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Sample records for neutron spectral fluence

  1. Spectral fluence of neutrons generated by radiotherapeutic Linacs

    International Nuclear Information System (INIS)

    Kralik, Miloslav; Solc, Jaroslav; Smoldasova, Jana; Vondracek, Vladimir; Farkasova, Estera; Ticha, Ivana

    2015-01-01

    Spectral fluences of neutrons generated in the heads of the radiotherapeutic linacs Varian Clinac 2100 C/D and Siemens ARTISTE were measured by means of the Bonner spheres spectrometer whose active detector of thermal neutrons was replaced by an activation detector, i.e. a tablet made of pure manganese. Measurements with different collimator settings reveal an interesting dependence of neutron fluence on the area defined by the collimator jaws. The determined neutron spectral fluences were used to derive ambient dose equivalent rate along the treatment coach. To clarify at which components of the linac neutrons are mainly created, the measurements were complemented with MCNPX calculations based on a realistic model of the Varian Clinac. (authors)

  2. Neutron fluence measurements

    International Nuclear Information System (INIS)

    1970-01-01

    For research reactor work dealing with such subjects as radiation effects on solids and such disciplines as radiochemistry and radiobiology, the radiation dose or neutron fluence is an essential parameter in evaluating results. Unfortunately it is very difficult to determine. Even when the measurements have been accurate, it is difficult to compare results obtained in different experiments because present methods do not always reflect the dependence of spectra or of different types of radiation on the induced processes. After considering the recommendations of three IAEA Panels, on 'In-pile dosimetry' held in July 1964, on 'Neutron fluence measurements' in October 1965, and on 'In-pile dosimetry' in November 1966, the Agency established a Working Group on Reactor Radiation Measurements. This group consisted of eleven experts from ten different Member States and two staff members of the Agency. In the measurement of energy absorbed by materials from neutrons and gamma rays, there are various reports and reviews scattered throughout the literature. The group, however, considered that the time was ripe for all relevant information to be evaluated and gathered together in the form of a practical guide, with the aim of promoting consistency in the measurement and reporting of reactor radiation. The group arranged for the material to be divided into two manuals, which are expected to be useful both for experienced workers and for beginners

  3. Neutron fluence measurement in nuclear facilities

    International Nuclear Information System (INIS)

    Camacho L, M.E.

    1997-01-01

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

  4. Neutron fluence spectrometry using disk activation

    International Nuclear Information System (INIS)

    Loevestam, Goeran; Hult, Mikael; Fessler, Andreas; Gasparro, Joel; Kockerols, Pierre; Okkinga, Klaas; Tagziria, Hamid; Vanhavere, Filip; Wieslander, J.S. Elisabeth

    2009-01-01

    A simple and robust detector for spectrometry of environmental neutrons has been developed. The technique is based on neutron activation of a series of different metal disks followed by low-level gamma-ray spectrometry of the activated disks and subsequent neutron spectrum unfolding. The technique is similar to foil activation but here the applied neutron fluence rates are much lower than usually in the case of foil activation. The detector has been tested in quasi mono-energetic neutron fields with fluence rates in the order of 1000-10000 cm -2 s -1 , where the obtained spectra showed good agreement with spectra measured using a Bonner sphere spectrometer. The detector has also been tested using an AmBe source and at a neutron fluence rate of about 40 cm -2 s -1 , again, a good agreement with the assumed spectrum was achieved

  5. Neutron fluence spectrometry using disk activation

    Energy Technology Data Exchange (ETDEWEB)

    Loevestam, Goeran [EC-JRC-Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium)], E-mail: goeran.loevestam@ec.europa.eu; Hult, Mikael; Fessler, Andreas; Gasparro, Joel; Kockerols, Pierre; Okkinga, Klaas [EC-JRC-Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium); Tagziria, Hamid [EC-JRC-Institute for the Protection and the Security of the Citizen (IPSC), Via E. Fermi 1, I-21020 Ispra (Vatican City State, Holy See,) (Italy); Vanhavere, Filip [SCK-CEN, Boeretang, 2400 Mol (Belgium); Wieslander, J.S. Elisabeth [EC-JRC-Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium); Department of Physics, P.O. Box 35 (YFL), FIN-40014, University of Jyvaeskylae (Finland)

    2009-01-15

    A simple and robust detector for spectrometry of environmental neutrons has been developed. The technique is based on neutron activation of a series of different metal disks followed by low-level gamma-ray spectrometry of the activated disks and subsequent neutron spectrum unfolding. The technique is similar to foil activation but here the applied neutron fluence rates are much lower than usually in the case of foil activation. The detector has been tested in quasi mono-energetic neutron fields with fluence rates in the order of 1000-10000 cm{sup -2} s{sup -1}, where the obtained spectra showed good agreement with spectra measured using a Bonner sphere spectrometer. The detector has also been tested using an AmBe source and at a neutron fluence rate of about 40 cm{sup -2} s{sup -1}, again, a good agreement with the assumed spectrum was achieved.

  6. Neutron dosimetry intercomparison run for verification of the neutron fluence

    International Nuclear Information System (INIS)

    Penev, I.; Kinova, L.

    2001-01-01

    For the neutron fluence verification the intercomparison runs Balakovo and KORPUS have been carried out. The participation in the international intercomparison runs shows that in order to more precisely verify the calculated values of the neutron fluence more intercomparison exercises are necessary. Due to such exercises the results improved after calibration of Nb performed and are in a very good agreement with RIIAR results in spite of the different approaches in the determination of its activity

  7. Phototransistor response under a neutron fluence

    International Nuclear Information System (INIS)

    Santos, Luiz A.P.; Barros, Fabio R.; Ursulino, Luciano C.; Silva Junior, Eronides F.; Antonio Filho, Joao

    2009-01-01

    The purpose of this communication is to show some effects on a bipolar phototransistor after it has been under a neutron fluence. Unlike a transistor, a phototransistor is designed so that the collector has a large area and consequently it has a higher radiation detection probability. Then, it is possible to have a certain number of interactions so that any changes in the internal structure of the phototransistor can be observed after a neutron irradiation. If a phototransistor is under a certain spectra of neutron fluence the interaction depends on the cross section of the either silicon chip or its encapsulation, and recoil protons could be the charged particle responsible for changes in the semiconductor structure. Furthermore, neutron irradiation could give to the device a state of vanishing in its electrical characteristic which can be performed tracing the current versus voltage curve (I x V). The experimental arrangement basically consists of a photonic device, a neutron-gamma radiation source and a Flip-Flop electrometer second generation (EFF-2G). One of the main parameters of evaluation was the phototransistor dark current. In fact, the first results demonstrate that when the phototransistor is neutron irradiated there is a significant variation in its I x V characteristic curve. (author)

  8. Passive detectors for neutron fluence measurement

    International Nuclear Information System (INIS)

    Holt, P.D.

    1985-01-01

    The use of neutron activation detectors (slow neutron detectors and threshold detectors) and fission track detectors for radiological protection purposes, principally in criticality dosimetry, dosimetry of pulsed accelerators and calibration of neutron fluxes is discussed. References are given to compilations of cross sections. For the determination of the activity induced, either beta ray or gamma ray counting may be used. For beta-ray counting, thin foils are usually necessary which result in low neutron sensitivity. When fission track detectors are used, it is necessary to know the efficiency of track registration. Alternatively, a detector-counter system may be calibrated by exposure to a known flux of monoenergetic neutrons. Usually, the sensitivity of activation detectors is low because small foils are used. For criticality dosimetry, calibration work and shielding studies on accelerators, low sensitivity is acceptable. However, there are some instances where, by the use of long integration times, or very large quantities of detector material with gamma ray detection, neutron fluences in operational areas have been measured. (author)

  9. Studies for improvement of WWER-440 neutron fluence determination

    International Nuclear Information System (INIS)

    Ilieva, Kr.; Belousov, S.; Apostolov, T.

    2001-01-01

    For assessment of radiation embrittlement and prediction of reactor vessel lifetime with reasonable conservatism a 'best estimated' neutron fluence is necessary. New studies purposed to improve the fluence determination are presented: 1) study on the reliability of multigroup presentation of the neutron cross sections, and 2) impact of negative gradient of reactor power in the periphery assemblies on the neutron fluence evaluation. The results of these studies are base for improvement of neutron fluence determination methodology applied by the INRNE, BAS at Kozloduy NPP. (author)

  10. A neutron source of variable fluence

    International Nuclear Information System (INIS)

    Brachet, Guy; Demichel, Pascal; Prigent, Yvon; Riche, J.C.

    1975-01-01

    The invention concerns a variable fluence neutron source, like those that use in the known way a reaction between a radioactive emitter and a target, particularly of type (α,n). The emitter being in powder form lies in a carrier fluid forming the target, inside a closed containment. Facilities are provided to cause the fluidisation of the emitter by the carrier fluid in the containment. The fluidisation of the emitting powder is carried out by a booster with blades, actuated from outside by a magnetic coupling. The powder emitter is a α emitter selected in the group of curium, plutonium, thorium, actinium and americium oxides and the target fluid is formed of compounds of light elements selected from the group of beryllium, boron, fluorine and oxygen 18. The target fluid is a gas used under pressure or H 2 O water highly enriched in oxygen 18 [fr

  11. The fluence research of filter material for fast neutron fluence measurement

    International Nuclear Information System (INIS)

    Tang Xiding

    2010-01-01

    When the fast neutron fluence is measured by radioactivation techniques in the nuclear reactor the fast neutron is also filtered a little by the thermal neutron filter material, and if the filter material thickness increase the filtered fast neutron increases therewith. For fast neutron fluenc measurement, there are only cadmium, boron and gadolinium three elements filtering fluence can be calculated ordinarily. In order to calculate the filtered fast neutron fluence of the all elements in the filter material, the many total cross sections of nuclides had checked out from nuclear cross section data library, converted them into the same energy group structure, then element's total cross section, compound's total cross section and multilayer filters' total cross section had calculated from these total cross sections with same energy group structure, a new cross section data library can be obtained lastly through merging these cross sections into the old cross section data library used for neutron fluence measurement. The calculation analysis indicates that the results of the unit 2 surveillance capsule U of DAYA Bay NPP and the unit 1 surveillance capsule A of the Second Nuclear Power Plant of Qinshan by considering the all elements subtracting iron are smaller about 1.5% and 2.6% respectively than the ones only to consider cadmium, boron. The old measured results accord with the new values under the measurement uncertainty, are reliable. The new results are more accuracy. (authors)

  12. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

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

  13. RAMA Methodology for the Calculation of Neutron Fluence

    International Nuclear Information System (INIS)

    Villescas, G.; Corchon, F.

    2013-01-01

    he neutron fluence plays an important role in the study of the structural integrity of the reactor vessel after a certain time of neutron irradiation. The NRC defined in the Regulatory Guide 1.190, the way must be estimated neutron fluence, including uncertainty analysis of the validation process (creep uncertainty is ? 20%). TRANSWARE Enterprises Inc. developed a methodology for calculating the neutron flux, 1,190 based guide, known as RAMA. Uncertainty values obtained with this methodology, for about 18 vessels, are less than 10%.

