-01-01

This standard applies to in-core (on-line) neutron detectors and instrumentation which is designed for safety, information or control purposes. It also applies to components in so far as these components are contained within the primary envelope of the reactor. The detector types usually used are dc ionization chambers and self-powered neutron detectors

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

Neutronics characteristics of space power reactors

International Nuclear Information System (INIS)

Little, W.; Barner, J.

1986-01-01

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

Reprocessing of fast neutron reactor fuel

International Nuclear Information System (INIS)

Bourgeois, M.

1981-05-01

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

Doping of monocrystalline silicon with phosphorus by means of neutron irradiation at the IEA-R1 research reactor

International Nuclear Information System (INIS)

Carbonari, A.W.; Puget, M.A.C.

1990-11-01

The first neutron irradiation experiments with monocrystal silicon in the IEA-R1 research reactor of IPEN are related. The silicon is irradiated with phosphorus producing a N type semiconductor with a very small resistivity variation throughout the crystal volume. The neutrons induce nuclear reactions in Si-30 isotope and these atoms are then transformed in to phosphorous atoms. This process is known as Neutron Transmutation Doping. In order to irradiate the silicon crystals in the reactor, a specific device has been constructed, and it permits the irradiation of up to 2.5'' diameter monocrystals. (author)

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

International Nuclear Information System (INIS)

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

1987-10-01

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

Neutron Beam Filters

International Nuclear Information System (INIS)

Adib, M.

2011-01-01

The purpose of filters is to transmit neutrons with selected energy, while remove unwanted ones from the incident neutron beam. This reduces the background, and the number of spurious. The types of commonly used now-a-day neutron filters and their properties are discussed in the present work. There are three major types of neutron filters. The first type is filter of selective thermal neutron. It transmits the main reflected neutrons from a crystal monochromate, while reject the higher order contaminations accompanying the main one. Beams coming from the moderator always contain unwanted radiation like fast neutrons and gamma-rays which contribute to experimental background and to the biological hazard potential. Such filter type is called filter of whole thermal neutron spectrum. The third filter type is it transmits neutrons with energies in the resonance energy range (En . 1 KeV). The main idea of such neutron filter technique is the use of large quantities of a certain material which have the deep interference minima in its total neutron cross-section. By transmitting reactor neutrons through bulk layer of such material, one can obtain the quasimonochromatic neutron lines instead of white reactor spectrum.

Description of the CAREM Reactor Neutronic Calculation Codes

International Nuclear Information System (INIS)

Villarino, Eduardo; Hergenreder, Daniel

2000-01-01

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

Sensitivity analysis of reflector types and impurities in a 10 MW MTR type nuclear research reactor

International Nuclear Information System (INIS)

Khattab, K.; Khamis, I.

2008-01-01

The 2-D and 3-D neutronics models for 10 MW nuclear research reactor of MTR type have been developed and presented in this paper. Our results agree very well with the results of seven countries mentioned in the IAEA-TECDOC-233. To study the effect of reflector types on the reactor effective multiplication factor, five types of reflectors such as pure beryllium, beryllium, heavy water, carbon and water are selected for this study. The pure beryllium is found to be the most efficient reflector in this group. The effect of the most important impurities, which exist on the beryllium reflector such as iron, silicon and aluminium on the reactor multiplication factor, have been analyzed as well. It is found that the iron impurity affects the reactor multiplication factor the most compared to silicon and aluminium impurities. (author)

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

International Nuclear Information System (INIS)

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

1967-01-01

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

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)

A new method to determine in situ the transmission of a neutron-guide system at a reactor source

CERN Document Server

Haan, V O D; Gommers, R M; Labohm, F; Well, A A V; De Leege, P F A; Schebetov, A; Pusenkov, V

2002-01-01

In this paper, a description of a new method to determine the transmission of neutron guides after they are installed in a beam-tube at a reactor source is given. The method is based on activation measurements of gold foils at the entrance of the beam-tube and at the exit of the neutron guides compared to Monte-Carlo calculations. In this method, a quality factor is defined as the ratio between the actual transmission and the theoretical maximum attainable transmission. This method is used to determine the quality of an optimised neutron-guide system developed for beam-tube R2 of the HOR. The HOR is a pool-type nuclear research reactor at the Interfaculty Reactor Institute of the Delft University of Technology. It is shown that the quality factors of the newly installed neutron guides are between 0.49 and 0.63.

