CFRMF neutron flux gradient and spectral determinations

International Nuclear Information System (INIS)

Rogers, J.W.; Turk, E.H.; Hogg, C.H.

1976-01-01

Recently more accurate and complete measurements of the flux gradient have been measured by the activation of 235 U and Au samples. Neutron spectrum characteristics were studied by making activation measurements with and without the ends of the CFRMF test region plugged with 10 B. These measurements define the flux gradient to +-1 to 2% and indicate there is no detectable streaming of thermal or resonance neutrons from the ends in the central 30 cm of the CFRMF test region. Measurements of the Cd ratio of Au foil activations were conducted and these results also indicate there is no streaming of thermal and resonance neutrons into the CFRMF test region

Compilation of neutron flux density spectra and reaction rates in different neutron fields

International Nuclear Information System (INIS)

Ertek, C.

1979-07-01

Upon the recommendation of International Working Group of Reactor Radiation Measurements (IWGRRM), the compilation of neutron flux density spectra and the reaction rates obtained by activation and fission foils in different neutron fields is presented. The neutron fields considered are as follows: 1/E; iron block; LWR core and pressure vessel; LMFBR core and blanket; CTR first wall and blanket; fission spectrum

First flux measurement in a SINQ supermirror neutron guide

Energy Technology Data Exchange (ETDEWEB)

Janssen, S.; Schlumpf, N.; Bauer, G. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1997-09-01

On Dec. 3, 1996, the Swiss spallation neutron source SINQ was taken into operation and produced its first neutrons successfully. The neutron spectrum within one of the supermirror guides was estimated by a chopper Time-of-Flight method. The result shows a 30% higher neutron intensity at the flux maximum than expected from previous Monte-Carlo simulations. (author) 1 fig., 4 refs.

Neutron flux monitor

International Nuclear Information System (INIS)

Oda, Naotaka.

1993-01-01

The device of the present invention greatly saves an analog processing section such as an analog filter and an analog processing circuit. That is, the device of the present invention comprises (1) a neutron flux detection means for detecting neutron fluxed in the reactor, (2) a digital filter means for dividing signals corresponding to the detected neutron fluxes into predetermined frequency band regions, (3) a calculation processing means for applying a calculation processing corresponding to the frequency band regions to the neutron flux detection signals divided by the digital filter means. With such a constitution, since the neutron detection signals are processed by the digital filter means, the accuracy is improved and the change for the property of the filter is facilitated. Further, when a neutron flux level is obtained, a calculation processing corresponding to the frequency band region can be conducted without the analog processing circuit. Accordingly, maintenance and accuracy are improved by greatly decreasing the number of parts. Further, since problems inherent to the analog circuit are solved, neutron fluxes are monitored at high reliability. (I.S.)

Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

International Nuclear Information System (INIS)

Angelone, M.; Klix, A.; Pillon, M.; Batistoni, P.; Fischer, U.; Santagata, A.

2014-01-01

Highlights: •Self powered neutron detector (SPND) is attractive neutron monitor for TBM in ITER. •In hard neutron spectra (e.g. TBM) there is the need to optimize their response. •Three state-of-the-art SPNDs were tested using fast and 14 MeV neutrons. •The response of SPNDs is much lower than in thermal neutron flux. •FISPACT calculations performed to find out candidate materials in hard spectra. -- Abstract: Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum. This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG). The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra

Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

Energy Technology Data Exchange (ETDEWEB)

Angelone, M., E-mail: maurizio.angelone@enea.it [Associazione ENEA-EURATOM sulla FusioneENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); Klix, A. [Association KIT-EURATOM, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pillon, M.; Batistoni, P. [Associazione ENEA-EURATOM sulla FusioneENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); Fischer, U. [Association KIT-EURATOM, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Santagata, A. [ENEA C.R. Casaccia, via Anguillarese Km. 1,300, 00100 Roma (Italy)

2014-10-15

Highlights: •Self powered neutron detector (SPND) is attractive neutron monitor for TBM in ITER. •In hard neutron spectra (e.g. TBM) there is the need to optimize their response. •Three state-of-the-art SPNDs were tested using fast and 14 MeV neutrons. •The response of SPNDs is much lower than in thermal neutron flux. •FISPACT calculations performed to find out candidate materials in hard spectra. -- Abstract: Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum. This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG). The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra.

Approximation for the adjoint neutron spectrum; Aproximacao para o espectro adjunto de neutrons

Energy Technology Data Exchange (ETDEWEB)

Suster, Luis Carlos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

2002-07-01

The proposal of this work is the determination of an analytical approximation which is capable to reproduce the adjoint neutron flux for the energy range of the narrow resonances (NR). In a previous work we developed a method for the calculation of the adjoint spectrum which was calculated from the adjoint neutron balance equations, that were obtained by the collision probabilities method, this method involved a considerable quantity of numerical calculation. In the analytical method some approximations were done, like the multiplication of the escape probability in the fuel by the adjoint flux in the moderator, and after these approximations, taking into account the case of the narrow resonances, were substituted in the adjoint neutron balance equation for the fuel, resulting in an analytical approximation for the adjoint flux. The results obtained in this work were compared to the results generated with the reference method, which demonstrated a good and precise results for the adjoint neutron flux for the narrow resonances. (author)

Determination of neutron flux with an arbitrary energy distribution by measurement of irradiated foils activity

International Nuclear Information System (INIS)

Ljubenov, V.; Milosevic, M.

2003-01-01

A procedure for the neutron flux determination in a neutron field with an arbitrary energy spectrum, based on the using of standard methods for the measurement of irradiated foils activity and on the application of the SCALE-4.4a code system for averaged cross section calculation is described in this paper. Proposed procedure allows to include the energy spectrum of neutron flux reestablished in the location of irradiated foils and the resonance self-shielding effects in the foils also. Example application of this procedure is given for the neutron flux determination inside the neutron filter with boron placed in the centre of heavy water critical assembly RB at the Vinca Institute (author)

OBJECT KINETIC MONTE CARLO SIMULATIONS OF RADIATION DAMAGE IN TUNGSTEN SUBJECTED TO NEUTRON FLUX WITH PKA SPECTRUM CORRESPONDING TO THE HFIR

Energy Technology Data Exchange (ETDEWEB)

Nandipati, Giridhar; Setyawan, Wahyu; Heinisch, Howard L.; Roche, Kenneth J.; Kurtz, Richard J.; Wirth, Brian D.

2015-12-31

The objective of this work is to study the damage accumulation in pure tungsten (W) subjected to neutron bombardment with a primary knock-on atom (PKA) spectrum corresponding to the High Flux Isotope Reactor (HFIR), using the object kinetic Monte Carlo (OKMC) method.

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)

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

Energy Technology Data Exchange (ETDEWEB)

Fukuda, Makoto, E-mail: makoto.fukuda@qse.tohoku.ac.jp [Tohoku University, Sendai, 980-8579 (Japan); Kiran Kumar, N.A.P.; Koyanagi, Takaaki; Garrison, Lauren M. [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA, 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Hasegawa, Akira [Tohoku University, Sendai, 980-8579 (Japan)

2016-10-15

Neutron irradiation to single crystal pure tungsten was performed in the mixed spectrum High Flux Isotope Reactor (HFIR). To investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ∼90–∼800 °C and fast neutron fluences were 0.02–9.00 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. The hardness and microstructure changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten. - Highlights: • The microstructure and irradiation hardening of single crystal pure W irradiated in HFIR was investigated. • The neutron energy spectrum influence was evaluated by comparing the HFIR results with previous work in Joyo and JMTR. • In the dose range up to ∼1 dpa, the neutron energy spectrum influence of irradiation hardening was not clear. • In the dose range above 1 dpa, the neutron energy influence on irradiation hardening and microstructural development was clearly observed. • The irradiation induced precipitates caused significant irradiation hardening of pure W irradiated in HFIR.

A neutron spectrum unfolding code based on iterative procedures

International Nuclear Information System (INIS)

Ortiz R, J. M.; Vega C, H. R.

2012-10-01

In this work, the version 3.0 of the neutron spectrum unfolding code called Neutron Spectrometry and Dosimetry from Universidad Autonoma de Zacatecas (NSDUAZ), is presented. This code was designed in a graphical interface under the LabVIEW programming environment and it is based on the iterative SPUNIT iterative algorithm, using as entrance data, only the rate counts obtained with 7 Bonner spheres based on a 6 Lil(Eu) neutron detector. The main features of the code are: it is intuitive and friendly to the user; it has a programming routine which automatically selects the initial guess spectrum by using a set of neutron spectra compiled by the International Atomic Energy Agency. Besides the neutron spectrum, this code calculates the total flux, the mean energy, H(10), h(10), 15 dosimetric quantities for radiation protection porpoises and 7 survey meter responses, in four energy grids, based on the International Atomic Energy Agency compilation. This code generates a full report in html format with all relevant information. In this work, the neutron spectrum of a 241 AmBe neutron source on air, located at 150 cm from detector, is unfolded. (Author)

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)

DIFFERENTIAL SPECTRUM OF NEUTRONS AT CHACALTAYA-BOLIVIA

International Nuclear Information System (INIS)

Mayta, R.; Zanini, A.; Ticona, R.; Velarde, A.

2009-01-01

We describe the Neutron Spectrometer Experiment installed at Chacaltaya Cosmic Rays Observatory (68 deg. O, 16.2 deg. S), located in Bolivia, at 5230 m.a.s.l. This experimental system is constituted by passive detectors which register the flux of neutrons, in an energy range of 10 KeV-20 MeV. Using the unfolding code BUNTO a peak around 1 MeV of the characteristic spectrum of neutrons was obtained. Experimental values, observed during April of 2008, are compared with similar ones carried out in 1997 at the same place, in order to look for eventual changes due to local atmosphere. A similar experiment was also carried up at the Laboratory of Testa Grigia-Italy (45.56 deg. N, 7.42 deg. E,. 3480 m.a.l.s). Data of both stations allow us to compare the spectra in order to explain the difference of neutron flux of these two stations.

Measurement of the fast neutron flux in the MNSR inner irradiation site

International Nuclear Information System (INIS)

Khattab, K.

2007-01-01

The WIMSD4 code was used to calculate the fast neutron flux spectrum and the fast neutron fission cross sections for 238 U, using six energy groups ranging from 0.5 to 10 MeV. These results, with the measured radioactivities of the 140 Ba, 131 I, 103 Ru, 95 Zr and 97 Zr fission products emerging from the fission of the 238 U foil covered with a cadmium filter, were used to measure the fast neutron flux in the Syrian MNSR inner irradiation site. (author)

Experimental Validation of Ex-Vessel Neutron Spectrum by Means of Dosimeter Materials Activation Method

Directory of Open Access Journals (Sweden)

S.A. Santa

2017-06-01

Full Text Available Neutron spectrum information in reactor core and around of ex-vessel reactor needs to be known with a certain degree of accuracy to support the development of fuels, materials, and other components. The most common method to determine neutron spectra is by utilizing the radioactivation of dosimeter materials. This report presents the evaluation of neutron flux incident on M3dosimeter sets which were irradiated outside the reactor vessel,as well as the validation of  neutron spectrum calculation. Al capsules containing both dosimeter set covered withCd and dosimeter set without Cd cover have been irradiated during the 35th operational cycle in the M3 ex-vessel irradiation hole position207 cmfrom core centerline at the space between the reactor vessel and the safety vessel. The capsules were positioned at Z=0.0 cm of core midplane. Each dosimeter set consists of Co-Al, Sc, Fe, Np, Nb, Ni, B, and Ta. The gamma-ray spectra of irradiated dosimeter materials were measured by 63 cc HPGe solid-state detector and photo-peak spectra were analyzed using BOB75 code. The reaction rates of each dosimeter materials and its uncertainty were analyzed based on 59Co (n,g 60Co, 237Np (n,f 95Zr-103Ru,  45Sc (n,g 46Sc, 58Fe (n,g 59Fe, 181Ta (n,g 182Ta, and 58Ni (n,p58Co reactions. The measured Cd ratios indicate that neutron spectrum at the irradiated dosimeter sets was dominated by low energy neutron. The experimental result shows that the calculated neutron spectra by DORT code at the ex-vessel positions need correction, especially in the fast neutron energy region, so as to obtain reasonable unfolding result consistent with the reaction rate measurement without any exception. Using biased DORT initial spectrum, the neutron spectrum and its integral quantity were unfolded by NEUPAC code. The result shows that total neutron flux, flux above 1.0 MeV, flux above 0.1 MeV, and the displacement rate of the dosimeter set not covered with Cd were 1.75× 1012 n cm2 s-1, 1

Neutron spectrum and dose-equivalent in shuttle flights during solar maximum

Energy Technology Data Exchange (ETDEWEB)

Keith, J E; Badhwar, G D; Lindstrom, D J [National Aeronautics and Space Administration, Houston, TX (United States). Lyndon B. Johnson Space Center

1992-01-01

This paper presents unambiguous measurements of the spectrum of neutrons found in spacecraft during spaceflight. The neutron spectrum was measured from thermal energies to about 10 MeV using a completely passive system of metal foils as neutron detectors. These foils were exposed to the neutron flux bare, covered by thermal neutron absorbers (Gd) and inside moderators (Bonner spheres). This set of detectors was flown on three U.S. Space Shuttle flights, STS-28, STS-36 and STS-31, during the solar maximum. We show that the measurements of the radioactivity of these foils lead to a differential neutron energy spectrum in all three flights that can be represented by a power law, J(E){approx equal}E{sup -0.765} neutrons cm{sup -2} day {sup -1} MeV{sup -1}. We also show that the measurements are even better represented by a linear combination of the terrestrial neutron albedo and a spectrum of neutrons locally produced in a aluminium by protons, computed by a previous author. We use both approximations to the neutron spectrum to produce a worst case and most probable case for the neutron spectra and the resulting dose-equivalents, computed using ICRP-51 neutron fluence-dose conversion tables. We compare these to the skin dose-equivalents due to charged particles during the same flights. (author).

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 stabilization in the NG-150 neutron generators

International Nuclear Information System (INIS)

Kuz'min, L.E.; Makarov, S.A.; Pronman, I.M.

1986-01-01

Problem of metal tritium target lifetime increase and neutron flux stabilization in the NG-150 neutron generators is studied. Possibility on neutron flux stabilization using the mass analyzer for low-angle (4 deg and 41 deg) mass separation of a beam in thre components, which fall on a target simultaneously, is confirmed experimentally. Basic generator parameters are: accelerating voltage of 150 kV, total beam current on a target of 1.5 mA, beam current density of 0.3-1.6 mA/cm 2 , beam diameter of 8 mm. The initial neutron flux on the targets of 0.73 mg/cm 2 thick constituted 1.1x10 11 ssup(-1). The neutron flux monitoring was accomplished from recoil proton recording by a plastic scintillator. Flux decrease by more than 5% served as a signel for measuring mass analyzer magnetic field providing beam displacement on a target and restoration of the given flux. The NG-150 generator neutron flux stabilization was attained during 2h

Experimental measurement of neutron spectrum in the reflector of a light water reactor

International Nuclear Information System (INIS)

Brethe, P.

1963-09-01

1. Thermal neutrons: The temperature of the thermal neutron spectrum was calculated using Au-Lu foils. This temperature varies from 300 deg. K (temperature of the moderator) at 30 cm of the core to 350 deg. K in a hole of the core. 2. Slowing down of neutron: Four resonance detectors have been used (Au, In, Co, Mn). We can write a 1/E form of the spectrum. The linking up energy E M between thermal neutron spectrum and slowing down spectrum is about 0.23 eV and is free from the Maxwell spectrum temperature. The decrease of slowing down flux regarding thermal flux, farther from the core, has been showed. 3. Fast neutrons: We used 3 threshold detectors (Ni, Al, Mg). We supposed a E 1/2 e -βE from of the spectrum above 3 MeV. The values of β are in a range from 0.775, at the centre of the core and in a loop-hole, to 0,64 at about 30 cm of the core. 4. Continuous shape of the spectrum: The following interpolations give useful informations between the field where measurements have been made: between 340 eV and 10 keV: 1/E form between 10 keV and 330 keV: 1/(E σ S (E)) form (σ S (E) elastic scattering section on hydrogen) between 330 keV and 3 MeV: calculated form by the moments method (ref. BSR). (author) [fr

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

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.)

Neutron flux characterization of californium-252 Neutron Research Facility at the University of Texas - Pan American by nuclear analytical technique

Science.gov (United States)

Wahid, Kareem; Sanchez, Patrick; Hannan, Mohammad

2014-03-01

In the field of nuclear science, neutron flux is an intrinsic property of nuclear reaction facilities that is the basis for experimental irradiation calculations and analysis. In the Rio Grande Valley (Texas), the UTPA Neutron Research Facility (NRF) is currently the only neutron facility available for experimental research purposes. The facility is comprised of a 20-microgram californium-252 neutron source surrounded by a shielding cascade containing different irradiation cavities. Thermal and fast neutron flux values for the UTPA NRF have yet to be fully investigated and may be of particular interest to biomedical studies in low neutron dose applications. Though a variety of techniques exist for the characterization of neutron flux, neutron activation analysis (NAA) of metal and nonmetal foils is a commonly utilized experimental method because of its detection sensitivity and availability. The aim of our current investigation is to employ foil activation in the determination of neutron flux values for the UTPA NSRF for further research purposes. Neutron spectrum unfolding of the acquired experimental data via specialized software and subsequent comparison for consistency with computational models lends confidence to the results.

Thermal neutron flux measurements using neutron-electron converters; Mesure de flux de neutrons thermiques avec des convertisseurs neutrons electrons

Energy Technology Data Exchange (ETDEWEB)

Le Meur, R; Lecomte, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

The operation of neutron-electron converters designed for measuring thermal neutron fluxes is examined. The principle is to produce short lived isotopes emitting beta particles, by activation, and to measure their activity not by extracting them from the reactor, but directly in the reactor using the emitted electrons to deflect the needle of a galvanometer placed outside the flux. After a theoretical study, the results of the measurements are presented; particular attention is paid to a new type of converter characterized by a layer structure. The converters are very useful for obtaining flux distributions with more than 10{sup 7} neutrons cm{sup -2}*sec{sup -1}. They work satisfactorily in pressurized carbon dioxide at 400 Celsius degrees. Some points still have to be cleared up however concerning interfering currents in the detectors and the behaviour of the dielectrics under irradiation. (authors) [French] On examine le fonctionnement de convertisseurs neutrons electrons destines a des mesures de flux de neutrons thermiques. Le principe est de former par activation des isotopes a periodes courtes et a emission beta et de mesurer leur activite non pas en les sortant du reacteur, mais directement en pile, utilisant les electrons emis pour faire devier l'aiguille d'un galvanometre place hors flux. Apres une etude theorique, on indique des resultats de mesures obtenus, en insistant particulierement sur un nouveau type de convertisseur, caracterise par sa structure stratifiee. Les convertisseurs sont tres interessants pour tracer, des cartes de flux a partir de 10{sup 7} neutrons cm{sup -2}*s{sup -1}. Ils sont utilisables pour des flux de 10{sup 14} neutrons cm{sup -2}*s{sup -1}. Ils fonctionnent correctement dans du gaz carbonique sous pression a 400 C. Des points restent cependant a eclaircir concernant les courants parasites dans les detecteurs et le comportement des dielectriques pendant leur irradiation. (auteur)

Measurement and evaluation of fast neutron flux of CT and OR5 irradiation hole in HANARO

International Nuclear Information System (INIS)

Yang, Seong Woo; Choo, Kee Nam; Lee, Seung-Kyu; Kim, Yong Kyun

2012-01-01

The irradiation test has been conducted to evaluate the irradiation performance of many materials by a material capsule at HANARO. Since the fast neutron fluence above 1 MeV is important for the irradiation test of material, it must be measured and evaluated exactly at each irradiation hole. Therefore, a fast neutron flux was measured and evaluated by a 09M-02K capsule irradiated in an OR5 irradiation hole and a 10M-01K capsule irradiated in a CT irradiation hole. Fe, Ni, and Ti wires as the fluence monitor were used for the detection of fast neutron flux. Before the irradiation test, the neutron flux and spectrum was calculated for each irradiation hole using an MCNP code. After the irradiation test, the activity of the fluence monitor was measured by an HPGe detector and the reaction rate was calculated. For the OR5 irradiation hole, the radial difference of the fast neutron flux was observed from a calculated data due to the OR5 irradiation hole being located outside the core. Furthermore, a control absorber rod was withdrawn from the core as the increase of the irradiation time at the same irradiation cycle, so the distribution of neutron flux was changed from the beginning to the end of the cycle. These effects were considered to evaluate the fast neutron flux. Neutron spectrums of the CT and OR5 irradiation hole were adjusted by the measured data. The fluxes of a fast neutron above 1 MeV were compared with calculated and measured value. Although the maximum difference was shown at 18.48%, most of the results showed good agreement. (author)

Measurements of thermal and fast neutron fluxes at the TRIGA reactor

International Nuclear Information System (INIS)

Zerdin, F.; Grabovsek, Z.; Klinc, T.; Solinc, H.

1966-01-01

Gold foils were placed at different positions in the TRIGA reactor core and in the experimental devices. Absolute values of the thermal neutron flux at these positions were obtained by coincidence method. Preliminary fast neutron spectrum was measured by threshold detector and by 'Li 6 sandwich' detector. A short description of the applied method and obtained measurements results are included [sl

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

International Nuclear Information System (INIS)

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

2006-01-01

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

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR

International Nuclear Information System (INIS)

Kurosawa, M.

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54 Mn and 60 Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. (authors)

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

Science.gov (United States)

Kurosawa, Masahiko

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

Notes on neutron flux measurement

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1984-01-01

The main purpose of this work is to get an useful guide to carry out topical neutron flux measurements. Although the foil activation technique is used in the majority of the cases, other techniques, such as those based on fission chambers and self-powered neutron detectors, are also shown. Special interest is given to the description and application of corrections on the measurement of relative and absolute induced activities by several types of detectors (scintillators, G-M and gas proportional counters). The thermal arid epithermal neutron fluxes, as determined in this work, are conventional or effective (West cots fluxes), which are extensively used by the reactor experimentalists; however, we also give some expressions where they are related to the integrated neutron fluxes, which are used in neutron calculations. (Author) 16 refs

Measurements of the energy spectrum of backscattered fast neutrons

International Nuclear Information System (INIS)

Segal, Y.

1976-03-01

Experimental measurements have been made of the energy spectra of neutrons transmitted through slabs of iron, lead and perspex for incident neutron energies of 0.5, 1.0, 1.5 and 1.8 MeV. The neutron energy measurements were made using a He-3 spectrometer. The dependence of the neutrons energy spectrum as a function of scattering thickness was determined. The neutrons source used was a 3MeV Van de Graaff accelerator with a tritium target using the H 3 (p,n) He 3 reaction. The results obtained by the investigator on energy dependence of transmitted neutrons as a function of thickness of scattering material were compared, where possible, with the results obtained by other workers. The comparisons indicated good agreement. The experiment's results are compared with MORSE Monte Carlo calculated values. It is worthwhile to note that direct comparison between measured cross section values and the recommended ones are very far from satisfactory. In almost all cases the calculated spectrum is harder than the experimental one, a situation common to the penetrating and the back-scattered flux

“Influence Method” applied to measure a moderated neutron flux

International Nuclear Information System (INIS)

Rios, I.J.; Mayer, R.E.

2016-01-01

The “Influence Method” is conceived for the absolute determination of a nuclear particle flux in the absence of known detector efficiency. This method exploits the influence of the presence of one detector, in the count rate of another detector when they are placed one behind the other and define statistical estimators for the absolute number of incident particles and for the efficiency. The method and its detailed mathematical description were recently published (Rios and Mayer, 2015 [1]). In this article we apply it to the measurement of the moderated neutron flux produced by an "2"4"1AmBe neutron source surrounded by a light water sphere, employing a pair of "3He detectors. For this purpose, the method is extended for its application where particles arriving at the detector obey a Poisson distribution and also, for the case when efficiency is not constant over the energy spectrum of interest. Experimental distributions and derived parameters are compared with theoretical predictions of the method and implications concerning the potential application to the absolute calibration of neutron sources are considered. - Highlights: • “Influence Method” applied to measure a moderated neutron flux. • Effective efficiency defined independently of calibration sources. • Neutron sources calibration discussion.

Development of high flux thermal neutron generator for neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Vainionpaa, Jaakko H., E-mail: hannes@adelphitech.com [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Chen, Allan X.; Piestrup, Melvin A.; Gary, Charles K. [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Jones, Glenn [G& J Jones Enterprice, 7486 Brighton Ct, Dublin, CA 94568 (United States); Pantell, Richard H. [Department of Electrical Engineering, Stanford University, Stanford, CA (United States)

2015-05-01

The new model DD110MB neutron generator from Adelphi Technology produces thermal (<0.5 eV) neutron flux that is normally achieved in a nuclear reactor or larger accelerator based systems. Thermal neutron fluxes of 3–5 · 10{sup 7} n/cm{sup 2}/s are measured. This flux is achieved using four ion beams arranged concentrically around a target chamber containing a compact moderator with a central sample cylinder. Fast neutron yield of ∼2 · 10{sup 10} n/s is created at the titanium surface of the target chamber. The thickness and material of the moderator is selected to maximize the thermal neutron flux at the center. The 2.5 MeV neutrons are quickly thermalized to energies below 0.5 eV and concentrated at the sample cylinder. The maximum flux of thermal neutrons at the target is achieved when approximately half of the neutrons at the sample area are thermalized. In this paper we present simulation results used to characterize performance of the neutron generator. The neutron flux can be used for neutron activation analysis (NAA) prompt gamma neutron activation analysis (PGNAA) for determining the concentrations of elements in many materials. Another envisioned use of the generator is production of radioactive isotopes. DD110MB is small enough for modest-sized laboratories and universities. Compared to nuclear reactors the DD110MB produces comparable thermal flux but provides reduced administrative and safety requirements and it can be run in pulsed mode, which is beneficial in many neutron activation techniques.

Neutron flux monitoring device

International Nuclear Information System (INIS)

Goto, Yasushi; Mitsubori, Minehisa; Ohashi, Kazunori.

1997-01-01

The present invention provides a neutron flux monitoring device for preventing occurrence of erroneous reactor scram caused by the elevation of the indication of a start region monitor (SRM) due to a factor different from actual increase of neutron fluxes. Namely, judgement based on measured values obtained by a pulse counting method and a judgment based on measured values obtained by a Cambel method are combined. A logic of switching neutron flux measuring method to be used for monitoring, namely, switching to an intermediate region when both of the judgements are valid is adopted. Then, even if the indication value is elevated based on the Cambel method with no increase of the counter rate in a neutron source region, the switching to the intermediate region is not conducted. As a result, erroneous reactor scram such as 'shorter reactor period' can be avoided. (I.S.)

Wide range neutron flux monitor

International Nuclear Information System (INIS)

Endo, Yorimasa; Fukushima, Toshiki.

1983-01-01

Purpose: To provide a wide range neutron-flux monitor adapted such that the flux monitoring function and alarming function can automatically by shifted from pulse counting system to cambel method system. Constitution: A wide range neutron-flux monitor comprises (la) pulse counting system and (lb) cambel-method system for inputting detection signals from neutron detectors and separating them into signals for the pulse measuring system and the cambel measuring system, (2) overlap detection and calculation circuit for detecting the existence of the overlap of two output signals from the (la) and (lb) systems, and (3) trip circuit for judging the abnormal state of neutron detectors upon input of the detection signals. (Seki, T.)

Compact neutron flux monitor

International Nuclear Information System (INIS)

Madhavi, V.; Phatak, P.R.; Bahadur, C.; Bayala, A.K.; Jakati, R.K.; Sathian, V.

2003-01-01

Full text: A compact size neutron flux monitor has been developed incorporating standard boards developed for smart radiation monitors. The sensitivity of the monitors is 0.4cps/nV. It has been tested up to 2075 nV flux with standard neutron sources. It shows convincing results even in high flux areas like 6m away from the accelerator in RMC (Parel) for 106/107 nV. These monitors have a focal and remote display, alarm function with potential free contacts for centralized control and additional provision of connectivity via RS485/Ethernet. This paper describes the construction, working and results of the above flux monitor

Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

International Nuclear Information System (INIS)

Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki; Torii, Kazutaka

2016-01-01

At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For this purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core. (author)

Neutron flux measurement utilizing Campbell technique

International Nuclear Information System (INIS)

Kropik, M.

2000-01-01

Application of the Campbell technique for the neutron flux measurement is described in the contribution. This technique utilizes the AC component (noise) of a neutron chamber signal rather than a usually used DC component. The Campbell theorem, originally discovered to describe noise behaviour of valves, explains that the root mean square of the AC component of the chamber signal is proportional to the neutron flux (reactor power). The quadratic dependence of the reactor power on the root mean square value usually permits to accomplish the whole current power range of the neutron flux measurement by only one channel. Further advantage of the Campbell technique is that large pulses of the response to neutrons are favoured over small pulses of the response to gamma rays in the ratio of their mean square charge transfer and thus, the Campbell technique provides an excellent gamma rays discrimination in the current operational range of a neutron chamber. The neutron flux measurement channel using state of the art components was designed and put into operation. Its linearity, accuracy, dynamic range, time response and gamma discrimination were tested on the VR-1 nuclear reactor in Prague, and behaviour under high neutron flux (accident conditions) was tested on the TRIGA nuclear reactor in Vienna. (author)

Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor

International Nuclear Information System (INIS)

Lerner, A.M.; Madariaga, M.R.

1998-01-01

Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm 2 sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm 2 .sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm 2 .sec). ((1) According to the reaction Au 197 (n,γ)Au 198 , having a cross section of σ 0 =98.8b for thermal neutrons. (2) According to the reaction In 115 (n,n')In 115m , with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [es

Analysis of neutron flux measurement systems using statistical functions

International Nuclear Information System (INIS)

Pontes, Eduardo Winston

1997-01-01

This work develops an integrated analysis for neutron flux measurement systems using the concepts of cumulants and spectra. Its major contribution is the generalization of Campbell's theorem in the form of spectra in the frequency domain, and its application to the analysis of neutron flux measurement systems. Campbell's theorem, in its generalized form, constitutes an important tool, not only to find the nth-order frequency spectra of the radiation detector, but also in the system analysis. The radiation detector, an ionization chamber for neutrons, is modeled for cylindrical, plane and spherical geometries. The detector current pulses are characterized by a vector of random parameters, and the associated charges, statistical moments and frequency spectra of the resulting current are calculated. A computer program is developed for application of the proposed methodology. In order for the analysis to integrate the associated electronics, the signal processor is studied, considering analog and digital configurations. The analysis is unified by developing the concept of equivalent systems that can be used to describe the cumulants and spectra in analog or digital systems. The noise in the signal processor input stage is analysed in terms of second order spectrum. Mathematical expressions are presented for cumulants and spectra up to fourth order, for important cases of filter positioning relative to detector spectra. Unbiased conventional estimators for cumulants are used, and, to evaluate systems precision and response time, expressions are developed for their variances. Finally, some possibilities for obtaining neutron radiation flux as a function of cumulants are discussed. In summary, this work proposes some analysis tools which make possible important decisions in the design of better neutron flux measurement systems. (author)

Neutron flux uncertainty and covariances for spectrum adjustment and estimation of WWER-1000 pressure vessel fluences

International Nuclear Information System (INIS)

Boehmer, Bertram

2000-01-01

Results of estimation of the covariance matrix of the neutron spectrum in the WWER-1000 reactor cavity and pressure vessel positions are presented. Two-dimensional calculations with the discrete ordinates transport code DORT in r-theta and r-z-geometry used to determine the neutron group spectrum covariances including gross-correlations between interesting positions. The new Russian ABBN-93 data set and CONSYST code used to supply all transport calculations with group neutron data. All possible sources of uncertainties namely caused by the neutron gross sections, fission sources, geometrical dimensions and material densities considered, whereas the uncertainty of the calculation method was considered negligible in view of the available precision of Monte Carlo simulation used for more precise evaluation of the neutron fluence. (Authors)

ATW neutron spectrum measurements at LAMPF

Energy Technology Data Exchange (ETDEWEB)

Butler, G.W.; Littleton, P.E.; Morgan, G.L. [Los Alamos National Laboratory, NM (United States)] [and others

1995-10-01

Accelerator transmutation of waste (ATW) is a proposal to use a high flux of accelerator-produced thermalized neutrons to transmute both fission product and higher actinide commercial nuclear waste into stable or short-lived radioactive species in order to avoid long-term storage of nuclear waste. At LAMPF the authors recently performed experiments that were designed to measure the spectrum of neutrons produced per incident proton for full-scale proposed ATW targets of lead and lithium. The neutrons produced in such targets have a spectrum of energies that extends up to the energy of the incident proton beam, but the distribution peaks between 1 and 5 MeV. Transmutation reactions and fission of actinides are most efficient when the neutron energy is below a few eV, so the target must be surrounded by a non-absorbing material (blanket) to produce additional neutrons and reduce the energy of high energy neutrons without loss. The experiments with the lead target, 25 cm diameter by 40 cm long, were conducted with 800 MeV protons, while those with the lithium target, 25 cm diameter by 175 cm long, were conducted with 400 MeV protons. The blanket in both sets of experiments was a 60 cm diameter by 200 cm long annulus of lead that surrounded the target. Surrounding the blanket was a steel water tank with dimensions of 250 cm diameter by 300 cm long that simulated the transmutation region. A small sample pipe penetrated the length of the lead blanket and other sample pipes penetrated the length of the water tank at different radii from the beam axis so that the neutron spectra at different locations could be measured by foil activation. After irradiation the activated foil sets were extracted and counted with calibrated high resolution germanium gamma ray detectors at the Los Alamos nuclear chemistry counting facility.

Measurement of thermal, epithermal and fast neutron flux in the IEA-R1 reactor by the foil activation method

International Nuclear Information System (INIS)

Koskinas, M.F.

1979-01-01

Experimental and theoretical details of the foil activation method applied to neutrons flux measurements at the IEA-R1 reactor are presented. The thermal - and epithermal - neutron flux were determined form activation measurements of gold, cobalt and manganese foils; and for the fast neutron flux determination, aluminum, iron and nickel foils were used. The measurements of the activity induced in the metal foils were performed using a Ge-Li gamma spectrometry system. In each energy range of the reactor neutron spectrum, the agreement among the experimental flux values obtained using the three kind of materials, indicates the consistency of the theoretical approach and of the nuclear parameters selected. (Author) [pt

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.)