  14. Nickel Foil as Transmutation Detector for Neutron Fluence Measurements

    Directory of Open Access Journals (Sweden)

    Klupák Vít

    2016-01-01

    Full Text Available Activation detectors are very often used for determination of the neutron fluence in reactor dosimetry. However, there are few disadvantages concerning these detectors; it is the demand of the knowledge of the irradiation history and a loss of information due to a radioactive decay in time. Transmutation detectors TMD could be a solution in this case. The transmutation detectors are materials in which stable or long-lived nuclides are produced by nuclear reactions with neutrons. From a measurement of concentration of these nuclides, neutron fluence can be evaluated regardless of the cooling time.

  15. Safety factors for neutron fluences in NPP safety assessment

    International Nuclear Information System (INIS)

    Demekhin, V.L.; Bukanov, V.N.; Il'kovich, V.V.; Pugach, A.M.

    2016-01-01

    In accordance with global practice and a number of existing regulations, the use of conservative approach is required for the calculations related to nuclear safety assessment of NPP. It implies the need to consider the determination of neutron fluence errors that is rather complicated. It is proposed to carry out the consideration by the way of multiplying the neutron fluences obtained with transport calculations by safety factors. The safety factor values are calculated by the developed technique based on the theory of errors, features of the neutron transport calculation code and the results obtained with the code. It is shown that the safety factor value is equal 1.18 with the confidence level of not less than 0.95 for the majority of VVER-1000 reactor places where neutron fluences are determined by MCPV code, and its maximum value is 1.25

  16. Development of a Secondary Neutron Fluence Standard at GELINA

    International Nuclear Information System (INIS)

    Heyse, Jan; Eykens, Roger; Moens, Andre; Plompen, Arjan J.M.; Schillebeeckx, Peter; Wynants, Ruud; Anastasiou, Maria

    2013-06-01

    The MetroFission project, a Joint Research Project within the European Metrology Research Program, aims at addressing a number of metrological problems involved in the design of proposed Generation IV nuclear reactors. One of the objectives of this multidisciplinary project is the improvement of neutron cross section measurement techniques in order to arrive at uncertainties as required for the design and safety assessment of new generation power plants and fuel cycles. This objective is in line with the 'Uncertainty and target accuracy assessment for innovative systems using recent covariance data evaluations' published by a working party of the OECD Nuclear Energy Agency in 2008. These requests are often very challenging, being at or beyond the state-of-the-art in neutron measurements, which is set by self-normalizing methods and the neutron data standards used at laboratories where the data are measured. A secondary neutron fluence standard has been developed and calibrated at the neutron time-of-flight facility GELINA of the JRC's Institute for Reference Materials and Measurements (IRMM). It consists of a flux monitor, a reference ionization chamber containing a 10 B layer and a 235 U layer, and a parallel plate ionization chamber with 8 well characterized 235 U deposits. These devices are used to determine the neutron fluence, based on the well-known neutron induced fission reaction on 235 U. All deposits have been prepared and characterized at the IRMM target preparation lab. The secondary fluence standard at the GELINA facility can be used for reliable determination of the efficiency of fluence measurement devices used in neutron data measurements at IRMM and elsewhere. It is an essential tool to reliably calibrate fluence normalization devices used in neutron time-of-flight cross section measurements. (authors)

  17. The development report of an intelligent neutron fluence integration monitor

    International Nuclear Information System (INIS)

    Jiang Zongbing; Wei Ying

    1996-10-01

    An intelligent neutron fluence integration monitor is introduced. It is used to measure the received neutron fluence of the monocrystalline silicon in reactor radiation channel. The significance of study and specifications of the instrument are briefly described. The emphasis is on the working principle, structure and characteristics of the instrument is intelligent due to use of monolithic microcomputer. It has many advantages proved in the actual practice, such as powerful function, high accuracy, diversity of application, high level automatization, easy to operate, high reliability, etc. After using this instrument the monocrystalline silicon radiation technology is improved and the efficiency of production is raised. (1 fig.)

  18. Neutron fluence produced in medical accelerators

    International Nuclear Information System (INIS)

    Castro, R.C.; Silva, A.X. da; Crispim, V.R.

    2004-01-01

    Radiotherapy with photon and electron beams still represents the most diffused technique to control and treat tumour diseases. To increase the treatment efficiency, accelerators of higher energy are used, the increase of electron and photon energy is joined with generation of undesired fast neutron that contaminated the therapeutic beam and give a non-negligible contribution to the patient dose. In this work we have simulated with the MCNP4B code the produced neutron spectra in the interaction between the beam and the head to the accelerator and estimating the equivalent dose for neutrons by x-ray dose for aims far from the targets. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Camacho L, M E

    1997-12-01

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

  20. ATR neutron spectral characterization

    Energy Technology Data Exchange (ETDEWEB)

    Rogers, J.W.; Anderl, R.A.

    1995-11-01

    The Advanced Test Reactor (ATR) at INEL provides intense neutron fields for irradiation-effects testing of reactor material samples, for production of radionuclides used in industrial and medical applications, and for scientific research. Characterization of the neutron environments in the irradiation locations of the ATR has been done by means of neutronics calculations and by means of neutron dosimetry based on the use of neutron activation monitors that are placed in the various irradiation locations. The primary purpose of this report is to present the results of an extensive characterization of several ATR irradiation locations based on neutron dosimetry measurements and on least-squares-adjustment analyses that utilize both neutron dosimetry measurements and neutronics calculations. This report builds upon the previous publications, especially the reference 4 paper. Section 2 provides a brief description of the ATR and it tabulates neutron spectral information for typical irradiation locations, as derived from the more historical neutron dosimetry measurements. Relevant details that pertain to the multigroup neutron spectral characterization are covered in section 3. This discussion includes a presentation on the dosimeter irradiation and analyses and a development of the least-squares adjustment methodology, along with a summary of the results of these analyses. Spectrum-averaged cross sections for neutron monitoring and for displacement-damage prediction in Fe, Cr, and Ni are given in section 4. In addition, section4 includes estimates of damage generation rates for these materials in selected ATR irradiation locations. In section 5, the authors present a brief discussion of the most significant conclusions of this work and comment on its relevance to the present ATR core configuration. Finally, detailed numerical and graphical results for the spectrum-characterization analyses in each irradiation location are provided in the Appendix.

  1. The activation method for determining neutron spectra and fluences

    International Nuclear Information System (INIS)

    Hogel, J.; Vespalec, R.

    1980-01-01

    3 mm thick foils of 4 and 17 mm in diameter were used for measurements. NaI scintillation detectors 45 mm in diameter by 50 mm thick and 40 mm in diameter by 1 mm thick, and a Ge-Li spectrometer of 53 cm 3 in volume were used for gamma detection. A photopeak or a certain part of the integral spectrum was measured for each radionuclide. Computer code PIKAR was applied in automatic calculation of a simple gamma spectrum obtained using the semiconductor spectrometer. The FACT code was used for calculating foil activity. Codes SAND II and RFSP were used for neutron spectra unfolding. Ge-Li detector spectrometry was used for determining neutron fluence. Code FLUE was used for determining the mean value of neutron flux density and fluence. (J.P.)

  2. Determination of fast neutron fluence at WWER-1000 pressure vessel

    International Nuclear Information System (INIS)

    Valenta, V. et al.

    1989-01-01

    The influence function method is an effective tool making it possible, by means of tabulated values to rapidly perform three-dimensional calculations of fast neutron fluences for various reactor core loadings and for various nuclear power plant units. The procedure for determining the spatial dependence of the fast neutron fluences in a WWER-1000 pressure vessel is described. For this, the reactor core is divided into sufficiently fine volume elements within which the neutron source can be regarded as coordinate-independent. The influence functions point to a substantial role of sources lying at the reactor core periphery. In WWER-1000 reactors, only 1 or 2 rows of peripheral assemblies are important. The influence function method makes possible a rapid and easy determination of preconditions for the assessment of the residual lifetime of the pressure vessel based on the actual reactor core loadings. (Z.M.). 7 figs., 8 refs

  3. Application of the adjoint function methodology for neutron fluence determination

    International Nuclear Information System (INIS)

    Haghighat, A.; Nanayakkara, B.; Livingston, J.; Mahgerefteh, M.; Luoma, J.

    1991-01-01

    In previous studies, the neutron fluence at a reactor pressure vessel has been estimated based on consolidation of transport theory calculations and experimental data obtained from in-vessel capsules and/or cavity dosimeters. Normally, a forward neutron transport calculation is performed for each fuel cycle and the neutron fluxes are integrated over the reactor operating time to estimate the neutron fluence. Such calculations are performed for a geometrical model which is composed of one-eighth (0 to 45 deg) of the reactor core and its surroundings; i.e., core barrel, thermal shield, downcomer, reactor vessel, cavity region, concrete wall, and instrumentation well. Because the model is large, transport theory calculations generally require a significant amount of computer memory and time; hence, more efficient methodologies such as the adjoint transport approach have been proposed. These studies, however, do not address the necessary sensitivity studies needed for adjoint function calculations. The adjoint methodology has been employed to estimate the activity of a cavity dosimeter and that of an in-vessel capsule. A sensitivity study has been performed on the mesh distribution used in and around the cavity dosimeter and the in-vessel capsule. Further, since a major portion of the detector response is due to the neutrons originated in the peripheral fuel assemblies, a study on the use of a smaller calculational model has been performed

  4. Investigation of neutron fluence using fluence monitors for irradiation test at WWR-K

    International Nuclear Information System (INIS)

    Romanova, N.K.; Takemoto, N.

    2013-01-01

    Irradiation test of a Si ingot is planned using WWR-K in Institute of Nuclear Physics Republic of Kazakhstan (INP RK) to develop an irradiation technology for Si semiconductor production by Neutron Transmutation Doping (NTD) method in the framework of an international cooperation between INP RK and Japan Atomic Energy Agency (JAEA), Japan. It is possible to irradiate the Si ingot of 6 inch in diameter at the K-23 irradiation channel in the WWR-K. The preliminary irradiation test using 4 Al ingots was performed to evaluate the actual neutronic irradiation field at the K-23 channel in the WWR-K. Each Al ingot has the same dimension as the Si ingot, and 15 fluence monitors are equipped in it. Iron wire and aluminum-cobalt wire are inserted into them, and it is possible to evaluate both fast and thermal neutron fluxes by measurement of these radiation activities after irradiation. This report described the results of the preliminary irradiation test and the neutronic calculations by Monte Carlo method in order to evaluate the neutronic irradiation field in the irradiation position for the silicon ingot at the channel in the WWR-K. (authors)

  5. Neutron Fluence Evaluation using an Am-Be Neutron Sources Assembly and P ADC Detectors

    International Nuclear Information System (INIS)

    Seddik, U.