A new method to determine in situ the transmission of a neutron-guide system at a reactor source

International Nuclear Information System (INIS)

Haan, V.O. de; Gibcus, H.P.M.; Gommers, R.M.; Labohm, F.; Well, A.A. van; Leege, P.F.A. de; Schebetov, A.; Pusenkov, V.

2002-01-01

In this paper, a description of a new method to determine the transmission of neutron guides after they are installed in a beam-tube at a reactor source is given. The method is based on activation measurements of gold foils at the entrance of the beam-tube and at the exit of the neutron guides compared to Monte-Carlo calculations. In this method, a quality factor is defined as the ratio between the actual transmission and the theoretical maximum attainable transmission. This method is used to determine the quality of an optimised neutron-guide system developed for beam-tube R2 of the HOR. The HOR is a pool-type nuclear research reactor at the Interfaculty Reactor Institute of the Delft University of Technology. It is shown that the quality factors of the newly installed neutron guides are between 0.49 and 0.63

Neutron shielding studies on an advanced molten salt fast reactor design

International Nuclear Information System (INIS)

Merk, Bruno; Konheiser, Jörg

2014-01-01

Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

Updated pipe break analysis for Advanced Neutron Source Reactor conceptual design

International Nuclear Information System (INIS)

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

1994-01-01

The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at the Oak Ridge National Laboratory that will supply the highest continuous neutron flux levels of any reactor in the world. It uses plate-type fuel with high-mass-flux and highly subcooled heavy water as the primary coolant. The Conceptual Safety Analysis for the ANSR was completed in June 1992. The thermal-hydraulic pipe-break safety analysis (performed with a specialized version of RELAP5/MOD3) focused primarily on double-ended guillotine breaks of the primary piping and some core-damage mitigation options for such an event. Smaller, instantaneous pipe breaks in the cold- and hot-leg piping were also analyzed to a limited extent. Since the initial analysis for the conceptual design was completed, several important changes to the RELAP5 input model have been made reflecting improvements in the fuel grading and changes in the elevation of the primary coolant pumps. Also, a new philosophy for pipe-break safety analysis (similar to that adopted for the New Production Reactor) accentuates instantaneous, limited flow area pipe-break accidents in addition to finite-opening-time, double-ended guillotine breaks of the major coolant piping. This paper discloses the results of the most recent instantaneous pipe-break calculations

Modeling delayed neutron monitoring systems for fast breeder reactors

International Nuclear Information System (INIS)

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

1983-10-01

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

Errors of absolute methods of reactor neutron activation analysis caused by non-1/E epithermal neutron spectra

International Nuclear Information System (INIS)

Erdtmann, G.

1993-08-01

A sufficiently accurate characterization of the neutron flux and spectrum, i.e. the determination of the thermal flux, the flux ratio and the epithermal flux spectrum shape factor, α, is a prerequisite for all types of absolute and monostandard methods of reactor neutron activation analysis. A convenient method for these measurements is the bare triple monitor method. However, the results of this method, are very imprecise, because there are high error propagation factors form the counting errors of the monitor activities. Procedures are described to calculate the errors of the flux parameters, the α-dependent cross-section ratios, and of the analytical results from the errors of the activities of the monitor isotopes. They are included in FORTRAN programs which also allow a graphical representation of the results. A great number of examples were calculated for ten different irradiation facilities in four reactors and for 28 elements. Plots of the results are presented and discussed. (orig./HP) [de

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-05-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-06-22

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

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

Neutron collar calibration for assay of LWR [light-water reactor] fuel assemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Pieper, J.E.

1987-03-01

The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the 235 U content, and the 238 U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities

Optimization study of ultracold neutron sources at TRIGA reactors using MCNP

International Nuclear Information System (INIS)

Pokotilovskij, Yu.N.; Rogov, A.D.

1997-01-01

Monte Carlo simulation for the optimization of ultracold and very cold neutron sources for TRIGA reactors is performed. The calculations of thermal and cold neutron fluxes from the TRIGA reactor for different positions and configurations of a very cold solid methane moderator were performed with using the MCNP program. The production of neutrons in the ultracold and very cold energy range was calculated for the most promising final moderators (converters): very cold solid deuterium and heavy methane. The radiation energy deposition was calculated for the optimized solid methane-heavy methane cold neutron moderator

Research reactor of the future: The advanced neutron source

International Nuclear Information System (INIS)

Appleton, B.; West, C.

1994-01-01

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

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)

RA-0 reactor. New neutronic calculations

International Nuclear Information System (INIS)

Rumis, D.; Leszczynski, F.