Fusion spectrum neutron source computation in "6LiD convertor for HFETR

International Nuclear Information System (INIS)

Sun Shouhua; Hu Yifei; Ye Bin

2014-01-01

A computation model of 14 MeV neutron from the "6LiD convertor has been established, the 14 MeV neutron sources and flux in the irradiation samples from the "6LiD convertor and the core have been computed separately, the neutron spectrum in the irradiation samples have been computed, too. The results show that the neutron sources that over 13 MeV account for 1 MeV above in the "6LiD convertor is 25.7%, 24.6% respectively, 14 MeV neutron sources get 4.31 × 10"1"3 n_T·s"-"1, 3.34 × 10"1"3 n_T·s"-"1, 14 MeV neutron flux get 2.66 × 10"1"0 n_T·cm"-"2·s"-"1, 3.53 × 10"1"0 n_T·cm"-"2·s"-"1, as He and H_2O charged in the irradiation capsule. (authors)

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

Development of Coincidence Method for Determination Thermal Neutron Flux on RSG-GAS

International Nuclear Information System (INIS)

Bakhri, Syaiful; Hamzah, Amir

2004-01-01

The research to develop detection radiation system using coincidence method has been done to determine thermal neutron flux in RS1 and RS2 irradiation facilities RSG-GAS. At this research has arranged beta-gamma coincidence equipment system and parameter of measurement according to Au-198 beta-gamma spectrum. Gold foils that have irradiated for period of time, counted, and the activities of radiation is analyzed to get neutron flux. Result of research indicate that systems measurement of absolute activity with gamma beta coincidence method functioning well and can be applied at activity measurement of gold foil for irradiation facility characterization. The results show that thermal neutron flux in RS1 and RS2, respectively is 2.007E+12 n/cm 2 s and 2.147E+12 n/cm 2 s. To examine the system performance, the result was compared to measure activity using high resolution of Hp Ge detector and achieved discrepancy is about 1.26% and 6.70%. (author)

Study of U235 neutron fission spectrum by the knowledge of cross sections average over that spectrum

International Nuclear Information System (INIS)

Suarez, P.M.

1997-01-01

A literature search of cross sections averaged over the fission neutron spectrum confirms inconsistencies between calculated and experimental values for high threshold reactions. Since, in this case, calculated averaged cross sections are systematically lower than measured values, it is concluded that the representations used to carry out these calculations underestimate the number of neutrons in the high energy region of the spectrum. A careful measurement of the averaged cross section for the 45 Sc(n,2n) 44g Sc and 45 Sc(n,2n) 44m Sc high threshold reactions had been performed in the RA-6 Neutron Activation Analysis Laboratory after carefully checking that the neutron flux at the core position where the samples were being irradiated was indeed an undisturbed fission spectrum. The experimental values are greater than those calculated with either, Watt type representations or the one based on the Madland and Nix model for the prompt fission spectrum. In many areas of nuclear engineering, like validation of nuclear data, reactor calculations, applied nuclear physics, shielding design, etc., it is of great practical importance to have a representation for the neutron flux that can be expressed in a closed analytical form and that agrees with experimental results, specially for the most widely fissile nuclide, 235 U. The results of the calculations mentioned above lead us to propose an analytical form for the 235 U fission neutron spectrum that better agrees with experimental results in the whole energy spectrum. We propose two different forms; both are a modification of the Watt-type form that has been adopted within the ENDF/B-V files. One of the new analytical representations is defined in two regions: below 9.5 MeV it is exactly the same formula as that used within the ENDF/B-V files, above this energy the parameters of this formula are changed. The other proposed analytical representation is expressed by a single formula in the whole energy range. These two new

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.)

Thick-foils activation technique for neutron spectrum unfolding with the MINUIT routine-Comparison with GEANT4 simulations

Science.gov (United States)

Vagena, E.; Theodorou, K.; Stoulos, S.

2018-04-01

Neutron activation technique has been applied using a proposed set of twelve thick metal foils (Au, As, Cd, In, Ir, Er, Mn, Ni, Se, Sm, W, Zn) for off-site measurements to obtain the neutron spectrum over a wide energy range (from thermal up to a few MeV) in intense neutron-gamma mixed fields such as around medical Linacs. The unfolding procedure takes into account the activation rates measured using thirteen (n , γ) and two (n , p) reactions without imposing a guess solution-spectrum. The MINUIT minimization routine unfolds a neutron spectrum that is dominated by fast neutrons (70%) peaking at 0.3 MeV, while the thermal peak corresponds to the 15% of the total neutron fluence equal to the epithermal-resonances area. The comparison of the unfolded neutron spectrum against the simulated one with the GEANT4 Monte-Carlo code shows a reasonable agreement within the measurement uncertainties. Therefore, the proposed set of activation thick-foils could be a useful tool in order to determine low flux neutrons spectrum in intense mixed field.

Measurement of the epithermal neutron flux of the Argonauta reactor by the Sandwich method

International Nuclear Information System (INIS)

Nascimento, H.M.

1973-01-01

A common method of obtaining information about the neutron spectrum in the energy range of 1 eV to a few keV is by using resonance sandwich detectors. A sandwich detector is usually made up of three foils placed one on top of the other, each having the same thickness and being made of the same material which has a pronounced absorption resonance. To make an adequate evaluation, the sandwich method was compared with one using an isolated detector. The results obtained from approximate theoretical calculations were checked experimentally, using In, Au and Mn foils, in an isotropic 1/E flux in the Argonaut Reactor at I.E.N. As practical application of this method, the deviation from a 1/E spectrum of the epithermal neutron flux in the core and external graphite reflector of the Argonaut Reactor has been measured with the sandwich foils previously calibrated in a 1/E spectrum. (author)

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

International Nuclear Information System (INIS)

Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J.P.

2011-01-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 x 10 21 m -2 s -1 , ion fluence: 4 x 10 25 m -2 ) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

Science.gov (United States)

Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J. P.

2011-08-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 × 1021 m-2 s-1, ion fluence: 4 × 1025 m-2) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

Use of new threshold detector 199Hg(n,n')/sup 199m/Hg for neutron spectrum unfolding

International Nuclear Information System (INIS)

Sakurai, K.

1982-01-01

The nuclear data for the 199 Hg(n,n')/sup 199m/Hg reaction are reviewed and the data are used for neutron spectrum unfolding. The neutron spectrum of the YAYOI glory-hole is unfolded by SAND II with 10 nuclear reactions including the 199 Hg(n,n')/sup 199m/Hg reaction. The ratio of the measured reaction rate to the calculated reaction rate is about 1:1.1 for the guess spectrum. The 199 Hg(n,n')/sup 199m/Hg, 115 In(n,n')/sup 115m/In, 103 Rh(n,n')/sup 103m/Rh reactions should be useful threshold detectors for the neutron dosimetry with low level fast neutron flux

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

International Nuclear Information System (INIS)

Aredes, Vitor Ottoni Garcia

2014-01-01

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

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

Energy Technology Data Exchange (ETDEWEB)

Shimada, Masashi, E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Hatano, Y. [Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555 (Japan); Calderoni, P. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Oda, T. [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo 113-8656 (Japan); Oya, Y. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Sokolov, M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Zhang, K. [Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555 (Japan); Cao, G. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI 53706 (United States); Kolasinski, R. [Hydrogen and Metallurgical Science Department, Sandia National Laboratories, Livermore, CA 94551 (United States); Sharpe, J.P. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

2011-08-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 x 10{sup 21} m{sup -2} s{sup -1}, ion fluence: 4 x 10{sup 25} m{sup -2}) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

Measurements of integral cross sections in the californium-252 fission neutron spectrum

International Nuclear Information System (INIS)

Alberts, W.G.; Guenther, E.; Matzke, M.; Rassl, G.

1977-01-01

In a low-scattering arrangement cross sections averaged over the californium-252 spontaneous fission neutron spectrum were measured. The reactions 27 Al(n,α) 46 Ti, 47 Ti, 48 Ti(n,p), 54 Fe, 56 Fe(n,p), 58 Ni(n,p), 64 Zn(n,p), 115 In(n,n') were studied in order to obtain a consistent set of threshold detectors used in fast neutron flux density measurements. Overall uncertainties between 2 and 2.5% could be achieved; corrections due to neutron scattering in source and samples are discussed

Neutron spectrum determination of d(20)+Be source reaction by the dosimetry foils method

Science.gov (United States)

Stefanik, Milan; Bem, Pavel; Majerle, Mitja; Novak, Jan; Simeckova, Eva

2017-11-01

The cyclotron-based fast neutron generator with the thick beryllium target operated at the NPI Rez Fast Neutron Facility is primarily designed for the fast neutron production in the p+Be source reaction at 35 MeV. Besides the proton beam, the isochronous cyclotron U-120M at the NPI provides the deuterons in the energy range of 10-20 MeV. The experiments for neutron field investigation from the deuteron bombardment of thick beryllium target at 20 MeV were performed just recently. For the neutron spectrum measurement of the d(20)+Be source reaction, the dosimetry foils activation method was utilized. Neutron spectrum reconstruction from resulting reaction rates was performed using the SAND-II unfolding code and neutron cross-sections from the EAF-2010 nuclear data library. Obtained high-flux white neutron field from the d(20)+Be source is useful for the intensive irradiation experiments and cross-section data validation.

Photoneutron Flux Measurement via Neutron Activation Analysis in a Radiotherapy Bunker with an 18 MV Linear Accelerator

Science.gov (United States)

Çeçen, Yiğit; Gülümser, Tuğçe; Yazgan, Çağrı; Dapo, Haris; Üstün, Mahmut; Boztosun, Ismail

2017-09-01

In cancer treatment, high energy X-rays are used which are produced by linear accelerators (LINACs). If the energy of these beams is over 8 MeV, photonuclear reactions occur between the bremsstrahlung photons and the metallic parts of the LINAC. As a result of these interactions, neutrons are also produced as secondary radiation products (γ,n) which are called photoneutrons. The study aims to map the photoneutron flux distribution within the LINAC bunker via neutron activation analysis (NAA) using indium-cadmium foils. Irradiations made at different gantry angles (0°, 90°, 180° and 270°) with a total of 91 positions in the Philips SLI-25 linear accelerator treatment room and location-based distribution of thermal neutron flux was obtained. Gamma spectrum analysis was carried out with high purity germanium (HPGe) detector. Results of the analysis showed that the maximum neutron flux in the room occurred at just above of the LINAC head (1.2x105 neutrons/cm2.s) which is compatible with an americium-beryllium (Am-Be) neutron source. There was a 90% decrease of flux at the walls and at the start of the maze with respect to the maximum neutron flux. And, just in front of the LINAC door, inside the room, neutron flux was measured less than 1% of the maximum.

Preliminary validation of computational model for neutron flux prediction of Thai Research Reactor (TRR-1/M1)

Science.gov (United States)

Sabaibang, S.; Lekchaum, S.; Tipayakul, C.

2015-05-01

This study is a part of an on-going work to develop a computational model of Thai Research Reactor (TRR-1/M1) which is capable of accurately predicting the neutron flux level and spectrum. The computational model was created by MCNPX program and the CT (Central Thimble) in-core irradiation facility was selected as the location for validation. The comparison was performed with the typical flux measurement method routinely practiced at TRR-1/M1, that is, the foil activation technique. In this technique, gold foil is irradiated for a certain period of time and the activity of the irradiated target is measured to derive the thermal neutron flux. Additionally, the flux measurement with SPND (self-powered neutron detector) was also performed for comparison. The thermal neutron flux from the MCNPX simulation was found to be 1.79×1013 neutron/cm2s while that from the foil activation measurement was 4.68×1013 neutron/cm2s. On the other hand, the thermal neutron flux from the measurement using SPND was 2.47×1013 neutron/cm2s. An assessment of the differences among the three methods was done. The difference of the MCNPX with the foil activation technique was found to be 67.8% and the difference of the MCNPX with the SPND was found to be 27.8%.

Photoneutron Flux Measurement via Neutron Activation Analysis in a Radiotherapy Bunker with an 18 MV Linear Accelerator

Directory of Open Access Journals (Sweden)

Çeçen Yiğit

2017-01-01

Full Text Available In cancer treatment, high energy X-rays are used which are produced by linear accelerators (LINACs. If the energy of these beams is over 8 MeV, photonuclear reactions occur between the bremsstrahlung photons and the metallic parts of the LINAC. As a result of these interactions, neutrons are also produced as secondary radiation products (γ,n which are called photoneutrons. The study aims to map the photoneutron flux distribution within the LINAC bunker via neutron activation analysis (NAA using indium-cadmium foils. Irradiations made at different gantry angles (0°, 90°, 180° and 270° with a total of 91 positions in the Philips SLI-25 linear accelerator treatment room and location-based distribution of thermal neutron flux was obtained. Gamma spectrum analysis was carried out with high purity germanium (HPGe detector. Results of the analysis showed that the maximum neutron flux in the room occurred at just above of the LINAC head (1.2x105 neutrons/cm2.s which is compatible with an americium-beryllium (Am-Be neutron source. There was a 90% decrease of flux at the walls and at the start of the maze with respect to the maximum neutron flux. And, just in front of the LINAC door, inside the room, neutron flux was measured less than 1% of the maximum.

Evaluation of the transmutation of transuranic using neutrons spectrum from the spallation reaction

Energy Technology Data Exchange (ETDEWEB)

Gilberti, Mauricio; Pereira, Claubia, E-mail: mgilber@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Angra dos Reis, RJ (Brazil); Veloso, Maria A. Fortini, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizante, MG (Brazil). Dept. de Engenharia Nuclear

2013-07-01

The transmutation of transuranic was analyzed by simulating the neutron flux from different spallation sources across arrays of fissile material with isotopic composition PWR reprocessing. A simplified model of Accelerator-Driven Systems (ADS) containing target, moderator graphite, lead-bismuth coolant or sodium coolant, is used. The simulation was made using the particles transport code MCNPX 2.6.0 which allowed to evaluate the rate of transmutation of actinides (Np, Pu, Am, and Cm) at different locations in the system. The objective of the study is to evaluate which the behavior and influences the spectrum of the spallation in the transmutation without the contribution or interference of multiplier, medium subcritical, which would add the contribution of fission neutrons generated, thus interfering in the analysis. The arrangement enable to infer the influence of hardened neutron flux from the spallation reaction in the transmutation, the results show that this is independent of the target material chosen, and the spectrum of spallation has a negligible importance compared to the influence of moderation and scattering generated by the coolant or moderator used. (author)

Evaluation of the transmutation of transuranic using neutrons spectrum from the spallation reaction

International Nuclear Information System (INIS)

Gilberti, Mauricio; Pereira, Claubia; Veloso, Maria A. Fortini

2013-01-01

The transmutation of transuranic was analyzed by simulating the neutron flux from different spallation sources across arrays of fissile material with isotopic composition PWR reprocessing. A simplified model of Accelerator-Driven Systems (ADS) containing target, moderator graphite, lead-bismuth coolant or sodium coolant, is used. The simulation was made using the particles transport code MCNPX 2.6.0 which allowed to evaluate the rate of transmutation of actinides (Np, Pu, Am, and Cm) at different locations in the system. The objective of the study is to evaluate which the behavior and influences the spectrum of the spallation in the transmutation without the contribution or interference of multiplier, medium subcritical, which would add the contribution of fission neutrons generated, thus interfering in the analysis. The arrangement enable to infer the influence of hardened neutron flux from the spallation reaction in the transmutation, the results show that this is independent of the target material chosen, and the spectrum of spallation has a negligible importance compared to the influence of moderation and scattering generated by the coolant or moderator used. (author)

Analysis of the fast-neutron spectrum inside the experimental cavity of the NRU Mk4 FN rod

International Nuclear Information System (INIS)

Leung, T.C.

1995-01-01

The fast-neutron (FN) rods in the NRU reactor provide a facility to study the effects of irradiation on CANDU reactor materials. The Mark 4 (Mk4) FN rods use natural uranium and supply fast-neutrons for experiments on irradiation creep and growth, and corrosion, for pressure- and calandria-tube materials. The neutron fluxes above 1 MeV are up to 2.7x10 17 n.m -2 .s -1 . This paper describes a calculation of the fast-neutron spectrum inside the NRU Mk4 FN rod cavity. The calculation was performed using the WIMS-AECL code, which is a multi-group transport code with two dimensional capabilities using the collision-probability method. Results for the fast-neutron spectrum above 1 MeV are presented in nine groups. The analysis confirms that the spectrum in the fast-neutron irradiation facility in NRU is representative of the actual irradiation spectrum for fast-neutron damage in a CANDU reactor. The effects of changes in specimen holder size, temperature, coolant density and fuel burnup on the fast neutron spectrum are also presented. (author). 9 refs., 3 tabs., 4 figs

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

International Nuclear Information System (INIS)

Nguyen Hoang Hai; Do Quang Binh

2011-01-01

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

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)

Fast neutron fluxes distribution in Egyptian ilmenite concrete

International Nuclear Information System (INIS)

Megahed, R.M.; Abou El-Nasr, T.Z.; Bashter, I.I.

1978-01-01

This work is concerned with the study of the distribution of fast neutron fluxes in a new type of heavy concrete made from Egyptian ilmenite ores. The neutron source used was a collimated beam of reactor neutrons emitted from one of the horizontal channels of the ET-RR-1 reactor. Measurements were carried-out using phosphorous activation detectors. Iso-flux curves were represented which give directly the shape and thickness required to attenuate the emitted fast neutron flux to a certain value. The relaxation lengths were also evaluated from the measured data for both disc monodirectional source and infinite plane monodirectional source. The obtained values were compared with that calculated using the derived values of relative number densities and microscopic removal cross-sections of the different constituents. The obtained data show that ilmenite concrete attenuates fast neutron flux more strongly than ordinary concrete. A semiemperical formula was derived to calculate the fast neutron flux at different thicknesses along the beam axis. Another semiemperical formula was also derived to calculate the fast neutron flux in ordinary concrete along the beam axis using the corresponding value in ilmenite concrete

Use of Germanium as comparator and integral monitor of neutron flux in activation analysis

International Nuclear Information System (INIS)

Furnari, Juan C.; Cohen, Isaac M.; Arribere, Maria A.; Kestelman, Abraham J.

1997-01-01

The possibility of using germanium as monitor of the thermal and epithermal components of the neutron flux, and comparator in parametric activation analysis, is discussed. The advantages and drawbacks associated to the use of this element are commented on, and the comparison with zirconium, in terms of the determination relative error, is performed. The utilisation of germanium as integral flux monitor, including the fast component of the neutron spectrum, is also discussed. Data corresponding to measurements of k 0 factor for the most relevant gamma transitions from Ge-75 and Be-77 are presented, as well as the results of the reference material analysis, employing germanium as flux monitor and comparator in a simultaneous way. (author). 8 refs., 3 figs., 2 tabs

Cadmium depletion impacts on hardening neutron spectrum for advanced fuel testing in ATR

International Nuclear Information System (INIS)

Chang, Gray S.

2011-01-01

For transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products effectively is in a fast neutron spectrum reactor. In the absence of a fast spectrum test reactor in the United States of America (USA), initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. A test region is achieved with a Cadmium (Cd) filter which can harden the neutron spectrum to a spectrum similar (although still somewhat softer) to that of the liquid metal fast breeder reactor (LMFBR). A fuel test loop with a Cd-filter has been installed within the East Flux Trap (EFT) of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). A detailed comparison analyses between the cadmium (Cd) filter hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum have been performed using MCWO. MCWO is a set of scripting tools that are used to couple the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2.2. The MCWO-calculated results indicate that the Cd-filter can effectively flatten the Rim-Effect and reduce the linear heat rate (LHGR) to meet the advanced fuel testing project requirements at the beginning of irradiation (BOI). However, the filtering characteristics of Cd as a strong absorber quickly depletes over time, and the Cd-filter must be replaced for every two typical operating cycles within the EFT of the ATR. The designed Cd-filter can effectively depress the LHGR in experimental fuels and harden the neutron spectrum enough to adequately flatten the Rim-Effect in the test region. (author)

Neutron energy spectrum in graphite blankets of fusion reactors

International Nuclear Information System (INIS)

Tsechanski, A.

1981-09-01

Neutron flux measurements were performed in a graphite stack and compared with calculations made with a two dimensional transport computer code. In the present work it is observed that the calculated spectrum in the elastic and inelastic scattering ranges (the first collision range in both cases), is sensitive to details of the angular distribution of these neutrons. Regarding the discrepancies in the elastic scattering range it is concluded that the microscopic cross section library ENDF/B-IV overestimates the large angle scattering (back scattering) as can be seen from comparison of measured and calculated spectra. The two most important conclusions of the present work are: 1. Inelastic scattering interaction of D-T neutrons in graphite cannot be calculated without a proper account of energy-angle correlation. 2. An experimental setup supplying monoenergetic collimated D-T neutrons constitutes a sensitive although indirect means for measuring angular distributions in inelastic and elastic scattering

Optimization of the testing volumes with respect to neutron flux levels in the two-target high flux D-Li neutron source for the international fusion materials irradiation facility

International Nuclear Information System (INIS)

Kelleher, W.P.; Varsamis, G.L.

1989-01-01

An economic and fusion-relevant source of high-energy neutrons is an essential element in the fusion nuclear technology and development program. This source can be generated by directing a high energy deuteron beam onto a flowing liquid lithium target, producing neutrons via the D-Lithium stripping reaction. Previous work on this type of source concentrated on a design employing one deuteron beam of modest amperage. This design was shown to have a relatively small testing volume with high flux gradients and was therefor considered somewhat unattractive from a materials testing standpoint. A design using two lithium targets and two high-amperage beams has recently been proposed. This two beam design has been examined in an effort to maximize the test volume while minimizing the flux gradients and minimizing the effect of radiation damage on one target due to the other. A spatial, energy and angle dependent neutron source modeling the D-Lithium source was developed. Using this source, a 3-dimensional map of uncollided flux within the test volume was calculated. The results showed that the target separation has little effect on the available experimental volume and that a testing volume of ∼35 liters is available with a volume averaged flux above 10 14 n/cm 2 /s. The collided flux within the test volume was then determined by coupling the source model with a Monte Carlo code. The spectral effects of the high-energy tail in the flux were examined and evaluated as to possible effects on materials response. Calculations comparing the radiation damage to materials from the D-Lithium source to that cause by a standard DT fusion first-wall neutron flux spectrum showed that the number of appm and dpa, as well as the ratio appm/dpa and dpa/MW/m 2 are within 30% for the two sources. 8 refs., 8 figs

Neutron spectrum in small iron pile surrounded by lead reflector

International Nuclear Information System (INIS)

Kimura, Itsuro; Hayashi, S.A.; Kobayashi, Katsuhei; Matsumura, Tetsuo; Nishihara, Hiroshi.

1978-01-01

In order to save the quantity of sample material, a possibility to assess group constants of a reactor material through measurement and analysis of neutron spectrum in a small sample pile surrounded by a reflector of heavy moderator, was investigated. As the sample and the reflector, we chose iron and lead, respectively. Although the time dispersion in moderation of neutrons was considerably prolonged by the lead reflector, this hardly interferes with the assessment of group constants. Theoretical calculation revealed that both the neutron flux spectrum and the sensitivity coefficient of group constants in an iron sphere, 35 cm in diameter surrounded by the lead reflector, 25 cm thick, were close to those of the bare iron sphere, 108 cm in diameter. The neutron spectra in a small iron pile surrounded by a lead reflector were experimentally obtained by the time-of-flight method with an electron linear accelerator and the result was compared with the predicted values. It could be confirmed that a small sample pile surrounded by a reflector, such as lead, was as useful as a much larger bulk pile for the assessment of group constants of a reactor material. (auth.)

Generating energy dependent neutron flux maps for effective ...

African Journals Online (AJOL)

For activation analysis and irradiation scheme of miniature neutron source reactor, designers or engineers usually require information on thermal neutron flux levels and other energy group flux levels (such as fast, resonance and epithermal). A methodology for readily generating such flux maps and flux profiles for any ...

Self-powered neutron flux detector

International Nuclear Information System (INIS)

Kroon, J.

1979-01-01

A self-powered neutron flux detector having an emitter electrode, at least a major portion of which is, 95 Mo encased in a tubular collector electrode and separated therefrom by dielectric material. The 95 Mo emitter electrode has experimentally shown a 98% prompt response, is primarily sensitive to neutron flux, has adequate sensitivity and has low burn up. Preferably the emitter electrode is molybdenum which has been enriched 75% to 99% by weight with 95 Mo

Study of the effect of the energy spectrum and of the total flux on the damage produced by neutrons in solids; Contribution a l'etude de l'influence du spectre et du flux integre sur les dommages crees par les neutrons dans les solides

Energy Technology Data Exchange (ETDEWEB)

Dulieu, P C [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-01-01

In the first part are studied the general relationships between the physical effects produced by neutrons in solids, and the total flux and neutron energy spectrum; some examples are given. The second part, describes the application to a silicon damage detector whose principle is to use the damage produced in a PIN Junction for measuring the neutron flux (intermediate and fast) received by the detector. Chapter I is devoted to the experimental determination of the energy given to the atoms by a primary in the silicon. The results and conclusions drawn from this determination make it possible to consider, in chapter II, the calculation of the detectors response characteristics.Chapter III deals with the measurement of the detectors response function and it is noted that good agreement is obtained between the calculation and experimental results. The whole of the second part constitutes a test of the methods presented in the first part. (author) [French] Dans la premiere partie, nous etudions les relations generales qui lient les effets physiques engendres par les neutrons dans les solides au flux integre et au spectre des neutrons et nous donnons des exemples d'utilisation. La deuxieme partie est une application au detecteur de dommages en silicium, dont le principe est d'utiliser les dommages crees dans une jonction PIN pour mesurer les flux de neutrons (intermediaires et rapides) recus par le detecteur. Le chapitre I est consacre a la determination experimentale de l'energie cedee aux atomes par un primaire dans le silicium. Les resultats et les conclusions que l'on peut tirer de cette determination permettent d'aborder, au chapitre II, le calcul de la fonction de reponse du detecteur. Le chapitre III porte sur la mesure de la fonction de reponse du detecteur et on constate qu'il y a un bon accord entre le calcul et l'experience. L'ensemble de la deuxieme partie constitue un test des methodes exposees dans la premiere partie. (auteur)

Real time neutron flux monitoring using Rh self powered neutron detector

Energy Technology Data Exchange (ETDEWEB)

Juna, Byung Jin; Lee, Byung Chul; Park, Sang Jun; Jung, Hoan Sung [KAERI, Daejeon (Korea, Republic of)

2012-10-15

Rhodium (Rh) self powered neutron detectors (SPNDs) are widely used for on line monitoring of local neutron flux. Its signal is slower than the actual variation of neutron flux owing to a delayed {beta} decay of the Rh activation product, but real time monitoring is possible by solving equations between the neutron reaction rate in the detector and its signal. While the measuring system is highly reliable, the accuracy depends on the method solving the equations and accuracy of the parameters in the equations. The uncertain parameters are the contribution of gamma rays to the signal, and the branching ratios of Rh 104 and Rh 104m after the neutron absorption of Rh 103. Real time neutron flux monitoring using Rh SPNDs has been quite successful for neutron transmutation doping (NTD) at HANARO. We revisited the initial data used for the verification of a real time monitoring system, to refine algorithm for a better solution and to check the parameters for correctness. As a result, we suggest an effective way to determine the prompt parameter.

Real time neutron flux monitoring using Rh self powered neutron detector

International Nuclear Information System (INIS)

Juna, Byung Jin; Lee, Byung Chul; Park, Sang Jun; Jung, Hoan Sung

2012-01-01

Rhodium (Rh) self powered neutron detectors (SPNDs) are widely used for on line monitoring of local neutron flux. Its signal is slower than the actual variation of neutron flux owing to a delayed β decay of the Rh activation product, but real time monitoring is possible by solving equations between the neutron reaction rate in the detector and its signal. While the measuring system is highly reliable, the accuracy depends on the method solving the equations and accuracy of the parameters in the equations. The uncertain parameters are the contribution of gamma rays to the signal, and the branching ratios of Rh 104 and Rh 104m after the neutron absorption of Rh 103. Real time neutron flux monitoring using Rh SPNDs has been quite successful for neutron transmutation doping (NTD) at HANARO. We revisited the initial data used for the verification of a real time monitoring system, to refine algorithm for a better solution and to check the parameters for correctness. As a result, we suggest an effective way to determine the prompt parameter

Fission neutron spectrum averaged cross sections for threshold reactions on arsenic

International Nuclear Information System (INIS)

Dorval, E.L.; Arribere, M.A.; Kestelman, A.J.; Comision Nacional de Energia Atomica, Cuyo Nacional Univ., Bariloche; Ribeiro Guevara, S.; Cohen, I.M.; Ohaco, R.A.; Segovia, M.S.; Yunes, A.N.; Arrondo, M.; Comision Nacional de Energia Atomica, Buenos Aires

2006-01-01

We have measured the cross sections, averaged over a 235 U fission neutron spectrum, for the two high threshold reactions: 75 As(n,p) 75 mGe and 75 As(n,2n) 74 As. The measured averaged cross sections are 0.292±0.022 mb, referred to the 3.95±0.20 mb standard for the 27 Al(n,p) 27 Mg averaged cross section, and 0.371±0.032 mb referred to the 111±3 mb standard for the 58 Ni(n,p) 58m+g Co averaged cross section, respectively. The measured averaged cross sections were also evaluated semi-empirically by numerically integrating experimental differential cross section data extracted for both reactions from the current literature. The calculations were performed for four different representations of the thermal-neutron-induced 235 U fission neutron spectrum. The calculated cross sections, though depending on analytical representation of the flux, agree with the measured values within the estimated uncertainties. (author)

Evaluation of neutron flux in the WWR-SM reactor channel and in the irradiating zone of U-150 cyclotron

International Nuclear Information System (INIS)

Sadikov, I.I.; Zinov'ev, V.G.; Sadikova, Z.O.; Salimov, M.I.

2006-01-01

Full text: For effective work of a reactor, and correct planning of experiments related to the reactor irradiation of various materials it is required to control a neutron flux in the given irradiation point for a long irradiation period. For realization of research works on topazes ennobling under irradiation by reactor neutrons as well as by secondary neutrons produced in a cyclotron it is necessary to know the total neutron flux and spectra. To resolve the problem a technique for registration of neutrons with different energy and calculation of a neutrons spectrum in the given irradiation points in reactor channels and in cyclotron behind the nickel target has been developed. Neutron flux density and energy spectra were monitored by use of the following nuclear reactions: 59 Co(n,γ) 60 Co, 197 Au(n,γ) 198 Au, 58 Ni(n,p) 58 Co, 24 Mg(n,p) 24 Na, 48 Ti(n,p) 48 Sc, 46 Ti(n,p) 46 Sc, 54 Fe(n,p) 54 Mn, 89 Y(n,2n) 88 Y, 60 Ni(np) 60 Co. Gamma spectrometer composed of HPGe detector (Rel. Eff. - 15%) and Digital Spectra Analyzer DSA-1000 (Canberra Ind., USA) was used to measure gamma activity of irradiated samples. Acquired gamma spectra were processed by means of Genie 2000 standard software package. The σ(E) functions and neutron spectra were calculated by using the least squares method and approximating the tabular and experimental data with power polynomials. The developed technique was applied for the adjustment of the topazes irradiation regimes in the reactor core and under secondary neutrons flux from a nickel target in the cyclotron. The given technique allows to calculate a logarithmic spectrum of neutrons in a energy range from 0,025 eV up to 12 MeV with the uncertainty of about 10 %. (author)

The Neutron Spectrum in a Uranium Tube

International Nuclear Information System (INIS)

Johansson, E.; Jonsson, E.; Lindberg, M.; Mednis, J.

1963-10-01

A series of experimental and theoretical investigations on neutron spectra in lattice cells has been started at the reactor R1. This report gives the results from the first one of these cells - one with a tube of natural -uranium surrounded by heavy water. In the measurements the cell was placed in the central, vertical channel of the reactor. The neutron spectrum from a lead scatterer in the uranium tube - outer diameter 49.2 mm, inner diameter 28.3 mm - was measured with a fast chopper in the energy region 0.01 to 100 eV. Subsidiary measurements indicated that the spectrum in the beam from the lead piece corresponds to the spectrum of the angular flux integrated over all angles. This correspondence is important for the interpretation of the experimental data. The thermal part of the spectrum was found to deviate significantly from a Maxwellian. However, the deviation is not very large, and one could use a Maxwellian, at least to give a rough idea of the hardness of the spectrum. For the present tube the temperature of this Maxwellian was estimated as 90 to 100 deg C above the moderator temperature (33 deg C). In the joining region the rise of the spectrum towards the thermal part is slower than for the cell boundary spectrum, measured earlier. In the epithermal region the limited resolution of the chopper has affected the measurements at the energies of the uranium resonances. However, the shape of the spectrum on the flanks of the first resonance in 238 U (6.68 eV) has been obtained accurately. In the theoretical treatment the THERMOS code with a free gas scattering model has been used. The energy region was 3.06 - 0.00025 eV. The agreement with the measurements is good for the thermal part - possibly the theoretical spectrum is a little softer than the experimental one. In the joining region the results from THERMOS are comparatively high - probably due to the scattering model used

The Neutron Spectrum in a Uranium Tube

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Jonsson, E; Lindberg, M; Mednis, J

1963-10-15

A series of experimental and theoretical investigations on neutron spectra in lattice cells has been started at the reactor R1. This report gives the results from the first one of these cells - one with a tube of natural -uranium surrounded by heavy water. In the measurements the cell was placed in the central, vertical channel of the reactor. The neutron spectrum from a lead scatterer in the uranium tube - outer diameter 49.2 mm, inner diameter 28.3 mm - was measured with a fast chopper in the energy region 0.01 to 100 eV. Subsidiary measurements indicated that the spectrum in the beam from the lead piece corresponds to the spectrum of the angular flux integrated over all angles. This correspondence is important for the interpretation of the experimental data. The thermal part of the spectrum was found to deviate significantly from a Maxwellian. However, the deviation is not very large, and one could use a Maxwellian, at least to give a rough idea of the hardness of the spectrum. For the present tube the temperature of this Maxwellian was estimated as 90 to 100 deg C above the moderator temperature (33 deg C). In the joining region the rise of the spectrum towards the thermal part is slower than for the cell boundary spectrum, measured earlier. In the epithermal region the limited resolution of the chopper has affected the measurements at the energies of the uranium resonances. However, the shape of the spectrum on the flanks of the first resonance in {sup 238}U (6.68 eV) has been obtained accurately. In the theoretical treatment the THERMOS code with a free gas scattering model has been used. The energy region was 3.06 - 0.00025 eV. The agreement with the measurements is good for the thermal part - possibly the theoretical spectrum is a little softer than the experimental one. In the joining region the results from THERMOS are comparatively high - probably due to the scattering model used.