    2008-01-01

    An assembly of four 241 Am-Be sources has been constructed at Nuclear Reactions Unit (NRU) of Nuclear Research Center (NRU) to perform analysis of different materials using thermal and fast neutrons. In the present paper, we measure the value of transmittance (T) in percentage of etched CR-39 detectors using a spectrophotometer at different neutron fluences ,to relate the transmittance of the detector with the neutron fluence values. The exposed samples to neutrons with accumulated fluence of order between 10 10 and 10 12 cm -2 were etched for 15 time intervals between 10-600 min in 6.25 N NaOH at 70 degree C. The etched samples were analyzed using Tech 8500 II spectrophotometer. A trend of the sample transmission and the etching time is observed which is different for each fluence value. A linear relation between the transmittance decay constant and the neutron fluence is observed which could be used as a calibration to determine unknown neutron fluence

  6. Estimates of neutron fluence for the SDC detector

    International Nuclear Information System (INIS)

    Job, P.K.; Price, L.E.; Handler, T.; Gabriel, T.A.

    1994-01-01

    The high energy and high luminosity of SSC cause radiation problems to detectors. Almost all the radiation in the SDC detector comes from the 20 TeV on 20 TeV pp collisions. The design luminosity corresponds to 10 8 collisions per second. This luminosity is maintained for 10 7 seconds (one SSC year). It is important to know the radiation fields experienced by the tracking volume, calorimeter, electronics and the phototubes. The loss of light due to the radiation damage to the scintillators can adversely affect the physics performance of the calorimeter. Studies have been carried out earlier to estimate the radiation dose in the SDC detector. In this note the authors use ISAJET in combination with CALOR89 to make an accurate prediction of neutron fluence at the various locations of the SDC detector. The low energy neutrons are important because they can produce radioactive nuclides in large quantities. In CALOR89 the low energy neutron fluence is accurately estimated by MORSE code

  7. Deduction of solar neutron fluences from large gamma-ray flares

    International Nuclear Information System (INIS)

    Yoshimori, Masato; Watanabe, Hiroyuki; Takahashi, Kazuyoshi.

    1986-01-01

    Solar neutron fluences from large gamma-ray flares are deduced from accelerated proton spectra and numbers derived from the gamma-ray observations. The deduced solar neutron fluences range from 1 to 200 neutrons cm -2 . The present result indicates a possibility that high sensitivity ground-based neutron monitors can detect solar neutron events, just as detected by the Jungfraujoch and Rome neutron monitors. (author)

  8. Monitoring of the Irradiated Neutron Fluence in the Neutron Transmutation Doping Process of Hanaro

    Science.gov (United States)

    Kim, Myong-Seop; Park, Sang-Jun

    2009-08-01

    Neutron transmutation doping (NTD) for silicon is a process of the creation of phosphorus impurities in intrinsic or extrinsic silicon by neutron irradiation to obtain silicon semiconductors with extremely uniform dopant distribution. HANARO has two vertical holes for the NTD, and the irradiation for 5 and 6 inch silicon ingots has been going on at one hole. In order to achieve the accurate neutron fluence corresponding to the target resistivity, the real time neutron flux is monitored by self-powered neutron detectors. After irradiation, the total irradiation fluence is confirmed by measuring the absolute activity of activation detectors. In this work, a neutron fluence monitoring method using zirconium foils with the mass of 10 ~ 50 mg was applied to the NTD process of HANARO. We determined the proportional constant of the relationship between the resistivity of the irradiated silicon and the neutron fluence determined by using zirconium foils. The determined constant for the initially n-type silicon was 3.126 × 1019 n·Ω/cm. It was confirmed that the difference between this empirical value and the theoretical one was only 0.5%. Conclusively, the practical methodology to perform the neutron transmutation doping of silicon was established.

  9. Effects of high thermal neutron fluences on Type 6061 aluminum

    International Nuclear Information System (INIS)

    Weeks, J.R.; Czajkowski, C.J.; Farrell, K.

    1992-01-01

    The control rod drive follower tubes of the High Flux Beam Reactor are contructed from precipitation-hardened 6061-T6 aluminum alloy and they operate in the high thermal neutron flux regions of the core. It is shown that large thermal neutron fluences up to ∼4 x 10 23 n/cm 2 at 333K cause large increases in tensile strength and relatively modest decreases in tensile elongation while significantly reducing the notch impact toughness at room temperature. These changes are attributed to the development of a fine distribution of precipitates of amorphous silicon of which about 8% is produced radiogenically. A proposed role of thermal-to-fast flux ratio is discussed

  10. Fluence-compensated down-scattered neutron imaging using the neutron imaging system at the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Casey, D. T., E-mail: casey21@llnl.gov; Munro, D. H.; Grim, G. P.; Landen, O. L.; Spears, B. K.; Fittinghoff, D. N.; Field, J. E.; Smalyuk, V. A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Volegov, P. L.; Merrill, F. E. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

    2016-11-15

    The Neutron Imaging System at the National Ignition Facility is used to observe the primary ∼14 MeV neutrons from the hotspot and down-scattered neutrons (6-12 MeV) from the assembled shell. Due to the strong spatial dependence of the primary neutron fluence through the dense shell, the down-scattered image is convolved with the primary-neutron fluence much like a backlighter profile. Using a characteristic scattering angle assumption, we estimate the primary neutron fluence and compensate the down-scattered image, which reveals information about asymmetry that is otherwise difficult to extract without invoking complicated models.

  11. Solid State Track Recorder fission rate measurements at high neutron fluence and high temperature

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Roberts, J.H.; Gold, R.

    1985-01-01

    Solid State Track Recorder (SSTR) techniques have been used to measure 239-Pu, 235-U, and 237-Np fission rates for total neutron fluences approaching 5 x 10 17 n/cm 2 at temperatures in the range 680 to 830 0 F. Natural quartz crystal SSTRs were used to withstand the high temperature environment and ultra low-mass fissionable deposits of the three isotopes were required to yield scannable track densities at the high neutron fluences. The results of these high temperature, high neutron fluence measurements are reported

  12. Measured Thermal and Fast Neutron Fluence Rates for ATF-1 Holders During ATR Cycle 157D

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Larry Don [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, David Torbet [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 157D which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order. This report contains measurements of the fluence rates corresponding to the particular elevations relative to the 80-ft. core elevation. The data in this report consist of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, and (3) plots of both the thermal and fast neutron fluence rates. The fluence rates reported are for the average power levels given in the table of power history and distribution.

  13. The determination of fast neutron fluence in radiation stability tests of steel samples

    International Nuclear Information System (INIS)

    Hogel, J.; Vespalec, R.

    1979-01-01

    The activation method is described of determining fast neutron fluence. Samples of steel designed for WWER type reactor pressure vessels were irradiated in the CHOUCA-rigs in the core of the WWR-S reactor. The neutron spectrum was measured by the multiple activation foil method and the effective cross sections of fluence monitors were calculated. The fluences obtained from the reactions 54 Fe(n,p) 54 Mn and 63 Cu(n,α) 60 Co are presented and the method is discussed. (author)

  14. A hybrid source-driven method to compute fast neutron fluence in reactor pressure vessel - 017

    International Nuclear Information System (INIS)

    Ren-Tai, Chiang

    2010-01-01

    A hybrid source-driven method is developed to compute fast neutron fluence with neutron energy greater than 1 MeV in nuclear reactor pressure vessel (RPV). The method determines neutron flux by solving a steady-state neutron transport equation with hybrid neutron sources composed of peripheral fixed fission neutron sources and interior chain-reacted fission neutron sources. The relative rod-by-rod power distribution of the peripheral assemblies in a nuclear reactor obtained from reactor core depletion calculations and subsequent rod-by-rod power reconstruction is employed as the relative rod-by-rod fixed fission neutron source distribution. All fissionable nuclides other than U-238 (such as U-234, U-235, U-236, Pu-239 etc) are replaced with U-238 to avoid counting the fission contribution twice and to preserve fast neutron attenuation for heavy nuclides in the peripheral assemblies. An example is provided to show the feasibility of the method. Since the interior fuels only have a marginal impact on RPV fluence results due to rapid attenuation of interior fast fission neutrons, a generic set or one of several generic sets of interior fuels can be used as the driver and only the neutron sources in the peripheral assemblies will be changed in subsequent hybrid source-driven fluence calculations. Consequently, this hybrid source-driven method can simplify and reduce cost for fast neutron fluence computations. This newly developed hybrid source-driven method should be a useful and simplified tool for computing fast neutron fluence at selected locations of interest in RPV of contemporary nuclear power reactors. (authors)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    Science.gov (United States)

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

    2018-02-01

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

  18. Neutron fluence rate and energy spectrum in SPRR-300 reactor thermal column

    International Nuclear Information System (INIS)

    Dou Haifeng; Dai Junlong

    2006-01-01

    In order to modify the simple one-dimension model, the neutron fluence rate distribution calculated with ANISN code ws checked with that calculated with MCNP code. To modify the error caused by ignoring the neutron landscape orientation leaking, the reflector that can't be modeled in a simple one-dimension model was dealt by extending landscape orientation scale. On this condition the neutron fluence rate distribution and the energy spectrum in the thermal column of SPRR-300 reactor were calculated with one-dimensional code ANISN, and the results of Cd ratio are well accorded with the experimental results. The deviation between them is less than 5% and it isn't above 10% in one or two special positions. It indicates that neutron fluence rate distribution and energy spectrum in the thermal column can be well calculated with one-dimensional code ANISN. (authors)

  19. Neutron fluence-to-dose conversion coefficients for embryo and fetus

    International Nuclear Information System (INIS)

    Chen, J.; Meyerhof, D.; Vlahovich, S.

    2004-01-01

    A problem of concern in radiation protection is the exposure of pregnant women to ionising radiation, because of the high radiosensitivity of the embryo and fetus. External neutron exposure is of concern when pregnant women travel by aeroplane. Dose assessments for neutrons frequently rely on fluence-to-dose conversion coefficients. While neutron fluence-to-dose conversion coefficients for adults are recommended in International Commission on Radiological Protection publications and International Commission on Radiological Units and Measurements reports, conversion coefficients for embryos and fetuses are not given in the publications. This study undertakes Monte Carlo calculations to determine the mean absorbed doses to the embryo and fetus when the mother is exposed to neutron fields. A new set of mathematical models for the embryo and fetus has been developed at Health Canada and is used together with mathematical phantoms of a pregnant female developed at Oak Ridge National Laboratory. Monoenergetic neutrons from 1 eV to 10 MeV are considered in this study. The irradiation geometries include antero-posterior (AP), postero-anterior (PA), lateral (LAT), rotational (ROT) and isotropic (ISO) geometries. At each of these standard irradiation geometries, absorbed doses to the fetal brain and body are calculated; for the embryo at 8 weeks and the fetus at 3, 6 or 9 months. Neutron fluence-to-absorbed dose conversion coefficients are derived for the four age groups. Neutron fluence-to-equivalent dose conversion coefficients are given for the AP irradiations which yield the highest radiation dose to the fetal body in the neutron energy range considered here. The results indicate that for neutrons <10 MeV more protection should be given to pregnant women in the first trimester due to the higher absorbed dose per unit neutron fluence to the fetus. (authors)

  20. Neutron fluence-to-dose conversion coefficients for embryo and fetus.

    Science.gov (United States)

    Chen, Jing; Meyerhof, Dorothy; Vlahovich, Slavica

    2004-01-01

    A problem of concern in radiation protection is the exposure of pregnant women to ionising radiation, because of the high radiosensitivity of the embryo and fetus. External neutron exposure is of concern when pregnant women travel by aeroplane. Dose assessments for neutrons frequently rely on fluence-to-dose conversion coefficients. While neutron fluence-to-dose conversion coefficients for adults are recommended in International Commission on Radiological Protection publications and International Commission on Radiological Units and Measurements reports, conversion coefficients for embryos and fetuses are not given in the publications. This study undertakes Monte Carlo calculations to determine the mean absorbed doses to the embryo and fetus when the mother is exposed to neutron fields. A new set of mathematical models for the embryo and fetus has been developed at Health Canada and is used together with mathematical phantoms of a pregnant female developed at Oak Ridge National Laboratory. Monoenergetic neutrons from 1 eV to 10 MeV are considered in this study. The irradiation geometries include antero-posterior (AP), postero-anterior (PA), lateral (LAT), rotational (ROT) and isotropic (ISO) geometries. At each of these standard irradiation geometries, absorbed doses to the fetal brain and body are calculated; for the embryo at 8 weeks and the fetus at 3, 6 or 9 months. Neutron fluence-to-absorbed dose conversion coefficients are derived for the four age groups. Neutron fluence-to-equivalent dose conversion coefficients are given for the AP irradiations which yield the highest radiation dose to the fetal body in the neutron energy range considered here. The results indicate that for neutrons <10 MeV more protection should be given to pregnant women in the first trimester due to the higher absorbed dose per unit neutron fluence to the fetus.