1990-01-01

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

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

International Nuclear Information System (INIS)

Uenlue, K.; Rios-Martinez, C.

2005-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Theoretical and experimental study of collectrons for epithermal neutron flux in reactors

International Nuclear Information System (INIS)

Agu, M.N.

1986-01-01

A theoretical study of nuclear reactions and electric charge displacements arising in sensitivity to thermal and epithermal neutrons in collectrons allowed a computer code conception. Collectrons in Rhodium, Silver, Cobalt, Hafnium, Erbium, Gadolinium and Holmium have been tested in different radiation fields given by neutron or gamma filters irradiated in different places of Melusine and Siloe reactors. Some emitters were covered with different steel, nickel or zircaloy thicknesses. Theoretical and experimental results are consistent; that validate the computer code and show possibilities and necessity of covering collectron emitters to reduce or cancel the gamma sensitivity and to improve response instantaneity. A selective measurement of epithermal neutron flux can by this way, made by associating two types of collectrons [fr

BWR type reactor core

International Nuclear Information System (INIS)

Tatemichi, Shin-ichiro.

1981-01-01

Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

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

International Nuclear Information System (INIS)

Zegarra, Manuel; Lopez, Alcides

2013-01-01

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

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.

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

Neutron spectra in two beam ports of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

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

2013-10-01

The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6 Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

Neutron spectra in two beam ports of the TRIGA Mark III reactor

Energy Technology Data Exchange (ETDEWEB)

Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Aguilar, F.; Paredes, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Rivera M, T., E-mail: fermineutron@yahoo.com [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Unidad Legaria, Av. Legaria 694, 11500 Mexico D. F. (Mexico)

2013-10-15

The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

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)

DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

Science.gov (United States)

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

1958-02-11

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

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

International Nuclear Information System (INIS)

Zylbersztejn, F.

2013-01-01

The Sodium cooled Fast Reactor (SFR) is one of the six reactor types selected by the Generation-IV international forum (GIF), and the building of an industrial prototype is planned in France. The safety standard of the future SFR has to be equivalent to the EPR's. The general improvement of the safety of the new reactor goes through the examination of all the potentially harmful scenarios and both the study and monitoring of early signs. The mechanical deformations of the core can have harmful consequences in sodium fast reactors, such as unexpected power variations due to the reactivity increase in case of core compaction, or the excessive deterioration of the mechanical structures. The monitoring of such phenomena and of their potential early signs is then needed. The monitoring of such phenomena can be done with neutron detectors placed inside and outside the tank. This PhD thesis deals with the study of the neutron noise generated by the periodic deformation of the SFR core, restricted to the so-called core compaction or core flowering phenomenon, a deformation consisting in the variation of the inter-assembly sodium width by a radial bending the assemblies (the assemblies in SFR are held by the base). The PhD thesis has been performed within collaboration between CEA (France) and Chalmers Institute of Technology (Sweden). The work realized during the thesis led to the publication of 3 articles as first author and another as second author. This work has embraced the following topics: A state of the art of the monitoring of the core deformation phenomenon by interpretation of the noise measurements in SFR has been done. The PHENIX reactor multi physics measurements database has been scrutinized to provide an interpretation of the neutron noise bringing out mechanical vibration phenomena. An important conclusion was that the lack of theoretical knowledge about the neutron noise induced by the vibration phenomenon and the ill positioning of the neutron detectors

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

Physical model study of neutron noise induced by vibration of reactor internals

International Nuclear Information System (INIS)

Liu Jinhui; Gu Fangyu

1999-01-01

The author presents a physical model of neutron noise induced by reactor internals vibration in frequency domain. Based on system control theory, the reactor dynamic equations are coupled with random vibration equation, and non-linear terms are also taken into accounted while treating the random vibration. Experiments carried out on a zero-power reactor show that the model can be used to describe dynamic character of neutron noise induced by internals' vibration. The model establishes a method to help to determine internals'vibration features, and to diagnosis anomalies through neutron noise

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

Resonance self-shielding effect in uncertainty quantification of fission reactor neutronics parameters

International Nuclear Information System (INIS)

Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

2014-01-01

In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-15

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

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

International Nuclear Information System (INIS)

Galo Rocha, F. del.

1989-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-12-15

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

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.

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)

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.

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.