An alternative method for the measurement of neutron flux

Indian Academy of Sciences (India)

A simple and easy method for measuring the neutron flux is presented. This paper deals with the experimental verification of neutron dose rate–flux relationship for a non-dissipative medium. Though the neutron flux cannot be obtained from the dose rate in a dissipative medium, experimental result shows that for ...

Neutron Flux Distribution on Neutron Radiography Facility After Fixing the Collimator

International Nuclear Information System (INIS)

Supandi; Parikin; Mohtar; Sunardi; Roestam, S

1996-01-01

The Radiography Neutron Facility consists of an inner collimator, outer collimator, main shutter, second shutter and the sample chamber with 300 mm in diameter. Neutron beam quality depends on the neutron flux intensities distribution, L/D ratio Cd ratio, neutron/gamma ratio. The results show that the neutron flux intensity was 2.83 x 107 n cm-2.s-1, with deviation of + 7.8 % and it was distributed homogeneously at the sample position of 200 mm diameter. The beam characteristics were L/D ratio 98 and Rod 8, and neutron gamma ratio 3.08 x 105n.cm-2.mR-1 and Reactor Power was 20 MW. This technique can be used to examine sample with diameter of < 200 mm

A neutron spectrum unfolding computer code based on artificial neural networks

International Nuclear Information System (INIS)

Ortiz-Rodríguez, J.M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J.M.; Vega-Carrillo, H.R.

2014-01-01

The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6 LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding

Spatial distribution of neutron flux for the A-711 neutron generator

International Nuclear Information System (INIS)

Essiet, A. E.; Owolabi, S. A.; Adesanmi, C. A.; Balogun, F. A.

1996-01-01

The spatial distribution of neutron flux for the Kaman sciences A-711 neutron generator recently installed at the Centre for Energy Research and Development (CERD), Ile-Ife Nigeria has been determined. At an operational tube current of 2.0 mA and high voltage power supply (HVPS) of 158 kV, the neutron flux increases from 1.608 ± 0.021*10 8 n/cm 2 s at the top of the irradiated plastic vial to 2.640 ± 0.022*10 8 n/cm 2 s at the centre, and then decreases to 1.943 ± 0.02* 8 n/cm 2 s at the bottom. The flux density is strongly dependent on the diameter of deuteron at the tritium target, and within this range a source strength of 10 8 n/s has been measured for the A-711 neutron generator

LANSCE steady state unperturbed thermal neutron fluxes at 100 μA

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

The ''maximum'' unperturbed, steady state thermal neutron flux for LANSCE is calculated to be 2 /times/ 10 13 n/cm 2 -s for 100 μA of 800-MeV protons. This LANSCE neutron flux is a comparable entity to a steady state reactor thermal neutron flux. LANSCE perturbed steady state thermal neutron fluxes have also been calculated. Because LANSCE is a pulsed neutron source, much higher ''peak'' (in time) neutron fluxes can be generated than at a steady state reactor source. 5 refs., 5 figs

A Novel Detector for High Neutron Flux Measurements

International Nuclear Information System (INIS)

Singo, T. D.; Wyngaardt, S. M.; Papka, P.; Dobson, R. T.

2010-01-01

Measuring alpha particles from a neutron induced break-up reaction with a mass spectrometer can be an excellent tool for detecting neutrons in a high neutron flux environment. Break-up reactions of 6 Li and 12 C can be used in the detection of slow and fast neutrons, respectively. A high neutron flux detection system that integrates the neutron energy sensitive material and helium mass spectrometer has been developed. The description of the detector configuration is given and it is soon to be tested at iThemba LABS, South Africa.

Study of the effect of the energy spectrum and of the total flux on the damage produced by neutrons in solids; Contribution a l'etude de l'influence du spectre et du flux integre sur les dommages crees par les neutrons dans les solides

Energy Technology Data Exchange (ETDEWEB)

Dulieu, P.C. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-01-01

In the first part are studied the general relationships between the physical effects produced by neutrons in solids, and the total flux and neutron energy spectrum; some examples are given. The second part, describes the application to a silicon damage detector whose principle is to use the damage produced in a PIN Junction for measuring the neutron flux (intermediate and fast) received by the detector. Chapter I is devoted to the experimental determination of the energy given to the atoms by a primary in the silicon. The results and conclusions drawn from this determination make it possible to consider, in chapter II, the calculation of the detectors response characteristics.Chapter III deals with the measurement of the detectors response function and it is noted that good agreement is obtained between the calculation and experimental results. The whole of the second part constitutes a test of the methods presented in the first part. (author) [French] Dans la premiere partie, nous etudions les relations generales qui lient les effets physiques engendres par les neutrons dans les solides au flux integre et au spectre des neutrons et nous donnons des exemples d'utilisation. La deuxieme partie est une application au detecteur de dommages en silicium, dont le principe est d'utiliser les dommages crees dans une jonction PIN pour mesurer les flux de neutrons (intermediaires et rapides) recus par le detecteur. Le chapitre I est consacre a la determination experimentale de l'energie cedee aux atomes par un primaire dans le silicium. Les resultats et les conclusions que l'on peut tirer de cette determination permettent d'aborder, au chapitre II, le calcul de la fonction de reponse du detecteur. Le chapitre III porte sur la mesure de la fonction de reponse du detecteur et on constate qu'il y a un bon accord entre le calcul et l'experience. L'ensemble de la deuxieme partie constitue un test des methodes exposees dans la premiere partie

In-core neutron flux measurements at PARR using self powered neutron detector

International Nuclear Information System (INIS)

Hussain, A.; Ansari, S.A.

1989-10-01

This report describes experimental reactor physics measure ments at PARR using the in-core neutron detectors. Rhodium self powered neutron detectors (SPND) were used in the PARR core and several measurements were made aimed at detector calibration, response time determination and neutron flux measurements. The detectors were calibrated at low power using gold foils and full power by the thermal channel. Based on this calibration it was observed that the detector response remains almost linear throughout the power range. The self powered detectors were used for on-line determination of absolute neutron flux in the core as well as the spatial distribution of neutron flux or reactor power. The experimental, axial and horizontal flux mapping results at certain locations in the core are presented. The total response time of rhodium detector was experimentally determined to be about 5 minutes, which agree well with the theoretical results. Because of longer response time of SPND of the detectors it is not possible to use them in the reactor protection system. (author). 10 figs

Cosmogenic excess of 40K and the flux of fast neutrons in meteorites

International Nuclear Information System (INIS)

Stegmann, W.; Begemann, F.

1975-01-01

Results are reported of a mass spectrometric investigation of the content and isotopic composition of potassium from the mesosiderite Emery (silicate phase) and the chondrite Elenovka (bulk). Normalized to the Nier value of 39 K/ 41 K = 13.47, the 40 K in Emery (K-content 220 +- 25 ppm) was found to be enriched by (4.03 +- 0.30)%, the potassium from Elenovka (760 +- 50 ppm) to be indistinguishable from terrestrial potassium. Evidence is presented that the excess 40 K in the silicates from Emery (Ca-content 6.06 weight %) has been produced essentially by secondary cosmic ray neutrons via the 40 Ca(n,p)-reaction. The total excess of (2.57 +- 0.39) x 10 14 40 K-atoms/gCa together with the excitation function of the 40 Ca(n,p)-reaction and the neutron flux spectrum of Arnold, Honda and Lal yields a dose of fast neutrons (2 MeV 16 neutrons/cm 2 and an average flux during the cosmic ray exposure age T = (134 +- 12) Myrs of PHI = (17.4 +- 3.1) neutrons/cm 2 sec. (orig./BJ) [de

Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping

Energy Technology Data Exchange (ETDEWEB)

Fourmentel, D.; Villard, J. F.; Lyoussi, A. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Reynard-Carette, C. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France); Bignan, G.; Chauvin, J. P.; Gonnier, C.; Guimbal, P.; Malo, J. Y. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Carette, M.; Janulyte, A.; Merroun, O.; Brun, J.; Zerega, Y.; Andre, J. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France)

2011-07-01

We study the combined analysis of nuclear measurements to improve the knowledge of the irradiation conditions in the experimental locations of the future Jules Horowitz Reactor (JHR). The goal of the present work is to measure more accurately neutron flux, photon flux and nuclear heating in the reactor. In a Material Testing Reactor (MTR), nuclear heating is a crucial parameter to design the experimental devices to be irradiated in harsh nuclear conditions. This parameter drives the temperature of the devices and of the samples. The numerical codes can predict this parameter but in-situ measurements are necessary to reach the expected accuracy. For this reason, one objective of the IN-CORE program [1] is to study the combined measurements of neutron and photon flux and their cross advanced interpretation. It should be reminded that both neutron and photon sensors are not totally selective as their signals are due to neutron and photon interactions. We intend to measure the neutron flux by three different kinds of sensors (Uranium Fission chamber, Plutonium Fission chamber and Self Powered Neutron Detector), the photon flux by two different sensors (Ionization chamber and Self Powered Gamma Detector) and the nuclear heating by two different ones (Differential calorimeter and Gamma Thermometer). For the same parameter, we expect that the use of different kinds of sensors will allow a better estimation of the aimed parameter by mixing different spectrum responses and different neutron and gamma contributions. An experimental test called CARMEN-1 is scheduled in OSIRIS reactor (CEA Saclay - France) at the end of 2011, with the goal to map irradiation locations in the reactor reflector to get a first validation of the analysis model. This article focuses on the sensor selection for CARMEN-1 experiment and to the way to link neutron and photon flux measurements in view to reduce their uncertainties but also to better assess the neutron and photon contributions to nuclear

Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping

International Nuclear Information System (INIS)

Fourmentel, D.; Villard, J. F.; Lyoussi, A.; Reynard-Carette, C.; Bignan, G.; Chauvin, J. P.; Gonnier, C.; Guimbal, P.; Malo, J. Y.; Carette, M.; Janulyte, A.; Merroun, O.; Brun, J.; Zerega, Y.; Andre, J.

2011-01-01

We study the combined analysis of nuclear measurements to improve the knowledge of the irradiation conditions in the experimental locations of the future Jules Horowitz Reactor (JHR). The goal of the present work is to measure more accurately neutron flux, photon flux and nuclear heating in the reactor. In a Material Testing Reactor (MTR), nuclear heating is a crucial parameter to design the experimental devices to be irradiated in harsh nuclear conditions. This parameter drives the temperature of the devices and of the samples. The numerical codes can predict this parameter but in-situ measurements are necessary to reach the expected accuracy. For this reason, one objective of the IN-CORE program [1] is to study the combined measurements of neutron and photon flux and their cross advanced interpretation. It should be reminded that both neutron and photon sensors are not totally selective as their signals are due to neutron and photon interactions. We intend to measure the neutron flux by three different kinds of sensors (Uranium Fission chamber, Plutonium Fission chamber and Self Powered Neutron Detector), the photon flux by two different sensors (Ionization chamber and Self Powered Gamma Detector) and the nuclear heating by two different ones (Differential calorimeter and Gamma Thermometer). For the same parameter, we expect that the use of different kinds of sensors will allow a better estimation of the aimed parameter by mixing different spectrum responses and different neutron and gamma contributions. An experimental test called CARMEN-1 is scheduled in OSIRIS reactor (CEA Saclay - France) at the end of 2011, with the goal to map irradiation locations in the reactor reflector to get a first validation of the analysis model. This article focuses on the sensor selection for CARMEN-1 experiment and to the way to link neutron and photon flux measurements in view to reduce their uncertainties but also to better assess the neutron and photon contributions to nuclear

Fast neutron flux and intracranial dose distribution at a neutron irradiation facility

International Nuclear Information System (INIS)

Matsumoto, Tetsuo; Aizawa, Otohiko; Nozaki, Tetsuya

1981-01-01

A head phantom filled with water was used to measure the fast neutron flux using 115 In(n, n')sup(115m)In and 103 Rh(n, n')sup(103m)Rh reactions. γ-ray from sup(115m)In and x-ray from sup(103m)Rh were detected by a Ge(Li) and a Na(Tl)I counter, respectively. TLD was used to investigate the γ-dose rate distribution inside the phantom. Flux of fast neutron inside the phantom was about 1 x 10 6 n/cm 2 sec, which was 3 order smaller than that of thermal neutron. The fast neutron flux decreased to 1/10 at 15 cm depth, and γ-dose rate was about 200 R/h at 100 kW inside the phantom. Total dose at the surface was 350 rad/h, to which, fast neutrons contributed more than γ-rays. The rate of fast neutron dose was about 10% of thermal neutron's in Kerma dose unit (rad), however, the rate was highly dependent on RBE value. (Nakanishi, T.)

Calculations of neutron spectra after neutron-neutron scattering

Energy Technology Data Exchange (ETDEWEB)

Crawford, B E [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Stephenson, S L [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); 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); Tornow, W [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Furman, W I [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Lychagin, E V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Muzichka, A Yu [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Nekhaev, G V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Strelkov, A 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)

2004-09-01

A direct neutron-neutron scattering length, a{sub nn}, measurement with the goal of 3% accuracy (0.5 fm) is under preparation at the aperiodic pulsed reactor YAGUAR. A direct measurement of a{sub nn} will not only help resolve conflicting results of a{sub nn} by indirect means, but also in comparison to the proton-proton scattering length, a{sub pp}, shed light on the charge-symmetry of the nuclear force. We discuss in detail the analysis of the nn-scattering data in terms of a simple analytical expression. We also discuss calibration measurements using the time-of-flight spectra of neutrons scattered on He and Ar gases and the neutron activation technique. In particular, we calculate the neutron velocity and time-of-flight spectra after scattering neutrons on neutrons and after scattering neutrons on He and Ar atoms for the proposed experimental geometry, using a realistic neutron flux spectrum-Maxwellian plus epithermal tail. The shape of the neutron spectrum after scattering is appreciably different from the initial spectrum, due to collisions between thermal-thermal and thermal-epithermal neutrons. At the same time, the integral over the Maxwellian part of the realistic scattering spectrum differs by only about 6 per cent from that of a pure Maxwellian nn-scattering spectrum.

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

International Nuclear Information System (INIS)

Perry, D.G.

1975-10-01

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

New fission-neutron-spectrum representation for ENDF

International Nuclear Information System (INIS)

Madland, D.G.

1982-04-01

A new representation of the prompt fission neutron spectrum is proposed for use in the Evaluated Nuclear Data File (ENDF). The proposal is made because a new theory exists by which the spectrum can be accurately predicted as a function of the fissioning nucleus and its excitation energy. Thus, prompt fission neutron spectra can be calculated for cases where no measurements exist or where measurements are not possible. The mathematical formalism necessary for application of the new theory within ENDF is presented and discussed for neutron-induced fission and spontaneous fission. In the case of neutron-induced fission, expressions are given for the first-chance, second-chance, third-chance, and fourth-chance fission components of the spectrum together with that for the total spectrum. An ENDF format is proposed for the new fission spectrum representation, and an example of the use of the format is given

Experimental measurement of neutron spectrum in the reflector of a light water reactor; Determination experimentale du spectre des neutrons dans le reflecteur d'une pile a eau legere

Energy Technology Data Exchange (ETDEWEB)

Brethe, P [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1963-09-15

1. Thermal neutrons: The temperature of the thermal neutron spectrum was calculated using Au-Lu foils. This temperature varies from 300 deg. K (temperature of the moderator) at 30 cm of the core to 350 deg. K in a hole of the core. 2. Slowing down of neutron: Four resonance detectors have been used (Au, In, Co, Mn). We can write a 1/E form of the spectrum. The linking up energy E{sub M} between thermal neutron spectrum and slowing down spectrum is about 0.23 eV and is free from the Maxwell spectrum temperature. The decrease of slowing down flux regarding thermal flux, farther from the core, has been showed. 3. Fast neutrons: We used 3 threshold detectors (Ni, Al, Mg). We supposed a E{sup 1/2} e{sup -{beta}}{sup E} from of the spectrum above 3 MeV. The values of {beta} are in a range from 0.775, at the centre of the core and in a loop-hole, to 0,64 at about 30 cm of the core. 4. Continuous shape of the spectrum: The following interpolations give useful informations between the field where measurements have been made: between 340 eV and 10 keV: 1/E form between 10 keV and 330 keV: 1/(E {sigma}{sub S}(E)) form ({sigma}{sub S}(E) elastic scattering section on hydrogen) between 330 keV and 3 MeV: calculated form by the moments method (ref. BSR). (author) [French] 1. Neutrons thermiques: La temperature du spectre des neutrons thermiques a ete determinee par la methode (or-lutecium). Cette temperature varie de 300 deg. K (temperature du moderateur) a 30 cm du coeur, a 350 deg. K dans une encoche du coeur. 2. Neutrons en ralentissement: 4 detecteurs resonnants ont ete employes (Au, In, Co, Mn). Le spectre peut etre mis sous la forme 1/E quelle que soit la distance a la limite coeur-reflecteur. L'energie de raccordement E{sub M} entre spectre des neutrons thermiques et spectre en ralentissement est environ 0,23 eV et independante de la temperature du spectre de Maxwell. La diminution relative du flux en ralentissement par rapport au flux thermique quand la distance au coeur

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)

Intermediate neutron spectrum problems and the intermediate neutron spectrum experiment

International Nuclear Information System (INIS)

Jaegers, P.J.; Sanchez, R.G.

1996-01-01

Criticality benchmark data for intermediate energy spectrum systems does not exist. These systems are dominated by scattering and fission events induced by neutrons with energies between 1 eV and 1 MeV. Nuclear data uncertainties have been reported for such systems which can not be resolved without benchmark critical experiments. Intermediate energy spectrum systems have been proposed for the geological disposition of surplus fissile materials. Without the proper benchmarking of the nuclear data in the intermediate energy spectrum, adequate criticality safety margins can not be guaranteed. The Zeus critical experiment now under construction will provide this necessary benchmark data

Sequential measurements of cosmic-ray neutron spectrum and dose rate at sea level in Sendai, Japan

International Nuclear Information System (INIS)

Nakamura, Takashi; Nunomiya, Tomoya; Abe, Shigeru; Terunuma, Kazutaka; Suzuki, Hiroyuki

2005-01-01

The cosmic-ray neutron energy spectrum and dose rate were measured sequentially for two years from April 2001 up to March 2003 by using three neutron detectors, a 3 He-loaded multi-moderator detector (Bonner ball), 12.7 cm diameter by 12.7 cm long NE213 organic liquid scintillator, and high-sensitivity rem (dose equivalent) counter at the Kawauchi campus of Tohoku University in Sendai, Japan of geomagnetic latitude, 29degN, and cutoff rigidity, 10.43 GV. The neutron spectrum has three major peaks, thermal energy peak, evaporation peak around 1 MeV and cascade peak around 100 MeV. The ambient neutron dose equivalent rates measured by the rem counter, and the Bonner ball keep almost constant values of 4.0 and 6.5 (nSv/h), respectively, throughout this time period, after atmospheric pressure correction, and it often decreased about 30% after a large Solar Flare, that is called as the Forbush decrease. The total neutron flux was also obtained by the Bonner ball measurements to be 7.5x10 -3 (ncm -2 ·s -1 ) in average. The altitude variation of neutron flux and dose was also investigated by comparing the measured results with other results measured at Mt. Fuji area and aboard an airplane, where the cutoff rigidities are similar. (author)

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)

Device for measuring neutron-flux distribution density

International Nuclear Information System (INIS)

Rozenbljum, N.D.; Mitelman, M.G.; Kononovich, A.A.; Kirsanov, V.S.; Zagadkin, V.A.

1977-01-01

An arrangement is described for measuring the distribution of neutron flux density over the height of a nuclear reactor core and which may be used for monitoring energy release or for detecting deviations of neutron flux from an optimal level so that subsequent balance can be achieved. It avoids mutual interference of detectors. Full constructional details are given. (UK)

Effects of silicon cross section and neutron spectrum on the radial uniformity in neutron transmutation doping

International Nuclear Information System (INIS)

Kim, Haksung; Ho Pyeon, Cheol; Lim, Jae-Yong; Misawa, Tsuyoshi

2012-01-01

The effects of silicon cross section and neutron spectrum on the radial uniformity of a Si-ingot are examined experimentally with various neutron spectrum conditions. For the cross section effect, the numerical results using silicon single crystal cross section reveal good agreements with experiments within relative difference of 6%, whereas the discrepancy is approximately 20% in free-gas cross section. For the neutron spectrum effect, the radial uniformity in hard neutron spectrum is found to be more flattening than that in soft spectrum. - Highlights: ► The effects of silicon cross section and neutron spectrum on the radial uniformity in NTD were experimentally investigated. ► The numerical results using silicon single crystal cross section reveal good agreements. ► The radial uniformity in hard neutron spectrum was more flat than that in soft spectrum. ► The silicon single crystal cross section and hard neutron spectrum are recommended for numerical analyses and radial uniformity flattening in NTD, respectively.

Dosimetry of mixed gamma - neutron fluxes in the active zone of working reactor and gamma-flux after quenching

International Nuclear Information System (INIS)

Mussaeva, M.A.; Zinov'ev, V.; Ibragimova, E.M.; Muminov, M.I.

2006-01-01

vacancy, varied within 0.57 - 2.8. Besides, pure SiO 2 samples in the Cd - can filled with water were irradiated in the thermal column of operating reactor for 6 hours. Under these conditions the fast neutron flux was estimated as weak as 6·10 10 n/cm 2 s, the fluence was 1.3·10 15 cm -2 . The optical density of band 215 nm was 2.5, while the neutron fluence was ∼30 times less. Thus, the concentration of E ' -centers does not correlate with a neutron fluence. To extract the contribution from gamma-rays into the induced optical absorption in the glass matrix, samples of pure SiO 2 were irradiated by gamma-rays in 4 hours after quenching the reactor at the ionization current of 50 nA during 30 minutes, 12 and 24 hours; next time in 9 hours after the quenching at 40 nA and for 120 hours at 10 nA. In this case the gamma-spectrum did not include 10 MeV line from oxygen due to the short life-time, which prevails in the spectrum of working reactor. Maximal dose of γ-radiation of the quenched reactor was shown to induce the band at 215 nm up to the density of 0.5. When the sample was in contact with water the efficiency of E'-center production was 2 times higher that in dry condition. Thus, the high efficiency of structure defect production in SiO 2 glass owes to the influence of 10 MeV γ-radiation of the working reactor. The work was carried out under the grant F2.1.2 from Center of Science and Technology of Uzbekistan and supported by NATO CBP.EAP.CLG.981765. (author)

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)

Production of a pulseable fission-like neutron flux using a monoenergetic 14 MeV neutron generator and a depleted uranium reflector

Science.gov (United States)

Koltick, D.; McConchie, S.; Sword, E.

2008-04-01

The design and performance of a pulseable neutron source utilizing a D-T neutron generator and a depleted uranium reflector are presented. Approximately half the generator's 14 MeV neutron flux is used to produce a fission-like neutron spectrum similar to 252Cf. For every 14 MeV neutron entering the reflector, more than one fission-like neutron is reflected back across the surface of the reflector. Because delayed neutron production is more than two orders of magnitude below the prompt neutron production, the source takes full advantage of the generator's pulsed mode capability. Applications include all elemental characterization systems using neutron-induced gamma-ray spectroscopy. The source simultaneously emits 14 MeV neutrons optimal to excite fast neutron-induced gamma-ray signals, such as from carbon and oxygen, and fission-like neutrons optimal to induce neutron capture gamma-ray signals, such as from hydrogen, nitrogen, and chlorine. Experiments were performed, which compare well to Monte Carlo simulations, showing that the uranium reflector enhances capture signals by up to a factor of 15 compared to the absence of a reflector.

Measuring planetary neutron albedo fluxes by remote gamma-ray sensing

International Nuclear Information System (INIS)

Haines, E.L.; Metzger, A.E.

1984-01-01

A remote-sensing γ-ray spectrometer (GRS) is capable of measuring planetary surface composition through the detection of characteristic gamma rays. In addition, the planetary neutron leakage flux may be detected by means of a thin neutron absorber surrounding the γ-ray detector which converts the neutron flux into a γ-ray flux having a unique energy signature. The γ rays representing the neutron flux are observed against interference consisting of cosmic γ rays, planetary continuum and line emission, and a variety of gamma rays arising from cosmic-ray particle interactions with the γ-ray spectrometer and spacecraft (SC). In this paper the amplitudes of planetary and non-planetary neutron fluxes are assessed and their impact on the sensitivity of measurement is calculated for a lunar orbiter mission and a comet nucleus rendezvous mission. For a 100 h observation period from an altitude of 100 km, a GRS on a lunar orbiter can detect a thermal neutron albedo flux as low as 0.002 cm -2 s -1 and measure the expected flux of approx.=0.6 cm -2 s -1 with an uncertainty of 0.001 cm -2 s -1 . A GRS rendezvousing with a comet at a distance equal to the radius of the comet's nucleus, again for a 100 h observation time, should detect a thermal neutron albedo flux at a level of 0.006 cm -2 s -1 and measure the expected flux of approx.=0.4 cm -2 s -1 with an uncertainty of 0.004 cm -2 s -1 . Mapping the planetary neutron flux jointly with the direct detection of H will not only provide a more accurate model for translating observed γ-ray fluxes into concentrations but will also extend the effective sampling depth and should provide a capability for simple stratigraphic modeling of hydrogen. (orig.)

Neutron flux control systems validation

International Nuclear Information System (INIS)

Hascik, R.

2003-01-01

In nuclear installations main requirement is to obtain corresponding nuclear safety in all operation conditions. From the nuclear safety point of view is commissioning and start-up after reactor refuelling appropriate period for safety systems verification. In this paper, methodology, performance and results of neutron flux measurements systems validation is presented. Standard neutron flux measuring chains incorporated into the reactor protection and control system are used. Standard neutron flux measuring chain contains detector, preamplifier, wiring to data acquisition unit, data acquisition unit, wiring to display at control room and display at control room. During reactor outage only data acquisition unit and wiring and displaying at reactor control room is verified. It is impossible to verify detector, preamplifier and wiring to data acquisition recording unit during reactor refuelling according to low power. Adjustment and accurate functionality of these chains is confirmed by start-up rate (SUR) measurement during start-up tests after refuelling of the reactors. This measurement has direct impact to nuclear safety and increase operational nuclear safety level. Briefly description of each measuring system is given. Results are illustrated on measurements performed at Bohunice NPP during reactor start-up tests. Main failures and their elimination are described (Authors)

Measured and Predicted Variations in Fast Neutron Spectrum when Penetrating Laminated Fe-D2O

International Nuclear Information System (INIS)

Aalto, E.; Sandlin, R.; Fraeki, R.

1965-09-01

Variations of the fast neutron spectrum in thin regions of alternating Fe and D O have been studied using threshold detectors (ln(n, n' ), S(n, p), Al(n, α)). The results have been compared to those calculated by two shielding codes (NRN and RASH D) of multigroup removal-diffusion type. The absolute fast spectrum calculated in our rather complicated configurations was found to agree with measurements within the same accuracy (a factor of two) as did the thermal flux. The calculated spectrum is slightly harder than the measured one, but the detailed variations (covering the range 1:5) in the form of the spectrum when penetrating Fe agree with observations to within 15-20 %. In and Al activities were found to be proportional to the integrated flux over 1 MeV throughout the whole configuration, while S showed the least proportionality

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

Measurement and calculation of spatial and energetic neutron flux in the IEA-R1 reactor core

International Nuclear Information System (INIS)

Bittelli, U.D.

1988-01-01

This work presents spatial and energetic flux distribution measured in the IEA-R1 reactor core. The thermal neutron flux was measured by gold activation foils (bare and covered with cadmium) in the fuel element number 108 (reaction: 197 Au(n,γ) 198 Au) at 451W overall reactor power. The fast neutron flux was measured by indium activation foils (reaction: 115 In(n,n') 115m In) in the fuel elements number 94 at 4510W overall reactor power. The neutron energy spectrum was adjusted by SAND II code with the data produced by the irradiation of seven activation detectors in the fuel element number 94 at 4510 W overall reactor power. The following reactions were used: 58 Fe(n,γ) 59 Fe, 232 Th(n,γ) 233 Th, 197 Au(n,γ) 198 Au, 59 Co(n,γ) 60 Co, 54 Fe(n,p) 54 Mn, 24 Mg(n,p) 24 Na, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc and 115 In(n,n') 115m In. The experimental results compared to those obtained by CITATION (spatial distribution flux) and HAMMER (energetic distribution flux) code, showed good agreement. The results presented in this work are a good contribution for a better knowledge of spatial and energetic neutron flux distribution in the IEA-R1 reactor core, besides that the experimental procedure is easily applicable to another situations. (autor) [pt

A study on the utilization of hyper-thermal neutrons for neutron capture therapy

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

1993-01-01

The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwellian distribution of a higher temperature than the room temperature of 300 K, was studied in order to improve the thermal neutron flux distribution at the deeper part in a living body for neutron capture therapy. Simulation calculations were carried out using MCNP-V3 in order to confirm the characteristics of hyper-thermal neutrons, i.e., (1) depth dependence of neutron energy spectrum, and (2) depth distribution of the reaction rate in a water phantom for materials with 1/v neutron absorption. It is confirmed that the hyper-thermal neutron irradiation can improve the thermal neutron flux distribution in the deeper and wider area in a living body compared with the thermal neutron irradiation. Practically, by the incidence of the hyper-thermal neutrons with a 3000 K Maxwellian distribution, the thermal neutron flux at 5 cm depth can be given about four times larger than by the incidence of the thermal neutrons of 300 K. (author)

Neutron activation analysis: Modelling studies to improve the neutron flux of Americium-Beryllium source

Energy Technology Data Exchange (ETDEWEB)

Didi, Abdessamad; Dadouch, Ahmed; Tajmouati, Jaouad; Bekkouri, Hassane [Advanced Technology and Integration System, Dept. of Physics, Faculty of Science Dhar Mehraz, University Sidi Mohamed Ben Abdellah, Fez (Morocco); Jai, Otman [Laboratory of Radiation and Nuclear Systems, Dept. of Physics, Faculty of Sciences, Tetouan (Morocco)

2017-06-15

Americium–beryllium (Am-Be; n, γ) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

Neutron spectrum for neutron capture therapy in boron

International Nuclear Information System (INIS)

Medina C, D.; Soto B, T. G.; Baltazar R, A.; Vega C, H. R.

2016-10-01

Glioblastoma multiforme is the most common and aggressive of brain tumors and is difficult to treat by surgery, chemotherapy or conventional radiation therapy. One treatment alternative is the Neutron Capture Therapy in Boron, which requires a beam modulated in neutron energy and a drug with 10 B able to be fixed in the tumor. When the patients head is exposed to the neutron beam, they are captured by the 10 B and produce a nucleus of 7 Li and an alpha particle whose energy is deposited in the cancer cells causing it to be destroyed without damaging the normal tissue. One of the problems associated with this therapy is to have an epithermal neutrons flux of the order of 10 9 n/cm 2 -sec, whereby irradiation channels of a nuclear research reactor are used. In this work using Monte Carlo methods, the neutron spectra obtained in the radial irradiation channel of the TRIGA Mark III reactor are calculated when inserting filters whose position and thickness have been modified. From the arrangements studied, we found that the Fe-Cd-Al-Cd polyethylene filter yielded a ratio between thermal and epithermal neutron fluxes of 0.006 that exceeded the recommended value (<0.05), and the dose due to the capture gamma rays is lower than the dose obtained with the other arrangements studied. (Author)

Study of the environmental neutron spectrum at Zacatecas city

International Nuclear Information System (INIS)

Vega C, H.R.

2003-01-01

The environmental neutron spectrum has been measured at Zacatecas City in Mexico. Neutron spectrum was unfolded from count rates obtained with a multisphere neutron spectrometer with a Li I(Eu) scintillator. With the spectrum information the ambient dose equivalent and the isotropic effective dose were calculated. A model based upon the geomagnetic latitude and the altitude above sea level, that allows to estimate the neutron fluence rate is proposed, the model results are compared with total neutron fluences measured at several locations worldwide. Environmental neutron spectrum shows peaks at 1 and 100 MeV as well as a relevant amount of low energy neutrons. The neutron fluence rate was 65 ± 3 cm -2 -h -1 , producing 13.7 ± 0.6 n Sv-h -1 due to ambient dose equivalent rate and an isotropic effective dose rate of 14.1 ± 0.6 n Sv-h -1 . Neutron fluence rates predicted with the model are in agreement with those reported in the literature. (Author)

Study of the environmental neutron spectrum at Zacatecas city

Energy Technology Data Exchange (ETDEWEB)

Vega C, H.R. [Universidad Autonoma de Zacatecas, Cuerpo Academico de Radiobiologia, A.P. 336, 98000 Zacatecas (Mexico)

2003-07-01

The environmental neutron spectrum has been measured at Zacatecas City in Mexico. Neutron spectrum was unfolded from count rates obtained with a multisphere neutron spectrometer with a Li I(Eu) scintillator. With the spectrum information the ambient dose equivalent and the isotropic effective dose were calculated. A model based upon the geomagnetic latitude and the altitude above sea level, that allows to estimate the neutron fluence rate is proposed, the model results are compared with total neutron fluences measured at several locations worldwide. Environmental neutron spectrum shows peaks at 1 and 100 MeV as well as a relevant amount of low energy neutrons. The neutron fluence rate was 65 {+-} 3 cm{sup -2}-h{sup -1}, producing 13.7 {+-} 0.6 n Sv-h{sup -1} due to ambient dose equivalent rate and an isotropic effective dose rate of 14.1 {+-} 0.6 n Sv-h{sup -1}. Neutron fluence rates predicted with the model are in agreement with those reported in the literature. (Author)

Measurement of 235U fission spectrum-averaged cross sections and neutron spectrum adjusted with the activation data

International Nuclear Information System (INIS)

Kobayashi, Katsuhei; Kobayashi, Tooru

1992-01-01

The 235 U fission spectrum-averaged cross sections for 13 threshold reactions were measured with the fission plate (27 cm in diameter and 1.1 cm thick) at the heavy water thermal neutron facility of the Kyoto University Reactor. The Monte Carlo code MCNP was applied to check the deviation from the 235 U fission neutron spectrum due to the room-scattered neutrons, and it was found that the resultant spectrum was close to that of 235 U fission neutrons. Supplementally, the relations to derive the absorbed dose rates with the fission plate were also given using the calculated neutron spectra and the neutron Kerma factors. Finally, the present values of the fission spectrum-averaged cross sections were employed to adjust the 235 U fission neutron spectrum with the NEUPAC code. The adjusted spectrum showed a good agreement with the Watt-type fission neutron spectrum. (author)

Measurement of the neutron spectrum of the Big Ten critical assembly by lithium-6 spectrometry

International Nuclear Information System (INIS)

De Leeuw-Gierts, G.; De Leeuw, S.; Hansen, G.E.; Helmick, H.H.