  1. Development of the processing software package for RPV neutron fluence determination methodology

    International Nuclear Information System (INIS)

    Belousov, S.; Kirilova, K.; Ilieva, K.

    2001-01-01

    According to the INRNE methodology the neutron transport calculation is carried out by two steps. At the first step reactor core eigenvalue calculation is performed. This calculation is used for determination of the fixed source for the next step calculation of neutron transport from the reactor core to the RPV. Both calculation steps are performed by state of the art and tested codes. The interface software package DOSRC developed at INRNE is used as a link between these two calculations. The package transforms reactor core calculation results to neutron source input data in format appropriate for the neutron transport codes (DORT, TORT and ASYNT) based on the discrete ordinates method. These codes are applied for calculation of the RPV neutron flux and its responses - induced activity, radiation damage, neutron fluence etc. Fore more precise estimation of the neutron fluence, the INRNE methodology has been supplemented by the next improvements: - implementation of more advanced codes (PYTHIA/DERAB) for neutron-physics parameter calculations; - more detailed neutron source presentation; - verification of neutron fluence by statistically treated experimental data. (author)

  2. Extraction of neutron spectral information from Bonner-Sphere data

    CERN Document Server

    Haney, J H; Zaidins, C S

    1999-01-01

    We have extended a least-squares method of extracting neutron spectral information from Bonner-Sphere data which was previously developed by Zaidins et al. (Med. Phys. 5 (1978) 42). A pulse-height analysis with background stripping is employed which provided a more accurate count rate for each sphere. Newer response curves by Mares and Schraube (Nucl. Instr. and Meth. A 366 (1994) 461) were included for the moderating spheres and the bare detector which comprise the Bonner spectrometer system. Finally, the neutron energy spectrum of interest was divided using the philosophy of fuzzy logic into three trapezoidal regimes corresponding to slow, moderate, and fast neutrons. Spectral data was taken using a PuBe source in two different environments and the analyzed data is presented for these cases as slow, moderate, and fast neutron fluences. (author)

  3. Development of neutron fluence measurement and evaluation technology for the test materials in the capsule

    Energy Technology Data Exchange (ETDEWEB)

    Hong, U.; Choi, S. H.; Kang, H. D. [Kyungsan University, Kyungsan (Korea)

    2000-03-01

    The four kinds of the fluence monitor considered by self-shielding are design and fabricated for evaluation of neutron irradiation fluence. They are equipped with dosimeters consisting of Ni, Fe and Ti wires and so forth. The nuclear reaction rate is obtained by measurement on dosimeter using the spectroscopic analysis of induced {gamma}-ray. We established the nuetron fluence evaluating technology that is based on the measurement of the reaction rate considering reactor's irradiation history, burn-out, self-shielding in fluence monitor, and the influence of impurity in dosimeter. The distribution of high energy neutron flux on the vertical axis of the capsule shows fifth order polynomial equation and is good agree with theoretical value in the error range of 30% by MCNP/4A code. 22 refs., 50 figs., 27 tabs. (Author)

  4. Superconductivity in irradiated A-15 compounds at low fluences. I. Neutron-irradiated V3Si

    International Nuclear Information System (INIS)

    Viswanathan, R.; Caton, R.; Pande, C.S.

    1978-01-01

    The behavior of the superconducting transition temperature T/sub c/ of single-crystal and polycrystalline V 3 Si was investigated as a function of low-fluence neutron irradiation. It is found that the initial degradation of T/sub c/ is sample-dependent, some specimens showing no degradation in T/sub c/ up to a fluence of 2 x 10 18 n/cm 2 . This and many other earlier observations on low-fluence behavior are explained in terms of a recently proposed model of radiation damage in A-15 compounds

  5. Calculation of neutron fluence to dose equivalent conversion coefficients using GEANT4

    International Nuclear Information System (INIS)

    Ribeiro, Rosane M.; Santos, Denison de S.; Queiroz Filho, Pedro P. de; Mauricio, CLaudia L.P.; Silva, Livia K. da; Pessanha, Paula R.

    2014-01-01

    Fluence to dose equivalent conversion coefficients provide the basis for the calculation of area and personal monitors. Recently, the ICRP has started a revision of these coefficients, including new Monte Carlo codes for benchmarking. So far, little information is available about neutron transport below 10 MeV in tissue-equivalent (TE) material performed with Monte Carlo GEANT4 code. The objective of this work is to calculate neutron fluence to personal dose equivalent conversion coefficients, H p (10)/Φ, with GEANT4 code. The incidence of monoenergetic neutrons was simulated as an expanded and aligned field, with energies ranging between thermal neutrons to 10 MeV on the ICRU slab of dimension 30 x 30 x 15 cm 3 , composed of 76.2% of oxygen, 10.1% of hydrogen, 11.1% of carbon and 2.6% of nitrogen. For all incident energy, a cylindrical sensitive volume is placed at a depth of 10 mm, in the largest surface of the slab (30 x 30 cm 2 ). Physic process are included for neutrons, photons and charged particles, and calculations are made for neutrons and secondary particles which reach the sensitive volume. Results obtained are thus compared with values published in ICRP 74. Neutron fluence in the sensitive volume was calculated for benchmarking. The Monte Carlo GEANT4 code was found to be appropriate to calculate neutron doses at energies below 10 MeV correctly. (author)

  6. Feasibility study on using imaging plates to estimate thermal neutron fluence in neutron-gamma mixed fields

    International Nuclear Information System (INIS)

    Fujibuchi, T.; Tanabe, Y.; Sakae, T.; Terunuma, T.; Isobe, T.; Kawamura, H.; Yasuoka, K.; Matsumoto, T.; Harano, H.; Nishiyama, J.; Masuda, A.; Nohtomi, A.

    2011-01-01

    In current radiotherapy, neutrons are produced in a photonuclear reaction when incident photon energy is higher than the threshold. In the present study, a method of discriminating the neutron component was investigated using an imaging plate (IP) in the neutron-gamma-ray mixed field. Two types of IP were used: a conventional IP for beta- and gamma rays, and an IP doped with Gd for detecting neutrons. IPs were irradiated in the mixed field, and the photo-stimulated luminescence (PSL) intensity of the thermal neutron component was discriminated using an expression proposed herein. The PSL intensity of the thermal neutron component was proportional to thermal neutron fluence. When additional irradiation of photons was added to constant neutron irradiation, the PSL intensity of the thermal neutron component was not affected. The uncertainty of PSL intensities was approximately 11.4 %. This method provides a simple and effective means of discriminating the neutron component in a mixed field. (authors)

  7. International intercomparison of fluence of fast neutrons using 115In(n,γ) activation

    International Nuclear Information System (INIS)

    Lesiecki, H.; Cosack, M.

    1985-07-01

    The Physikalisch-Technische Bundesanstalt (PTB) has participated in an international intercomparison of fluence measurements of fast neutrons. This was organized under the auspices of the ''Comite Consultatif pour les Etalons de Mesure des Rayonnements Ionisants (CCEMRI)'', Sect. 3 (Mesures Neutronique). The National Physical Laboratory (NPL), Teddington, UK volunteered to assume responsibility for the experimental realization and final evaluation. This report deals with the measurements performed at the PTB for the neutron fluence intercomparison at neutron energies of Esub(n) = 144 keV and 570 keV which was based on the 115 In(n,γ) 116 Insup(m) reaction. The count rate of a 4πβ-counter which had to be used to determine the activation of the In sample was to be compared with the neutron fluence by which the sample was irradiated. A description of the neutron production, the fluence determination, the 4πβ-counting, and the evaluation of the results will be given. (orig.) [de

  8. Study on measurement technique contrast of 14 MeV neutron fluence

    International Nuclear Information System (INIS)

    Jiang Li; Hu Jun; Wen Dezhi

    2005-10-01

    The stability and repetition of the associated-particle method to measure DT neutron fluence was tested. The neutron activation iron method was contrasted with the associated-particle method, the preparatory experiment was done. The neutron fluence measured with associated-particle method was contrasted with neutron activation Al method, the Al activated foil was measured with 4πβ (PC)-γ coincidence standard device. The contrast result's standard deviation of the two method was less than the expand uncertainty of the associated-particle method. Therein, the uncertainty of the associated-particle method is 1.6%, the uncertainty of the activation Al method is 1.8%. (authors)

  9. The plasma focus as a large fluence neutron source

    International Nuclear Information System (INIS)

    Zucker, O.; Bostick, W.; Long, J.; Luce, J.; Sahlin, H.

    1977-01-01

    A continuously operated, 1 pps, dense-plasma-focus device capable of delivering a minimum of 10 15 neutrons per pulse for material testing purposes is described. With I 5 scaling, predicted from analysis of existing machines, yields of 10 16 -10 17 neutrons per pulse are postulated. The average power consumption, which has become a major issue as a result of the energy crisis is shown to be highly favorable. A novel approach to the capacitor bank and switch design allowing repetitive operation is discussed. (Auth.)

  10. Absolute measurement and international intercomparison of 0.1-0.8 MeV monoenergetic neutron fluence rate

    International Nuclear Information System (INIS)

    Ma Hongchang; Lu Hanlin; Rong Chaofan

    1988-01-01

    The methods for absolute measurement of 0.1-18MeV monoenergetic neutron fluence rate are described. Which include proton recoil telescope, semicoducetor telescope, hydrogen filled proportional counter and associated particale method. A long counter used as secondary recent international intercomparison of neutron fluence rate organized by BIPM, and the results were given

  11. Measured thermal and fast neutron fluence rates for ATF-1 holders during ATR cycle 160A

    International Nuclear Information System (INIS)

    Walker, B. J.; Miller, D. T.