Application of neutron noise analysis to a swimming pool research reactor

International Nuclear Information System (INIS)

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

1982-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-05-06

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

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

International Nuclear Information System (INIS)

Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

1988-04-01

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

Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

International Nuclear Information System (INIS)

Zheng, Meiyin; Tian, Wenxi; Zhang, Dalin; Qiu, Suizheng; Su, Guanghui

2015-01-01

Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

Characterization of a fast to thermal neutron spectrum converter on PROSPERO reactor

Energy Technology Data Exchange (ETDEWEB)

Jacquet, X.; Authier, N.; Casoli, P.; Combacon, S. [CEA, Valduc Center, 21120 Is sur Tille (France); Calzavarra, Y. [ILL, Institut Laue Langevin, 38000 Grenoble (France)

2009-07-01

The PROSPERO reactor is located at CEA Valduc Center in France. The reactor is composed of an internal core made of High Enriched Uranium metal alloy surrounded by a reflector of depleted uranium. The reactor is used as a fast neutron spectrum source and is operated in delayed critical state with a continuous and steady power for several hours, which can vary from 3 mW to 3 kW, which is the nominal power. The flux at nominal power varies from 5.10{sup +10} n.cm{sup -2}/s at the reflector surface to 10{sup +7} n.cm{sup -2}/s at 5 meters from reactor axis. It has been decided to build a neutron energy converter allowing the production of a neutron thermal spectrum. As the core produces fast neutrons spectrum, we built a hollow cubic box of 50 cm x 50 cm x 50 cm with 10-cm-thick polyethylene bricks and placed one meter away from central reactor axis to moderate as much as possible neutrons to lower energies (E<0.6 eV). Analysis of the moderated flux inside the converter was performed using different activation foils such as indium or gold. We have developed a model of the experiment in the Monte Carlo neutron transport code TRIPOLI-4. A non-analogous transport calculation scheme was necessary to reproduce properly the experimental activities. The results of the calculated activations are within 4% of the experimental measurements given with 10% uncertainty (2 sigma). We show that the converter realizes thermalization of 80 % of the PROSPERO reactor fast neutrons below the cadmium threshold of 0.6 eV. Epithermal neutrons represent 15% of the spectrum and only 5% are in the fast neutron range above 1 MeV. The total flux at the center of the converter is 1.4 10{sup +9} n.cm{sup -2}/s at 3000 W

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

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

Design, construction and characterization of a new neutron beam for neutron radiography at the Tehran Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Choopan Dastjerdi, M.H., E-mail: mdastjerdi@aeoi.org.ir [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Department of Energy Engineering and Physics, Amirkabir University of Technology, Tehran (Iran, Islamic Republic of); Khalafi, H.; Kasesaz, Y.; Mirvakili, S.M.; Emami, J.; Ghods, H.; Ezzati, A. [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of)

2016-05-11

To obtain a thermal neutron beam for neutron radiography applications, a neutron collimator has been designed and implemented at the Tehran Research Reactor (TRR). TRR is a 5 MW open pool light water moderated reactor with seven beam tubes. The neutron collimator is implemented in the E beam tube of the TRR. The design of the neutron collimator was performed using MCNPX Monte Carlo code. In this work, polycrystalline bismuth and graphite have been used as a gamma filter and an illuminator, respectively. The L/D parameter of the facility was chosen in the range of 150–250. The thermal neutron flux at the image plane can be varied from 2.26×10{sup 6} to 6.5×10{sup 6} n cm{sup −2} s{sup −1}. Characterization of the beam was performed by ASTM standard IQI and foil activation technique to determine the quality of neutron beam. The results show that the obtained neutron beam has a good quality for neutron radiography applications.

The neutron utilization and promotion program of TRR-II research reactor project in Taiwan

International Nuclear Information System (INIS)

Gone, J.K.; Huang, Y.H.

2001-01-01

The objective of the Taiwan research reactor system improvement and utilization promotion project is to reconstruct the old Taiwan research reactor (TRR), which was operated by the Institute of Nuclear Energy Research (INER) between 1973 and 1988, into a multi-purpose medium flux research reactor (TRR-II). The project started in 1998, and the new reactor is scheduled to have its first critical in June of 2006. The estimated maximum unperturbed thermal neutron flux (E 14 n/cm 2 sec, and it is about one order of magnitude higher than other operating research reactors in Taiwan. The new reactor will equip with secondary neutron sources to provide neutrons with different energies, which will be an essential tool for advanced material researches in Taiwan. One of the major tasks of TRR-II project is to promote domestic utilization of neutrons generated at TRR-II. The traditional uses of neutrons in fuel/material research, trace element analysis, and isotope production has been carried out at INER for many years. On the other hand, it is obvious that promotions of neutron spectrometric technique will be a major challenge for the project team. The limited neutron flux from operating research reactors had discouraged domestic users in developing neutron spectrometric technique for many years, and only few researchers in Taiwan are experienced in using spectrometers. It is important for the project team to encourage domestic researchers to use neutron spectrometers provided by TRR-II as a tool for their future researches in various fields. This paper describes the current status of TRR-II neutron utilization and promotion program. The current status and future plans for important issues such as staff recruiting, personnel training, international collaboration, and promotion strategy will be described. (orig.)