1979-01-01

The central neutron-flux spectrum of the Los Alamos Scientific Laboratory's critical assembly, Big Ten, was measured with a 6 Li spectrometer and techniques developed at the Centre d'Etude de L'Energie Nucleaire, Mol, as part of an experimental program to establish the characteristics of Big Ten

Measurement of the neutron spectrum of the Big Ten critical assembly by lithium-6 spectrometry

International Nuclear Information System (INIS)

Leeuw-Gierts, G. de; Leeuw, S. de

1980-01-01

The central neutron-flux spectrum of the Los Alamos Scientific Laboratory's critical assembly, Big Ten, was measured with a 6 Li spectrometer and techniques developed at the Centre d'Etude de l'Energie Nucleaire, Mol, as part of an experimental program to establish the characteristics of Big Ten

Measurement and simulation of thermal neutron flux distribution in the RTP core

Science.gov (United States)

Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

2018-01-01

The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

Using thermalizers in measuring 'Ukryttia' object's FCM neutron fluxes

CERN Document Server

Krasnyanskaya, O G; Odinokin, G I; Pavlovich, V N

2003-01-01

The results of research of a thermalizer (heater) width influence on neutron thermalization efficiency during FCM neutron flux measuring in the 'Ukryttia' are described. The calculations of neutron flux densities were performed by the Monte-Carlo method with the help of computer code MCNP-4C for FCM different models.Three possible installations of detectors were considered: on FCM surface,inside the FCM, and inside the concrete under the FCM layer. It was shown,that in order to increase the sensitivity of neutron detectors in intermediate and fast neutrons field,and consequently, to decrease the dependence of the readings of spectral distribution of neutron flux,it is necessary to position the detector inside the so-called thermalizer or heater. The most reasonable application of thick 'heaters' is the situation, when the detector is placed on FCM surface.

Neutron Spectrum Parameters In Inner Irradiation Channel Of The Nigeria Research Reactor-1 (NIRR-1) For Use In Absolute And KO-NAA Methods

International Nuclear Information System (INIS)

Jonah, S.A; Balogun, G.I; Mayaki, M.C.

2004-01-01

In Nigeria, the first Nuclear Reactor achieved critically on February 03, 2004 at about 11:35 GMT and has been commissioned or training and research. It is a Miniature Neutron Source Reactor (MNSR), code-named Nigeria Research Reactor-1 (NIRR-1). NIRR-1 has a tan-in-pool structural configuration and a nominal thermal power rating of 30 Kw. With a built-in clean old core excess reactivity of 3.77 mk determined during the on-site zero and critically experimental, the reactor can operate for a n.cm-2 .s-1 in the inner irradiation channels). Under these conditions, the reactor can operate with the same fuel loading for over ten years with a burn-up of <1%. A detailed description of operating characteristics for NIRR-1, measured during the on-site zero-power and criticality experiments has been given elsewhere. In order to extend its utilization to include absolute and ko-NAA methods, the neutron spectrum parameters in the irradiation channels: power and critically experiments has been given elsewhere. In order to extend it's the irradiation channels: thermal-to-epithermal flux ration, F; and epithermal flux shape factor, a in both the inner and outer irradiation channels must be determined experimentally. In this work, we have developed and experimental procedure for monitoring the neutron spectrum parameters in an inner irradiation channel based on irradiation and gamma-ray counting of detector foils via (n,y), (n,p) and (n,a) dosimetry reactions. Results obtained indicate that a thermal neutron flux of (5.14+-0.02) x 1011 n/c m2.s determined by foil activation method in the inner irradiation channel, B2, at a power level of 15.5 kw corresponds to the flux indicators on the control console and the micro-computer control system respectively. Other parameters of the neutron spectrum determined for inner irradiation channel B2, are: a -0.0502+0.003; 18.92+-0.14; F = 3.87=0.23. The method was validated through the comparison of our result with published neutron spectrum

Evaluation of the Neutron Detector Response for Cosmic Ray Energy Spectrum by Monte Carlo Transport Simulation

International Nuclear Information System (INIS)

Pazianotto, Mauricio T.; Carlson, Brett V.; Federico, Claudio A.; Gonzalez, Odair L.

2011-01-01

Neutrons generated by the interaction of cosmic rays with the atmosphere make an important contribution to the dose accumulated in electronic circuits and aircraft crew members at flight altitude. High-energy neutrons are produced in spallation reactions and intranuclear cascade processes by primary cosmic-ray particle interactions with atoms in the atmosphere. These neutrons can produce secondary neutrons and also undergo a moderation process due to atmosphere interactions, resulting in a wider energy spectrum, ranging from thermal energies (0.025 eV) to energies of several hundreds of MeV. The Long-Counter (LC) detector is a widely used neutron detector designed to measure the directional flux of neutrons with about constant response over a wide energy range (thermal to 20 MeV). ). Its calibration process and the determination of its energy response for the wide-energy of cosmic ray induced neutron spectrum is a very difficult process due to the lack of installations with these capabilities. The goal of this study is to assess the behavior of the response of a Long Counter using the Monte Carlo (MC) computational code MCNPX (Monte Carlo N-Particle eXtended). The dependence of the Long Counter response on the angle of incidence, as well as on the neutron energy, will be carefully investigated, compared with the experimental data previously obtained with 241 Am-Be and 252 Cf neutron sources and extended to the neutron spectrum produced by cosmic rays. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Neutron flux and gamma dose measurement in the BNCT irradiation facility at the TRIGA reactor of the University of Pavia

Science.gov (United States)

Bortolussi, S.; Protti, N.; Ferrari, M.; Postuma, I.; Fatemi, S.; Prata, M.; Ballarini, F.; Carante, M. P.; Farias, R.; González, S. J.; Marrale, M.; Gallo, S.; Bartolotta, A.; Iacoviello, G.; Nigg, D.; Altieri, S.

2018-01-01

University of Pavia is equipped with a TRIGA Mark II research nuclear reactor, operating at a maximum steady state power of 250 kW. It has been used for many years to support Boron Neutron Capture Therapy (BNCT) research. An irradiation facility was constructed inside the thermal column of the reactor to produce a sufficient thermal neutron flux with low epithermal and fast neutron components, and low gamma dose. In this irradiation position, the liver of two patients affected by hepatic metastases from colon carcinoma were irradiated after borated drug administration. The facility is currently used for cell cultures and small animal irradiation. Measurements campaigns have been carried out, aimed at characterizing the neutron spectrum and the gamma dose component. The neutron spectrum has been measured by means of multifoil neutron activation spectrometry and a least squares unfolding algorithm; gamma dose was measured using alanine dosimeters. Results show that in a reference position the thermal neutron flux is (1.20 ± 0.03) ×1010 cm-2 s-1 when the reactor is working at the maximum power of 250 kW, with the epithermal and fast components, respectively, 2 and 3 orders of magnitude lower than the thermal component. The ratio of the gamma dose with respect to the thermal neutron fluence is 1.2 ×10-13 Gy/(n/cm2).

Prompt neutron spectrum of the spontaneous fission of californium-252

International Nuclear Information System (INIS)

Zamyatnin, Yu.S.; Kroshkin, N.I.; Korostylev, V.A.; Nefedov, V.N.; Ryazanov, D.K.; Starostov, B.I.; Semenov, A.F.

1976-01-01

The californium-252 spontaneous fission neutron spectrum was measured in the energy range of 0.01 to 10 MeV by the time-of-flight technique using various neutron detectors. The measurements of 252 Cf neutron spectrum at energies of 0.01 to 5 MeV were performed as a function of fission fragment kinetic energy. The mean neutron spectrum energy in the range of 0.7 to 10 MeV was found from the results of measurements. The irregularity in the 252 Cf neutron spectrum in the neutron energy range of less than 0.7 MeV compared to theoretical values is discussed. The mechanism of 252 Cf neutron emission is also discussed on the basis of neutron yield angle measurements. 12 references

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

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.

Calculation of the thermal neutron flux depression in the loop VISA-1

International Nuclear Information System (INIS)

Martinc, R.

1961-01-01

Among other applications, the VISA-1 loop is to be used for thermal load testing of materials. For this type of testing one should know the maximum power generated in the loop. This power is determined from the maximum thermal neutron flux in the VK-5 channel and mean flux depression in the fissile component of the loop. Thermal neutron flux depression is caused by neutron absorption in the components of the loop, shape of the components and neutron leaking through gaps as well as properties of the surrounding medium of the core. All these parameters were taken into account for calculating the depression of thermal neutron flux in the VISA-1 loop. Two group diffusion theory was used. Fast neutron from the fission in the loop and slowed down were taken into account. Depression of the thermal neutron flux is expressed by depression factor which represents the ratio of the mean thermal neutron flux in the fissile loop component and the thermal neutron flux in the VK-5 without the loop. Calculation error was estimated and it was recommended to determine the depression factor experimentally as well [sr

Effects of silicon cross section and neutron spectrum on the radial uniformity in neutron transmutation doping.

Science.gov (United States)

Kim, Haksung; Ho Pyeon, Cheol; Lim, Jae-Yong; Misawa, Tsuyoshi

2012-01-01

The effects of silicon cross section and neutron spectrum on the radial uniformity of a Si-ingot are examined experimentally with various neutron spectrum conditions. For the cross section effect, the numerical results using silicon single crystal cross section reveal good agreements with experiments within relative difference of 6%, whereas the discrepancy is approximately 20% in free-gas cross section. For the neutron spectrum effect, the radial uniformity in hard neutron spectrum is found to be more flattening than that in soft spectrum. Copyright © 2011 Elsevier Ltd. All rights reserved.

Measured and Predicted Variations in Fast Neutron Spectrum when Penetrating Laminated Fe-D{sub 2}O

Energy Technology Data Exchange (ETDEWEB)

Aalto, E; Sandlin, R; Fraeki, R

1965-09-15

Variations of the fast neutron spectrum in thin regions of alternating Fe and D{sub O} have been studied using threshold detectors (ln(n, n' ), S(n, p), Al(n, {alpha})). The results have been compared to those calculated by two shielding codes (NRN and RASH D) of multigroup removal-diffusion type. The absolute fast spectrum calculated in our rather complicated configurations was found to agree with measurements within the same accuracy (a factor of two) as did the thermal flux. The calculated spectrum is slightly harder than the measured one, but the detailed variations (covering the range 1:5) in the form of the spectrum when penetrating Fe agree with observations to within 15-20 %. In and Al activities were found to be proportional to the integrated flux over 1 MeV throughout the whole configuration, while S showed the least proportionality.

Use of CITATION code for flux calculation in neutron activation analysis with voluminous sample using an Am-Be source

International Nuclear Information System (INIS)

Khelifi, R.; Idiri, Z.; Bode, P.

2002-01-01

The CITATION code based on neutron diffusion theory was used for flux calculations inside voluminous samples in prompt gamma activation analysis with an isotopic neutron source (Am-Be). The code uses specific parameters related to the energy spectrum source and irradiation system materials (shielding, reflector). The flux distribution (thermal and fast) was calculated in the three-dimensional geometry for the system: air, polyethylene and water cuboidal sample (50x50x50 cm). Thermal flux was calculated in a series of points inside the sample. The results agreed reasonably well with observed values. The maximum thermal flux was observed at a distance of 3.2 cm while CITATION gave 3.7 cm. Beyond a depth of 7.2 cm, the thermal flux to fast flux ratio increases up to twice and allows us to optimise the detection system position in the scope of in-situ PGAA

A neutron spectrum unfolding computer code based on artificial neural networks

Science.gov (United States)

Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

2014-02-01

The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding in

A theoretical model for predicting neutron fluxes for cyclic Neutron ...

African Journals Online (AJOL)

A theoretical model has been developed for prediction of thermal neutron fluxes required for cyclic irradiations of a sample to obtain the same activity previously used for the detection of any radionuclide of interest. The model is suitable for radiotracer production or for long-lived neutron activation products where the ...

Neutron flux monitor

International Nuclear Information System (INIS)

Seki, Eiji; Tai, Ichiro.

1984-01-01

Purpose: To maintain the measuring accuracy and the reponse time within an allowable range in accordance with the change of neutron fluxes in a nuclear reactor pressure vessel. Constitution: Neutron fluxes within a nuclear reactor pressure vessel are detected by detectors, converted into pulse signals and amplified in a range switching amplifier. The amplified signals are further converted through an A/D converter and digital signals from the converter are subjected to a square operation in an square operation circuit. The output from the circuit is inputted into an integration circuit to selectively accumulate the constant of 1/2n, 1 - 1/2n (n is a positive integer) respectively for two continuing signals to perform weighing. Then, the addition is carried out to calculate the integrated value and the addition number is changed by the chane in the number n to vary the integrating time. The integrated value is inputted into a control circuit to control the value of n so that the fluctuation and the calculation time for the integrated value are within a predetermined range and, at the same time, the gain of the range switching amplifier is controlled. (Seki, T.)

Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

Science.gov (United States)

Ali, Nur Syazwani Mohd; Hamzah, Khaidzir; Mohamad Idris, Faridah; Hairie Rabir, Mohamad

2018-01-01

The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

Spectrum and H(10) of secondary neutrons around Linacs

International Nuclear Information System (INIS)

Ortiz H, A.; Hernandez A, B.; Vega C, H. R.; Hernandez D, V. M.; Rivera M, T.

2009-10-01

Neutron spectrum and ambient dose equivalent has been measured around two 10 MV linear accelerators. Accelerators are Siemens, one is a Mevatron model while another is the Primus. Main differences between those models are the beam collimator and the vault room. Here, Bonner sphere spectrometer with a passive thermal neutron detector has been utilized to measure the neutron spectrum inside the vault. Using an active detector the neutron spectrum was measured by the vaults door of both accelerators. With a neutron area monitor the dose equivalent was measured by the doors. Neutron strength, total fluence rate and ambient dose equivalent were compared, from this was found that shielding conditions are better in the Primus model. (Author)

Neutron spectrum unfolding using neural networks

International Nuclear Information System (INIS)

Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E.

2004-01-01

An artificial neural network has been designed to obtain the neutron spectra from the Bonner spheres spectrometer's count rates. The neural network was trained using a large set of neutron spectra compiled by the International Atomic Energy Agency. These include spectra from iso- topic neutron sources, reference and operational neutron spectra obtained from accelerators and nuclear reactors. The spectra were transformed from lethargy to energy distribution and were re-binned to 31 energy groups using the MCNP 4C code. Re-binned spectra and UTA4 matrix were used to calculate the expected count rates in Bonner spheres spectrometer. These count rates were used as input and correspondent spectrum was used as output during neural network training. The network has 7 input nodes, 56 neurons as hidden layer and 31 neurons in the output layer. After training the network was tested with the Bonner spheres count rates produced by twelve neutron spectra. The network allows unfolding the neutron spectrum from count rates measured with Bonner spheres. Good results are obtained when testing count rates belong to neutron spectra used during training, acceptable results are obtained for count rates obtained from actual neutron fields; however the network fails when count rates belong to monoenergetic neutron sources. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-10-15

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

Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III

International Nuclear Information System (INIS)

Parra M, M. A.

2014-01-01

This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au 197 , Ni 58 , In 115 , Mg 24 , Al 27 , Fe 58 , Co 59 and Cu 63 , they were selected to cover the full range the energies (10 -10 to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or research projects. (Author)

Neutron flux enhancement in the NRAD reactor

International Nuclear Information System (INIS)

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

1988-01-01

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

Average cross section measurements in U-235 fission neutron spectrum for some threshold reactions

International Nuclear Information System (INIS)

Maidana, N.L.

1993-01-01

The average cross section in the 235 U fission spectrum has been measured by the activation technique, for the following thresholds reactions: 115 In(n,n') 115m In, 232 Th(n,f) P.F., 46 , 47 , 48 Ti(n,p) 46,47 , 48 Sc, 55 Mn(n,2 n) 54 Mn, 51 V(n,α) 48 Sc, 90 Zr(n,2 n) 89 Zr, 93 Nb(n,2 n) 92m Nb, 58 Ni(n,2 n) 57 Ni, 24 Mg(n,p) 24 Na, 56 Fe(n,p) 56 Mn, 59 Co(n,α) 56 Mn and 63 Cu(n,α) 60 Co. The activation foils were irradiated close (∼ 4 mm) to the core of the IEA-R1 research reactor in the IPEN-CNEN/SP. The reactor was operated at 2 MW yielding a fast neutron flux around 5 x 10 12 n.cm -2 . s -1 . The neutron flux density was monitored by activation reactions with well known averaged cross sections and with effective thresholds above 1 MeV. The foil activities were measured in a calibrated HPGe spectrometer. The neutron spectrum has been calculated using the SAIPS unfolding system applied to the activation data. A detailed error analysis was performed using the covariance matrix methodology. The results were compared with those from other authors. (author)

Neutron activation analysis: Modelling studies to improve the neutron flux of Americium–Beryllium source

Directory of Open Access Journals (Sweden)

Abdessamad Didi

2017-06-01

Full Text Available Americium–beryllium (Am-Be; n, γ is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci, yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

Numeric modeling of HfO2 neutron flux sensor parameters during sensor burnup in the RBMK-1500 reactor

International Nuclear Information System (INIS)

Jurkevicius, A.; Remeikis, V.

2001-01-01

The isotopic composition of hafnium in the radial neutron flux sensor of the RBMK-1500 reactor, the rates of the neutron absorption on Hf isotopes and the neutron spectrum in the sensor were numerically modeled. The sequence SAS2 (Shielding Analysis Sequence) from the package SCALE 4.3 was used for calculations. It has been obtained that the main neutron absorber 167 Er isotope practically burns up completely at the 18 MW d/kgU burnup depth, and at that time the capture rate of thermal neutrons in erbium decreases ten-fold. The average neutron flux density was calculated 7.6*10 13 neutrons. Cm -2 S -1 in the RBMK-1500 reactor grating, when the nuclear fuel enriched with 235 U by 2.4% and with Er by 0.4% is used in a fuel assembly. When the sensor burnup reaches 28 MW d/kgU, the neutron absorption rate of 178 Hf exceeds the rate of 177 Hf. The overall neutron absorption rate in hafnium decreases 2.53 times due to the sensor burnup to 56 MW d/kgU. The corrective factors ξ d (I) at different integral flux I of the sensors were calculated. The obtained dependence ξ d (I) calculated numerically was compared to the experimental one determined by processing repeated calibration results of Hf sensors in RBMK-1500 reactors, as well as compared to the theoretical one currently used in the Ignalina NPP special mathematical algorithms. (author)

Neutron flux measurement in the central channel (XC-1) of TRIGA 14 MW LEU core

International Nuclear Information System (INIS)

BARBOS, D.; BUSUIOC, P.; ROTH, Cs.; PAUNOIU, C.

2008-01-01

The TRIGA 14 MW reactor, operated by Institute for Nuclear Research Pitesti, Romania, is a pool type reactor, and has a rectangular shape which holds fuel bundles and is surrounded with beryllium reflectors. Each fuel bundle is composed of 25 nuclear fuel rods. The TRIGA 14 MW reactor was commissioned 28 years ago with HEU fuel rods. The conversion was gradually achieved, starting in February 1992 and completed in March 2006. The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel, replaced by LEU fuel, accompanied by a large set of theoretical evaluation and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC in the XC-1 water 1 water-filled channel were performed using multi multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral neutron flux value for LEU core is greater of 13% than for the standard HEU core. Also thermal neutron flux value for converted LEU core is smaller by 0.38% than for the standard HEU core. These differences appear because the foil activation detectors have been irradiated using a pneumatic rabbit having a diameter of 32 mm, whereas foil irradiations in standard HEU core has been performed with a pneumatic rabbit having a diameter of 14 mm, and therefore the neutron spectra in LEU core is less thermalized and the weight of fast neutron is greater

Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum

Energy Technology Data Exchange (ETDEWEB)

Koyanagi, Takaaki, E-mail: koyanagit@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kumar, N.A.P. Kiran [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hwang, Taehyun [Tohoku University, Sendai, 980-8579 (Japan); Garrison, Lauren M.; Hu, Xunxiang [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2017-07-15

Microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ∼90–800 °C to 0.03–4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ∼90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ∼1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- and Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.

Least-squares adjustment of a 'known' neutron spectrum: The importance of the covariance matrix of the input spectrum

International Nuclear Information System (INIS)

Mannhart, W.

1986-01-01

Based on the responses of 25 different neutron activation detectors, the neutron spectrum of Cf-252 hs been adjusted with least-squares methods. For a fixed input neutron spectrum, the covariance matrix of this spectrum has been systematically varied to investigate the influence of this matrix on the final result. The investigation showed that the adjusted neutron spectrum is rather sensitive to the structure of the covariance matrix for the input spectrum. (author)

Neutron Flux Characterization of Irradiation Holes for Irradiation Test at HANARO

Directory of Open Access Journals (Sweden)

Yang Seong Woo

2016-01-01

Full Text Available The High flux Advanced Neutron Application ReactOr (HANARO is a unique research reactor in the Republic of Korea, and has been used for irradiation testing since 1998. To conduct irradiation tests for nuclear materials, the irradiation holes of CT and OR5 have been used due to a high fast-neutron flux. Because the neutron flux must be accurately calculated to evaluate the neutron fluence of irradiated material, it was conducted using MCNP. The neutron flux was measured using fluence monitor wires to verify the calculated result. Some evaluations have been conducted, however, more than 20% errors have frequently occurred at the OR irradiation hole, while a good agreement between the calculated and measured data was shown at the CT irradiation hole.

The new high flux neutron source FRM-2 in Munich

International Nuclear Information System (INIS)

Roegler, H.J.; Wierheim, G.

2002-01-01

Quite some years ago in 1974 to be exact, the first consideration on a new neutron source started at the technical university of Munich (Germany). 27 years later the new high flux neutron source (FRM-2) was read for hot operation, now delayed by a refused approval for its third partial license by the federal government of Germany despite a wide support from the scientific community. FRM-2 is a tank-type research reactor cooled by water, moderated by heavy water and whose thermal power was limited to 20 MW maximum. The extreme compact core together with the applied inverse flux principle led to a neutron flux design value of 8.10 18 n/m 2 .s at the reflector peak. 10 beam tubes will allow an optimized use of the high neutron flux. A hot neutron source with graphite at about 2200 Celsius degrees and a cold neutron source with liquid D 2 at about 25 K will provide shifted energy spectra. The utilization of FRM-2 is many-fold: neutronography and tomography, medical irradiation, radio-nuclide production, doping of pure silicon, neutron activation analysis. (A.C.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

Fusion neutron yield and flux calculation on HT-7 superconducting tokamak

International Nuclear Information System (INIS)

Fu Yanzhang; Zhu Yubao; Chen Juequan

2006-01-01

Neutron yield and flux have been numerically estimated on HT-7 tokamak. The total fusion neutron yield and neutron flux distribution on different positions and azimuth angles of the device are presented. Analyses on the errors induced by ion temperature and density distribution factors are given in detail. The results of the calculations provide a useful database for neutron diagnostics and neutron radiation protection. (authors)

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

The measurements of thermal neutron flux distribution in a paraffin

Indian Academy of Sciences (India)

The term `thermal flux' implies a Maxwellian distribution of velocity and energy corresponding to the most probable velocity of 2200 ms-1 at 293.4 K. In order to measure the thermal neutron flux density, the foil activation method was used. Thermal neutron flux determination in paraffin phantom by counting the emitted rays of ...

Fast neutron flux in heavy water reactors; Flux de neutrons rapides dans les piles a eau lourde

Energy Technology Data Exchange (ETDEWEB)

Brisbois, J; Katz, S [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France)

1966-07-01

The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author) [French] La possibilite de calculer le flux de neutrons rapides dans un reseau d'uranium naturel a eau lourde par superposition des apports des divers barreaux, a ete verifiee en utilisant un code Monte-Carlo monodimensionel. Les resultats obtenus concordent avec des mesures experimentales effectuees dans le coeur et reacteur de la pile Aquilon. (auteurs)

Effect of time dependence of neutron flux on the plutonium 238 production

International Nuclear Information System (INIS)

Rudik, A.P.

1975-01-01

An analytical treatment is given of the dependence of a plutonium-238 yield when irradiating neptunium-237 on the time variation of the neutron flux. This dependence is governed by the two physical factors: the competition between the beta decay and the neptunium-238 depletion, and the differences in the depletion rate of neptunium-237, neptunium-238 and plutonium-238 with thermal and resonance neutrons. The role of the neptunium-238 nonprompt decay has been studied and the order of the effect determined, and it has been established that it is advantageous to irradiate neptunium-237 in the lowest central neutron flux possible for thermal and resonance neutron reactors. A perturbation theory has been developed to determine the role of the time variation of the thermal neutron flux and it has been established that, the integral neutron flux being the same, it is advantageous to irradiate the material in a drooping rather than a constant flux of neutrons. The simplest example of great time variations of a neutron flux has been considered and it has been established that in this particular case the conclusions made in the framework of the perturbation theory find a qualitative support. Convenient formulas are given for determining the number of neutrons used to produce plutonium-238

Notes on neutron flux measurement; Notas sobre medida de flujos neutronicos

Energy Technology Data Exchange (ETDEWEB)

Alcala Ruiz, F

1984-07-01

The main purpose of this work is to get an useful guide to carry out topical neutron flux measurements. Although the foil activation technique is used in the majority of the cases, other techniques, such as those based on fission chambers and self-powered neutron detectors, are also shown. Special interest is given to the description and application of corrections on the measurement of relative and absolute induced activities by several types of detectors (scintillators, G-M and gas proportional counters). The thermal arid epithermal neutron fluxes, as determined in this work, are conventional or effective (West cots fluxes), which are extensively used by the reactor experimentalists; however, we also give some expressions where they are related to the integrated neutron fluxes, which are used in neutron calculations. (Author) 16 refs.

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

Energy Technology Data Exchange (ETDEWEB)

Aredes, Vitor Ottoni Garcia

2014-07-01

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

Albedo analytical method for multi-scattered neutron flux calculation in cavity

International Nuclear Information System (INIS)

Shin, Kazuo; Selvi, S.; Hyodo, Tomonori

1986-01-01

A simple formula which describes multi-scattered neutron flux in a spherical cavity was derived based on the albedo concept. The formura treats a neutron source which has an arbitrary energy-angle distribution and is placed at any point in the cavity. The derived formula was applied to the estimation of neutron fluxes in two cavities, i.e. a spherical concrete cell with a 14-MeV neutron source at the center and the ''YAYOI'' reactor cavity with a pencil beam of reactor neutrons. The results of the analytical formula agreed very well with the reference data in the both problems. It was concluded that the formula is applicable to estimate the neutron fluxes in a spherical cell except for special cases that tangential source neutrons are incident to the cavity wall. (author)

Determination of Neutron Flux Parameter f and α and k0 Factor in Irradiation Facility of RSG GA Siwabessy reactor

International Nuclear Information System (INIS)

Amir Hamzah

2004-01-01

Determination of neutron flux thermal to epithermal ratio f and parameter α and k 0 factor has been done in irradiation facility of RSG G.A. Siwabessy reactor. Those parameters are needed to determine the concentration of an element in a sample using k 0 NAA method. Parameters f was measured using foil activation method and α parameter was obtained from power function fitting at epithermal neutron spectrum. Based on the fitting method the a parameter was determined of 0.0267,0.0255 and -0.0346 at system rabbit, IP2 and CIP irradiation position. The k 0 factor is depended on absolute gamma fraction. The neutron flux thermal to epithermal ratio f at all rabbit system is closed to 40. (author)

Design features of HANARO Neutron Flux Monitoring System and its operating experiences

International Nuclear Information System (INIS)

Kim, Young-Ki; Ahn, Guk-Hoon

1999-01-01

The Neutron Flux Monitoring System for HANARO provides reliable neutron flux measurement from reactor shutdown to reactor full power level ranging 10 decades from 10 0 nv to 10 10 nv. The neutron flux monitoring system consists of a guarded fission chamber, amplifier and signal processor. The neutron flux as the measure of reactor power is continuously monitored by six(6) fission chambers mounted on the courtside wall of the reflector tank in the pool. Three(3) of the fission chambers are used for reactor power control, while the other three(3) are used for tripping the reactor in case of power excursion. Only the wide range fission chamber-based neutron monitoring system is employed for neutron power measurement thereby source range and intermediate range detectors are not necessary and the number of neutron monitoring channels are minimized at HANARO. (author)

Analysis of the Photoneutron Yield and Thermal Neutron Flux in an Unreflected Electron Accelerator-Driven Neutron Source

International Nuclear Information System (INIS)

Dale, Gregory E.; Gahl, John M.

2005-01-01

There are several potential uses for a high-flux thermal neutron source in both industrial and clinical applications. The viable commercial implementation of these applications requires a low-cost, high-flux thermal neutron generator suitable for installation in industrial and clinical environments. This paper describes the Monte Carlo for N-Particle modeling results of a high-flux thermal neutron source driven with an electron accelerator. An electron linear accelerator (linac), fitted with a standard X-ray converter, can produce high neutron yields in materials with low photonuclear threshold energies, such as D and 9 Be. Results indicate that a 10-MeV, 10-kW electron linac can produce on the order of 10 12 n/s in a heavy water photoneutron target. The thermal neutron flux in an unreflected heavy water target is calculated to be on the order of 10 10 n.cm -2 .s. The sensitivity of these answers to heavy water purity is also investigated, specifically the dilution of heavy water with light water. It is shown that the peak thermal neutron flux is not adversely effected by dilution up to a light water weight fraction of 35%

The influence of the number of activation detectors on the Seibersdorf - Milano intercomparison of neutron flux density spectra by WINDOWS code

International Nuclear Information System (INIS)

Ertek, C.

1981-02-01

This work is a continuation of the work performed within the IAEA programme on standardization of reactor radiation measurements, one of the important objectives of which is the assistance to laboratories in Member States to implement or intercompare the multiple foil activation techniques for different neutron field measurements. The importance of these techniques is well recognized. In CESNEF-FERMI Politecnico di Milano, Italy, they have installed near the core of a water boiler of 50kW, a neutron filter made of B 4 C in order to obtain a neutron flux density spectrum that could be of utility in intercalibration problems connected with irradiation in fast assemblies. Dr. V. Sangiust from CESNEF kindly sent the input guess neutron flux density spectrum and a series of measured reaction rates to be treated by the IAEA Seibersdorf laboratory using the SAND-II and the WINDOWS unfolding codes. The meaningful comparison using partly the same ENDF/B IV cross section data is performed. In the present work we extended the investiga tion using WINDOWS unfolding code for different numbers of activation fo ils or reaction rates

Online In-Core Thermal Neutron Flux Measurement for the Validation of Computational Methods

International Nuclear Information System (INIS)

Mohamad Hairie Rabir; Muhammad Rawi Mohamed Zin; Yahya Ismail

2016-01-01

In order to verify and validate the computational methods for neutron flux calculation in RTP calculations, a series of thermal neutron flux measurement has been performed. The Self Powered Neutron Detector (SPND) was used to measure thermal neutron flux to verify the calculated neutron flux distribution in the TRIGA reactor. Measurements results obtained online for different power level of the reactor. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and measured thermal neutron flux in the core are in very good agreement indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux distribution in the reactor core. Since the computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of RTP utilization. (author)

Measurement of angular distribution of neutron flux for the 6 MeV race-track microtron based pulsed neutron source

Energy Technology Data Exchange (ETDEWEB)

Patil, B.J., E-mail: bjp@physics.unipune.ernet.i [Department of Physics, University of Pune, Pune 411 007 (India); Chavan, S.T.; Pethe, S.N.; Krishnan, R. [SAMEER, IIT Powai Campus, Mumbai 400 076 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ernet.i [Department of Physics, University of Pune, Pune 411 007 (India)

2010-09-15

The 6 MeV race track microtron based pulsed neutron source has been designed specifically for the elemental analysis of short lived activation products, where the low neutron flux requirement is desirable. Electrons impinges on a e-{gamma} target to generate bremsstrahlung radiations, which further produces neutrons by photonuclear reaction in {gamma}-n target. The optimisation of these targets along with their spectra were estimated using FLUKA code. The measurement of neutron flux was carried out by activation of vanadium at different scattering angles. Angular distribution of neutron flux indicates that the flux decreases with increase in the angle and are in good agreement with the FLUKA simulation.

Gas loop - continuous measurement of thermal and fast neutron fluxes; Boucle a gaz - mesure continue de flux de neutrons thermiques et rapides

Energy Technology Data Exchange (ETDEWEB)

Droulers, Y; Pleyber, G; Sciers, P; Maurin, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1964-07-01

The measurement method described in this report can be applied both to thermal and fast neutron fluxes. A description is given of two practical applications in each of these two domains. This method is particularly suitable for measurements carried out on 'loop' type equipment. The measurement of the relative flux variations are carried out with an accuracy of 5 per cent. The choice of the shape of the gas circuit leaves a considerable amount of liberty for the adaptation of the measurement circuit to the experimental conditions. (authors) [French] La methode de mesure defrite dans ce rapport s1 applique aussi bien au flux de neutrons thermiques, qu'au flux de neutrons rapides. On donne la description de deux realisations pratiques dans chacun de ces domaines. Cette methode est particulierement adaptee a des mesures effectuees sur des dispositifs du type 'boucle'. La mesure des variations relatives de flux se fait avec une precision de 5 pour cent. Le choix de la configuration du circuit gazeux donne une grande souplesse dans l'adaptation du circuit de mesure aux conditions experimentales. (auteurs)

Methods and applications in high flux neutron imaging

International Nuclear Information System (INIS)

Ballhausen, H.