    2017-01-01

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 160A which were measured by the Radiation Measurements Laboratory (RML).

  12. Measured thermal and fast neutron fluence rates for ATF-1 holders during ATR cycle 160A

    Energy Technology Data Exchange (ETDEWEB)

    Walker, B. J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, D. T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-06-06

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 160A which were measured by the Radiation Measurements Laboratory (RML).

  13. Spectral distribution of particle fluence in small field detectors and its implication on small field dosimetry.

    Science.gov (United States)

    Benmakhlouf, Hamza; Andreo, Pedro

    2017-02-01

    photoabsorption in the high-Z shielding material. For the ionization chambers and the diamond detector, the electron fluence spectra were found to be similar to that in water, for both field sizes. In contrast, electron spectra in the silicon diodes were much higher than that in water for both field sizes. The estimated perturbations of the fluence spectra for the silicon diodes were 11-21% for the large fields and 14-27% for the small fields. These perturbations are related to the atomic number, density and mean excitation energy (I-value) of silicon, as well as to the influence of the "extracameral"' components surrounding the detector sensitive volume. For most detectors the fluence perturbation was also found to increase when the field size was decreased, in consistency with the increased small-field effects observed for the smallest field sizes. The present work improves the understanding of small-field effects by relating output correction factors to spectral fluence perturbations in small field detectors. It is shown that the main reasons for the well-known small-field effects in silicon diodes are the high-Z and density of the "extracameral" detector components and the high I-value of silicon relative to that of water and diamond. Compared to these parameters, the density and atomic number of the radiation sensitive volume material play a less significant role. © 2016 American Association of Physicists in Medicine.

  14. About reliability of WWER pressure vessel neutron fluence calculation

    Energy Technology Data Exchange (ETDEWEB)

    Belousov, S; Ilieva, K; Antonov, S [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    This reliability study was carried out under a Research Contracts F111 and TH-324 of the Bulgarian Ministry of Higher Education and the IAEA. The effect of geometry approximation and the choice of neutron cross-sections on the calculation model is estimated. The neutron flux onto reactor pressure vessel at locations, important for metal embrittlement surveillance, has been calculated using the codes TORT and DORT. The flux values calculated for both WWER-440 and WWER-1000 show good consistency within the limits of solution accuracy. It is concluded that the synthesis method (DORT) can be used for calculations at a reasonable cost whenever metal embrittlement surveillance is considered. Using an iron sphere benchmark measurement, a comparison of an experimental leakage spectrum with spectrum calculated using multigroup neutron cross-sections based on ENDF/B-4 and ENDF/B-6 data is performed. In the energy region above 1 MeV the best agreement with the experiment is achieved for ENDF/B-6 in VITAMIN-E group structure. 7 refs., 1 fig., 4 tabs.

  15. About reliability of WWER pressure vessel neutron fluence calculation

    International Nuclear Information System (INIS)

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

    1995-01-01

    This reliability study was carried out under a Research Contracts F111 and TH-324 of the Bulgarian Ministry of Higher Education and the IAEA. The effect of geometry approximation and the choice of neutron cross-sections on the calculation model is estimated. The neutron flux onto reactor pressure vessel at locations, important for metal embrittlement surveillance, has been calculated using the codes TORT and DORT. The flux values calculated for both WWER-440 and WWER-1000 show good consistency within the limits of solution accuracy. It is concluded that the synthesis method (DORT) can be used for calculations at a reasonable cost whenever metal embrittlement surveillance is considered. Using an iron sphere benchmark measurement, a comparison of an experimental leakage spectrum with spectrum calculated using multigroup neutron cross-sections based on ENDF/B-4 and ENDF/B-6 data is performed. In the energy region above 1 MeV the best agreement with the experiment is achieved for ENDF/B-6 in VITAMIN-E group structure. 7 refs., 1 fig., 4 tabs

  16. Electrical and optical analyses of low fluence fast neutron damage to JFETs

    International Nuclear Information System (INIS)

    Hoffmann, A.; Charles, J.P.; Kerns, S.E.; Kerns, D.V. Jr.; Bardonnie, M. de la; Mialhe, P.

    1999-01-01

    The effects of fast neutron irradiation (30 MeV) on silicon n-channel JFETs are studied. Electrical parameters of the gate-channel junction are analysed at 3 fluences: 4,06*10 10 , 8,12*10 10 and 1,22*10 11 n/cm 2 for a flux of 2,82*10 6 n/s*cm 2 and using a custom software. Electrical parameter changes are attributed to bulk semi-conductor defects. Irradiation effects on passivation overlayers are evacuate using analysis of gate-channel junction electroluminescence. This study shows that even for low neutron fluences (10 11 n/cm 2 ), n-channel JFETs, characterized in direct conducting mode and submitted to neutron radiation, present a decrease in the reverse saturation current associated with recombination. (A.C.)

  17. Neutron irradiation effects on intermetallic precipitates in Zircaloy as a function of fluence

    International Nuclear Information System (INIS)

    Etoh, Y.; Shimada, S.

    1993-01-01

    Intermetallic precipitates in Zircaloy-2 and -4, recrystallized at the α-phase temperature, have been examined using analytical electron microscopy. The specimens were irradiated in BWRs up to a fast neutron fluence of 1.4x10 26 n/m 2 (E>1 MeV). Neutron irradiation induces a crystalline-to-amorphous transition, depleting Fe in the amorphous phase of Zr(Fe, Cr) 2 precipitates in the alloys. Amorphization starts from the periphery of the precipitates and all of them are totally amorphized at higher fluences than 1.2x10 26 n/m 2 . The width of the Fe-depleted zone increases in proportion to the 0.45 power of fluence. This result indicates that diffusion of Fe is the rate-controlling process for Fe depletion in Zr(Fe, Cr) 2 precipitates. Dissolution of Zr 2 (Fe, Ni) precipitates in Zircaloy-2 occurs during neutron irradiation. At a high fluence, such as 1.2x10 26 n/m 2 , Zr 2 (Fe, Ni) precipitates are almost completely dissolved into the matrix and the dissolution rate of Fe is faster than that of Ni. (orig.)

  18. Measurement of low neutron-fluences using electrochemically etched PC and PET track detectors

    International Nuclear Information System (INIS)

    Somogyi, G.; Dajko, G.; Turek, K.; Spurny, F.

    1979-01-01

    Systematic investigations have been carried out to study different properties of electrochemically etched (ECE) polycarbonate (PC) and polyethylene-terephthalate (PET) foils. The dependence of the density of background discharge spots on surface-thickness removal, electrical field strength and frequency of voltage is given. The effect of these parameters on the neutron sensitivity of polycarbonate and polyethylene-terephthalate foils irradiated at right angles to 14.7 MeV, 241 Am-Be and 252 Cf neutrons is also studied. With knowledge of the background and sensitivity data, the etching and electrical parameters are optimized for low neutron-fluence measurements. (author)

  19. Neutron fluence-to-dose equivalent conversion factors: a comparison of data sets and interpolation methods

    International Nuclear Information System (INIS)

    Sims, C.S.; Killough, G.G.

    1983-01-01

    Various segments of the health physics community advocate the use of different sets of neutron fluence-to-dose equivalent conversion factors as a function of energy and different methods of interpolation between discrete points in those data sets. The major data sets and interpolation methods are used to calculate the spectrum average fluence-to-dose equivalent conversion factors for five spectra associated with the various shielded conditions of the Health Physics Research Reactor. The results obtained by use of the different data sets and interpolation methods are compared and discussed. (author)

  20. Measured thermal and fast neutron fluence rates ATR Cycle 101-B, October 11, 1993--November 27, 1993

    International Nuclear Information System (INIS)

    Murray, R.K.; Rogers, J.W.

    1994-01-01

    This report contains the thermal (2200 m/s) and fast (E>lMeV) neutron fluence rate data for ATR Cycle 101-B which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order. This report contains fluence rate values corresponding to the particular elevations (relative to the 80 ft. core elevation) where the measurements were taken. The data in this report consists of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, (3) plots of both the thermal and fast neutron fluence rates, and (4) a magnetic record (3.5 inch diskette) containing a listing of only the fast neutron fluence rates, their assigned elevations proper header identification of all monitor positions contained herein

  1. A new expression for determination of fluences from a spherical moderator neutron source for the calibration of spherical neutron measuring devices

    International Nuclear Information System (INIS)

    Khoshnoodi, M.; Sohrabi, M.

    1997-01-01

    A new expression modifying the inverse square law for determination of neutron fluences from spherical moderator neutron sources is reported. The formalism is based on the neutron fluence at a point outside the moderator as the summation of fluxes of two groups of neutrons: direct neutrons from the central region of the moderator, and moderated neutrons which, to a first approximation, are scattered from the outermost layers of the spherical moderator. The expression has been further developed for spherical neutron measuring devices with an appropriate geometry factor which corrects the reading of the device for non-uniform irradiation of the detector. The combination of the new fluence function and those of the air and room scattered components introduce a calibration model. The fluence relationship obtained for moderated sources may conveniently be used for calculating the more rapid change of neutron dose at close distances than that which is based on the inverse square dependence. (author)

  2. A new method for the determination of unknown neutron fluence for 14.0 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Malik, Fariha [Physics Reasearch Division, PINSTECH, Nilore, Islamabad (Pakistan)]. E-mail: fariha@pinstech.org.pk; Khan, Ehsan U. [Department of Physics, CIIT, Islamabad (Pakistan); Qureshi, Imtinan [Physics Reasearch Division, PINSTECH, Nilore, Islamabad (Pakistan); Husaini, Syed N. [Physics Reasearch Division, PINSTECH, Nilore, Islamabad (Pakistan); Ahmad, Waqar [Physics Reasearch Division, PINSTECH, Nilore, Islamabad (Pakistan); Rajput, Usman [Physics Reasearch Division, PINSTECH, Nilore, Islamabad (Pakistan); Raza, Qaiser [Applied Physics Division, PINSTECH, Nilore, Islamabad (Pakistan)

    2006-11-15

    Measuring the correct neutron fluence in various energy intervals in and around the neutron sources is important for the purpose of personnel and environmental neutron dosimetry. In this paper, we present a new method for the measurement of the fluence of mono-energetic neutrons having the energy of 14.0 MeV. The samples exposed to neutrons from the 14.0 MeV neutron generator at PINSTECH with various fluence values ranging from 10{sup 7} to 10{sup 10} n cm{sup -2} were etched for 10 min in 6 N NaOH at 70.0{+-}1.0 {sup o}C and the transmittance of UV radiation was measured using a spectrophotometer. This procedure was repeated 20 times after etching the same sample each time for increasing time intervals till the stage when transmittance reached the constant minimum value. An exponential decay of the transmittance has been observed with respect to the increasing etching time interval in each of the samples exposed to various neutron fluence. Further, it has also been observed that there is a linear relationship between the transmittance decay constant and neutron fluence. Hence, the linear graph can be used as a calibration for measuring the unknown fluence of 14.0 MeV neutrons.