Neutron and gamma characterization within the FFTF reactor cavity

International Nuclear Information System (INIS)

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

1980-08-01

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

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.

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.

Neutronic and thermal-hydraulic analysis of devices for irradiation of LEU targets type of UALx-Al and U-Ni to production of 99Mo in reactor IEA-R1 and RMB

International Nuclear Information System (INIS)

Domingos, Douglas Borges

2014-01-01

In this work neutronic and thermal-hydraulic analyses were made to compare three types of targets (UAl 2 -Al, U-Ni cylindrical and U-Ni plate) used for the production of 99 Mo by fission of 235 U. Some experiments were conducted to validate the neutronic and thermal-hydraulics methodologies used in this work. For the neutronic calculations the computational programs NJOY99.0, AMPX-II and HAMMERTECHNION were used to generate the cross sections. SCALE 6.0 and CITATION computational programs were used for three-dimensional calculations of the reactor cores, fuel burning and the production of 99 Mo. The computational programs MTRCR-IEAR1 and ANSYS CFX were used to calculate the thermal and hydraulic parameters of the irradiation devices and for comparing them to limits and design criteria. First were performed neutronic and thermal-hydraulic analyzes for the reactor IEA-R1 with the targets of UAl 2 -Al (10 mini plates). Analyses have shown that the total activity obtained for 99 Mo on the mini plates does not meet the demand of Brazilian hospitals (450 Ci/week) and that no limit of thermo-hydraulic design is overtaken. Next, the same calculations were performed for the three target types in Multipurpose Brazilian Reactor (MBR). The neutronic analyzes demonstrated that the three targets meet the demand of Brazilian hospitals. The thermal hydraulic analysis shows that a minimum speed of 7 m/s for the target UAl 2 -Al, 8 m/s for the cylindrical target U-Ni and 9 m/s for the target U-Ni plate will be necessary in the irradiation device to not exceed the design limits. Were performed experiments using a test bench for validate the methodologies for the thermal-hydraulic calculation. The experiments performed to validate the neutronic calculations were made in the reactor IPEN/MB-01. All experiments were simulated with the methodologies described above and the results compared. The simulations results showed good agreement with experimental results. (author)

Neutron detector

International Nuclear Information System (INIS)

Endo, Hiroshi.

1993-01-01

The device of the present invention detects neutrons in a reactor container under a high temperature and reduces the noise level in an FBR type reactor. That is, the detection section comprises a high heat resistant vessel containing a scintillator therein for detecting neutrons. Neutron signals sent from the detection section are inputted to a neutron measuring section by way of a signal transmission section. The detection section is disposed at the inside of the reactor container. Further, the signal transmission section is connected optically to the detection section. With such a constitution, since the detection section comprising the high temperature resistant vessel is disposed at the inside of the reactor container, neutron fluxes can be detected and measured at high sensitivity even under a high temperature circumstance. Since the signal transmission section is optically connected to the detection section, influence of radiation rays upon transmission of the neutron detection signals can be reduced. Accordingly, the noise level can be kept low. (I.S.)

RESONANCE SELF-SHIELDING EFFECT IN UNCERTAINTY QUANTIFICATION OF FISSION REACTOR NEUTRONICS PARAMETERS

Directory of Open Access Journals (Sweden)

GO CHIBA

2014-06-01

Full Text Available In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

Monte Carlo neutronics analysis of the ANS reactor three-element core design

International Nuclear Information System (INIS)

Wemple, C.A.