2007-01-01

This treatise develops new methods for high flux neutron radiography and high flux neutron tomography and describes some of their applications in actual experiments. Instead of single images, time series can be acquired with short exposure times due to the available high intensity. To best use the increased amount of information, new estimators are proposed, which extract accurate results from the recorded ensembles, even if the individual piece of data is very noisy and in addition severely affected by systematic errors such as an influence of gamma background radiation. The spatial resolution of neutron radiographies, usually limited by beam divergence and inherent resolution of the scintillator, can be significantly increased by scanning the sample with a pinhole-micro-collimator. This technique circumvents any limitations in present detector design and, due to the available high intensity, could be successfully tested. Imaging with scattered neutrons as opposed to conventional total attenuation based imaging determines separately the absorption and scattering cross sections within the sample. For the first time even coherent angle dependent scattering could be visualized space-resolved. New applications of high flux neutron imaging are presented, such as materials engineering experiments on innovative metal joints, time-resolved tomography on multilayer stacks of fuel cells under operation, and others. A new implementation of an algorithm for the algebraic reconstruction of tomography data executes even in case of missing information, such as limited angle tomography, and returns quantitative reconstructions. The setup of the world-leading high flux radiography and tomography facility at the Institut Laue-Langevin is presented. A comprehensive appendix covers the physical and technical foundations of neutron imaging. (orig.)

Neutron spectrum measurement using rise-time discrimination method

International Nuclear Information System (INIS)

Luo Zhiping; Suzuki, C.; Kosako, T.; Ma Jizeng

2009-01-01

PSD method can be used to measure the fast neutron spectrum in n/γ mixed field. A set of assemblies for measuring the pulse height distribution of neutrons is built up,based on a large volume NE213 liquid scintillator and standard NIM circuits,through the rise-time discrimination method. After that,the response matrix is calculated using Monte Carlo method. The energy calibration of the pulse height distribution is accomplished using 60 Co radioisotope. The neutron spectrum of the mono-energetic accelerator neutron source is achieved by unfolding process. Suggestions for further improvement of the system are presented at last. (authors)

Comparison of Thermal Neutron Flux Measured by Uranium 235 Fission Chamber and Rhodium Self-Powered Neutron Detector in MTR

International Nuclear Information System (INIS)

Fourmentel, D.; Filliatre, P.; Barbot, L.; Villard, J.-F.; Lyoussi, A.; Geslot, B.; Malo, J.-Y.; Carcreff, H.; Reynard-Carette, C.

2013-06-01

Thermal neutron flux is one of the most important nuclear parameter to be measured on-line in Material Testing Reactors (MTRs). In particular two types of sensors with different physical operating principles are commonly used: self-powered neutron detectors (SPND) and fission chambers with uranium 235 coating. This work aims to compare on one hand the thermal neutron flux evaluation given by these two types of sensors and on the other hand to compare these evaluations with activation dosimeter measurements, which are considered as the reference for absolute neutron flux assessment. This study was conducted in an irradiation experiment, called CARMEN-1, performed during 2012 in OSIRIS reactor (CEA Saclay - France). The CARMEN-1 experiment aims to improve the neutron and photon flux and nuclear heating measurements in MTRs. In this paper we focus on the thermal neutron flux measurements performed in CARMEN-1 experiment. The use of fission chambers to measure the absolute thermal neutron flux in MTRs is not very usual. An innovative calibration method for fission chambers operated in Campbell mode has been developed at the CEA Cadarache (France) and tested for the first time in the CARMEN-1 experiment. The results of these measurements are discussed, with the objective to measure with the best accuracy the thermal neutron flux in the future Jules Horowitz Reactor. (authors)

Study of the flux attenuation and energy degradation of 14.8 MeV neutrons in different materials

International Nuclear Information System (INIS)

Boufraqech, A.

1981-01-01

The attenuation of 14.8 MeV neutrons has been determined using the 63-Cu(n,2n)62-Cu threshold reaction for the detection of the primary neutrons. The attenuation of primary flux in different materials can be well described by a simple exponential relation based on the removal cross section. The microscopic removal cross sections determined for graphite, aluminium, iron and lead are as follows 0.73 +- 0.04, 1.04 +-0.04, 1.41 +- 0.02 and 2.63 +- 0.12 barn respectively. The dependance of secondary neutron spectrum on the thickness of slabs has also been investigated by threshold detectors. 15 refs., 38 figs., 23 tabs. (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.

Neutron flux measurement by mobile detectors

International Nuclear Information System (INIS)

Verchain, M.

1987-01-01

Various incore instrumentation systems and their technological evolution are first reviewed. Then, for 1300 MWe PWR nuclear power plant, temperature and neutron flux measurement are described. Mobile fission chambers, with their large measuring range and accurate location allow a good knowledge of the core. Other incore measures are possible because of flux detector thimble tubes inserted in the reactor core [fr

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

A Wide Spectrum Neutron Polarizer for a Pulsed Neutron Source

International Nuclear Information System (INIS)

Nikitenko, Yu.V.

1994-01-01

A wide spectrum neutron polarizer for a pulsed neutron source is considered. The polarizer is made in a form of a set of magnetized mirrors placed on a drum. Homogeneous rotation of the polarizer is synchronized with the power pulses of the neutron source. The polarizer may be utilized in a collimated neutron beam with cross section of the order of magnitude of 100 cm 2 within a wavelength from 2 up to 20 A on sources with a pulse repetition frequency up to 50 Hz. (author). 5 refs.; 3 figs

Neutron flux enhancement in the NRAD reactor

International Nuclear Information System (INIS)

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

1988-01-01

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

Measurement of absolute neutron flux in LWSCR based on the nuclear track method

International Nuclear Information System (INIS)

Sadeghzadeh, J.; Nassiri Mofakham, N.; Khajehmiri, Z.

2012-01-01

Highlights: ► Up to now the spectral parameters of thermal neutrons are measured with activation foils that are not always reliable in low flux systems. ► We applied a solid state nuclear track detector to measure the absolute neutron flux in the light water sub-critical reactor (LWSCR). ► Experiments concerning fission track detecting were performed and were investigated using the Monte Carlo code MCNP. ► The neutron fluxes obtained in experiment are in fairly good agreement with the results obtained by MCNP. - Abstract: In the present paper, a solid state nuclear track detector is applied to measure the absolute neutron flux in the light water sub-critical reactor (LWSCR) in Nuclear Science and Technology Research Institute (NSTRI). Up to now, the spectral parameters of thermal neutrons have been measured with activation foils that are not always reliable in low flux systems. The method investigated here is the irradiation method. Experiments concerning fission track detecting were performed. The experiment including neutron flux calculation method has also been investigated using the Monte Carlo code MCNP. The analysis shows that the values of neutron flux obtained by experiment are in fairly good agreement with the results obtained by MCNP. Thus, this method may be able to predict the absolute value of neutron flux at LWSCR and other similar reactors.

Study of the influence of the fast neutron spectrum on the production of defects in solids and liquids

International Nuclear Information System (INIS)

Mas, P.; Droulers, Y.

1964-01-01

In the first part of this work a calculation has been made of the number of defects formed in graphite by a given neutron flux having various spectral distributions. The defect formation function is that of KINCHIN and PEASE; its formulation is briefly given. An efficiency function is then defined for a fast neutron spectrum. This defects produced in a light water reactor and those produced in a graphite reactor. Finally an application of this method is given for comparing the defect forming tendency in graphite in the case of the reactor Melusine and of the reactor G-2 and G-3. In the second part are calculated the integrals for the energy release brought about by fast neutrons in carbon oxygen and hydrogen. In a region of 25 cm around the core of a swimming-pool type reactor these energy release integrals are approximately proportional to the neutron flux above 1 MeV. The determination of the energy released as a result of the passage of neutrons in organic liquids can therefore be reduced to the measurement of the flux above 1 MeV for the real spectral distribution. A calorimetric verification has been carried out in the case of water. (authors) [fr

SHREDI, Neutron Flux and Neutron Activation in 2-D Shields by Removal Diffusion

International Nuclear Information System (INIS)

Daneri, A.; Toselli, G.

1976-01-01

1 - Nature of physical problem solved: SHREDI is a removal - diffusion neutron shielding code. The program computes neutron fluxes and activations in bidimensional sections (x,y or r,z) of the shield. It is also possible to consider shielding points with the same y or z coordinate (mono-dimensional problems). 2 - Method of solution: The integrals which define the removal fluxes are computed in some shield points by means of a particular algorithm based on the Simpson's and trapezoidal rules. For the diffusion calculation the finite difference method is used. The removal sources are interpolated in all diffusion points by Chebyshev polynomials. 3 - Restrictions on the complexity of the problem: Maxima: number of removal energy groups NGR = 40; number of diffusion energy groups NGD = 40; number of the reactor core and shield materials NCMP = 50; number of core mesh points in r (or x) direction for integral calculation = 75; number of core mesh points in z (or y) direction for integral calculation = 75; number of core mesh points in theta (or z) direction for integral calculation = 75; number of shield mesh points for the neutron flux calculation in r (or x) direction NPX = 200; number of shield mesh points for the neutron flux calculation in z (or y) direction NPY = 200; n.b. (NPX * NPY) le 12000

Self-powered neutron and gamma-ray flux detector

International Nuclear Information System (INIS)

Allan, C.J.; Shields, R.B.; Lynch, G.F.; Cuttler, J.M.

1980-01-01

A new type of self-powered neutron detector was developed which is sensitive to both the neutron and gamma-ray fluxes. The emitter comprises two parts. The central emitter core is made of materials that generate high-energy electrons on exposure to neutrons. The outer layer acts as a gamma-ray/electron converter, and since it has a higher atomic number and higher back-scattering coefficient than the collector, increases the net outflow or emmission of electrons. The collector, which is around the emitter outer layer, is insulated from the outer layer electrically with dielectric insulation formed from compressed metal-oxide powder. The fraction of electrons given off by the emitter that is reflected back by the collector is less than the fraction of electrons emitted by the collector that is reflected back by the emitter. The thickness of the outer layer needed to achieve this result is very small. A detector of this design responds to external reactor gamma-rays as well as to neutron capture gamma-rays from the collector. The emitter core is either nickel, iron or titanium, or alloys based on these metals. The outer layer is made of platinum, tantalum, osmium, molybdenum or cerium. The detector is particularly useful for monitoring neutron and gamma ray flux intensities in nuclear reactor cores in which the neutron and gamma ray flux intensities are closely proportional, are unltimately related to the fission rate, and are used as measurements of nuclear reactor power. (DN)

Characterization of the neutron flux in the Hohlraum of the thermal column of the TRIGA Mark III reactor of the ININ

International Nuclear Information System (INIS)

Delfin L, A.; Palacios, J.C.; Alonso, G.

2006-01-01

Knowing the magnitude of the neutron flux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutron flux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutron flux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutron flux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutron flux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutron flux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutron flux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

Neutron flux calculations for criticality safety analysis using the narrow resonance approximations. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

Hathout, A M [National Center for Nuclear Safety and Radiation Control, NC-NSRC, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

The narrow resonance approximation is applicable for all low-energy resonances and the heaviest nuclides. It is of great importance in neutron calculations, hence, fertile isotopes do not undergo fission at resonance energies. The effect of overestimating the self shielded group averaged cross-section data for a given resonance nuclide can be fairly serious. In the present work, a detailed study, and derivation of the problem of self-shielding are carried-out through the information of Hansen-roach library which is used for criticality safety analysis. The intermediate neutron flux spectrum is analyzed, using the narrow resonance approximation. The resonance self-shielded values of various cross-sections are determined. 4 figs., 3 tabs.

Neutron flux enhancement at LASREF

International Nuclear Information System (INIS)

Sommer, W.F.; Ferguson, P.D.; Wechsler, M.S.

1991-01-01

The accelerator at the Los Alamos Meson Physics Facility produces a 1-mA beam of protons at an energy of 800 MeV. Since 1985, the Los Alamos Spallation Radiation Effects Facility (LASREF) has made use of the neutron flux that is generated as the incident protons interact with the nuclei in targets and a copper beam stop. A variety of basic and applied experiments in radiation damage and radiation effects have been completed. Recent studies indicate that the flux at LASREF can be increased by at least a factor of ten from the present level of about 5 E+17 m -2 s -1 . This requires changing the beam-stop material from Cu to W and optimizing the geometry of the beam-target interaction region. These studies are motivated by the need for a large volume, high energy, and high intensity neutron source in the development of materials for advanced energy concepts such as fusion reactors. 18 refs., 7 figs., 2 tabs

Neutron flux enhancement at LASREF

International Nuclear Information System (INIS)

Sommer, W.F.; Ferguson, P.D.; Wechsler, M.S.

1992-01-01

The accelerator at the Los Alamos Meson Physiscs Facility produces a 1 mA beam of protons at an energy of 800 MeV. Since 1985, the Los Alamos Spallation Radiation Effects Facility (LASREF) has made use of the neutron flux that is generated as the incident protons interact with the targets and a copper beam stop. A variety of basic and applied experiments in radiation damage and radiation effects have been completed. Recent studies indicate that the flux at LASREF can be increased by at least a factor of 10 from the present level of about 5 E + 17 m -2 s -1 . This requires changing the beam stop material from Cu to W and optimizing the geometry of the beam-target interaction region. These studies are motivated by the need for a large volume, high energy, and high intensity neutron source in the development of materials for advanced energy concepts such as fusion reactors. (orig.)

Neutron flux enhancement at LASREF

Energy Technology Data Exchange (ETDEWEB)

Sommer, W.F. (Los Alamos National Lab., Los Alamos, NM (United States)); Ferguson, P.D. (Univ. of Missouri, Rolla, MO (United States)); Wechsler, M.S. (Iowa State Univ., Ames, IA (United States))

1992-09-01

The accelerator at the Los Alamos Meson Physiscs Facility produces a 1 mA beam of protons at an energy of 800 MeV. Since 1985, the Los Alamos Spallation Radiation Effects Facility (LASREF) has made use of the neutron flux that is generated as the incident protons interact with the targets and a copper beam stop. A variety of basic and applied experiments in radiation damage and radiation effects have been completed. Recent studies indicate that the flux at LASREF can be increased by at least a factor of 10 from the present level of about 5 E + 17 m[sup -2] s[sup -1]. This requires changing the beam stop material from Cu to W and optimizing the geometry of the beam-target interaction region. These studies are motivated by the need for a large volume, high energy, and high intensity neutron source in the development of materials for advanced energy concepts such as fusion reactors. (orig.).

Feasibility of sealed D-T neutron generator as neutron source for liver BNCT and its beam shaping assembly.

Science.gov (United States)

Liu, Zheng; Li, Gang; Liu, Linmao

2014-04-01

This paper involves the feasibility of boron neutron capture therapy (BNCT) for liver tumor with four sealed neutron generators as neutron source. Two generators are placed on each side of the liver. The high energy of these emitted neutrons should be reduced by designing a beam shaping assembly (BSA) to make them useable for BNCT. However, the neutron flux decreases as neutrons pass through different materials of BSA. Therefore, it is essential to find ways to increase the neutron flux. In this paper, the feasibility of using low enrichment uranium as a neutron multiplier is investigated to increase the number of neutrons emitted from D-T neutron generators. The neutron spectrum related to our system has a proper epithermal flux, and the fast and thermal neutron fluxes comply with the IAEA recommended values. Copyright © 2014 Elsevier Ltd. All rights reserved.

Use of the associated particle technique in the fast neutron spectroscopy

International Nuclear Information System (INIS)

Aquirre O, G.A.

1978-01-01

Selecting a neutrons monoenergetic source it was found that the nuclear reaction D(d,n) 3 He can be used to measure nuclear sections and differentials in elastic nuclear reactions through the associated particle technique; the neutron beam energy is directly determined in time of flight spectrum of the neutron. The flux is determined by the number of 3 He ions observed in the charged particle spectrum. The neutron flux can be increased increasing the solid angle of the neutrons beam in two magnitude orders according to the results of neutrons beam profile measures. (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)

The effect of neutron spectrum on the mechanical and physical properties of pure copper and copper alloys

International Nuclear Information System (INIS)

Fabritsiev, S.A.; Pokrovsky, A.S.; Sandakov, V.A.; Zinkle, S.J.; Rowcliffe, A.F.; Edwards, D.J.; Garner, F.A.; Singh, B.N.; Barabash, V.R.

1996-01-01

The electrical resistivity and tensile properties of copper and oxide dispersion strengthened (DS) copper alloys have been measured before and after fission neutron irradiation to damage levels of 0.5 to 5 displacements per atom (dps) at ∼100 to 400 degrees C. Some of the specimens were irradiated inside a 1.5 mm Cd shroud in order to reduce the thermal neutron flux. The electrical resistivity data could be separated into two components, a solid transmutation component Δρ tr which was proportional to thermal neutron fluence and a radiation defect component Δρ rd which was independent of the displacement dose. The saturation value for Δρ rd was ∼1.2 nanohm-meters for pure copper and ∼1.6 nanohm-meters for the DS copper alloys irradiated at 100 degrees C in positions with a fast-to-thermal neutron flux ratio of 5. Considerable radiation hardening was observed in all specimens at irradiation temperatures below 200 degrees C. The yield strength was relatively insensitive to neutron spectrum in specimens strengthened by dispersoids or cold- working. 17 refs., 7 figs., 1 tab

Reactor AQUILON. The hardening of neutron spectrum in natural uranium rods, with a computation of epithermal fissions (1961); Pile AQUILON. Durcissement du spectre des neutrons dans les barreaux d'uranium et calcul des fissions epithermiques (1961)

Energy Technology Data Exchange (ETDEWEB)

Durand -Smet, R; Lourme, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

- Microscopic flux measurements in reactor Aquilon have allowed to investigate the thermal and epithermal flux distribution in natural uranium rods, then to obtain the neutron spectrum variations in uranium, Wescott '{beta}' term of the average spectrum in the rod, and the ratio of epithermal to therma fissions. A new definition for the infinite multiplication factor is proposed in annex, which takes into account epithermal parameters. (authors) [French] - Un certain nombre de mesures effectuees dans la pile Aquilon ont permis d'etablir la distribution fine des flux thermique et epithermique dans les barreaux d'uranium, et d'en deduire les variations du spectre des neutrons dans l'uranium, le terme {beta} du spectre de Wescott moyen dans le barreau et le nombre de fissions epithermiques. En annexe, il est propose une definition nouvelle du coefficient de multiplication infini, qui fait intervenir les parametres epithermiques. (auteurs)

Facility for fast neutron irradiation tests of electronics at the ISIS spallation neutron source

International Nuclear Information System (INIS)

Andreani, C.; Pietropaolo, A.; Salsano, A.; Gorini, G.; Tardocchi, M.; Paccagnella, A.; Gerardin, S.; Frost, C. D.; Ansell, S.; Platt, S. P.

2008-01-01

The VESUVIO beam line at the ISIS spallation neutron source was set up for neutron irradiation tests in the neutron energy range above 10 MeV. The neutron flux and energy spectrum were shown, in benchmark activation measurements, to provide a neutron spectrum similar to the ambient one at sea level, but with an enhancement in intensity of a factor of 10 7 . Such conditions are suitable for accelerated testing of electronic components, as was demonstrated here by measurements of soft error rates in recent technology field programable gate arrays

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

Accounting for the thermal neutron flux depression in voluminous samples for instrumental neutron activation analysis

International Nuclear Information System (INIS)

Overwater, R.M.W.; Hoogenboom, J.E.

1994-01-01

At the Delft University of Technology Interfaculty Reactor Institute, a facility has been installed to irradiate cylindrical samples with diameters up to 15 cm and weights up to 50 kg for instrumental neutron activation analysis (INAA) purposes. To be able to do quantitative INAA on voluminous samples, it is necessary to correct for gamma-ray absorption, gamma-ray scattering, neutron absorption, and neutron scattering in the sample. The neutron absorption and the neutron scattering are discussed. An analytical solution is obtained for the diffusion equation in the geometry of the irradiation facility. For samples with known composition, the neutron flux--as a function of position in the sample--can be calculated directly. Those of unknown composition require additional flux measurements on which least-squares fitting must be done to obtain both the thermal neutron diffusion coefficient D s and the diffusion length L s of the sample. Experiments are performed to test the theory

The effect of an iron plug on the neutron flux distributions in water

International Nuclear Information System (INIS)

Lotfi, A.; Maayouf, R.M.A.; Megahid, R.

1978-01-01

This work is concerned with studying both fast and thermal neutron fluxes distribution in water and its perturbation due to the presence of a cylindrical iron plug. The measurements were carried out using a collimated neutron beam emitted from one of the horizontal channels of the ET-RR-1 reactor. The fast neutron fluxes were measured using phosphorus activation detectors, while the thermal neutron ones were measured using fission fragment track detectors from glass. The results show that the presence of an iron plug causes a remarkable change in the intensities of both the fast and thermal neutron fluxes distribution in the water medium surrounding the iron plug. The flux intensities at the peaks, formed beyond the iron plug in case of thermal neutrons, are also compared with values calculated using the available emperical formula

Dependence of the Ratio between the Resonance Integral and Thermal Neutron Cross Section on the Deviation of the Epithermal Neutron Spectrum from the 1/E Law

International Nuclear Information System (INIS)

Soliman, N.F.

2012-01-01

In k 0 - Neutron Activation Analysis (k 0 -NAA), the conversion from the tabulated Q 0 (ratio of the resonance integral to thermal neutron cross-section)to Q 0 (α) (α is the shape factor of the epithermal neutron flux, indicating the deviation of the epithermal neutron spectrum from the ideal 1/E shape) are calculated using a FORTRAN program. The calculations are done for most elements that can be detected by neutron activation using different values of the parameter (α) ranging from -0.1≤α≤+0.1. The obtained data are used to study the dependence of the values (α) on the irradiation position factor in (k 0 -NAA)equation for some selected isotopes differ in their resonance energy and its Q 0 values. The results show that, the irradiation factor is affective mainly for low thermal tro epithermal flux ratio f especially for Q 0 value greater than 50. so consequently determining the irradiation parameters α value is not needed for irradiation positions that rich with thermal neutron. But for high f values the irradiation position factor should be taken into account. On the other hand the constructed FORTRAN program can be used to calculate the value Q 0 (α) directly for different value of α

Neutron flux response to regulating rod random vibrations

International Nuclear Information System (INIS)

Dach, K.; Nemec, J.; Pecinka, L.

The relation is presented for the mean square value of the deflection of the rod for the n-th vibration shape on an arbitrary site. The relation may serve the obtaining of a variable which may be used both in a mechanical, i.e., stress analysis and in the determination of neutron flux fluctuations. It is demonstrated that the vibration frequency introduced in the reactor by the regulating rod has the same response in the neutron flux. This effect was used in the localization of an enormously vibrating regulating rod. (J.P.)

Gamma-ray-spectroscopy following high-flux 14-MeV neutron activation

International Nuclear Information System (INIS)

Williams, R.E.

1981-01-01

The Rotating Target Neutron Source (RTNS-I), a high-intensity source of 14-MeV neutrons at the Lawrence Livermore National Laboratory (LLNL), has been used for applications in activation analysis, inertial-confinement-fusion diagnostic development, and fission decay-heat studies. The fast-neutron flux from the RTNS-I is at least 50 times the maximum fluxes available from typical neutron generators, making these applications possible. Facilities and procedures necessary for gamma-ray spectroscopy of samples irradiated at the RTNS-I were developed

Gamma-ray-spectroscopy following high-flux 14-MeV neutron activation

Energy Technology Data Exchange (ETDEWEB)

Williams, R.E.

1981-10-12

The Rotating Target Neutron Source (RTNS-I), a high-intensity source of 14-MeV neutrons at the Lawrence Livermore National Laboratory (LLNL), has been used for applications in activation analysis, inertial-confinement-fusion diagnostic development, and fission decay-heat studies. The fast-neutron flux from the RTNS-I is at least 50 times the maximum fluxes available from typical neutron generators, making these applications possible. Facilities and procedures necessary for gamma-ray spectroscopy of samples irradiated at the RTNS-I were developed.

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

International Nuclear Information System (INIS)

Ljubenov, V.; Milosevic, M.

2004-11-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Effect of Different Thermal Neutron Fluxes on Blood of Male Mice

International Nuclear Information System (INIS)

Abd El-Latif, A.A.; Saeid, Kh. S.; Abd El-Latif, A.A.; Emara, N.M.; Emara, N.M.

2010-01-01

This work deals with the exposing of male mice to different fluxes of thermal neutron .Investigation has been performed by calculating of thermal neutron fluxes(0.27x10 8 N/cm 2 . 1h , 0.54x10 8 N/cm 2 . 1h, 1.08x10 8 N/cm 2 . 1h, 2.16x10 8 N/cm 2 . 3h and 4.32x10 8 N/cm 2 . 6h) which emitted from neutron irradiation cell with source Ra - Be (α,n) have activity 3 m. Ci made by leybold(55930) . The number and differential leucocytes counts types of white blood cells in million per cubic millimeter (W. B. Cs. mm -3 ) ,the number of platelets mm -3 ,the number of red blood cells in million per cubic millimeter (R. B. Cs. mm -3 ), the hemoglobin in Blood (mg/dl), the lymphocytes ,and the eosiniphil leucocytes in blood decrease with increasing thermal neutron fluxes. But neutrophile and monocytes in blood increase with increasing the thermal neutron fluxes

Study of U{sup 235} neutron fission spectrum by the knowledge of cross sections average over that spectrum; Estudio del espectro de neutrones de fision del {sup 235}U a traves del conocimiento de secciones eficaces promediadas sobre dicho espectro

Energy Technology Data Exchange (ETDEWEB)

Suarez, P M [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche

1998-12-31

A literature search of cross sections averaged over the fission neutron spectrum confirms inconsistencies between calculated and experimental values for high threshold reactions. Since, in this case, calculated averaged cross sections are systematically lower than measured values, it is concluded that the representations used to carry out these calculations underestimate the number of neutrons in the high energy region of the spectrum. A careful measurement of the averaged cross section for the {sup 45}Sc(n,2n) {sup 44g}Sc and {sup 45}Sc(n,2n) {sup 44m}Sc high threshold reactions had been performed in the RA-6 Neutron Activation Analysis Laboratory after carefully checking that the neutron flux at the core position where the samples were being irradiated was indeed an undisturbed fission spectrum. The experimental values are greater than those calculated with either, Watt type representations or the one based on the Madland and Nix model for the prompt fission spectrum. In many areas of nuclear engineering, like validation of nuclear data, reactor calculations, applied nuclear physics, shielding design, etc., it is of great practical importance to have a representation for the neutron flux that can be expressed in a closed analytical form and that agrees with experimental results, specially for the most widely fissile nuclide, {sup 235}U. The results of the calculations mentioned above lead us to propose an analytical form for the {sup 235}U fission neutron spectrum that better agrees with experimental results in the whole energy spectrum. We propose two different forms; both are a modification of the Watt-type form that has been adopted within the ENDF/B-V files. One of the new analytical representations is defined in two regions: below 9.5 MeV it is exactly the same formula as that used within the ENDF/B-V files, above this energy the parameters of this formula are changed. The other proposed analytical representation is expressed by a single formula in the whole

Some neutronic studies on flux-trap type moderator

International Nuclear Information System (INIS)

Kiyanagi, Y.; Watanabe, N.

1991-01-01

The neutronic performance of a flux-trap type moderator was studied by computer simulation in connection with the KENS-II target-moderator system. It was confirmed that this system can provide 1.3-1.4 times higher neutron intensity than a traditional wing-geometry moderator system. (author)

The calculation of neutron flux using Monte Carlo method

Science.gov (United States)

Günay, Mehtap; Bardakçı, Hilal

2017-09-01

In this study, a hybrid reactor system was designed by using 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2 fluids, ENDF/B-VII.0 evaluated nuclear data library and 9Cr2WVTa structural material. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion-fission hybrid reactor system. The neutron flux was calculated according to the mixture components, radial, energy spectrum in the designed hybrid reactor system for the selected fluids, library and structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.

Fission ionisation chamber for the measurement of low fluxes of slow neutrons; Chambre d'ionisation a fission pour la mesure des faibles flux de neutrons lents

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

The ionisation chamber described is designed for the measurement of slow neutron fluxes of average or low intensity, in the presence, eventually, of very high gamma fluxes. The capture of a slow neutron by a fissile material, in this case {sup 235}U, gives rise to fission fragments, high-energy particles which ionise the gas contained in the chamber. The neutrons are detected by virtue of the potential pulses, on the collecting electrode of the chamber, deriving from the collection of the ions produced by the fission fragments. The pulses are counted by means of a measuring system consisting of a preamplifier, a 2 Mc amplifier, a discriminator and an electronic scale with numerator or integrator. The general characteristics are as follows: sensitivity to neutrons: 0.07 kicks/n/cm{sup 2}.s, sensitivity to {gamma} rays: zero up to 3.10{sup 4} R/H, a background noise at the normal discrimination voltage: 0.01 kicks/s, working H.T.: -500 V, capacity: 40 {mu}{mu}F, average height of pulse: 8 mV, limits of use: from several neutrons to 10{sup 6} n/cm{sup 2}.s. This chamber may be used in all cases where low fluxes of slow neutrons must be measured, especially in the presence of high gamma fluxes, for example in the checking of Pu concentrations in an extraction plant or for the starting up of reactors. (author) [French] La chambre d'ionisation decrite est destinee a la mesure des flux de neutrons lents d'intensite moyenne ou faible, en presence eventuelle de flux gamma tres importants. La capture d'un neutron lent par un materiau fissile, en l'occurrence {sup 235}U, donne naissance a des fragments de fission, particules de grande energie qui ionisent le gaz contenu dans la chambre. Les neutrons sont detectes grace aux impulsions de potentiel, sur l'electrode collectrice de la chambre, provenant de la collection des ions produits par les fragments de fission. Une chaine de mesure comprenant un preamplificateur, un amplificateur 2 Mc, un discriminateur, une echelle

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)

The measurement of tripartition alpha particle low energy spectrum in 235U fission induced by thermal neutrons

International Nuclear Information System (INIS)

El Hage Sleiman, F.

1980-01-01

The energy spectrum of the α particles emitted in the thermal neutron induced fission of 235 U was measured from 11.5 MeV down to 2 MeV using the parabola mass spectrometer Lohengrin at the ILL high flux reactor. A Monte Carlo program, that simulates the α particle motion to the spectrometer, has been developed. Numerical results of Monte Carlo calculations for differents values of parameter are reported. The overall energy spectrum is slightly asymmetric at low energy. The possible reasons for the existence of this asymmetry are discussed [fr

Thermal, intermediate and fast neutron flux measurements using activation detectors; Mesure des flux de neutrons thermiques, intermediaires et rapides au moyen de detecteurs par activation

Energy Technology Data Exchange (ETDEWEB)

Brisbois, J; Lott, M; Manent, G [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

The problem of neutron flux measurements using activation detectors is studied in the particular case of protection research. It is shown how it possible, it is possible, using a known thermal flux, to organise a coherent calibration system for all the detectors. The rapid neutron detectors are calibrated with respect to a reference detector (phosphorus) in a natural uranium converter; the intermediate neutron detectors with respect to gold in the axial channel of ZOE. This method makes it possible to minimise the errors due to the activation cross-sections. A brief description is given of the counting room of the Pile Safety Study Service, as well of the practical utilisation characteristics of the counters employed. (authors) [French] Le probleme de la mesure des flux de neutrons au moyen de detecteurs par activation est etudie dans le cas particulier des etudes de protections. On montre comment, a partir d'un flux thermique connu, on peut organiser un systeme coherent d'etalonnage de tous les detecteurs. Les detecteurs de neutrons rapides sont etalonnes par rapport a un detecteur de reference (phosphore) dans un convertisseur en uranium naturel; les detecteurs de neutrons intermediaires, par rapport a l'or dans le canal axial de ZOE, Cette methode permet de minimiser les erreurs dues aux sections efficaces d'activation. On decrit sommairement la salle de comptage du Service d'Etudes de Protections de Piles et on indique les caracteristiques d'emploi pratique des detecteurs utilises. (auteurs)

Feasibility of laser pumping with neutron fluxes from present-day large tokamaks

Energy Technology Data Exchange (ETDEWEB)

Jassby, D.L.

1986-08-01

The minimum fusion-neutron flux needed to observe nuclear-pumped lasing with tokamaks can be reduced substantially by optimizing neutron scattering into the laser cell, located between adjacent toroidal-field coils. The laser lines most readily pumped are probably the /sup 3/He-Ne lines at 0.633 ..mu.. and in the infrared, where the /sup 3/He-Ne gas is excited by energetic ions produced in the /sup 3/He(n,p)T reaction. These lines are expected to lase at the levels of D-T neutron flux foreseen for the TFTR in 1989 (>>10/sup 12/ n/cm/sup 2//s), while amplification should be observable at the existing levels of D-D neutron flux (greater than or equal to 5 x 10/sup 9/ n/cm/sup 2//s). Lasing on the 1.73 ..mu.. and 2.63 ..mu.. transitions of Xe may be observable at the maximum expected levels of D-T neutron flux in TFTR enhanced by scattering.

Feasibility of laser pumping with neutron fluxes from present-day large tokamaks

International Nuclear Information System (INIS)

Jassby, D.L.