  3. 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)

  4. Thickness optimization of various moderator materials for maximization of thermal neutron fluence

    International Nuclear Information System (INIS)

    Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

    2015-01-01

    Plasma focus device is widely being used as pulsed neutron source for variety of applications. Measurements of neutron yield by largely preferred Helium-3 proportional counter and Silver activation counter are mainly sensitive to thermal neutrons and are typically used with a neutron moderator. Thermalization of neutron is based on scattering reaction and hydrogenous materials are the best thermalizing medium. The efficiency of aforementioned neutron detectors is considerably affected by physical and geometrical properties of thermalizing medium i.e. moderator material, its thickness and shape. In view of the same, simulations have been performed to explore the effective utilization of Polyethylene, Perspex and Light water as moderating mediums for cylindrical and spherical geometry. In this study, estimated thermal fluence value up to 0.5 eV has been considered as the benchmark factor for comparing efficient thermalization by specific material, its thickness and shape. In either of the shapes being cylindrical or spherical, use of Polyethylene as moderating medium has resulted in minimum optimum thickness along with highest thermal fluence. (author)

  5. Neutron fluence measurement in the cavity of Balakovo nuclear power plant, unit 3

    International Nuclear Information System (INIS)

    Voorbraak, W.P.; Baard, J.H.; Paardekooper, A.; Nolthenius, H.J.

    1996-12-01

    An international benchmark exercise has been organized by the Russian GOSATOMNADZOR. The aim was to reduce the uncertainty of fluence measurements in Nuclear Power Plants in particular VVER-1000 reactors. The benchmark was set up in the cavity of the Balakovo NPP 3. Eight institutes were involved. This report presents the results obtained by ECN. From this report, it can be concluded that the results of the relative large monitor set (13 different reaction rates with overlapping response regions) point to possible imperfections in the calculated neutron spectra. However the experimental information is not powerful enough to reduce the uncertainty of the neutron fluence rate especially in the energy region between 0.1 and 0.5 MeV below 50 percent. (orig.)

  6. Fast neutron fluence calculations as support for a BWR pressure vessel and internals surveillance program

    International Nuclear Information System (INIS)

    Lucatero, Marco A.; Palacios-Hernandez, Javier C.; Ortiz-Villafuerte, Javier; Xolocostli-Munguia, J. Vicente; Gomez-Torres, Armando M.

    2010-01-01

    Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1. The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are

  7. Effect of high fluence neutron irradiation on transport properties of thermoelectrics

    Science.gov (United States)

    Wang, H.; Leonard, K. J.

    2017-07-01

    Thermoelectric materials were subjected to high fluence neutron irradiation in order to understand the effect of radiation damage on transport properties. This study is relevant to the NASA Radioisotope Thermoelectric Generator (RTG) program in which thermoelectric elements are exposed to radiation over a long period of time in space missions. Selected n-type and p-type bismuth telluride materials were irradiated at the High Flux Isotope Reactor with a neutron fluence of 1.3 × 1018 n/cm2 (E > 0.1 MeV). The increase in the Seebeck coefficient in the n-type material was partially off-set by an increase in electrical resistivity, making the power factor higher at lower temperatures. For the p-type materials, although the Seebeck coefficient was not affected by irradiation, electrical resistivity decreased slightly. The figure of merit, zT, showed a clear drop in the 300-400 K range for the p-type material and an increase for the n-type material. Considering that the p-type and n-type materials are connected in series in a module, the overall irradiation damages at the device level were limited. These results, at neutron fluences exceeding a typical space mission, are significant to ensure that the radiation damage to thermoelectrics does not affect the performance of RTGs.

  8. Standard Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 The purpose of this test method is to define a general procedure for determining an unknown thermal-neutron fluence rate by neutron activation techniques. It is not practicable to describe completely a technique applicable to the large number of experimental situations that require the measurement of a thermal-neutron fluence rate. Therefore, this method is presented so that the user may adapt to his particular situation the fundamental procedures of the following techniques. 1.1.1 Radiometric counting technique using pure cobalt, pure gold, pure indium, cobalt-aluminum, alloy, gold-aluminum alloy, or indium-aluminum alloy. 1.1.2 Standard comparison technique using pure gold, or gold-aluminum alloy, and 1.1.3 Secondary standard comparison techniques using pure indium, indium-aluminum alloy, pure dysprosium, or dysprosium-aluminum alloy. 1.2 The techniques presented are limited to measurements at room temperatures. However, special problems when making thermal-neutron fluence rate measurements in high-...

  9. Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ∼ 25 effective full power years of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M and calculated (C results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE/C ratios of 1.10 for both neutron (E >1.0 MeV flux and iron atom displacement rate.

  10. RAMA Methodology for the Calculation of Neutron Fluence; Metodologia RAMA para el Calculo de la Fluencia Neutronica

    Energy Technology Data Exchange (ETDEWEB)

    Villescas, G.; Corchon, F.

    2013-07-01

    he neutron fluence plays an important role in the study of the structural integrity of the reactor vessel after a certain time of neutron irradiation. The NRC defined in the Regulatory Guide 1.190, the way must be estimated neutron fluence, including uncertainty analysis of the validation process (creep uncertainty is ? 20%). TRANSWARE Enterprises Inc. developed a methodology for calculating the neutron flux, 1,190 based guide, known as RAMA. Uncertainty values obtained with this methodology, for about 18 vessels, are less than 10%.

  11. A new Recoil Proton Telescope for energy and fluence measurement of fast neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Lebreton, Lena; Bachaalany, Mario [IRSN / LMDN (Institut de Radioprotection et de Surete nucleaire / Laboratoire de Metrologie et de dosimetrie des neutrons), Cadarache Bat.159, 13115 Saint Paul-lez-Durance, (France); Husson, Daniel; Higueret, Stephane [IPHC / RaMsEs (Institut Pluridisciplinaire Hubert Curien / Radioprotection et Mesures Environnementales), 23 rue du loess - BP28, 67037 Strasbourg cedex 2, (France)

    2015-07-01

    The spectrometer ATHENA (Accurate Telescope for High Energy Neutron metrology Applications), is being developed at the IRSN / LMDN (Institut de Radioprotection et de Surete nucleaire / Laboratoire de Metrologie et de dosimetrie des neutrons) and aims at characterizing energy and fluence of fast neutron fields. The detector is a Recoil Proton Telescope and measures neutron fields in the range of 5 to 20 MeV. This telescope is intended to become a primary standard for both energy and fluence measurements. The neutron detection is achieved by a polyethylene radiator for n-p conversion, three 50{sub m} thick silicon sensors that use CMOS technology for the proton tracking and a 3 mm thick silicon diode to measure the residual proton energy. This first prototype used CMOS sensors called MIMOSTAR, initially developed for heavy ion physics. The use of CMOS sensors and silicon diode increases the intrinsic efficiency of the detector by a factor of ten compared with conventional designs. The first prototype has already been done and was a successful study giving the results it offered in terms of energy and fluence measurements. For mono energetic beams going from 5 to 19 MeV, the telescope offered an energy resolution between 5 and 11% and fluence difference going from 5 to 7% compared to other home standards. A second and final prototype of the detector is being designed. It will hold upgraded CMOS sensors called FastPixN. These CMOS sensors are supposed to run 400 times faster than the older version and therefore give the telescope the ability to support neutron flux in the order of 107 to 108cm{sup 2}:s{sup 1}. The first prototypes results showed that a 50 m pixel size is enough for a precise scattering angle reconstruction. Simulations using MCNPX and GEANT4 are already in place for further improvements. A DeltaE diode will replace the third CMOS sensor and will be installed right before the silicon diode for a better recoil proton selection. The final prototype with

  12. Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

    International Nuclear Information System (INIS)

    Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

    2011-01-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

  13. Fluence measurement at the neutron time of flight experiment at CERN

    CERN Document Server

    Weiss, Christina; Jericha, Erwin

    At the neutron time of flight facility n_TOF at CERN a new spallation target was installed in 2008. In 2008 and 2009 the commissioning of the new target took place. During the summer 2009 a fission chamber of the Physikalisch Technische Bundesanstalt (PTB) Braunschweig was used for the neutron fluence measurement. The evaluation of the data recorded with this detector is the primary topic of this thesis. Additionally a neutron transmission experiment with air has been performed at the TRIGA Mark II reactor of the Atomic Institute of the Austrian Universities (ATI). The experiment was implemented to clarify a question about the scattering cross section of molecular gas which could not be answered clearly via the literature. This problem came up during the evaluations for n_TOF.

  14. Measurement of thermal neutron fluence with CaSO4 thermoluminescent phosphors

    International Nuclear Information System (INIS)

    Liu Jinhua; Su Jingling; Wei Zemin

    1984-01-01

    During neutron irradiation, some TL phosphors were activated. After leaving the irradiation field the TL phosphor produced self-irradiation. The TL output of self-dose was only related to the original neutron fluence and independent of the γ-radiation. Several CaSO 4 TL phosphors were made. They were CaSO 4 :Dy, CaSO 4 :Dy-Teflon, CaSO 4 :Dy mixed with Dy 2 O 3 , CaSO 4 :Mn mixed with Dy 2 O 3 . The linearity, and lower detection limits of these TL phosphors were measured. The thermal neutron response of CaSO 4 :Mn mixed with Dy 2 O 3 was 64 R/(10 10 cm -2 ) and the lower detection limit was 1.3x10 5 cm -2

  15. Validation of neutron-transport calculations in benchmark facilities for improved damage-fluence predictions

    International Nuclear Information System (INIS)

    Williams, M.L.; Stallmann, F.W.; Maerker, R.E.; Kam, F.B.K.

    1983-01-01

    An accurate determination of damage fluence accumulated by reactor pressure vessels (RPV) as a function of time is essential in order to evaluate the vessel integrity for both pressurized thermal shock (PTS) transients and end-of-life considerations. The desired accuracy for neutron exposure parameters such as displacements per atom or fluence (E > 1 MeV) is of the order of 20 to 30%. However, these types of accuracies can only be obtained realistically by validation of nuclear data and calculational methods in benchmark facilities. The purposes of this paper are to review the needs and requirements for benchmark experiments, to discuss the status of current benchmark experiments, to summarize results and conclusions obtained so far, and to suggest areas where further benchmarking is needed

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  17. Measuring neutron fluences and gamma/x-ray fluxes with CCD cameras

    International Nuclear Information System (INIS)

    Yates, G.J.; Smith, G.W.; Zagarino, P.; Thomas, M.C.

    1991-01-01

    The capability to measure bursts of neutron fluences and gamma/x-ray fluxes directly with charge coupled device (CCD) cameras while being able to distinguish between the video signals produced by these two types of radiation, even when they occur simultaneously, has been demonstrated. Volume and area measurements of transient radiation-induced pixel charge in English Electric Valve (EEV) Frame Transfer (FT) charge coupled devices (CCDs) from irradiation with pulsed neutrons (14 MeV) and Bremsstrahlung photons (4--12 MeV endpoint) are utilized to calibrate the devices as radiometric imaging sensors capable of distinguishing between the two types of ionizing radiation. Measurements indicate ∼.05 V/rad responsivity with ≥1 rad required for saturation from photon irradiation. Neutron-generated localized charge centers or ''peaks'' binned by area and amplitude as functions of fluence in the 10 5 to 10 7 n/cm 2 range indicate smearing over ∼1 to 10% of CCD array with charge per pixel ranging between noise and saturation levels

  18. Calculation of neutron fluence-to-dose conversion factors for extremities

    International Nuclear Information System (INIS)

    Stewart, R.D.; Harty, R.; McDonald, J.C.; Tanner, J.E.