1995-01-01

The advanced neutron source (ANS) is a world-class research reactor and experimental center for neutron research, currently being designed at the Oak Ridge National Laboratory (ORNL). The reactor consists of a 330-MW(fission) highly enriched uranium core, which is cooled, moderated, and reflected with heavy water. It was designed to be the preeminent ultrahigh neutron flux reactor in the world, with facilities for research programs in biology, materials science, chemistry, fundamental and nuclear physics, and analytical chemistry. Irradiation facilities are provided for a variety of isotope production capabilities, as well as materials irradiation. This paper summarizes the neutronics efforts at the Idaho National Engineering Laboratory in support of the development and analysis of the three-element core for the advanced conceptual design phase

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

International Nuclear Information System (INIS)

Funahashi, Satoru; Morii, Yukio; Minakawa, Nobuaki

1992-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-12-15

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

Livermore pool-type reactor

International Nuclear Information System (INIS)

Mann, L.G.

1977-01-01

The Livermore Pool-Type Reactor (LPTR) has served a dual purpose since 1958--as an instrument for fundamental research and as a tool for measurement and calibration. Our early efforts centered on neutron-diffraction, fission, and capture gamma-ray studies. During the 1960's it was used for extensive calibration work associated with radiochemical and physical measurements on nuclear-explosive tests. Since 1970 the principal applications have been for trace-element measurements and radiation-damage studies. Today's research program is dominated by radiochemical studies of the shorter-lived fission products and by research on the mechanisms of radiation damage. Trace-element measurement for the National Uranium Resource Evaluation (NURE) program is the major measurement application today

Neutronic parameters calculations of a CANDU reactor

International Nuclear Information System (INIS)

Zamonsky, G.

1991-01-01

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

Determination of the effective delayed neutron fraction in the Coral-I Reactor; Determinacion de la fraccion efectiva de neutrones retardados en el Reactor Coral-1

Energy Technology Data Exchange (ETDEWEB)

Francisco, J L. de; Perez-Navarro, A; Rodriguez-Mayquez, E

1973-07-01

The effective delayed neutron fraction, {beta} eff, has been determined from the measurement of E / {beta}{sup 2}, by means of reactor noise analysis in the time domain, and the neutron detector efficiency, {epsilon}. For the {epsilon} measurement it is necessary to determine the fission rate in the reactor. This value can be obtained from the absolute measurement of the fission rate per cm{sup 3}, at a certain point of the reactor, and the determination of these two values ratio, which has been calculated by the Monte Cario method and also measured with results in good agreement. (Author)

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

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)

Progress in study of a medical reactor for boron neutron capture therapy

International Nuclear Information System (INIS)

Sasaki, Makoto; Hirota, Jitsuya; Tamao, Shigeo; Kanda, Keiji; Mishima, Yutaka.

1993-01-01

A design study of a medical reactor for Boron Neutron Capture Therapy has made progress. Main specifications of the reactor are as follows; thermal power of 2 MW, water cooling by natural convection, semitight core of hexagonal lattice, UO 2 fuel rod of 9.5 mm diameter and no refueling in the reactor-life. Three horizontal and one vertical neutron beam holes are to be provided for simultaneous treatments by thermal and epithermal neutrons and for further biomedical research. The design objectives for the beam holes are to deliver the therapeutic doses in a modest time (30 to 60 min) with minimal fast neutron and gamma contaminants. The n-γ coupling Sn transport calculations have been carried out using n-21 and γ-9 group cross sections on 2-dim. practical models. The calculated results indicate that the design objectives will be achievable even if the thermal power of the reactor is reduced to 1 MW. (author)

Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

International Nuclear Information System (INIS)

Delafosse, Jacques.

1976-01-01

French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

Measurement and analysis of the neutron noise of the pool research reactor at IPEN

International Nuclear Information System (INIS)

Simoes, Graciete Pedro

1979-01-01

Variations in the neutron density or power of a nuclear reactor (the neutron noise) operating at nominally constant power are generally random and can only be described in terms of statistical parameters. Random variations in the power of a power reactor are produced by one or more driving functions. In this work the neutron noise of the pool reactor IEAR-1 (2 MW nominal power) has been studied using two compensated ionization chambers ( Westinghouse VJL6377) and related to three possible-driving functions, namely vibration of the control bar and reactor support bridge and the temperature of the water entering the core. The CIC detectors were located in rigid tubes in turn positively located in the reactor lattice plate. Conventional accelerometers were used. Temperature measurements were made with a NiCr/Ni thermocouple (wire diam ∼ 0.2mm) located 10 mm above the top of a fuel element. Although the correlation between the measured neutron signals was high ( > 0,4) for frequencies in the range 0 to 10 Hz no resonances were identified in the neutron noise. A significant correlation (> 0,4) between the control bar acceleration and the neutron flux was obtained in the frequency range 0 to 10 Hz. The measured correlation between the neutron noise and both the bridge vibration and the reactor water inlet temperature was insignificant. (author)

Fine 3D neutronic characterization of a gas-cooled fast reactor based on plate-type sub-assemblies

International Nuclear Information System (INIS)

Bosq, J. C.; Peneliau, Y.; Rimpault, G.; Vanier, M.