1986-08-01

The minimum fusion-neutron flux needed to observe nuclear-pumped lasing with tokamaks can be reduced substantially by optimizing neutron scattering into the laser cell, located between adjacent toroidal-field coils. The laser lines most readily pumped are probably the 3 He-Ne lines at 0.633 μ and in the infrared, where the 3 He-Ne gas is excited by energetic ions produced in the 3 He(n,p)T reaction. These lines are expected to lase at the levels of D-T neutron flux foreseen for the TFTR in 1989 (>>10 12 n/cm 2 /s), while amplification should be observable at the existing levels of D-D neutron flux (≥ 5 x 10 9 n/cm 2 /s). Lasing on the 1.73 μ and 2.63 μ transitions of Xe may be observable at the maximum expected levels of D-T neutron flux in TFTR enhanced by scattering

Measurement of the neutron flux distributions, epithermal index, Westcott thermal neutron flux in the irradiation capsules of hydraulic conveyer (Hyd) and pneumatic tubes (Pn) facilities of the KUR

International Nuclear Information System (INIS)

Chatani, Hiroshi

2001-05-01

The reactions of Au(n, γ) 198 Au and Ti(n, p) 47 or 48 Sc were used for the measurements of the thermal and epithermal (thermal + epithermal) and the fast neutron flux distributions, respectively. In the case of Hyd (Hydraulic conveyer), the thermal + epithermal and fast neutron flux distributions in the horizontal direction in the capsule are especially flat; the distortion of the fluxes are 0.6% and 5.4%, respectively. However, these neutron fluxes in the vertical direction are low at the top and high at the bottom of the capsule. These differences between the top and bottom are 14% for both distributions. On the other hand, in polyethylene capsules of Pn-1, 2, 3 (Pneumatic tubes Nos. 1, 2, 3), in contrast with Hyd, these neutron flux distributions in the horizontal direction have gradients of 8 - 18% per 2.5 cm diameter, and those on the vertical axis have a distortion of approximately 5%. The strength of the epithermal dE/E component relative to the neutron density including both thermal and epithermal neutrons, i.e., the epithermal index, for the hydraulic conveyer (Hyd) and pneumatic tube No.2 (Pn-2), in which the irradiation experiments can be achieved, are determined by the multiple foil activation method using the reactions of Au(n, γ) 198 Au and Co(n, γ) 60(m+g) Co. The epithermal index observed in an aluminum capsule of Hyd is 0.034-0.04, and the Westcott thermal neutron flux is 1.2x10 14 cm -2 sec -1 at approximately 1 cm above the bottom. The epithermal index in a Pn-2 polyethylene capsule was measured by not only the multiple foil activation method but also the Cd-ratio method in which the Au(n, γ) 198 Au reaction in a cadmium cover is also used. The epithermal index is 0.045 - 0.055, and the thermal neutron flux is 1.8x10 13 cm -2 sec -1 . (J.P.N.)

Connection factor calculation for isotopic neutron flux measurements with foil detectors

International Nuclear Information System (INIS)

Avila L, J.

1987-01-01

Thermal and resonance neutron self-shielding factors, neutron flux distortion and edge effects as well as a connection factor for neutron flux profile around a foil detector have been calculated. A general expression for resonance self shielding factor is presented in order to take into account the most important resonances for a given isotope. A computer program SPRESYTER.BAS was written and results for In-115 and Au-197 foils are given

A fission ionization detector for neutron flux measurements at a spallation source

Energy Technology Data Exchange (ETDEWEB)

Wender, S.A. (Los Alamos National Lab., Los Alamos, NM (United States)); Balestrini, S. (Los Alamos National Lab., Los Alamos, NM (United States)); Brown, A. (Los Alamos National Lab., Los Alamos, NM (United States)); Haight, R.C. (Los Alamos National Lab., Los Alamos, NM (United States)); Laymon, C.M. (Los Alamos National Lab., Los Alamos, NM (United States)); Lee, T.M. (Los Alamos National Lab., Los Alamos, NM (United States)); Lisowski, P.W. (Los Alamos National Lab., Los Alamos, NM (United States)); McCorkle, W. (Los Alamos National Lab., Los Alamos, NM (United States)); Nelson, R.O. (Los Alamos National Lab., Los Alamos, NM (United States)); Parker, W. (Los Alamos National Lab., Los Alamos, NM (United States)); Hill, N.W. (Oak Ridge National Lab., Oak Ridge, TN (United States))

1993-11-15

The construction of a neutron flux monitor that can measure absolute neutron intensities in the neutron energy range from below 1 MeV to over 500 MeV is described. The detector consists of an ionization chamber with several thin deposits of fissionable material. The ionization chamber is thin enough that it does not significantly affect the neutron beam and may be left in the neutron flight path during experimental measurements to continuously monitor the beam flux. The use of this monitor at the continuous-energy spallation neutron source at the WNR target area at LAMPF is described. (orig.)

A fission ionization detector for neutron flux measurements at a spallation source

International Nuclear Information System (INIS)

Wender, S.A.; Balestrini, S.; Brown, A.; Haight, R.C.; Laymon, C.M.; Lee, T.M.; Lisowski, P.W.; McCorkle, W.; Nelson, R.O.; Parker, W.; Hill, N.W.

1993-01-01

The construction of a neutron flux monitor that can measure absolute neutron intensities in the neutron energy range from below 1 MeV to over 500 MeV is described. The detector consists of an ionization chamber with several thin deposits of fissionable material. The ionization chamber is thin enough that it does not significantly affect the neutron beam and may be left in the neutron flight path during experimental measurements to continuously monitor the beam flux. The use of this monitor at the continuous-energy spallation neutron source at the WNR target area at LAMPF is described. (orig.)

Design basis for creep of zirconium alloy components in a fast neutron flux

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Fidleris, V.

1975-01-01

The chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behaviour of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

Design basis for creep of zirconium alloy components in a fast neutron flux

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Fidleris, V.

1974-01-01

The Chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behavior of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time, and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

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.

Experimental techniques for the consolidation of the neutron spectrum

International Nuclear Information System (INIS)

Chiaraviglio, N.; Bazzana, S.

2013-01-01

Unfolding techniques are widely known but their use is not widespread due to their complexity. Such procedure consists in the adjustment of calculated quantities to experimental results by the modification of the neutron spectrum, getting correction factors for the calculated quantities. In this work we describe the general procedure that must be executed for a neutron spectrum unfolding. (author) [es

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

International Nuclear Information System (INIS)

Mura, Luiz Ernesto Credidio

2011-01-01

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

Calculation of neutron flux distribution of thermal neutrons from microtron converter in a graphite moderator with water reflector

International Nuclear Information System (INIS)

Andrejsek, K.

1977-01-01

The calculation is made of the thermal neutron flux in the moderator and reflector by solving the neutron diffusion equation using the four-group theory. The correction for neutron absorption in the moderator was carried out using the perturbation theory. The calculation was carried out for four groups with the following energy ranges: the first group 2 MeV to 3 keV, the second group 3 keV to 5 eV, the third group 5 eV to 0.025 eV and the fourth group 0.025 eV. The values of the macroscopic cross section of capture and scattering, of the diffusion coefficient, the macroscopic cross section of the moderator, of the neutron age and the extrapolation length for the water-graphite moderator used in the calculations are given. The spatial distribution of the thermal neutron flux is graphically represented for graphite of a 30, 40, and 50 cm radius and for graphite of a 30 and 40 cm radius with a 10 cm water reflector; a graphic comparison is made of the distribution of the thermal neutron flux in water and in graphite, both 40 cm in radius. The system of graphite with reflector proved to be the best and most efficient system for raising the flux density of thermal neutrons. (J.P.)

Recent developments on micrometric fission chambers for high neutron fluxes

International Nuclear Information System (INIS)

Letourneau, A.; Bringer, O.; Dupont, E.; Marie, F.; Panebianco, S.; Toussaint, J. C.; Veyssiere, C.; Chabod, S.; Breaud, S.; Oriol, L.

2009-01-01

With the development of innovative nuclear systems and new generation neutron sources, the nuclear instrumentation should be adapted. Since several years, we developed microscopic fission chambers to study the transmutation of minor actinides in high thermal-neutron fluxes. The recent developments done to fulfill the drastic conditions of irradiations are described in this paper together with the feedback from the measurements. Two installations were used: the HFR of the ILL for its highest thermal neutron flux of the world and the MEGAPIE target which was the first 1 MW liquid Pb-Bi spallation target in the world. (authors)

Recent developments on micrometric fission chambers for high neutron fluxes

Energy Technology Data Exchange (ETDEWEB)

Letourneau, A. [Irfu, Service de Physique Nucleaire, CEA-Saclay, 91191 Gif-sur-Yvette (France); Bringer, O.; Dupont, E.; Marie, F.; Panebianco, S.; Toussaint, J. C.; Veyssiere, C. [Irfu, CEA-Saclay, 91191 Gif-sur-Yvette (France); Chabod, S. [LPSC, Universite Joseph Fourier Grenoble 1, CNRS/IN2P3 (France); Breaud, S.; Oriol, L. [DEN/DER/SPEX, CEA-Cadarache, Saint-Paul-lez-Durances (France)

2009-07-01

With the development of innovative nuclear systems and new generation neutron sources, the nuclear instrumentation should be adapted. Since several years, we developed microscopic fission chambers to study the transmutation of minor actinides in high thermal-neutron fluxes. The recent developments done to fulfill the drastic conditions of irradiations are described in this paper together with the feedback from the measurements. Two installations were used: the HFR of the ILL for its highest thermal neutron flux of the world and the MEGAPIE target which was the first 1 MW liquid Pb-Bi spallation target in the world. (authors)

Monte Carlo simulation study of the muon-induced neutron flux at LNGS

International Nuclear Information System (INIS)

Persiani, R.; Garbini, M.; Massoli, F.; Sartorelli, G; Selvi, M.

2011-01-01

Muon-induced neutrons are ultimate background for all the experiments searching for rare events in underground laboratories. Several measurements and simulations were performed concerning the neutron production and propagation but there are disagreements between experimental data and simulations. In this work we present our Monte-Carlo simulation study, based on Geant4, to estimate the muon-induced neutron flux at LNGS. The obtained integral flux of neutrons above 1 MeV is 2.31 x 10 -10 n/cm 2 /s.

The secondary neutron sources for generation of particular neutron fluxes

International Nuclear Information System (INIS)

Tracz, G.

2007-07-01

The foregoing paper presents the doctor's thesis entitled '' The secondary neutron sources for generation of particular neutron fluxes ''. Two secondary neutron sources have been designed, which exploit already existing primary sources emitting neutrons of energies different from the desired ones. The first source is devoted to boron-neutron capture therapy (BNCT). The research reactor MARIA at the Institute of Atomic Energy in Swierk (Poland) is the primary source of the reactor thermal neutrons, while the secondary source should supply epithermal neutrons. The other secondary source is the pulsed source of thermal neutrons that uses fast 14 MeV neutrons from a pulsed generator at the Institute of Nuclear Physics PAN in Krakow (Poland). The physical problems to be solved in the two mentioned cases are different. Namely, in order to devise the BNCT source the initial energy of particles ought to be increased, whilst in the other case the fast neutrons have to be moderated. Slowing down of neutrons is relatively easy since these particles lose energy when they scatter in media; the most effective moderators are the materials which contain light elements (mostly hydrogen). In order to increase the energy of neutrons from thermal to epithermal (the BNCT case) the so-called neutron converter should be exploited. It contains a fissile material, 235 U. The thermal neutrons from the reactor cause fission of uranium and fast neutrons are emitted from the converter. Then fissile neutrons of energy of a few MeV are slowed down to the required epithermal energy range. The design of both secondary sources have been conducted by means of Monte Carlo simulations, which have been carried out using the MCNP code. In the case of the secondary pulsed thermal neutron source, some of the calculated results have been verified experimentally. (author)

Measurement of neutron flux by semiconductor detector; Merenje raspodele neutronskog fluksa poluprovodnickim detektorom

Energy Technology Data Exchange (ETDEWEB)

Obradovic, D; Bosevski, T [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-07-01

Using semiconductor detectors for measuring the neutron flux distribution is considered suitable and faster than using activation foils. Results of radial neutron flux distribution obtained by semiconductor detectors are presented.

Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

Energy Technology Data Exchange (ETDEWEB)

Barnett, P. C. [Canadian General Electric (Canada)

1968-07-15

The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

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)

Flux distribution in phantom for biomedical use of beam-type thermal neutrons

International Nuclear Information System (INIS)

Aoki, Kazuhiko; Kobayashi, Tooru; Kanda, Keiji; Kimura, Itsuro

1985-01-01

For boron neutron capture therapy, the thermal neutron beam is worth using as therapeutic neutron irradiation without useless and unfavorable exposure of normal tissues around tumor and for microanalysis system to measure ppm-order 10 B concentrations in tissue and to search for the location of the metastasis of tumor. In the present study, the thermal neutron flux distribution in a phantom, when beam-type thermal neutrons were incident on it, was measured at the KUR Neutron Guide Tube. The measurements were carried out by two different methods using indium foil. The one is an ordinary foil activation technique by using the 115 In(n, γ) 116m 1 In reactions, while the other is to detect γ-rays from the 115 In(n, γ) 116m 2 In reactions during neutron irradiations with a handy-type Ge detector. The calculations with DOT 3.5 were performed to examine thermal neutron flux in the phantom for various beam size and phantom size. The experimental and calculated results are in good agreement and it is shown that the second type measurement has a potential for practical application as a new monitoring system of the thermal neutron flux in a living body for boron neutron capture therapy. (author)

Compilation of neutron flux density spectra and reaction rates in different neutron fields. V.3

International Nuclear Information System (INIS)

Ertek, C.

1980-04-01

Upon the recommendation of the International Working Group of Reactor Radiation Measurements (IWGRRM) a compilation of documents containing neutron flux density spectra and the reaction rates obtained by activiation and fission foils in different neutron fields is presented

Measurements of neutron spectrum from uranium converter

International Nuclear Information System (INIS)

Ninkovic, M.; Sotic, O.; Marinkovic, S.

1978-01-01

The procedure for determination of energetic distribution of neutrons by the multisphere technique is given. The theoretical basis and features of the method are explained. The spectral distribution of neutrons emerging from the neutron converter constructed at the bare reactor assembly RB, has been determined applying the existing computer programme and literature data for the energetic dependence functions of spheres of various diameters. The obtained spectral distribution has a specific maximum in the domain of fast neutrons, justifying thus the reacton for the construction of the converter. The neutron spectrum data obtained and given in this report are very important for the use of the converter in neutron dosimetry and radiation protection, as well as in the radiobiology, shielding, reactor physics etc. (author)

Epithermal neutron flux in the experimental channels of the RA reactor; Fluks epitermalnih 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

Epithermal neutron flux was determined by measuring the cadmium ratio from activation analysis of gold and indium foils. Irradiation was done in experimental channels VK-5, VK-2, VK-7, VK-0, VK-HS next to the core, and next to the fuels elements. Activation of bare foils and foils covered by 0.8 and 1.0 mm thick cadmium foils was done simultaneously. Activity was measured by GM counter. Corrections were done for resonant neutrons self-absorption since the foils used were too thick concerning most important resonances of gold and indium. Final results were presented as spectrum factor r for their direct use in determining the effective neutron cross sections.

Neutron flux measurements at the Wendelstein VII-A stellarator

International Nuclear Information System (INIS)

Weller, A.; Maassberg, H.

1985-10-01

In addition to charge exchange analysis (CX) and charge exchange recombination spectroscopy (CXRS), the time evolution of the central ion temperature during neutral beam heated plasma discharges in the Wendelstein VII-A stellarator is derived from the neutron flux from thermal D-D reactions. In general, good quantitative agreement between the different methods is obtained. Neutron flux measurements also permit to investigate the slowing down of fast D + -ions from neutral beam injection (NBI). The results agree well with the predictions based on the assumption of a collisional slowing down mechanism. (orig.)

Study of neutronic flux in IPR-R1 reactor with MCNPX; Estudo do fluxo neutronico no reator IPR-R1 com o MCNPX

Energy Technology Data Exchange (ETDEWEB)

Melo, J.A.S.; Castrillo, L.S., E-mail: julio.angelo@poli.br, E-mail: lazara@poli.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica; Oliveira, R.M.B.M., E-mail: romero.matias@educacao.pe.gov.br [Secretaria Executiva de Educacao do Estado de Pernambuco (SEE), Recife, PE (Brazil)

2016-11-01

MCNPX computer code, one of the latest versions of code MCNP transport were used to study the flux distribution and its neutronic fluence as a function of energy in two research reactor irradiation IPR-R1. The model developed was validated with research conducted by Dalle (2005). Initially, in the simulation is considered fresh fuel whose core configuration contained three neutron rods control, being two of them 100% ejected while the other inserted 3,1 x 10{sup -1} m deep, as adopted in the literature situation. The neutron source used was the critical type, through KSRC card. The results of the neutron flow and neutronic fluence were obtained in the central tube and the turntable on a range of energy spectrum that ranged from 1.0 x 10{sup -9} MeV to 10 MeV, showing good correlations with the model used in validation. Finally, a hypothetical situation wherein the three reactor control rods are ejected simultaneously was simulated. The simulation results showed an increase in the neutron flux of 7% in the central tube and 5% on the turntable.

Evaluation of the current fast neutron flux monitoring instrumentation applied to LFR demonstrator ALFRED. Capabilities and limitations

International Nuclear Information System (INIS)

Lepore, Luigi; Remetti, Romolo; Cappelli, Mauro

2015-01-01

Among Gen IV projects for future nuclear power plants, Lead Fast Reactors (LFR) seem to be a very interesting solution due to their benefits in terms of fuel cycle, coolant-safety and waste management. The novelty of the matter causes some open issues about coolant chemical aspect, structural aspects, monitoring instrumentation, etc. Particularly hard neutron flux spectra would make traditional neutron instrumentation unfit to all reactor conditions, i.e. source, intermediate, and power range. Identification of new models of nuclear instrumentation specialized for LFR neutron flux monitoring asks for an accurate evaluation of the environment the sensor will work in. In this study, thermal-hydraulics and chemical conditions for LFR core environment will be assumed, as the neutron flux will be studied extensively by means of the Monte Carlo transport code MCNPX. The core coolant’s high temperature drastically reduces the candidate instrumentation, because only some kind of fission chambers and Self Powered Neutron Detectors can be operated in such an environment. This work aims to evaluate the capabilities of the available instrumentation (usually designed for Sodium Fast Reactors, SFRs) when exposed to the neutron spectrum derived from ALFRED, a pool-type small-power LFR project to demonstrate the feasibility of this technology into the European framework. This paper shows that such instruments do follow the power evolution, but they are not completely suitable to detect the whole range of reactor power. Some improvements are then possible in order to increase the signal-to-noise ratio, by optimizing each instrument in the range of reactor power, such to get the best solution. Some new detector designs are here proposed, and the possibilities for prototyping and testing by means of a fast reactor investigated. (author)

NULIF: neutron spectrum generator, few-group constant calculator, and fuel depletion code

International Nuclear Information System (INIS)

Wittkopf, W.A.; Tilford, J.M.; Andrews, J.B. II; Kirschner, G.; Hassan, N.M.; Colpo, P.N.

1977-02-01

The NULIF code generates a microgroup neutron spectrum and calculates spectrum-weighted few-group parameters for use in a spatial diffusion code. A wide variety of fuel cells, non-fuel cells, and fuel lattices, typical of PWR (or BWR) lattices, are treated. A fuel depletion routine and change card capability allow a broad range of problems to be studied. Coefficient variation with fuel burnup, fuel temperature change, moderator temperature change, soluble boron concentration change, burnable poison variation, and control rod insertion are readily obtained. Heterogeneous effects, including resonance shielding and thermal flux depressions, are treated. Coefficients are obtained for one thermal group and up to three epithermal groups. A special output routine writes the few-group coefficient data in specified format on an output tape for automated fitting in the PDQ07-HARMONY system of spatial diffusion-depletion codes

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

Criticality experiments to provide benchmark data on neutron flux traps

International Nuclear Information System (INIS)

Bierman, S.R.

1988-06-01

The experimental measurements covered by this report were designed to provide benchmark type data on water moderated LWR type fuel arrays containing neutron flux traps. The experiments were performed at the US Department of Energy Hanford Critical Mass Laboratory, operated by Pacific Northwest Laboratory. The experimental assemblies consisted of 2 /times/ 2 arrays of 4.31 wt % 235 U enriched UO 2 fuel rods, uniformly arranged in water on a 1.891 cm square center-to-center spacing. Neutron flux traps were created between the fuel units using metal plates containing varying amounts of boron. Measurements were made to determine the effect that boron loading and distance between the fuel and flux trap had on the amount of fuel required for criticality. Also, measurements were made, using the pulse neutron source technique, to determine the effect of boron loading on the effective neutron multiplications constant. On two assemblies, reaction rate measurements were made using solid state track recorders to determine absolute fission rates in 235 U and 238 U. 14 refs., 12 figs., 7 tabs

Analysis of neutron flux increase in the horizontal experimental channels of Ra reactor - masters thesis

International Nuclear Information System (INIS)

Strugar, P.

1964-12-01

Calculation and experimental results shown in this paper show that higher thermal neutron flux is obtained in the reactor core with central horizontal reflector at the same power level. The flux is increased when the moderation capability of the core is decreased. Apart from increase of the thermal component of the neutron flux in the experimental channels, the central reflector causes decrease of the epithermal neutron flux and gamma radiation intensity. This is very useful for studying (n, γ) reaction, neutron diffraction, etc. [sr

Fast neutron flux analyzer with real-time digital pulse shape discrimination

Energy Technology Data Exchange (ETDEWEB)

Ivanova, A.A., E-mail: a.a.ivanova@inp.nsk.su [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Zubarev, P.V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Novosibirsk State Technical University, 630092 Novosibirsk (Russian Federation); Ivanenko, S.V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Khilchenko, A.D. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Novosibirsk State Technical University, 630092 Novosibirsk (Russian Federation); Kotelnikov, A.I. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Polosatkin, S.V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Novosibirsk State Technical University, 630092 Novosibirsk (Russian Federation); Novosibirsk State University, 630090 Novosibirsk (Russian Federation); Puryga, E.A.; Shvyrev, V.G. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Novosibirsk State Technical University, 630092 Novosibirsk (Russian Federation); Sulyaev, Yu.S. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Novosibirsk State University, 630090 Novosibirsk (Russian Federation)

2016-08-11

Investigation of subthermonuclear plasma confinement and heating in magnetic fusion devices such as GOL–3 and GDT at the Budker Institute (Novosibirsk, Russia) requires sophisticated equipment for neutron-, gamma- diagnostics and upgrading data acquisition systems with online data processing. Measurement of fast neutron flux with stilbene scintillation detectors raised the problem of discrimination of the neutrons (n) from background cosmic particles (muons) and neutron-induced gamma rays (γ). This paper describes a fast neutron flux analyzer with real-time digital pulse-shape discrimination (DPSD) algorithm FPGA-implemented for the GOL–3 and GDT devices. This analyzer was tested and calibrated with the help of {sup 137}Cs and {sup 252}Cf radiation sources. The Figures of Merit (FOM) calculated for different energy cuts are presented. - Highlights: • Electronic equipment for measurement of fast neutron flux with stilbene scintillator is presented. • FPGA-implemented digital pulse-shape discrimination algorithm by charge comparison method is shown. • Calibration of analyzer was carried out with {sup 137}Cs and {sup 252}Cf. • Figures of Merit (FOM) values for energy cuts from 1/8 Cs to 2 Cs are from 1.264 to 2.34 respectively.

Self-powered neutron flux detector assembly

International Nuclear Information System (INIS)

Allan, C.J.; McIntyre, I.L.

1980-01-01

A self-powered neutron flux detector has both the central emitter electrode and its surrounding collector electrode made of inconel 600. The lead cables may also be made of inconel. Other nickel alloys, or iron, nickel, titamium, chromium, zirconium or their alloys may also be used for the electrodes

Modulation of the neutron spectrum for NCTB; Modulacion del espectro de neutrones para TCNB

Energy Technology Data Exchange (ETDEWEB)

Medina C, D.; Letechipia de L, C.; Vega C, H. R., E-mail: dmedina_c@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No.10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

2016-09-15

Cancer is characterized by the uncontrolled growth of differentiated cells in a part of the organism. Currently in the world there are millions of people living with cancer. Glioblastoma multiform e is the most common and most aggressive of brain tumors and is very difficult to treat by conventional surgery, chemotherapy or radiation. The only viable alternative is its treatment through Neutron Capture Therapy in Boron (NCTB), since is a selective therapy that requires a drug with {sup 10}B (a non-radioactive isotope of boron) and a modulated neutron beam. Thermal neutrons are captured by {sup 10}B, because has a large effective section of thermal neutron absorption, in an exothermic reaction forming the nucleus composed of {sup 11}B in an excited state that induces its cleavage in two nuclei: {sup 7}Li and alpha particle ({sup 4}He). This process causes the destruction of cancer cells by direct DNA damage, without damaging normal tissue. One of the problems associated with this therapy is to have a neutron beam with adequate flow and spectrum. The neutron spectrum must be moderated and filtered from the characteristics of the source. To this end, the main sources of neutrons are nuclear research reactors and particle accelerators. The intensity of the flow should be 2 x 10 E{sup 9} n/cm{sup 2}.s, in order to treat the patient in a reasonable time; thus, is interesting to design filters for a radial beam of a TRIGA reactor, where materials such as Cd, Al, Fe and polyethylene are being implemented in the interest of having a spectrum with which the therapy can be implemented. For this design is being played with the position of the materials, to be able to see the behavior of the spectrum and thus choose some arrangement as indicated, of course taking into account the doses of both neutrons and gammas. (Author)

A Method for the neutron flux determination during the activation process

International Nuclear Information System (INIS)

Maayouf, R.M.A.; Khalil, M.I.

2000-01-01

The present work deals with an accurate method for determining the neutron flux coming out from a neutron source during the experimental measurements. Accordingly, a suitable detector, followed by preamplifier and amplifier, is connected to a data acquisition system designed specially for this purpose; and the number of neutrons detected during every sampling period is stored in the PC. The historical file can be used to compute the average or the integral flux during any time period; considering the detector efficiency, geometrical arrangement and the amplification gain

Characterization of magnetic degradation mechanism in a high-neutron-flux environment

Energy Technology Data Exchange (ETDEWEB)

Samin, Adib; Qiu, Jie [Nuclear Engineering Program, Department of Mechanical and Aerospace, The Ohio State University, Columbus, OH 43210 (United States); Hattrick-Simpers, Jason; Dai-Hattrick, Liyang [Department of Chemical Engineering, University of South Carolina, Columbia, SC 29208 (United States); Zheng, Yuan F. [Department of Electrical and Computer Engineering, The Ohio State University, Columbus, OH 43210 (United States); Cao, Lei, E-mail: Cao.152@osu.edu [Nuclear Engineering Program, Department of Mechanical and Aerospace, The Ohio State University, Columbus, OH 43210 (United States)

2014-09-01

Radiation-induced demagnetization of permanent magnets can result in the failure of magnet-based devices operating in high-radiation environments. To understand the mechanism underlying demagnetization, Nd-Fe-B magnets were irradiated with fast and fast plus thermal neutrons at fluences of 10{sup 12}, 10{sup 13}, 10{sup 14}, and 10{sup 15} n/cm{sup 2}, respectively. After irradiation, magnetic flux losses were shown to increase with the fluence. Compared with samples irradiated only with fast neutrons, the samples exposed to the fast plus thermal neutrons have higher magnetic flux losses, which is attributed to the thermal neutron capture reaction of boron. Hysteresis loops of the Nd-Fe-B magnets reveal a slightly increase in the coercivity after irradiation. Full remagnetization of the samples after irradiation was possible, which indicates that structural damage is unlikely an important factor in the demagnetization process at these levels of neutron flux and fluence. Finally, we performed a preliminary Molecular Dynamic (MD) simulation on a cube of ions to obtain a better understanding of the thermal spike mechanism.

Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code

Energy Technology Data Exchange (ETDEWEB)

Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.

2014-04-15

In this study, the molten salt-heavy metal mixtures 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % NpO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion-fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library. (orig.)

Design and fabrication of self-powered in-core neutron flux monitor assembly

International Nuclear Information System (INIS)

Chung, M.K.; Cho, S.W.; Kang, H.D.; Cho, K.K.; Cho, B.S.; Kang, S.S.

1980-01-01

This is the final report on the prototypical fabrication of an in-core neutron flux monitor detector assembly for a specific power reactor conducted by KAERI from July 1, 1978 to December 31, 1979. It is well known that power reactors require a large number of in-core neutron flux detector for reactor regulation and the structures of detector assemblies are different from reactor to reactor. Therefore, from the nature of this project, it should be noted here that the target model of the prototypical farbrication of an in-core neutron flux monitor detector assembly is a VFD-2 System for Wolsung CANDU. It is concluded that fabrication of in-core neutron flux monitor detector assembly for CANDU reactor is technically feasible and will bring economical benefit as much as 50 % of the unit price if they are fabricated in Korea by using partially materials which are available from local market. (author)

Computational uncertainties in silicon dioxide/plutonium intermediate neutron spectrum systems

International Nuclear Information System (INIS)

Jaegers, P.J.

1997-01-01

In the past several years, several proposals have been made for the long-term stabilization and storage of surplus fissile materials. Many of these proposed scenarios involve systems that have an intermediate neutron energy spectrum. Such intermediate-energy systems are dominated by scattering and fission events induced by neutrons ranging in energy from 1 eV to 100keV. To ensure adequate safety margins and cost effectiveness, it is necessary to have benchmark data for these intermediate-energy spectrum systems; however, a review of the nuclear criticality benchmarks indicates that no formal benchmarks are available. Nuclear data uncertainties have been reported for some types of intermediate-energy spectrum systems. Using a variety of Monte Carlo computer codes and cross-section sets, reported significant variations in the calculated k ∞ of intermediate-energy spectrum metal/ 235 U systems. We discuss the characteristics of intermediate neutron spectrum systems and some of the computational differences that can occur in calculating the k eff of these systems

Pulsed White Spectrum Neutron Generator for Explosive Detection

International Nuclear Information System (INIS)

King, Michael J.; Miller, Gill T.; Reijonen, Jani; Ji, Qing; Andresen, Nord; Gicquel, Frederic; Kavlas, Taneli; Leung, Ka-Ngo; Kwan, Joe

2008-01-01

Successful explosive material detection in luggage and similar sized containers is a critical issue in securing the safety of all airline passengers. Tensor Technology Inc. has recently developed a methodology that will detect explosive compounds with pulsed fast neutron transmission spectroscopy. In this scheme, tritium beams will be used to generate neutrons with a broad energy spectrum as governed by the T(t,2n)4He fission reaction that produces 0-9 MeV neutrons. Lawrence Berkeley National Laboratory (LBNL), in collaboration with Tensor Technology Inc., has designed and fabricated a pulsed white-spectrum neutron source for this application. The specifications of the neutron source are demanding and stringent due to the requirements of high yield and fast pulsing neutron emission, and sealed tube, tritium operation. In a unique co-axial geometry, the ion source uses ten parallel rf induction antennas to externally couple power into a toroidal discharge chamber. There are 20 ion beam extraction slits and 3 concentric electrode rings to shape and accelerate the ion beam into a titanium cone target. Fast neutron pulses are created by using a set of parallel-plate deflectors switching between +-1500 volts and deflecting the ion beams across a narrow slit. The generator is expected to achieve 5 ns neutron pulses at tritium ion beam energies between 80-120 kV. First experiments demonstrated ion source operation and successful beam pulsing

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

International Nuclear Information System (INIS)

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

2013-06-01

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

Development of network communication function for digitalized neutron flux monitoring instrument

International Nuclear Information System (INIS)

Li Kai; Zhang Liangju; Chen Xiaojun; Li Baoxiang

2002-01-01

It is essential for a digitalized Neutron Flux Monitoring Instrument to communicate with other parts of Instrumentation and Control System in a network environment, and it is fairly different from the case of traditional analogue nuclear instrumentation. How to satisfy all the requirements of different network structure and communication protocol, which might be adopted in different target nuclear power plant, is a key issue in the design and development of a digitalized neutron flux monitoring instrument. The author describes the overall communication scheme, mainly discusses the design idea and the requirements of the communication interfaces and the implementation of the RS-485 interface as an example of the digitalized neutron flux monitoring instrument, which is under development in the institute

Histological and Physiological Alterations Induced by Thermal Neutron Fluxes in Male Swiss Albino Mice

International Nuclear Information System (INIS)

Alzergy, A.A.; Emara, N.M.; Abd El-Latif, A.A.; El-Saady, S.M.M.; Emara, N.M.; Abd El-Latif, A.A.

2010-01-01

This work was performed to investigate the biological effects of different thermal neutron fluxes (0.27x10 8 , 0.52X10 8 , 1.089X10 8 , 2.16X10 8 and 4.32X10 8 ) on liver and kidney of male mice using neutron irradiation cell with Ra-Be(α,n) 3 mCi neutron source Leybold (55930). Exposed to various fluxes of thermal neutron induced a dramatic alterations in hepatic and renal functions as indicated by biochemical estimation of several parameters (bilirubin, SGT, and alkaline phosphate .Urea , total protein, and albumin) and confirmed by histological examinations Thermal neutron exposure induces marked increase in the serum activities of total bilirubin, alanine amino transaminase (ALT or GPT), and alkaline phosphate, whereas, urea, total protein and albumin showed marked decline as compared to control group. The physiological changes induced in thermal neutron fluxes dependent manner. Histopathological results revealed mild to severe type of necrosis, and degenerative changes in liver and kidney of male mice exposed to thermal neutron fluxes. Also it was found that the histopathological alterations induced in thermal neutron fluxes dependent manner. It was found that exposed to thermal neutron fluxes irradiation plays prominent role in the development of the physiological alterations in male Swiss albino mice. The Former up normalities as a result of the sequence events followed interaction of radiation with the former biological mater (liver and kidney) of male Swiss albino mice, which are, physical, physicochemical, chemical, and biological stages.