    1993-04-01

    The Pacific Northwest Laboratory is developing a standard for the performance testing of personnel extremity dosimeters for the US Department of Energy. Part of this effort requires the calculation of neutron fluence-to-dose conversion factors for finger and wrist extremities. This study focuses on conversion factors for two types of extremity models: namely the polymethyl methacrylate (PMMA) phantom (as specified in the draft standard for performance testing of extremity dosimeters) and more realistic extremity models composed of tissue-and-bone. Calculations for each type of model are based on both bare and D 2 O-moderated 252 Cf sources. The results are then tabulated and compared with whole-body conversion factors. More appropriate energy-averaged quality factors for the extremity models have also been computed from the neutron fluence in 50 equally spaced energy bins with energies from 2.53 x 10 -8 to 15 MeV. Tabulated results show that conversion factors for both types of extremity phantom are 15 to 30% lower than the corresponcung whole-body phantom conversion factors for 252 Cf neutron sources. This difference in extremity and whole-body conversion factors is attributable to the proportionally smaller amount of back-scattering that occurs in the extremity phantoms compared with whole-body phantoms

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

    International Nuclear Information System (INIS)

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

    2015-09-01

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

  20. Formation of austenite in high Cr ferritic/martensitic steels by high fluence neutron irradiation

    Science.gov (United States)

    Lu, Z.; Faulkner, R. G.; Morgan, T. S.

    2008-12-01

    High Cr ferritic/martensitic steels are leading candidates for structural components of future fusion reactors and new generation fission reactors due to their excellent swelling resistance and thermal properties. A commercial grade 12%CrMoVNb ferritic/martensitic stainless steel in the form of parent plate and off-normal weld materials was fast neutron irradiated up to 33 dpa (1.1 × 10 -6 dpa/s) at 400 °C and 28 dpa (1.7 × 10 -6 dpa/s) at 465 °C, respectively. TEM investigation shows that the fully martensitic weld metal transformed to a duplex austenite/ferrite structure due to high fluence neutron irradiation, the austenite was heavily voided (˜15 vol.%) and the ferrite was relatively void-free; whilst no austenite phases were detected in plate steel. Thermodynamic and phase equilibria software MTDATA has been employed for the first time to investigate neutron irradiation-induced phase transformations. The neutron irradiation effect is introduced by adding additional Gibbs free energy into the system. This additional energy is produced by high energy neutron irradiation and can be estimated from the increased dislocation loop density caused by irradiation. Modelling results show that neutron irradiation reduces the ferrite/austenite transformation temperature, especially for high Ni weld metal. The calculated results exhibit good agreement with experimental observation.

  1. Fast neutron fluence evaluation of the smart reactor pressure vessel by using the GEOSHIELD code

    International Nuclear Information System (INIS)

    Kim, K.Y.; Kim, K.S.; Kim, H.Y.; Lee, C.C.; Zee, S.Q.

    2007-01-01

    In Korea, the design of an advanced integral reactor system called SMART has been developed by KAERI to supply energy for seawater desalination as well as an electricity generation. A fast neutron fluence distribution at the SMART reactor pressure vessel was evaluated to confirm the integrity of the vessel by using the GEOSHIELD code. The GEOSHIELD code was developed by KAERI in order to prepare an input list including a geometry modeling of the DORT code and to process results from the DORT code output list. Results by a GEOSHIELD code processing and by a manual processing of the DORT show a good agreement. (author)

  2. Neutron fluence in a 18 MeV Electron Accelerator for Therapy

    International Nuclear Information System (INIS)

    Paredes G, L.C.

    2001-01-01

    An investigation was made on the theoretical fundamentals for the determination of the neutron fluence in a linear electron accelerator for radiotherapy applications and the limit values of leakage neutron radiation established by guidelines and standards in radiation protection for these type of accelerators. This investigation includes the following parts: a) Exhaustive bibliographical review on the topics mentioned above, in order to combine and to update the necessary basic information to facilitate the understanding of this subject; b) Analysis of the accelerator operation and identification of its main components, specially in the accelerator head; c) Study of different types of targets and its materials for the Bremsstrahlung production which is based on the electron initial energy, the thickness of the target, and its angular distribution and energy, which influences in the neutron generation by means of the photonuclear and electro disintegration reactions; d) Analysis of the neutron yield based on the target type and its thickness, the energy of electrons and photons; e) Analysis of the neutron energy spectra generated in the accelerator head, inside and outside the treatment room; f) Study of the dosimetry fundamentals for neutron and photon mixed fields, the dosimeter selection criteria and standards applied for these applications, specially the Panasonic U D-809 thermoluminescent dosemeter and C R-39 nuclear track dosimeter; g) Theoretical calculation of the neutron yield using a simplified geometric model for the accelerator head with spherical cell, which considers the target, primary collimator, flattener filter, movable collimators and the head shielding as the main components for radiation production. The cases with W and Pb shielding for closed movable collimators and an irradiation field of 20 x 20 cm 2 were analyzed and, h) Experimental evaluation of the leakage neutron radiation from the patient and head planes, observing that the accelerator

  3. Measuring neutron fluences and gamma/x-ray fluxes with CCD cameras

    International Nuclear Information System (INIS)

    Yates, G.J.; Smith, G.W.; Zagarino, P.; Thomas, M.C.

    1991-01-01

    Volume and area measurements of transient radiation-induced pixel charge in English Electric Valve (EEV) Frame Transfer (FT) charge coupled devices (CCDs) from irradiation with pulsed neutrons (14 MeV) and Bremsstrahlung photons (16-MeV endpoint) are utilized to calibrate the devices as radiometric imaging sensors capable of distinguishing between the two types of ionizing radiation. Measurements indicate ∼0.5 V/rad responsivity with ≥1 rad required for saturation from photon irradiation. Neutron-generated localized charge centers or ''peaks'' binned by area and amplitude as functions of fluence in the 10 5 to 10 7 n/cm 2 range indicate smearing over ∼1 to 10% of CCD array with charge per pixel ranging between noise and saturation levels. 9 refs., 12 figs., 4 tabs

  4. Experiments for neutron fluence assessment on WWER-440 and WWER-1000 pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ilieva, K; Apostolov, T; Penev, I; Trifonov, A; Taskaev, E; Belousov, S; Antonov, S; Petrova, T; Stoeva, L [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Boyadzhiev, Z; Nelov, N; Tsocheva, V; Andreeva, I; Lilkov, B; Velichkov, V; Monev, M [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    The activity of shavings sampled out from the expected maximum embrittlement location (weld 4) on the inner pressure vessel wall of the Kozloduy-1 Unit after the 14-th cycle has been measured. The experiment was carried out along the INEI channel using Fe and Cu string and foil detectors. The axial neutron flux distribution at the Unit 3 after the cycle 11 has been measured and compared to the calculated values. The calculations of the expected activities have been carried out taking into account the local power distribution. A comparison between measured and calculated values using ACTIVAT code is made. It shows a discrepancy of about 20%. It is recommended to carry out ex-vessel neutron fluence measurements using a rack device with activation detectors in order to verify the calculation results. 8 refs., 3 figs., 2 tabs.

  5. Neutron fluence determination for operation effectiveness assessment and prediction of WWER pressure vessel lifetime at the Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Apostolov, T; Ilieva, K; Belousov, S; Petrova, T; Antonov, S; Ivanov, K; Prodanova, R; Penev, I; Taskaev, E [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Ivanov, I; Tsokov, P; Nelov, N; Lilkov, B; Tsocheva, V; Monev, M; Velichkov, V; Kharalampieva, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    Embrittlement processes in reactor pressure vessel (RPV) metal have been investigated by neutron dosimetry. A software package for fluence calculations has been developed and used for evaluation of the accumulated neutron fluence, the critical temperature of radiation embrittlement and the RPV lifetime. A digital reactivity meter DR-8 has been introduced for continuous neutron fluence monitoring. Estimates of the neutron fluence and the radiation state of all 6 units of the Kozloduy NPP are presented. The Unit 4 RPV is in the best state regarding metal embrittlement, while the Units 2 and 3 can be safely operated up to the end of their design lifetime only using dummy cassettes. The neutron fluence accumulation in the Unit 1 RPV is quite big and can not be reduced with annealing. Activity measurements of the Unit 1 internal wall shavings are made after the 14-th cycle which show a good agreement with calculated values (1.10{sup 5} Bq/g). The critical embrittlement temperature of the Units 1 - 4 is estimated as a function of the working cycles. 11 figs., 1 tab.

  6. Influence of the number of energy groups on the accuracy of neutron fluence calculations

    International Nuclear Information System (INIS)

    Barz, H.U.; Konheiser, J.

    1999-01-01

    The question how many groups are necessary to obtain all needed integral quantities for the neutron load of pressure vessels and detector positions outside the vessel with sufficient accuracy is of general interest. Until now, there are no systematic investigations on this question. In principle 3-dimensional consideration is required for such neutron load calculations. Therefore, an estimation of the needed number of groups can be of interest to minimize calculation time. One general problem is the P L -approximation of the angular distributions for the transfers between different groups. For elastic scattering this P L -approximation becomes poorer with increasing number of groups. As deterministic methods generally use the P L -approximation they cannot be used for investigations of the errors caused by the group approximation. We have investigated this problem applying group Monte-Carlo but nearly exact representation of this elastic slowing down without P L -approximation. The calculations were directed to assess the neutron fluence of a Russian WWER-1000 reactor. For that a simplified geometrical model of this reactor type has been used. (orig.)

  7. Electrical characterization of 10B doped diamond irradiated with low thermal neutron fluence

    International Nuclear Information System (INIS)

    Reed, M.L.; Reed, M.J.; Jagannadham, K.; Verghese, K.; Bedair, S.M.; El-Masry, N.; Butler, J.E.

    2004-01-01

    A sample of 10 B isotope doped diamond was neutron irradiated to a thermal fluence of 1.3x10 19 neutron cm -2 . The diamond sample was cooled continuously during irradiation in a nuclear reactor. 7 Li is formed by nuclear transmutation reaction from 10 B. Characterization for electrical conductance in the temperature range of 160 K 10 B doped sample and the 10 B doped and irradiated sample. The unirradiated diamond sample showed p-type conductance at higher temperature (T>200 K) and p-type surface conductance at lower temperature (T 7 Li that is formed by nuclear transmutation reaction from 10 B atoms. Also, compensation of n-type carriers from 7 Li by p-type carriers from 10 B is used to interpret the conductance above 400 K. A low concentration of radiation induced defects, absence of defect complexes, and the low activation energy of n-type 7 Li are thought responsible for the observed variation of conductance in the irradiated diamond. The present results illustrate that neutron transmutation from 10 B doped diamond is a useful method to achieve n-type conductivity in diamond

  8. EFFECTS OF IRRADIATION ON THERMAL CONDUCTIVITY OF ALLOY 690 AT LOW NEUTRON FLUENCE

    Directory of Open Access Journals (Sweden)

    WOO SEOG RYU

    2013-04-01

    Full Text Available Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7−28 × 1019n/cm2 (E>0.1MeV at 250°C, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at 250°C did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1% at fluences of (0.7∼28 × 1019n/cm2 (E>0.1MeV.