2006-01-01

CEA neutronic studies have allowed the definition of a first 2400 MWth reference gas-cooled fast reactor core using plate-type sub-assemblies, for which the main neutronic characteristics were calculated by the so-called ERANOS 'design calculation scheme' relying on several method approximations. The last stage has consisted in a new refine characterization, using the reference calculation scheme, in order to confirm the impact of the approximations of the design route. A first core lay-out taking into account control rods was proposed and the reactivity penalty due to the control rod introduction in this hexagonal core lay-out was quantified. A new adjusted core was defined with an increase of the plutonium content. This leads to a significant decrease of the breeding gain which needs to be recovered in future design evolutions in order to achieve the self breeding goal. Finally, the safety criteria associated to the control rods were calculated with a first estimation of the uncertainties. All these criteria are respected, even if the safety analysis of GFR concepts and the determination of these uncertainties should be further studied and improved. (authors)

FBR type reactor

International Nuclear Information System (INIS)

Yamaoka, Mitsuaki

1988-01-01

Purpose: To enable to increase the burning period by enabling to decrease the reduction of burning reactivity and unifying the irradiation amount of fast neutrons. Constitution: A cylindrical reactor core made of fissile material-enriched fuel is constituted so as to form a plurality of layer-like enriched regions in which the enrichment degree of the fissile material is increased from the center to the radial and axial directions. Then, the ratio between the average enrichment degree for all of the enrichment regions other than the region at the reactor core center with the lowest enrichment degree and the enrichment degree of the enriched region formed at the center of the reactor core is made greater by 5 % or 20 % than the ratio at the initial burning stage where the power distribution of the reactor core is most flattened. (Kawakami, Y.)

Use of code DTF-4 for determining the coefficient of back-reflection of the neutron within the thermonuclear plasma of a thermonuclear reactor controlled by the rate of the fission reactions. Pt. 1

International Nuclear Information System (INIS)

Cristea, G.

1975-01-01

The neutron problems are discussed of the thermonuclear reactor controlled by the rate of the fission reactions. The results obtained by rolling the DTF-4 program in a spherical geometry in the case of an ''external source'' problem permit to draw conclusions concerning the problems of the neutronics system of this thermonuclear reactor type. A relation is deduced for estimating the coefficient of back-reflection of the neutrons within the thermonuclear plasma and the focussion system is discussed of the neutronics of this reactor type

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)

Reactor neutron activation for multielemental analysis

International Nuclear Information System (INIS)

Reddy, A.V.R.

1999-01-01

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

Numerical effects in the neutron flux calculations into WWER-type reactor vessels using the Monte Carlo

International Nuclear Information System (INIS)

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

2000-01-01

In the present paper, a PC version of the Monte Carlo 3-D HEXANN-EVALU system is used for the estimation of the WWER reactor pressure vessel irradiation. It was selected on the basis of its flexible options that, however, need to be quantified in connection with the desired magnitudes. The parameters that control the random walk of neutrons, as well as the efficiency increasing options included in the code, are studied in order to identify their impact on the final results for fluxes and fluence in the reactor pressure vessel. As a result, an optimal set of parameters is suggested. (authors)

Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

International Nuclear Information System (INIS)

Guo, Zhangpeng; Zhou, Jianjun; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

2013-01-01

Highlights: ► A multiple-channel analysis code (MAC) is developed to be coupled with MCNP. ► 1/8 of core is simulated in MCNP and thermal-hydraulic code. ► The coupling calculation can achieve stable state after a few iterations. ► The coupling calculation results are in reasonable agreement with the analytic solutions of the ORNL. ► Parametric studies of MSR are performed to provide valuable information for future design MSR. -- Abstract: The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail

Critical survey of the neutron-induced creep behaviour of steel alloys for the fusion reactor materials programme

International Nuclear Information System (INIS)

Hausen, H.