Neutron spectrum for neutron capture therapy in boron; Espectro de neutrones para terapia por captura de neutrones en boro

Energy Technology Data Exchange (ETDEWEB)

Medina C, D.; Soto B, T. G. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Programa de Doctorado en Ciencias Basicas, 98068 Zacatecas, Zac. (Mexico); Baltazar R, A. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Programa de Doctorado en Ingenieria y Tecnologia Aplicada, 98068 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: dmedina_c@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

2016-10-15

Glioblastoma multiforme is the most common and aggressive of brain tumors and is difficult to treat by surgery, chemotherapy or conventional radiation therapy. One treatment alternative is the Neutron Capture Therapy in Boron, which requires a beam modulated in neutron energy and a drug with {sup 10}B able to be fixed in the tumor. When the patients head is exposed to the neutron beam, they are captured by the {sup 10}B and produce a nucleus of {sup 7}Li and an alpha particle whose energy is deposited in the cancer cells causing it to be destroyed without damaging the normal tissue. One of the problems associated with this therapy is to have an epithermal neutrons flux of the order of 10{sup 9} n/cm{sup 2}-sec, whereby irradiation channels of a nuclear research reactor are used. In this work using Monte Carlo methods, the neutron spectra obtained in the radial irradiation channel of the TRIGA Mark III reactor are calculated when inserting filters whose position and thickness have been modified. From the arrangements studied, we found that the Fe-Cd-Al-Cd polyethylene filter yielded a ratio between thermal and epithermal neutron fluxes of 0.006 that exceeded the recommended value (<0.05), and the dose due to the capture gamma rays is lower than the dose obtained with the other arrangements studied. (Author)

Research on measurement of neutron flux in irradiation channels of research reactor

International Nuclear Information System (INIS)

Yin Zhitao; Lv Zheng; Wang Yulin; Zheng Wuqin

2014-01-01

Relative distribution of thermal neutron flux in the irradiation channel is measured by classical activation foil method. After that, on a representative point in the irradiation channel, neutron temperature and absolute neutron flux are also measured. Cadmium ratio correction method is used to check the experiment result in the end. Comparative analysis shows that the results from two different methods are agreed pretty well, which adds the credibility of experiment results. (authors)

NEUTRON SPECTRUM MEASUREMENTS USING MULTIPLE THRESHOLD DETECTORS

Energy Technology Data Exchange (ETDEWEB)

Gerken, William W.; Duffey, Dick

1963-11-15

From American Nuclear Society Meeting, New York, Nov. 1963. The use of threshold detectors, which simultaneously undergo reactions with thermal neutrons and two or more fast neutron threshold reactions, was applied to measurements of the neutron spectrum in a reactor. A number of different materials were irradiated to determine the most practical ones for use as multiple threshold detectors. These results, as well as counting techniques and corrections, are presented. Some materials used include aluminum, alloys of Al -Ni, aluminum-- nickel oxides, and magesium orthophosphates. (auth)

A deuteron linac for a high-intensity neutron source

International Nuclear Information System (INIS)

Staples, J.; Clark, D.; Grunder, H.; Lancaster, H.; Main, R.; Selph, F.; Smith, L.; Voelker, F.; Yourd, R.

1976-01-01

The preliminary design of an accelerator suitable to meet the flux and neutron energy requirements of a CTR materials test facility is presented. The specifications of such a facility call for a neutron flux of 10 14 n/cm 2 -sec distributed over an area of about 10 2 cm 2 with a neutron spectrum similar to that anticipated from a fusion reactor. A 30 MeV deuteron linac producing a CW beam of 125 mA, upgradable to 40 MeV at 250 mA at a later date, would produce the relatively broad spectrum of neutrons at the required intensity. Attention to the low-energy beam intercept on the drift tubes and diffusive losses producing neutrons and attendant activation problems are discussed

A technique of measuring neutron spectrum

International Nuclear Information System (INIS)

Sarkar, P.K.; Kirthi, K.N.; Ganguly, A.K.

1975-01-01

Plastic scintillators have been used to measure fast neutron spectrum from various sources. Gamma background discrimination has been done by selecting thin scintillators and thereby achieving near 100% transmission of Compton-edge electrons. The measured distribution has been unfolded by using an iterative least square technique. This gives minimum variance and maximum likelihood estimate with error minimised. Smoothening of the observed distribution has been done by Fourier and time series analyses. The method developed is applicable in principle for the determination of spectra of high energy neutrons ranging from 1 MeV to 70 MeV and beyond. However, practical application of the method is limited by the non-availability of cross-section data for various neutron induced reactions with carbon and hydrogen present in the polymerised polystyrene scintillator. This procedure has been adopted in the present work for spectral determination up to 14 MeV neutrons using the published value of reaction and scattering cross-sections. The spectra of Po-Be, Pu-Be, Am-Be and Ra-Be arrived at agree well with the published spectra obtained by other methods. Spectrum from spontaneous fission of Cf-252 have also been measured and fitted to the expression N(E)=Esup(1/2)exp(-E/T). The fitted parameter T and spectral details agree well with those in published literature

The epithermal neutron-flux distribution in the reactor RA - Vinca

International Nuclear Information System (INIS)

Marinkov, V.; Bikit, I.; Martinc, R.; Veskovic, M.; Slivka, J.; Vaderna, S.

1987-01-01

The distribution of the epithermal neutron flux in the reactor RA - Vinca has been measured by means of Zr - activation detectors. In the channel VK-8 non-homogeneous flux distribution was observed (author) [sr

Three new nondestructive evaluation tools based on high flux neutron sources

International Nuclear Information System (INIS)

Hubbard, C.R.; Raine, D.; Peascoe, R.; Wright, M.

1997-01-01

Nondestructive evaluation methods and systems based on specific attributes of neutron interactions with materials are being developed. The special attributes of neutrons are low attenuation in most engineering materials, strong interaction with low Z elements, and epithermal neutron absorption resonances. The three methods under development at ORNL include neutron based tomography and radiography; through thickness, nondestructive texture mapping; and internal, noninvasive temperature measurement. All three techniques require high flux sources such as the High Flux Isotope Reactor, a steady state source, or the Oak Ridge Electron Linear Accelerator, a pulsed neutron source. Neutrons are quite penetrating in most engineering materials and thus can be useful to detect internal flaws and features. Hydrogen atoms, such as in a hydrocarbon fuel, lubricant, or a metal hydride, are relatively opaque to neutron transmission and thus neutron based tomography/radiography is ideal to image their presence. Texture, the nonrandom orientation of crystalline grains within materials, can be mapped nondestructively using neutron diffraction methods. Epithermal neutron resonance absorption is being studied as a noncontacting temperature sensor. This paper highlights the underlying physics of the methods, progress in development, and the potential benefits for science and industry of the three facilities

The energy spectrum of delayed neutrons from thermal neutron induced fission of 235U and its analytical approximation

International Nuclear Information System (INIS)

Doroshenko, A.Yu.; Tarasko, M.Z.; Piksaikin, V.M.

2002-01-01

The energy spectrum of the delayed neutrons is the poorest known of all input data required in the calculation of the effective delayed neutron fractions. In addition to delayed neutron spectra based on the aggregate spectrum measurements there are two different approaches for deriving the delayed neutron energy spectra. Both of them are based on the data related to the delayed neutron spectra from individual precursors of delayed neutrons. In present work these two different data sets were compared with the help of an approximation by gamma-function. The choice of this approximation function instead of the Maxwellian or evaporation type of distribution is substantiated. (author)

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Localisation of a neutron source using measurements and calculation of the neutron flux and its gradient

CERN Document Server

Linden, P; Dahl, B; Pázsit, I; Por, G

1999-01-01

We have performed laboratory measurements of the neutron flux and its gradient in a static model experiment, similar to a model problem proposed in Pazsit (Ann. Nucl. Energy 24 (1997) 1257). The experimental system consists of a radioactive neutron source located in a water tank. The measurements are performed using a recently developed very small optical fibre detector. The measured values of the flux and its gradient are then used to test the possibility of localising the source. The results show that it is possible to measure the flux on the circumference of a circle and from this calculate the flux gradient vector. Then, by comparison of the measured quantities with corresponding MCNP calculations, both the direction and the distance to the source are found and thus the position of the source can be determined.

Fission ionisation chamber for the measurement of low fluxes of slow neutrons

International Nuclear Information System (INIS)

Weill, J.; Duchene, J.P.

1958-01-01

The ionisation chamber described is designed for the measurement of slow neutron fluxes of average or low intensity, in the presence, eventually, of very high gamma fluxes. The capture of a slow neutron by a fissile material, in this case 235 U, gives rise to fission fragments, high-energy particles which ionise the gas contained in the chamber. The neutrons are detected by virtue of the potential pulses, on the collecting electrode of the chamber, deriving from the collection of the ions produced by the fission fragments. The pulses are counted by means of a measuring system consisting of a preamplifier, a 2 Mc amplifier, a discriminator and an electronic scale with numerator or integrator. The general characteristics are as follows: sensitivity to neutrons: 0.07 kicks/n/cm 2 .s, sensitivity to γ rays: zero up to 3.10 4 R/H, a background noise at the normal discrimination voltage: 0.01 kicks/s, working H.T.: -500 V, capacity: 40 μμF, average height of pulse: 8 mV, limits of use: from several neutrons to 10 6 n/cm 2 .s. This chamber may be used in all cases where low fluxes of slow neutrons must be measured, especially in the presence of high gamma fluxes, for example in the checking of Pu concentrations in an extraction plant or for the starting up of reactors. (author) [fr

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

International Nuclear Information System (INIS)

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

1958-01-01

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

High-sensitive spectrometer of fast neutrons and the results of fast neutron background flux measurements at the Gallium-Germanium Solar Neutrino Experiment

CERN Document Server

Abdurashitov, J N; Kalikhov, A V; Matushko, V L; Shikhin, A A; Yants, V E; Zaborskaia, O S

2002-01-01

The principle of operation, design, registration system and main characteristics of a fast neutron spectrometer are described. The spectrometer is intended for direct measurements of ultra low fluxes of fast neutrons. It is sensitive to neutron fluxes of 10 sup - sup 7 cm sup - sup 2 s sup - sup 1 and lower. The detection efficiency of fast neutrons with simultaneous energy measurement was determined from Monte-Carlo simulation to be equal to 0.11+-0.01. The background counting rate in the detector corresponds to a neutron flux of (6.5+-2.1)x10 sup - sup 7 cm sup - sup 2 s sup - sup 1 in the range 1.0-11.0 MeV. The natural neutron flux from the surrounding mine rock at the depth of 4600 m of water equivalent was measured to be (7.3+-2.4)x10 sup - sup 7 cm sup - sup 2 s sup - sup 1 in the range 1.0-11.0 MeV. The flux of fast neutrons in the SAGE main room was measured to be <2.3x10 sup - sup 7 cm sup - sup 2 s sup - sup 1 in 1.0-11.0 MeV energy range.

Analysis of neutron spectrum effects on primary damage in tritium breeding blankets

Energy Technology Data Exchange (ETDEWEB)

Choi, Yong Hee, E-mail: cyh871@snu.ac.kr [School of Energy Systems Engineering, Seoul National University, 599 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Joo, Han Gyu [School of Energy Systems Engineering, Seoul National University, 599 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of)

2012-07-15

The effect of neutron spectrum on primary damages in a structural material of a tritium breeding blanket is investigated with a newly established recoil spectrum estimation system. First, a recoil spectrum generation code is developed to obtain the energy spectrum of primary knock-on atoms (PKAs) for a given neutron spectrum utilizing the latest ENDF/B data. Secondly, a method for approximating the high energy tail of the recoil spectrum is introduced to avoid expensive molecular dynamics calculations for high energy PKAs using the concept of recoil energy of the secondary knock-on atoms originated by the INtegration of CAScades (INCAS) model. Thirdly, the modified spectrum is combined with a set of molecular dynamics calculation results to estimate the primary damage parameters such as the number of surviving point defects. Finally, the neutron spectrum is varied by changing the material of the spectral shifter and the result in primary damage parameters is examined.

Analysis of neutron spectrum effects on primary damage in tritium breeding blankets

Science.gov (United States)

Choi, Yong Hee; Joo, Han Gyu

2012-07-01

The effect of neutron spectrum on primary damages in a structural material of a tritium breeding blanket is investigated with a newly established recoil spectrum estimation system. First, a recoil spectrum generation code is developed to obtain the energy spectrum of primary knock-on atoms (PKAs) for a given neutron spectrum utilizing the latest ENDF/B data. Secondly, a method for approximating the high energy tail of the recoil spectrum is introduced to avoid expensive molecular dynamics calculations for high energy PKAs using the concept of recoil energy of the secondary knock-on atoms originated by the INtegration of CAScades (INCAS) model. Thirdly, the modified spectrum is combined with a set of molecular dynamics calculation results to estimate the primary damage parameters such as the number of surviving point defects. Finally, the neutron spectrum is varied by changing the material of the spectral shifter and the result in primary damage parameters is examined.

A comparison in the reconstruction of neutron spectrums using classical iterative techniques

International Nuclear Information System (INIS)

Ortiz R, J. M.; Martinez B, M. R.; Vega C, H. R.; Gallego, E.

2009-10-01

One of the key drawbacks to the use of BUNKI code is that the process begins the reconstruction of the spectrum based on a priori knowledge as close as possible to the solution that is sought. The user has to specify the initial spectrum or do it through a subroutine called MAXIET to calculate a Maxwellian and a 1/E spectrum as initial spectrum. Because the application of iterative procedures by to resolve the reconstruction of neutron spectrum needs an initial spectrum, it is necessary to have new proposals for the election of the same. Based on the experience gained with a widely used method of reconstruction, called BUNKI, has developed a new computational tools for neutron spectrometry and dosimetry, which was first introduced, which operates by means of an iterative algorithm for the reconstruction of neutron spectra. The main feature of this tool is that unlike the existing iterative codes, the choice of the initial spectrum is performed automatically by the program, through a neutron spectra catalog. To develop the code, the algorithm was selected as the routine iterative SPUNIT be used in computing tool and response matrix UTA4 for 31 energy groups. (author)

Transmutation and neutron flux studies with fission chambers in the MEGAPIE target

International Nuclear Information System (INIS)

Chabod, S.; Foucher, Y.; Letourneau, A.; Marie, F.; Toussaint, J.C.; Blandin, Ch.; Chartier, F.; Fioni, G.

2005-01-01

Eight fission micro chambers will be inserted inside the central rod of the 1 MW liquid Pb-Bi MEGAPIE target in order to study the transmutation of two major actinides and to measure the neutron flux at a level of 5%. These chambers were developed for high neutron fluxes and tested at Laue Langevin Institute. (authors)

Neutron spectrum measurement by TOF

International Nuclear Information System (INIS)

Aizawa, Otohiko

1982-01-01

The TOF experiments by using various facilities are described. The steady neutron spectra in light water which contains non-1/V absorbing materials were measured by the TOF method at a LINAC facility. The results were compared with the calculations based on the Koppel-Haywood model and two others. The leakage neutron spectra from a heavy-water assembly were measured and compared with model calculations. The time-dependent energy spectra in a small graphite assembly were measured. For this measurement, a chopper system was also used. The two-region calculation explains the spectrum just after the neutron burst. The time-dependent spectra in a small Be assembly and in an assembly of coolant-moderator containing hydrogen were also measured. The calculations based on various models are in progress. The TOF experiments at the reactor-chopper facility were carried out for measuring the total cross sections of crystalline moderators, the thermal neutron total cross section of high temperature beryllium, the thermal neutron total cross sections of granular lead and high temperature liquid lead, and the angle-dependent scattering spectra. A pseudo-chopper was designed and constructed. The spectra of the neutron field for medical use were measured by the chopper-TOF system. The thermal neutron total cross sections of Fe, Zr, Nb and Mg were measured, and the results were compared with the calculations by THRUSH and UNCLE-TOM codes. The random-trigger TOF experiments were made by using Cf-252. (Kato, T.)

An evaluation of the spontaneous fission prompt neutron spectrum of 252Cf

International Nuclear Information System (INIS)

Bojkov, G.S.; Yurevich, V.I.

1987-01-01

An evaluation of the spontaneous fission prompt neutron spectrum of 252 Cf from 1 keV to 20 MeV is described. Variance-covariance matrices for a number of recent experimental data sets were constructed and used to evaluate the neutron spectrum following a Bayesian procedure. The evaluated spectrum is compared with various experimental and theoretical representations. (author)

Evaluation of thermal margin during BWR neutron flux oscillation

International Nuclear Information System (INIS)

Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

1992-01-01

Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

High-sensitive spectrometer of fast neutrons and the results of fast neutron background flux measurements at the gallium-germanium solar neutrino experiment (SAGE)

International Nuclear Information System (INIS)

Abdurashitov, D.N.; Gavrin, V.N.; Kalikhov, A.V.; Matushko, V.L.; Shikhin, A.A.; Yants, V.E.; Zaborskaya, O.S.

2001-01-01

The principle of operation, design, registration system and main characteristics of a fast neutron spectrometer are described. The spectrometer is intended for direct measurements of ultra-low fluxes of fast neutrons. It is sensitive to neutron fluxes of 10 -7 cm -2 · s -1 and lower. The detection efficiency of fast neutrons with simultaneous energy measurement was determined from Monte-Carlo simulation to be equal to 0.11 ± 0.01. The background counting rate in the detector corresponds to a neutron flux of (6.5 ± 2.1) · 10 -7 cm -2 · s -1 in the range of 1.0-11.0 MeV. The natural neutron flux from the surrounding mine rock at the depth of 4600 meters of water equivalent was measured to be (7.3 ± 2.4) · 10 -7 cm -2 · s -1 in the interval 1.0 -11.0 MeV. The flux of fast neutrons in the SAGE main room was measured to be 2.3 · 10 -7 cm -2 · s -1 in 1.0 - 11.0 MeV energy range

Comparison of Americium-Beryllium neutron spectrum obtained using activation foil detectors and NE-213 spectrometer

International Nuclear Information System (INIS)

Sunny, Sunil; Subbaiah, K.V.; Selvakumaran, T.S.

1999-01-01

Neutron spectrum of Americium - Beryllium (α,n) source is measured with two different spectrometers vis-a-vis activation foils (foil detectors) and NE-213 organic scintillator. Activity induced in the foils is measured with 4π-β-γ sodium iodide detector by integrating counts under photo peak and the saturation activity is found by correcting to elapsed time before counting. The data on calculated activity is fed into the unfolding code, SAND-II to obtain neutron spectrum. In the case of organic scintillator, the pulse height spectrum is obtained using MCA and this is processed with unfolding code DUST in order to get neutron spectrum. The Americium - Beryllium (α,n) neutron spectrum thus obtained by two different methods is compared. It is inferred that the NE-213 scintillator spectrum is in excellent agreement with the values beyond 1MeV. Neutron spectrum obtained by activation foils depends on initial guess spectrum and is found to be in reasonable agreement with NE-213 spectrum. (author)

Determination of photon detector coefficient in neutron flux study

International Nuclear Information System (INIS)

Jedol Dayou; Azali Muhammad; Abd Aziz Mohamed; Abdul Razak Daud; Elias Saniman

1995-01-01

The efficiency of photon detector which is normally used in neutron flux measurement has been studied. The data obtain have been plotted using mathematical models in the form of reciprocal, exponential and semilog equation and subsequently efficiency coefficient of the detector has been determined. Beside that, energy quadratic equation model has also been used. It has been found that equation model selection is very important in the detector efficiency coefficient determination. In the case of energy quadratic equation, it has been found that the selection of energy set influenced the result. It can be concluded that energy quadratic equation is the best model in the neutron flux determination

Frequency spectrum analysis of 252Cf neutron source based on LabVIEW

International Nuclear Information System (INIS)

Mi Deling; Li Pengcheng

2011-01-01

The frequency spectrum analysis of 252 Cf Neutron source is an extremely important method in nuclear stochastic signal processing. Focused on the special '0' and '1' structure of neutron pulse series, this paper proposes a fast-correlation algorithm to improve the computational rate of the spectrum analysis system. And the multi-core processor technology is employed as well as multi-threaded programming techniques of LabVIEW to construct frequency spectrum analysis system of 252 Cf neutron source based on LabVIEW. It not only obtains the auto-correlation and cross correlation results, but also auto-power spectrum,cross-power spectrum and ratio of spectral density. The results show that: analysis tools based on LabVIEW improve the fast auto-correlation and cross correlation code operating efficiency about by 25% to 35%, also verify the feasibility of using LabVIEW for spectrum analysis. (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.

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Prompt neutron energy spectrum for the spontaneous fission of Cf-252

International Nuclear Information System (INIS)

Blinov, M.V.; Boykov, G.S.; Vitenko, V.A.

1985-06-01

The prompt neutron spectrum for the spontaneous fission of Cf-252 has been measured in 0.01-10 MeV region by the time-of-flight technique using a fast ionization chamber with U-235 layers as the neutron detector. Numerical data for the spectrum are presented, with an error file. (author)

Neutron flux density data acquisition system based on LabVIEW

International Nuclear Information System (INIS)

Zhao Yanhui; Zhao Xiuliang; Li Zonglun; Liang Fengyan; Liu Liyan

2011-01-01

In the LabVIEW software, combined with PCI-6251 data acquisition card, VI of neutron flux density data acquisition is realized by DAQmx data acquisition functions. VI is composed of front panel and block diagram. The data collected can be displayed in the forms of the data curve and the data control, and saved in the form of files. Test results show that the frequency of output signal in NI ELVIS can be accurately measured by the system, realizing neutron flux density data acquisition based on LabVIEW. (authors)

On line local measurement of thermal neutron flux on BNCT patient using SPND

International Nuclear Information System (INIS)

Miller, M.E.; Sztejnberg Goncalves-Carralves, M.L.; Gonzalez, S.J.

2006-01-01

The first on-line neutron flux measurement on a patient using a self-powered neutron detector (SPND) was assessed during the fourth clinical trial of the Boron Neutron Capture Therapy (BNCT) Project carried out at the National Atomic Energy Commission of Argentina (CNEA) and the medical center Angel H. Roffo. The SPND was specially developed and assembled for BNCT by CNEA. Its small size, 1 cm sensible length and 1.9 mm diameter, allowed performing a localized measurement. Since the treated tumors were cutaneous melanomas of nodular type, the SPND was located on the patient's skin. The patient was exposed to three different and consecutive fields and in each of them the SPND was used to measure local thermal neutron fluxes at selected dosimetric reference points. The values of the measured fluxes agreed with the ones estimated by calculation. This trial also demonstrated the usefulness of the SPND for assessing flux on-line. (author)

Study of the neutron flux distribution in acylindrical reactor

Directory of Open Access Journals (Sweden)

A. Vidal-Ferràndiz

2017-08-01

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

Incident spectrum determination for time-of-flight neutron powder diffraction data analysis

International Nuclear Information System (INIS)

Hodges, J. P.

1998-01-01

Accurate characterization of the incident neutron spectrum is an important requirement for precise Rietveld analysis of time-of-flight powder neutron diffraction data. Without an accurate incident spectrum the calculated model for the measured relative intensities of individual Bragg reflections will possess systematic errors. We describe a method for obtaining an accurate numerical incident spectrum using data from a transmitted beam monitor

Simulation of e-{gamma}-n targets by FLUKA and measurement of neutron flux at various angles for accelerator based neutron source

Energy Technology Data Exchange (ETDEWEB)

Patil, B.J., E-mail: bjp@physics.unipune.ernet.i [Department of Physics, University of Pune, Pune 411 007 (India); Chavan, S.T.; Pethe, S.N.; Krishnan, R. [SAMEER, IIT Powai Campus, Mumbai 400 076 (India); Bhoraskar, V.N. [Department of Physics, University of Pune, Pune 411 007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ernet.i [Department of Physics, University of Pune, Pune 411 007 (India)

2010-10-15

A 6 MeV Race track Microtron (an electron accelerator) based pulsed neutron source has been designed specifically for the elemental analysis of short lived activation products where the low neutron flux requirement is desirable. The bremsstrahlung radiation emitted by impinging 6 MeV electron on the e-{gamma} primary target, was made to fall on the {gamma}-n secondary target to produce neutrons. The optimisation of bremsstrahlung and neutron producing target along with their spectra were estimated using FLUKA code. The measurement of neutron flux was carried out by activation of vanadium and the measured fluxes were 1.1878 x 10{sup 5}, 0.9403 x 10{sup 5}, 0.7428 x 10{sup 5}, 0.6274 x 10{sup 5}, 0.5659 x 10{sup 5}, 0.5210 x 10{sup 5} n/cm{sup 2}/s at 0{sup o}, 30{sup o}, 60{sup o}, 90{sup o}, 115{sup o}, 140{sup o} respectively. The results indicate that the neutron flux was found to be decreased as increase in the angle and in good agreement with the FLUKA simulation.

Corrections on energy spectrum and scattering for fast neutron radiography at NECTAR facility

International Nuclear Information System (INIS)

Liu Shuquan; Thomas, Boucherl; Li Hang; Zou Yubin; Lu Yuanrong; Guo Zhiyu

2013-01-01

Distortions caused by the neutron spectrum and scattered neutrons are major problems in fast neutron radiography and should be considered for improving the image quality. This paper puts emphasis on the removal of these image distortions and deviations for fast neutron radiography performed at the NECTAR facility of the research reactor FRM-Ⅱ in Technische Universitaet Mounchen (TUM), Germany. The NECTAR energy spectrum is analyzed and established to modify the influence caused by the neutron spectrum, and the Point Scattered Function (PScF) simulated by the Monte-Carlo program MCNPX is used to evaluate scattering effects from the object and improve image quality. Good analysis results prove the sound effects of the above two corrections. (authors)

Corrections on energy spectrum and scatterings for fast neutron radiography at NECTAR facility

Science.gov (United States)

Liu, Shu-Quan; Bücherl, Thomas; Li, Hang; Zou, Yu-Bin; Lu, Yuan-Rong; Guo, Zhi-Yu

2013-11-01

Distortions caused by the neutron spectrum and scattered neutrons are major problems in fast neutron radiography and should be considered for improving the image quality. This paper puts emphasis on the removal of these image distortions and deviations for fast neutron radiography performed at the NECTAR facility of the research reactor FRM- II in Technische Universität München (TUM), Germany. The NECTAR energy spectrum is analyzed and established to modify the influence caused by the neutron spectrum, and the Point Scattered Function (PScF) simulated by the Monte-Carlo program MCNPX is used to evaluate scattering effects from the object and improve image quality. Good analysis results prove the sound effects of the above two corrections.

Neutron spectrum measurement inside containment vessel at Kori nuclear power plant unit 1

Energy Technology Data Exchange (ETDEWEB)

Han, J. M.; Kim, T. W.; Kim, K. D.; Youn, C. H. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of)

2003-10-01

There would be a case for the radiation worker have to work inside of the containment vessel to inspect or repair reactor facilities. In this case, the information about distribution of neutron field is needed to estimate neutron exposure dose of worker. Neutron spectra were measured by BMS(Bonner Multisphere Spectrometer) at 4 points of 6 ft and 20 ft, 2 points of 44 ft, 5 points of 70 ft in containment vessel of Kori unit 1. From the calculation, the following results were obtained. Neutron fluxes of 6 ft were between 2.623 x 10{sup 2} and 2.746 x 10{sup 4} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 9.209 x 10{sup -6} and 3.377 x 10{sup -2} MeV, equivalent doses of neutron were between 0.025 and 2.675 mSv/hr. Neutron fluxes of 20 ft were between 1.771 x 10{sup 1} and 1.682 x 10{sup 3} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 6.084 x 10{sup -6} and 2.988 x 10{sup -1} MeV, equivalent doses of neutron were between 0.004 and 0.228 mSv/hr. Neutron fluxes of 44 ft were between 3.367 x 10{sup 2} and 3.483 x 10{sup 2} neutron / cm{sup 2}{center_dot}sec, average neutron energies were between 3.962 x 10{sup -2} and 7.360 x 10{sup -2} MeV, equivalent doses of neutron were between 0.069 and 0.089 mSv/hr. Neutron fluxes of 70 ft were between 4.553 x 10{sup 3} and 1.407 x 10{sup 4} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 3.668 x 10{sup -4} and 6.764 x 10{sup -2} MeV, equivalent doses of neutron were between 0.449 and 2.660 mSv/hr.

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

VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation

International Nuclear Information System (INIS)

Teuchert, E.; Haas, K.A.

1995-01-01

1 - Description of problem or function: VSOP (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D and 3-D diffusion calculation, depletion and shut-down features, in- core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (steady state and transient). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of thermal reactors, their fuel cycles, thermal transients, and safety assessment. Besides its use in research and development work for the Gas Cooled High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors, MAGNOX, and RBMK. 2 - Method of solution: The nuclear data for 184 isotopes are contained in two libraries. Fast and epithermal data in a 68 group GAM-I structure have been prepared mainly from ENDF/B-V and JEF-1. Resonance cross section data are given as input. Thermal data in a 30 group THERMOS structure have been collapsed from a 96 group THERMALIZATION (GATHER) library by a relevant neutron energy spectrum generated by the THERMALIZATION code. Graphite scattering matrices are based on the Young phonon spectrum in graphite. The neutron spectrum is calculated by a combination of the GAM and THERMOS codes. They can simultaneously be employed for many core regions differing in temperature, burnup, and fuel element lay-out. The thermal cell code THERMOS has been extended to treat the grain structure of the coated particles inside the fuel elements, and the epithermal GAM code uses modified cross sections for the resonance absorbers prepared from double heterogeneous ZUT-DGL calculations. The diffusion module of the code is CITATION with 2 - 8 energy groups. It provides the neutron

Simulations of muon-induced neutron flux at large depths underground

International Nuclear Information System (INIS)

Kudryavtsev, V.A.; Spooner, N.J.C.; McMillan, J.E.

2003-01-01

The production of neutrons by cosmic-ray muons at large depths underground is discussed. The most recent versions of the muon propagation code MUSIC, and particle transport code FLUKA are used to evaluate muon and neutron fluxes. The results of simulations are compared with experimental data

Broad Energy Range Neutron Spectroscopy using a Liquid Scintillator and a Proportional Counter: Application to a Neutron Spectrum Similar to that from an Improvised Nuclear Device.

Science.gov (United States)

Xu, Yanping; Randers-Pehrson, Gerhard; Marino, Stephen A; Garty, Guy; Harken, Andrew; Brenner, David J

2015-09-11

A novel neutron irradiation facility at the Radiological Research Accelerator Facility (RARAF) has been developed to mimic the neutron radiation from an Improvised Nuclear Device (IND) at relevant distances (e.g. 1.5 km) from the epicenter. The neutron spectrum of this IND-like neutron irradiator was designed according to estimations of the Hiroshima neutron spectrum at 1.5 km. It is significantly different from a standard reactor fission spectrum, because the spectrum changes as the neutrons are transported through air, and it is dominated by neutron energies from 100 keV up to 9 MeV. To verify such wide energy range neutron spectrum, detailed here is the development of a combined spectroscopy system. Both a liquid scintillator detector and a gas proportional counter were used for the recoil spectra measurements, with the individual response functions estimated from a series of Monte Carlo simulations. These normalized individual response functions were formed into a single response matrix for the unfolding process. Several accelerator-based quasi-monoenergetic neutron source spectra were measured and unfolded to test this spectroscopy system. These reference neutrons were produced from two reactions: T(p,n) 3 He and D(d,n) 3 He, generating neutron energies in the range between 0.2 and 8 MeV. The unfolded quasi-monoenergetic neutron spectra indicated that the detection system can provide good neutron spectroscopy results in this energy range. A broad-energy neutron spectrum from the 9 Be(d,n) reaction using a 5 MeV deuteron beam, measured at 60 degrees to the incident beam was measured and unfolded with the evaluated response matrix. The unfolded broad neutron spectrum is comparable with published time-of-flight results. Finally, the pair of detectors were used to measure the neutron spectrum generated at the RARAF IND-like neutron facility and a comparison is made to the neutron spectrum of Hiroshima.

Neutron background measurements in the underground laboratory of Modane

International Nuclear Information System (INIS)

Chazal, V.; Chambon, B.; De Jesus, M.; Drain, D.; Pastor, C.; Vagneron, L.; Brissot, R.; Cavaignac, J.F.; Stutz, A.; Giraud-Heraud, Y.

1997-07-01

Measurements of the background neutron environment, at a depth of 1780 m (4800 mWe) in the Underground Laboratory of Modane (L.S.M) are reported. Using a 6 Li liquid scintillator, the energy spectrum of the fast neutron flux has been determined. Monte-Carlo calculations of the (α,n) and spontaneous fission processes in the surrounding rock has been performed and compared to the experimental result. In addition, using two 3 He neutron counters, the thermal neutron flux has been measured. (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

Development of a position-sensitive fission counter and measurement of neutron flux distributions

International Nuclear Information System (INIS)

Yamagishi, Hideshi; Soyama, Kazuhiko; Kakuta, Tsunemi

2001-08-01

A position-sensitive fission counter (PSFC) that operates in high neutron flux and high gamma-ray background such as at the side of a power reactor vessel has been developed. Neutron detection using the PSFC with a solenoid electrode is based on a delay-line method. The PSFC that has the outer diameter of 25 mm and the sensitive length of 1000 mm was manufactured for investigation of the performances. The PSFC provided output current pulses that were sufficiently higher than the alpha noise, though the PSFC has a solenoid electrode and large electrode-capacitance. The S/N ratio of PSFC outputs proved to be higher than that of ordinary fission counters with 200 mm sensitive length. A performance test to measure neutron flux distributions by a neutron measuring system with the PSFC was carried out by the side of a graphite pile, W2.4 x H1.4 x L1.2 m, with neutron sources, Am-Be 370 GBq x 2. It was confirmed that the neutron flux distribution was well measured with the system. (author)

Spectrum of neutrons leaking from an iron sphere with a central 14 MeV neutron source

International Nuclear Information System (INIS)

Borisov, A.A.; Zagryadskij, V.A.; Chuvilin, D.Yu.; Kralik, M.; Pulpan, J.; Tichy, M.

1991-01-01

Following a review of the present state of nuclear data requisite for the calculation of the transport of 14 MeV neutrons through iron of natural isotopic composition, the results are given of the calculation of the energy spectrum of such neutrons after their passage through an iron sphere 240 mm o.d. and 90 mm i.d., the neutron source being accommodated in the centre of the sphere. The calculations were made using the one-dimensional code BLANK working with the nuclear data libraries ENDL-75, ENDL-83, ENDL/B-IV, JENDL-2 and BROND, and using the three-dimensional code BRAND with the library ENDL-78. The calculated spectra were compared with the experimental spectrum measured at a distance of 3 m from the sphere by means of an NE-213 scintillator, which records reflected protons. The reflected proton spectrum was processed by the matrix method (program FORIST), and the result was normalized to one neutron emitted by the source, as were the calculated spectra. The comparison demonstrates that the experiment is best fitted by the spectrum calculated by using the library JENDL-2, where the integrals of the observed and calculated spectra over the 1-15 MeV range differ as little as approximately 10%. (author). 3 figs., 5 tabs., 16 refs

Pursuing nuclear energy with no nuclear contamination - from neutron flux reactor to deuteron flux reactor

International Nuclear Information System (INIS)

Li, X. Z.; Wei, Q. M.; Liu, B.; Zhu, X. G.; Ren, S. L.

2007-01-01

Pursuing nuclear energy with no nuclear contamination has been a long endeavor since the first fission reactor in 1942. Four major concepts have been the key issues: i.e. resonance, negative feed back, self-sustaining, nuclear radiation. When nuclear energy was just discovered in laboratory, the key issue was to enlarge it from the micro-scale to the macro-scale. Slowing-down the neutrons was the key issue to enhance the fission cross-section in order to build-up the neutron flux through the chain-reactions using resonance between neutron and fissile materials. Once the chain-reaction was realized, the negative feed-back was the key issue to keep the neutron flux at the allowable level. The negative reaction coefficient was introduced by the thermal expansion, and the resonant absorption in cadmium or boron was used to have a self-sustaining fission reactor with neutron flux. Then the strong neutron flux became the origin of all nuclear contamination, and a heavy shielding limits the application of the nuclear energy. The fusion approach to nuclear energy was much longer; nevertheless, it evolved with the similar issues. The resonance between deuteron and triton was resorted to enlarge the fusion cross section in order to keep a self-sustaining hot plasma. However, the 14 MeV neutron emission became the origin of all nuclear contamination again. Deuteron plus helium-3 fusion reaction was proposed to avoid neutron emission although there are two more difficulties: the helium-3 is supposed to be carried back from the moon; and much more higher temperature plasma has to be confined while 50 years needed to realized the deuteron-triton plasma already. Even if deuteron plus helium-3 fusion plasma might be realized in a much higher temperature plasma, we still have the neutron emission from the deuteron-deuteron fusion reaction in the deuteron plus helium-3 fusion plasma. Polarized deuteron-deuteron fusion reaction was proposed early in 1980's to select the neutron

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

Science.gov (United States)

Khattab, K; Sulieman, I

2009-04-01

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

Self-Powered Neutron Detector Qualification for Absolute On-Line In-Pile Neutron Flux Measurements in BR2

Science.gov (United States)

Vermeeren, L.; Wéber, M.

2003-06-01

A set of ten Self-Powered Neutron Detectors with Co, Rh and Ag emitters has been irradiated in several channels of the BR2 research reactor at SCK•CEN aiming at a comparison of their performance as thermal neutron flux detectors under various conditions. To allow for a correct interpretation of their signals, all detector sensitivity contributions (prompt and delayed) were calculated using a dedicated Monte Carlo model. The various contributions were also measured separately; the agreement between calculated and experimental data, including data from activation dosimetry, was excellent. Detailed neutron flux profiles were obtained from the SPND data, after correction for the finite detector lengths and for the slow response of delayed SPNDs.

Theoretical analysis of integral neutron transport equation using collision probability method with quadratic flux approach

International Nuclear Information System (INIS)

Shafii, Mohammad Ali; Meidianti, Rahma; Wildian,; Fitriyani, Dian; Tongkukut, Seni H. J.; Arkundato, Artoto

2014-01-01

Theoretical analysis of integral neutron transport equation using collision probability (CP) method with quadratic flux approach has been carried out. In general, the solution of the neutron transport using the CP method is performed with the flat flux approach. In this research, the CP method is implemented in the cylindrical nuclear fuel cell with the spatial of mesh being conducted into non flat flux approach. It means that the neutron flux at any point in the nuclear fuel cell are considered different each other followed the distribution pattern of quadratic flux. The result is presented here in the form of quadratic flux that is better understanding of the real condition in the cell calculation and as a starting point to be applied in computational calculation

Theoretical analysis of integral neutron transport equation using collision probability method with quadratic flux approach

Energy Technology Data Exchange (ETDEWEB)

Shafii, Mohammad Ali, E-mail: mashafii@fmipa.unand.ac.id; Meidianti, Rahma, E-mail: mashafii@fmipa.unand.ac.id; Wildian,, E-mail: mashafii@fmipa.unand.ac.id; Fitriyani, Dian, E-mail: mashafii@fmipa.unand.ac.id [Department of Physics, Andalas University Padang West Sumatera Indonesia (Indonesia); Tongkukut, Seni H. J. [Department of Physics, Sam Ratulangi University Manado North Sulawesi Indonesia (Indonesia); Arkundato, Artoto [Department of Physics, Jember University Jember East Java Indonesia (Indonesia)

2014-09-30

Theoretical analysis of integral neutron transport equation using collision probability (CP) method with quadratic flux approach has been carried out. In general, the solution of the neutron transport using the CP method is performed with the flat flux approach. In this research, the CP method is implemented in the cylindrical nuclear fuel cell with the spatial of mesh being conducted into non flat flux approach. It means that the neutron flux at any point in the nuclear fuel cell are considered different each other followed the distribution pattern of quadratic flux. The result is presented here in the form of quadratic flux that is better understanding of the real condition in the cell calculation and as a starting point to be applied in computational calculation.

Calculation of neutron and gamma-ray flux-to-dose-rate conversion factors

International Nuclear Information System (INIS)

Kwon, S.G.; Lee, S.Y.; Yook, C.C.

1981-01-01

This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute (ANSI) N666. These data are used to calculate the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoenergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions. (author)

Measurement of the North-South asymmetry in the solar proton albedo neutron flux

International Nuclear Information System (INIS)

Ifedili, S.O.

1979-01-01

The solar proton albedo neutron flux in the range 10 -2 --10 7 eV measured by a neutron detector on board the Ogo 6 satellite was examined for north-south asymmetry. For the solar proton event of December 19, 1969, the S/N ratio of the solar proton albedo neutron rate at geomagnetic latitude lambda>70 0 was 1.61 +- 0.27 during the event, while for the November 2, 1969, event at 40 0 0 and altitudes ranging from 700 km to 800 km the solar proton albedo neutron rate was 0.40 +- 0.10 count/s in the north and 0.00 +- 0.10 count/s in the south. During the solar proton event of December 18, 1969, the N/S ratio of the solar proton albedo neutron rate at lambda>70 0 was 1.00 +- 0.26. The results are consistent with the expected N-S asymmetry in the solar proton flux. An interplanetary proton anisotropy with the interplanetary magnetic field polarity away from the sun corresponded to larger fluxes of solar proton albedo neutrons at the north polar cap than at the south, while an interplanetary proton anisotropy with the interplanetary magnetic field polarity toward the sun corresponded to larger fluxes of solar proton albedo neutrons at the south polar cap than at the north. This evidence favors the direct access of solar protons to the earth's polar caps via the merged interplanetary and geomagnetic field lines

Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Parra M, M. A.

2014-07-01

This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au{sup 197}, Ni{sup 58}, In{sup 115}, Mg{sup 24}, Al{sup 27}, Fe{sup 58}, Co{sup 59} and Cu{sup 63}, they were selected to cover the full range the energies (10{sup -10} to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or

Contributions to the study of fast neutron spectrum in the 10 keV - 3 MeV range

International Nuclear Information System (INIS)

Garlea, I.

1979-01-01

The main objective of the work presented in this thesis was to create a fast neutron spectrum corresponding to the conditions required for a reference neutron field. The reference system for the fast neutron dosimetry in reactors, which the author promoted, is referred to as ΣΣ-ITN in the books. The conditions for introducing the ΣΣ systems into the thermal columns have been determined. The original contribution consists in determining the Westcott parameters of reactions 151 Eu(n,γ) 152 Eu and 176 Lu(n,γ) 17 +H7Lu used as thermal spectrum factors. The neutron description of the spectrum in cavity ΣΣ revealed that it is a Maxwell thermal spectrum displaying a temperature of 305+-7 deg C and a very small epithermal component (phisub(epi)/phisub(thermal) =4,5.10 -4 ). Better methods for determining reaction absolute rates resulted in less errors in calculating the microscopic integral sections mediated on the ΣΣ spectrum; there are under 5% errors for the fission cross sections and between 3% and 8% errors for the activating ones. The section values determined by the author have been included into the EXFOR library (IAEA); they are considered as reference measuremtns for the nuclear data improvement program. Testing the proposed method for the TRIGA on the ΣΣ-INT system proved that the multiple foil method provides correct results for both describing the spectral shape and for obtaining absolute values of the flux. Taking into account that the ΣΣ-ITN spectrum is a rapid one, the proposed method could not be tested within the low energy thermal and epithermal domain. For testing the method on an operational reactor, the core of the VVR-S IFIN reactor was employed. Due to the spectral structure of this reactor, it was possible to test the procedure within the whole energy range. In this view, the 5/10 core channel was selected which is similar to the channel required for measurements in the TRIGA-ROMANIA reactor. The absolute spectrum values are given in

Simultaneous neutron and gamma spectrum adjustment

International Nuclear Information System (INIS)

Remec, I.

1996-01-01

The spectrum adjustment procedure was extended to simultaneous neutron and gamma spectrum adjustment, and the feasibility of this technique is demonstrated in the analysis of HFIR dosimetry experiments. Conditions in which gamma rays may contribute considerably to radiation damage in steels are discussed. Beryllium helium accumulation fluence monitors (HAFMs) were found to be good monitors in gamma fields of intensities high enough to contribute to steel embrittlement. Use of 237 Np, 238 U, and 9 Be HAFM as gamma dosimeters is proposed for high-dose irradiations in high-energy, high-intensity gamma fields

Effects of neutron spectrum and external neutron source on neutron multiplication parameters in accelerator-driven system

International Nuclear Information System (INIS)

Shahbunder, Hesham; Pyeon, Cheol Ho; Misawa, Tsuyoshi; Lim, Jae-Yong; Shiroya, Seiji

2010-01-01

The neutron multiplication parameters: neutron multiplication M, subcritical multiplication factor k s , external source efficiency φ*, play an important role for numerical assessment and reactor power evaluation of an accelerator-driven system (ADS). Those parameters can be evaluated by using the measured reaction rate distribution in the subcritical system. In this study, the experimental verification of this methodology is performed in various ADS cores; with high-energy (100 MeV) proton-tungsten source in hard and soft neutron spectra cores and 14 MeV D-T neutron source in soft spectrum core. The comparison between measured and calculated multiplication parameters reveals a maximum relative difference in the range of 6.6-13.7% that is attributed to the calculation nuclear libraries uncertainty and accuracy for energies higher than 20 MeV and also dependent on the reaction rate distribution position and count rates. The effects of different core neutron spectra and external neutron sources on the neutron multiplication parameters are discussed.

Little Boy neutron spectrum below 3 MeV

International Nuclear Information System (INIS)

Evans, A.E.; Bennett, E.F.; Yule, T.J.

1984-01-01

The leakage neutron spectrum from the Little Boy replica has been measured from 12 keV to 3 MeV using a high-resolution 3 He ionization chamber, and from 1 keV to 3 MeV using proton-recoil proportional counters. The 3 He-spectrometer measurements were made at distances of 0.75 and 2.0 m from the active center and at angles of 0 0 , 45 0 , and 90 0 with respect to the axis of the assembly. Proton-recoil measurments were made at 90 0 to the assembly axis at distances of 0.75 and 2.0 m, with a shielded measurement made at 2.0 m to estimate background due to scattering. The 3 He spectrometer was calibrated at Los Alamos using monoenergetic 7 Li(p,n) 7 Be neutrons to generate a family of response functions. The proton-recoil counters were calibrated at Argonne by studying the capture of thermal neutrons by nitrogen in the counters, by observation of the 24-keV neutron resonance in iron, and by relating to the known hydrogen content of the counters. The neutron spectrum from Little Boy was found to be highly structured, with peaks corresponding to minima in the iron total neutron cross section. In particular, influence of the 24-keV iron window was evident in both sets of spectra. The measurements provide information for dosimetry calculations and also a valuable intercomparison of neutron spectrometry using the two different detector types. Spectra measured with both detectors are in essential agreement. 8 references, 7 figures, 2 tables

APPLE, Plot of 1-D Multigroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN

International Nuclear Information System (INIS)

Kawasaki, Hiromitsu; Seki, Yasushi

1983-01-01

A - Description of problem or function: The APPLE-2 code has the following functions: (1) It plots multi-group energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT-3.5, and MORSE. (2) It gives an overview plot of multi-group neutron fluxes calculated by ANISN and DOT-3.5. The scalar neutron flux phi(r,E) is plotted with the spatial parameter r linear along the Y-axis, logE along the X-axis and log phi(r,E) in the Z direction. (3) It calculates the spatial distribution and region volume integrated values of reaction rates using the scalar flux calculated with ANISN and DOT-3.5. (4) Reaction rate distribution along the R or Z direction may be plotted. (5) An overview plot of reaction rates or scalar fluxes summed over specified groups may be plotted. R(ri,zi) or phi(ri,zi) is plotted with spatial parameters r and z along the X- and Y-axes in an orthogonal coordinate system. (6) Angular flux calculated by ANISN is rearranged and a shell source at any specified spatial mesh point may be punched out in FIDO format. The shell source obtained may be employed in solving deep penetration problems with ANISN, when the entire reactor system is divided into two or more parts and the neutron fluxes in two adjoining parts are connected by using the shell source. B - Method of solution: (a) The input data specification is made as simple as possible by making use of the input data required in the radiation transport code. For example, geometry related data in ANISN and DOT are transmitted to APPLE-2 along with scalar flux data so as to reduce duplicity and errors in reproducing these data. (b) Most the input data follow the free form FIDO format developed at Oak Ridge National Laboratory and used in the ANISN code. Furthermore, the mixture specifying method used in ANISN is also employed by APPLE-2. (c) Libraries for some standard response functions required in fusion reactor design have been prepared and are made available to users of the 42-group neutron

Monitoring the fast neutrons in a high flux: The case for 242Pu fission chambers

International Nuclear Information System (INIS)

Filliatre, P.; Jammes, C.; Oriol, L.; Geslot, B.; Vermeeren, L.

2009-01-01

Fission chambers are widely used for on-line monitoring of neutron fluxes in irradiation reactors. A selective measurement of a component of interest of the neutron flux is possible in principle thanks to a careful choice of the deposit material. However, measuring the fast component is challenging when the flux is high (up to 10 15 n/cm 2 /s) with a significant thermal component. The main problem is that the isotopic content of a material selected for its good response to fast neutrons evolves with irradiation, so that the material is more and more sensitive to thermal neutrons. Within the framework of the FNDS (Fast Neutron Detector System) project, we design tools that simulate the evolution of the isotopic composition and fission rate for several deposits under any given flux. In the case of a high flux with a significant thermal component, 242 Pu is shown after a comprehensive study of all possibilities to be the best choice for measuring the fast component, as long as its purity is sufficient. If an estimate of the thermal flux is independently available, one can correct the signal for that component. This suggests a system of two detectors, one of which being used for such a correction. It is of very high interest when the detectors must be operated up to a high neutron fluence. (authors)

Mitigation of end flux peaking in CANDU fuel bundles using neutron absorbers

Energy Technology Data Exchange (ETDEWEB)

Pierce, D.; Chan, P.K., E-mail: dylan.pierce@rmc.ca [Royal Military College of Canada, Kingston ON, (Canada); Shen, W. [Canadian Nuclear Safety Commission, Ottawa ON, (Canada)

2015-07-01

End flux peaking (EFP) is a phenomenon where a region of elevated neutron flux occurs between two adjoining fuel bundles. These peaks lead to an increase in fission rate and therefore greater heat generation. It is known that addition of neutron absorbers into fuel bundles can help mitigate EFP, yet implementation in Canada Deuterium Uranium (CANDU) type reactors using natural uranium fuel has not been pursued. Monte Carlo N-Particle code (MCNP) 6.1 was used to simulate the addition of a small amount of neutron absorbers strategically within the fuel pellets. This paper will present some preliminary results collected thus far. (author)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

The Neutron-Gamma Pulse Shape Discrimination Method for Neutron Flux Detection in the ITER

International Nuclear Information System (INIS)

Xu Xiufeng; Li Shiping; Cao Hongrui; Yin Zejie; Yuan Guoliang; Yang Qingwei

2013-01-01

The neutron flux monitor (NFM), as a significant diagnostic system in the International Thermonuclear Experimental Reactor (ITER), will play an important role in the readings of a series of key parameters in the fusion reaction process. As the core of the main electronic system of the NFM, the neutron-gamma pulse shape discrimination (n-γ PSD) can distinguish the neutron pulse from the gamma pulse and other disturbing pulses according to the thresholds of the rising time and the amplitude pre-installed on the board, the double timing point CFD method is used to get the rising time of the pulse. The n-γ PSD can provide an accurate neutron count. (magnetically confined plasma)

DEMONR, Monte-Carlo Shielding Calculation for Neutron Flux and Neutron Spectra, Teaching Program

International Nuclear Information System (INIS)

Courtney, J. C.

1987-01-01

1 - Description of problem or function: DEMONR treats the behavior of neutrons in a slab shield. It is frequently used as a teaching tool. 2 - Method of solution: An unbiased Monte Carlo code calculates the number, energy, and direction of neutrons that penetrate or are reflected from a shield. 3 - Restrictions on the complexity of the problem: Only one shield may be used in each problem. The shield material may be a single element or a homogeneous mixture of elements with a single effective atomic weight. Only elastic scattering and neutron capture processes are allowed. The source is a point located on one face of the slab. It provides a cosine distribution of current. Monoenergetic or fission spectrum neutrons may be selected

A novel neutron energy spectrum unfolding code using particle swarm optimization

International Nuclear Information System (INIS)

Shahabinejad, H.; Sohrabpour, M.

2017-01-01

A novel neutron Spectrum Deconvolution using Particle Swarm Optimization (SDPSO) code has been developed to unfold the neutron spectrum from a pulse height distribution and a response matrix. The Particle Swarm Optimization (PSO) imitates the bird flocks social behavior to solve complex optimization problems. The results of the SDPSO code have been compared with those of the standard spectra and recently published Two-steps Genetic Algorithm Spectrum Unfolding (TGASU) code. The TGASU code have been previously compared with the other codes such as MAXED, GRAVEL, FERDOR and GAMCD and shown to be more accurate than the previous codes. The results of the SDPSO code have been demonstrated to match well with those of the TGASU code for both under determined and over-determined problems. In addition the SDPSO has been shown to be nearly two times faster than the TGASU code. - Highlights: • Introducing a novel method for neutron spectrum unfolding. • Implementation of a particle swarm optimization code for neutron unfolding. • Comparing results of the PSO code with those of recently published TGASU code. • Match results of the PSO code with those of TGASU code. • Greater convergence rate of implemented PSO code than TGASU code.

Self-shielding coefficient and thermal flux depression factor of voluminous sample in neutron activation analysis

International Nuclear Information System (INIS)

Noorddin Ibrahim; Rosnie Akang

2009-01-01

Full text: One of the major problems encountered during the irradiation of large inhomogeneous samples in performing activation analysis using neutron is the perturbation of the neutron field due to absorption and scattering of neutron within the sample as well as along the neutron guide in the case of prompt gamma activation analysis. The magnitude of this perturbation shown by self-shielding coefficient and flux depression depend on several factors including the average neutron energy, the size and shape of the sample, as well as the macroscopic absorption cross section of the sample. In this study, we use Monte Carlo N-Particle codes to simulate the variation of neutron self-shielding coefficient and thermal flux depression factor as a function of the macroscopic thermal absorption cross section. The simulation works was carried out using the high performance computing facility available at UTM while the experimental work was performed at the tangential beam port of Reactor TRIGA PUSPATI, Malaysia Nuclear Agency. The neutron flux measured along the beam port is found to be in good agreement with the simulated data. Our simulation results also reveal that total flux perturbation factor decreases as the value of absorption increases. This factor is close to unity for low absorbing sample and tends towards zero for strong absorber. In addition, sample with long mean chord length produces smaller flux perturbation than the shorter mean chord length. When comparing both the graphs of self-shielding factor and total disturbance, we can conclude that the total disturbance of the thermal neutron flux on the large samples is dominated by the self-shielding effect. (Author)

Measurement of Neutron Energy Spectrum Emitted by Cf-252 Source Using Time-of-Flight Method

Energy Technology Data Exchange (ETDEWEB)

Lee, Cheol Ho; Son, Jaebum; Kim, Tae Hoon; Lee, Sangmin; Kim, Yong-Kyun [Hanyang University, Seoul (Korea, Republic of)

2016-10-15

The techniques proposed to detect the neutrons usually require the detection of a secondary recoiling nucleus in a scintillator (or other type of detector) to indicate the rare collision of a neutron with a nucleus. This is the same basic technique, in this case detection of a recoil proton that was used by Chadwick in the 1930 s to discover and identify the neutron and determine its mass. It is primary technique still used today for detection of fast neutron, which typically involves the use of a hydrogen based organic plastic or liquid scintillator coupled to a photo-multiplier tube. The light output from such scintillators is a function of the cross section and nuclear kinematics of the n + nucleus collision. With the exception of deuterated scintillators, the scintillator signal does not necessarily produce a distinct peak in the scintillator spectrum directly related to the incident neutron energy. Instead neutron time-of-flight (TOF) often must be utilized to determine the neutron energy, which requires generation of a prompt start signal from the nuclear source emitting the neutrons. This method takes advantage of the high number of prompt gamma rays. The Time-of-Flight method was used to measure neutron energy spectrum emitted by the Cf-252 neutron source. Plastic scintillator that has a superior discrimination ability of neutron and gamma-ray was used as a stop signal detector and liquid scintillator was used as a stat signal detector. In experiment, neutron and gamma-ray spectrum was firstly measured and discriminated using the TOF method. Secondly, neutron energy spectrum was obtained through spectrum analysis. Equation of neutron energy spectrum that was emitted by Cf-252 source using the Gaussian fitting was obtained.

Neutron flux in a periodical slab geometry (1960)

International Nuclear Information System (INIS)

Lamare, J. de; Mathelot, P.; Cadhilac, M.

1960-01-01

In the present report, we explain an original method to perform exact calculations of neutron flux in either of two geometries: a slab surrounded by an infinite multiplying medium or a periodical, one dimensional array of two different media. (author) [fr

Refinements in the Los Alamos model of the prompt fission neutron spectrum

Energy Technology Data Exchange (ETDEWEB)

Madland, D.G., E-mail: dgm@lanl.gov; Kahler, A.C.

2017-01-15

This paper presents a number of refinements to the original Los Alamos model of the prompt fission neutron spectrum and average prompt neutron multiplicity as derived in 1982. The four refinements are due to new measurements of the spectrum and related fission observables many of which were not available in 1982. They are also due to a number of detailed studies and comparisons of the model with previous and present experimental results including not only the differential spectrum, but also integral cross sections measured in the field of the differential spectrum. The four refinements are (a) separate neutron contributions in binary fission, (b) departure from statistical equilibrium at scission, (c) fission-fragment nuclear level-density models, and (d) center-of-mass anisotropy. With these refinements, for the first time, good agreement has been obtained for both differential and integral measurements using the same Los Alamos model spectrum.

Continuous energy Neutron Transport Monte Carlo Simulator Project: Decomposition of the neutron energy spectrum by target nuclei tagging

Energy Technology Data Exchange (ETDEWEB)

Barcellos, Luiz Felipe F.C.; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B., E-mail: luizfelipe.fcb@gmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Grupo de Estudos Nucleares; Leite, Sergio Q. Bogado, E-mail: sbogado@ibest.com.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

In this work a Monte Carlo simulator with continuous energy is used. This simulator distinguishes itself by using the sum of three probability distributions to represent the neutron spectrum. Two distributions have known shape, but have varying population of neutrons in time, and these are the fission neutron spectrum (for high energy neutrons) and the Maxwell-Boltzmann distribution (for thermal neutrons). The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. It is common practice in neutron transport calculations, e.g. multi-group transport, to consider that the neutrons only lose energy with each scattering reaction and then to use a thermal group with a Maxwellian distribution. Such an approximation is valid due to the fact that for fast neutrons up-scattering occurrence is irrelevant, being only appreciable at low energies, i.e. in the thermal energy region, in which it can be regarded as a Maxwell-Boltzmann distribution for thermal equilibrium. In this work the possible neutron-matter interactions are simulated with exception of the up-scattering of neutrons. In order to preserve the thermal spectrum, neutrons are selected stochastically as being part of the thermal population and have an energy attributed to them taken from a Maxwellian distribution. It is then shown how this procedure can emulate the up-scattering effect by the increase in the neutron population kinetic energy. Since the simulator uses tags to identify the reactions it is possible not only to plot the distributions by neutron energy, but also by the type of interaction with matter and with the identification of the target nuclei involved in the process. This work contains some preliminary results obtained from a Monte Carlo simulator for neutron transport that is being developed at Federal University of Rio Grande do Sul. (author)

Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux

Science.gov (United States)

Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; Boukhal, H.; Jaï, Otman

2018-06-01

The MNSR is a pool type research reactor, which is difficult to model because of the importance of neutron leakage. The aim of this study is to evaluate a 2-D transport model for the reactor compatible with the latest release of the DRAGON code and 3-D diffusion of the DONJON code. DRAGON code is then used to generate the group macroscopic cross sections needed for full core diffusion calculations. The diffusion DONJON code, is then used to compute the effective multiplication factor (keff), the feedback reactivity coefficients and neutron flux which account for variation in fuel and moderator temperatures as well as the void coefficient have been calculated using the DRAGON and DONJON codes for the MNSR research reactor. The cross sections of all the reactor components at different temperatures were generated using the DRAGON code. These group constants were used then in the DONJON code to calculate the multiplication factor and the neutron spectrum at different water and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. Finally, Good agreements between the calculated and measured have been obtained for every of the feedback reactivity coefficients and neutron flux.

The increase of the subthermal neutron flux by using a cold neutron source at the FRG-1

International Nuclear Information System (INIS)

Krueger, A.; Turgut, M.H.

1986-01-01

The increase of the subthermal neutron flux (wavelength range 4-6 A by a cold neutron source (CNS) at a radial beam tube of the FRG-1 reactor is investigated in combination with different reflectors (H 2 O, C, Be, D 2 O). Advantage factors on the basis of the directed neutron flux, resulting from the use of the CNS, are calculated for various configurations. In addition, the influence of different scattering models (gas, Koppel/Young) for the CNS, group structure, and structural materials are described. Finally, the CNS assembly which is going to be installed at the FRG-1 is treated in detail. For the calculations the transport code NEUTRA and the spectral code GGC-4 are used. (orig.) [de

Spectrum shaping of accelerator-based neutron beams for BNCT

CERN Document Server

Montagnini, B; Esposito, J; Giusti, V; Mattioda, F; Varone, R

2002-01-01

We describe Monte Carlo simulations of three facilities for the production of epithermal neutrons for Boron Neutron Capture Therapy (BNCT) and examine general aspects and problems of designing the spectrum-shaping assemblies to be used with these neutron sources. The first facility is based on an accelerator-driven low-power subcritical reactor, operating as a neutron amplifier. The other two facilities have no amplifier and rely entirely on their primary sources, a D-T fusion reaction device and a conventional 2.5 MeV proton accelerator with a Li target, respectively.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Fast neutron dosimetry in research reactors; Dosimetrie en neutrons rapides dans les reacteurs de recherche

Energy Technology Data Exchange (ETDEWEB)

Eckert, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

This work chiefly concerns the measurement of fast neutron fluxes by means of threshold detectors. It is shown first that the cross sections to use for measurements by threshold detectors depend largely on the neutron spectrum, that is the position in which the measurement is performed. The spectrum is determined by calculation for several positions in the piles EL2 and EL3; from this can be deduced the cross-sections to be used for the measurements carried out in these positions. In the last part of the report, possible methods for the experimental determination of the spectrum are indicated. (author) [French] On etudie principalement la mesure des flux de neutrons rapides a l'aide de detecteurs a seuil. On montre d'abord que les sections efficaces a utiliser pour les mesures par detecteurs a seuil, dependent grandement du spectre des neutrons, c'est-a-dire de l'emplacement ou s'effectue la mesure. La determination du spectre est effectuee par le calcul pour plusieurs emplacements des piles EL2 et EL3; on en deduit les sections efficaces a utiliser pour les mesures effectuees a ces emplacements. Dans la derniere partie du rapport, on indique quelles methodes sont possibles pour la determination experimentale du spectre. (auteur)

Comparison of neutron spectrum unfolding codes

International Nuclear Information System (INIS)

Zijp, W.

1979-02-01

This final report contains a set of four ECN-reports. The first is dealing with the comparison of the neutron spectrum unfolding codes CRYSTAL BALL, RFSP-JUL, SAND II and STAY'SL. The other three present the results of calculations about the influence of statistical weights in CRYSTAL BALL, SAND II and RFSP-JUL

High-accuracy determination of the neutron flux at n{sub T}OF

Energy Technology Data Exchange (ETDEWEB)

Barbagallo, M.; Colonna, N.; Mastromarco, M.; Meaze, M.; Tagliente, G.; Variale, V. [Sezione di Bari, INFN, Bari (Italy); Guerrero, C.; Andriamonje, S.; Boccone, V.; Brugger, M.; Calviani, M.; Cerutti, F.; Chin, M.; Ferrari, A.; Kadi, Y.; Losito, R.; Versaci, R.; Vlachoudis, V. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Tsinganis, A. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); National Technical University of Athens (NTUA), Athens (Greece); Tarrio, D.; Duran, I.; Leal-Cidoncha, E.; Paradela, C. [Universidade de Santiago de Compostela, Santiago (Spain); Altstadt, S.; Goebel, K.; Langer, C.; Reifarth, R.; Schmidt, S.; Weigand, M. [Johann-Wolfgang-Goethe Universitaet, Frankfurt (Germany); Andrzejewski, J.; Marganiec, J.; Perkowski, J. [Uniwersytet Lodzki, Lodz (Poland); Audouin, L.; Leong, L.S.; Tassan-Got, L. [Centre National de la Recherche Scientifique/IN2P3 - IPN, Orsay (France); Becares, V.; Cano-Ott, D.; Garcia, A.R.; Gonzalez-Romero, E.; Martinez, T.; Mendoza, E. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas (CIEMAT), Madrid (Spain); Becvar, F.; Krticka, M.; Kroll, J.; Valenta, S. [Charles University, Prague (Czech Republic); Belloni, F.; Fraval, K.; Gunsing, F.; Lampoudis, C.; Papaevangelou, T. [Commissariata l' Energie Atomique (CEA) Saclay - Irfu, Gif-sur-Yvette (France); Berthoumieux, E.; Chiaveri, E. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Commissariata l' Energie Atomique (CEA) Saclay - Irfu, Gif-sur-Yvette (France); Billowes, J.; Ware, T.; Wright, T. [University of Manchester, Manchester (United Kingdom); Bosnar, D.; Zugec, P. [University of Zagreb, Department of Physics, Faculty of Science, Zagreb (Croatia); Calvino, F.; Cortes, G.; Gomez-Hornillos, M.B.; Riego, A. [Universitat Politecnica de Catalunya, Barcelona (Spain); Carrapico, C.; Goncalves, I.F.; Sarmento, R.; Vaz, P. [Universidade Tecnica de Lisboa, Instituto Tecnologico e Nuclear, Instituto Superior Tecnico, Lisboa (Portugal); Cortes-Giraldo, M.A.; Praena, J.; Quesada, J.M.; Sabate-Gilarte, M. [Universidad de Sevilla, Sevilla (Spain); Diakaki, M.; Karadimos, D.; Kokkoris, M.; Vlastou, R. [National Technical University of Athens (NTUA), Athens (Greece); Domingo-Pardo, C.; Giubrone, G.; Tain, J.L. [CSIC-Universidad de Valencia, Instituto de Fisica Corpuscular, Valencia (Spain); Dressler, R.; Kivel, N.; Schumann, D.; Steinegger, P. [Paul Scherrer Institut, Villigen PSI (Switzerland); Dzysiuk, N.; Mastinu, P.F. [Laboratori Nazionali di Legnaro, INFN, Rome (Italy); Eleftheriadis, C.; Manousos, A. [Aristotle University of Thessaloniki, Thessaloniki (Greece); Ganesan, S.; Gurusamy, P.; Saxena, A. [Bhabha Atomic Research Centre (BARC), Mumbai (IN); Griesmayer, E.; Jericha, E.; Leeb, H. [Technische Universitaet Wien, Atominstitut, Wien (AT); Hernandez-Prieto, A. [European Organization for Nuclear Research (CERN), Geneva (CH); Universitat Politecnica de Catalunya, Barcelona (ES); Jenkins, D.G.; Vermeulen, M.J. [University of York, Heslington, York (GB); Kaeppeler, F. [Institut fuer Kernphysik, Karlsruhe Institute of Technology, Campus Nord, Karlsruhe (DE); Koehler, P. [Oak Ridge National Laboratory (ORNL), Oak Ridge (US); Lederer, C. [Johann-Wolfgang-Goethe Universitaet, Frankfurt (DE); University of Vienna, Faculty of Physics, Vienna (AT); Massimi, C.; Mingrone, F.; Vannini, G. [Universita di Bologna (IT); INFN, Sezione di Bologna, Dipartimento di Fisica, Bologna (IT); Mengoni, A.; Ventura, A. [Agenzia nazionale per le nuove tecnologie, l' energia e lo sviluppo economico sostenibile (ENEA), Bologna (IT); Milazzo, P.M. [Sezione di Trieste, INFN, Trieste (IT); Mirea, M. [Horia Hulubei National Institute of Physics and Nuclear Engineering - IFIN HH, Bucharest - Magurele (RO); Mondalaers, W.; Plompen, A.; Schillebeeckx, P. [Institute for Reference Materials and Measurements, European Commission JRC, Geel (BE); Pavlik, A.; Wallner, A. [University of Vienna, Faculty of Physics, Vienna (AT); Rauscher, T. [University of Basel, Department of Physics and Astronomy, Basel (CH); Roman, F. [European Organization for Nuclear Research (CERN), Geneva (CH); Horia Hulubei National Institute of Physics and Nuclear Engineering - IFIN HH, Bucharest - Magurele (RO); Rubbia, C. [European Organization for Nuclear Research (CERN), Geneva (CH); Laboratori Nazionali del Gran Sasso dell' INFN, Assergi (AQ) (IT); Weiss, C. [European Organization for Nuclear Research (CERN), Geneva (CH); Johann-Wolfgang-Goethe Universitaet, Frankfurt (DE)

2013-12-15

The neutron flux of the n{sub T}OF facility at CERN was measured, after installation of the new spallation target, with four different systems based on three neutron-converting reactions, which represent accepted cross sections standards in different energy regions. A careful comparison and combination of the different measurements allowed us to reach an unprecedented accuracy on the energy dependence of the neutron flux in the very wide range (thermal to 1 GeV) that characterizes the n{sub T}OF neutron beam. This is a pre-requisite for the high accuracy of cross section measurements at n{sub T}OF. An unexpected anomaly in the neutron-induced fission cross section of {sup 235}U is observed in the energy region between 10 and 30keV, hinting at a possible overestimation of this important cross section, well above currently assigned uncertainties. (orig.)