1985-01-01

The differences between the irradiation environment of a fission reactor and that of a fusion reactor are respectively described in relation to the radiation damage found and expected in the two types of nuclear reactor. It is shown that the microstructure developing for instance in stainless steel alloys is almost invariant to whether the production rate of helium is high or low. The finding is valid up to neutron doses corresponding to about 60 dpa. For this reason, irradiation creep data obtained in fission reactors may be used, with caution, for predicting creep behaviour in fusion reactors.It was further recognized that irradiation creep performed with high energy particles from an accelerator, yields results which are comparable to those obtained in fission reactors. For this reason, simulation creep experiments are found to be valuable for the development of irradiation creep resistant materials using, for example, high energy electrons or protons. Such kind of experiments are performed in many laboratories. For irradiation doses larger than 60 dpa, predictions with respect to creep rates in fission and fusion reactors are difficult. In end-of-life tests, which concern swelling, ductility, tensile properties, rupture, fatigue and embrittlement, the presence of helium, due to its production rate being much higher in most materials exposed to 14 MeV neutrons than to fission neutrons, may be of great importance

Optimization of the neutron calculation model for the RA-6 reactor

International Nuclear Information System (INIS)

Coscia, G.A.

1981-01-01

A model for the neutronic calculation of the RA-6 reactor which includes the codes ANISN and EQUIPOSE is analyzed. Starting with a brief description of the reactor, the core and its parts, the general scheme of calculation applied is presented. The fuel elements used were those which are utilized in the RA-3 reactor; this is of the MTR type with 90% enriched uranium. With the approximations used, an analysis of such model of calculation was made, trying to optimize it by reducing, if possible, the calculation time without loosing accuracy. In order to improve the calculation model, it is recomended a cross section data library specific for the enrichment of the fuel considered 90% and the incorporation of a more advanced code than EQUIPOISE which would be DIXYBAR. (M.E.L.) [es

Neutronic design of a traveling wave reactor core

International Nuclear Information System (INIS)

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

2010-10-01

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

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

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

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)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

The reactor and cold neutron research facility at NIST

Energy Technology Data Exchange (ETDEWEB)

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

1992-07-01

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

The reactor and cold neutron research facility at NIST

International Nuclear Information System (INIS)

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

1992-01-01

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

Design and construction of a preamplifier for research reactor control system using Russia’s neutron detectors

International Nuclear Information System (INIS)

Trinh Dinh Hai; Vo Van Tai; Le Van Diep; Nguyen Nhi Dien

2016-01-01

This paper presents the design and construction of a preamplifier device for Research Reactor Control System, using Russia’s Neutron Detectors of ionization and fission chambers. In this work, the preamplifier device which consists of a wide range Current to Frequency Converter block used with a compensation ionization chamber type KNK-3 to measure the thermal neutron flux in the range of 1x10"6 - 1x10"1"1 n/cm"2.s, a Pulse Preamplifier block used with a fission chamber type KNK-15 to measure the thermal neutron flux in the range of 1x10"0 - 1x10"6 n/cm"2.s, and a Power Supply block, was designed and tested in different conditions in the laboratory and at Dalat Nuclear Research Reactor (DNRR). Obtained results show that, the above blocks have almost design specifications as equivalent or better in comparison with the same function blocks of the DNRR Control System which were designed by the former Soviet Union. They also meet the utilization requirements as well as the experimental and training purposes. (author)

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

International Nuclear Information System (INIS)

Rimpler, A.; Kneschke, H.

1985-01-01

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

Long-term performance of the CANDU-type of vanadium self-powered neutron detectors in NRU

Energy Technology Data Exchange (ETDEWEB)

Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)]. E-mail: leungt@aecl.ca

2007-07-01

The CANDU-type of in-core vanadium self-powered neutron flux detectors have been installed in NRU to monitor the axial neutron flux distributions adjacent to the loop fuel test sites since 1996. This paper describes how the thermal neutron fluxes were measured at two monitoring sites, and presents a method of correcting the vanadium burn-up effect, which can be up to 2 to 3% per year, depending on the detector locations in the reactor. It also presents the results of measurements from neutron flux detectors that have operated for over eight-years in NRU. There is good agreement between the measured and simulated neutron fluxes, to within {+-} 6.5%, and the long-term performance of the CANDU-type of vanadium neutron flux detectors in NRU is satisfactory. (author)

Utilization of fast reactor excess neutrons for burning long-lived fission products

International Nuclear Information System (INIS)

Kawashima, K.; Kobayashi, K.; Kaneto, K.

1995-01-01

An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc-99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction. The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-06-22

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

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)

Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

Energy Technology Data Exchange (ETDEWEB)

Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

2006-01-01

The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

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

International Nuclear Information System (INIS)

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

1997-01-01

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