  9. Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

    International Nuclear Information System (INIS)

    Lee, V.D.; Gohar, Y.

    1985-01-01

    A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of builtup, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

  10. Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

    International Nuclear Information System (INIS)

    Lee, V.D.; Gohar, Y.

    1985-01-01

    A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of built-up, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

  11. Tritium release kinetics in lithium orthosilicate ceramic pebbles irradiated with low thermal-neutron fluence

    International Nuclear Information System (INIS)

    Xiao, Chengjian; Gao, Xiaoling; Kobayashi, Makoto; Kawasaki, Kiyotaka; Uchimura, Hiromichi; Toda, Kensuke; Kang, Chunmei; Chen, Xiaojun; Wang, Heyi; Peng, Shuming; Wang, Xiaolin; Oya, Yasuhisa; Okuno, Kenji

    2013-01-01

    Tritium release kinetics in lithium orthosilicate (Li 4 SiO 4 ) ceramic pebbles irradiated with low thermal-neutron fluence was studied by out-of-pile annealing experiments. It was found that the tritium produced in Li 4 SiO 4 pebbles was mainly released as tritiated water vapor (HTO). The apparent desorption activation energy of tritium on the pebble surface was consistent with the diffusion activation energy of tritium in the crystal grains, indicating that tritium release was mainly controlled by diffusion process. The diffusion coefficients of tritium in the crystal grains at temperatures ranging from 450 K to 600 K were obtained by isothermal annealing tests, and the Arrhenius relation was determined to be D = 1 × 10 −7.0 exp (−40.3 × 10 3 /RT) cm 2 s −1

  12. Tritium release kinetics in lithium orthosilicate ceramic pebbles irradiated with low thermal-neutron fluence

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Chengjian; Gao, Xiaoling [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Box 919-214, Mian Yang 621900 (China); Kobayashi, Makoto; Kawasaki, Kiyotaka; Uchimura, Hiromichi; Toda, Kensuke [China Academy of Engineering Physics, Box 919-1, Mian Yang 621900 (China); Kang, Chunmei; Chen, Xiaojun; Wang, Heyi; Peng, Shuming [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Box 919-214, Mian Yang 621900 (China); Wang, Xiaolin, E-mail: xlwang@caep.ac.cn [China Academy of Engineering Physics, Box 919-1, Mian Yang 621900 (China); Oya, Yasuhisa; Okuno, Kenji [Radiochemistry Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Shizuoka 422-8529 (Japan)

    2013-07-15

    Tritium release kinetics in lithium orthosilicate (Li{sub 4}SiO{sub 4}) ceramic pebbles irradiated with low thermal-neutron fluence was studied by out-of-pile annealing experiments. It was found that the tritium produced in Li{sub 4}SiO{sub 4} pebbles was mainly released as tritiated water vapor (HTO). The apparent desorption activation energy of tritium on the pebble surface was consistent with the diffusion activation energy of tritium in the crystal grains, indicating that tritium release was mainly controlled by diffusion process. The diffusion coefficients of tritium in the crystal grains at temperatures ranging from 450 K to 600 K were obtained by isothermal annealing tests, and the Arrhenius relation was determined to be D = 1 × 10{sup −7.0} exp (−40.3 × 10{sup 3}/RT) cm{sup 2} s{sup −1}.

  13. High Neutron Fluence Survivability Testing of Advanced Fiber Bragg Grating Sensors

    International Nuclear Information System (INIS)

    Fielder, Robert S.; Klemer, Daniel; Stinson-Bagby, Kelly L.

    2004-01-01

    The motivation for the reported research was to support NASA space nuclear power initiatives through the development of advanced fiber optic sensors for space-based nuclear power applications. The purpose of the high-neutron fluence testing was to demonstrate the survivability of fiber Bragg grating (FBG) sensors in a fission reactor environment. 520 FBGs were installed in the Ford reactor at the University of Michigan. The reactor was operated for 1012 effective full power hours resulting in a maximum neutron fluence of approximately 5x1019 n/cm2, and a maximum gamma dose of 2x103 MGy gamma. This work is significant in that, to the knowledge of the authors, the exposure levels obtained are approximately 1000 times higher than for any previously published experiment. Four different fiber compositions were evaluated. An 87% survival rate was observed for fiber Bragg gratings located at the fuel centerline. Optical Frequency Domain Reflectometry (OFDR), originally developed at the NASA Langley Research Center, can be used to interrogate several thousand low-reflectivity FBG strain and/or temperature sensors along a single optical fiber. A key advantage of the OFDR sensor technology for space nuclear power is the extremely low mass of the sensor, which consists of only a silica fiber 125μm in diameter. The sensors produced using this technology will fill applications in nuclear power for current reactor plants, emerging Generation-IV reactors, and for space nuclear power. The reported research was conducted by Luna Innovations and was funded through a Small Business Innovative Research (SBIR) contract with the NASA Glenn Research Center

  14. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core

    International Nuclear Information System (INIS)

    Lucatero, M.A.; Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C.

    2006-01-01

    With the purpose of defining the inspection frequency, in function of the embrittlement of the materials that compose the welding of the 'Core Shroud' or encircling of the core of a BWR type reactor, is necessary to know the neutron fluence received for this welding. In the work the calculated values of neutron fluence accumulated maxim (E > 1 MeV) during the first 8 operation cycles of the reactor are presented. The calculations were carried out according to the NRC Regulatory Guide 1.190, making use of the DORT code, which solves the transport equation in discreet ordinate in two dimensions (xy, rΘ, and rz). The results in 3D were obtained applying the Synthesis method according to the guide before mentioned. Results are presented for the horizontal welding H3, H4, and H5, showing the corresponding curves to the fluence accumulated to the cycle 8 and a projection for the cycle 14 is presented. (Author)

  15. Measured thermal and fast neutron fluence rates, ATR Cycle 102-A, 11/28/93 thru 1/16/94

    International Nuclear Information System (INIS)

    Murray, R.K.; Rogers, J.W.

    1994-02-01

    This report contains the thermal (2,200 m/s) and fast (E > 1MeV) neutron fluence rate data for ATR Cycle 102-A which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order. This report contains fluence rate values corresponding to the particular elevations (relative to the 80 ft. core elevation) where the measurements were taken. The data in this report consists of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, (3) plots of both the thermal and fast neutron fluence rates, and (4) a magnetic record (3.5 inch diskette) containing a listing of only the fast neutron fluence rates, their assigned elevations and proper header identification of all monitor positions contained herein. The fluence rates reported are for the average power levels given in the table of power history and distribution. All ''H'' holder monitoring wires for this cycle are 54 inches long. All ''SR'' holder monitor wires for this cycle are 55 inches long. This length allows measurement of the full core region and makes the first count elevation 24.73 inches above core midplane. Due to the safety rod problems in the west lobe, ''BR'' holders were used in the W-1, 2, 3, and 4 positions. All ''BR'' holder monitor wires for this cycle are 56.25 inches long. The distance from the end of the wires to the first count position was 4.25 inches for all wires counted from this cycle

  16. Measured thermal and fast neutron fluence rates ATR Cycle 99-A, November 23, 1992--January 23, 1993

    International Nuclear Information System (INIS)

    Murray, R.K.; Rogers, J.W.

    1993-03-01

    This report contains the thermal (2200 m/s) and fast (E>me) neutron fluence rate data for ATR Cycle 99-A which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power ReactorPrograms (ATR Experiments) Radiation Measurements Work Order. This report contains fluence rate values corresponding to the particular elevations (relative to the 80 ft. core elevation) where the measurements were taken. The data in this report consists of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, (3) plots of both the thermal and fast neutron fluence rates, and (4) a magnetic record (3.5 inch diskette) containing a listing of only the fast neutron fluence rates, their assigned elevations and proper header identification of all monitor positions contained herein. The fluence rates reported are for the average power levels given in the table of power history and distribution. All ''H'' holder monitor wires for this cycle are 54 inches long. All ''SR'' holder monitor wires for this cycle are 55 inches long. This length allows measurement of the full core region and makes the first count elevation 24.73 inches above core midplane. Due to the safety rod problems in the west lobe, ''BR'' holders were used in the W-1, 2, 3, and 4 positions. All ''BR'' holder monitor wires for this cycle are 56.25 inches long. The distance from the end of the wires to the first count position was 4.25 inches for all wires counted from this cycle

  17. Measured thermal and fast neutron fluence rates for ATF-1 holders during ATR cycle 158B/159A

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Larry Don [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, David Torbet [Idaho National Lab. (INL), Idaho Falls, ID (United States); Walker, Billy Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-11-01

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 158B/159A which were measured by the Radiation Measurements Laboratory (RML).

  18. Test of Fibre Bragg Gratings samples under High Fast Neutrons Fluence

    Science.gov (United States)

    Cheymol, G.; Remy, L.; Gusarov, A.; Kinet, D.; Mégret, P.; Laffont, G.; Blanchet, T.; Morana, A.; Marin, E.; Girard, S.

    2018-01-01

    Optical fibre sensors (OFS) are worthy of interest for measurements in nuclear reactor thanks to their unique features, particularly compact size and remote multi-point sensing for some of them. But besides non negligible constraints associated with the high temperature environment of the experiments of interest, it is well known that the performances of OFS can be severely affected by high level of radiations. The Radiation Induced Attenuation (RIA) in the fibre is probably most known effect, which can be to some extent circumvented by using rad hard fibres to limit the dynamic loss. However, when the fast neutron fluence reaches 1018 to 1019 n/cm2, the density and index variations associated to structural changes may deteriorate drastically the performances of OFS even if they are based on rad hard fibres, by causing direct errors in the measurements of temperature and/or strain changes. The aim of the present study is to access the effect of nuclear radiations on the Fabry Perot (FP) and of Fibre Bragg Grating (FBG) sensors through the comparison of measurements made on these OFS - or part of them - before and after irradiation [1]. In the context of development of OFS for high irradiation environment and especially for Material Testing Reactors (MTRs), Sake 2 experiment consists in an irradiation campaign at high level of gamma and neutron fluxes conducted on samples of fibre optics - bare or functionalised with FBG. The irradiation was performed at two levels of fast neutron fluence: 1 and 3.1019 n/cm2 (E>1MeV), at 250°± 25°C, in the SCK•CEN BR2 reactor (Mol Belgium). An irradiation capsule was designed to allow irradiation at the specified temperature without active control. The neutron fluence was measured with activation dosimeters and the results were compared with MCPN computations. Investigation of bare samples gives information on the density changes, while for the FBGs both density and refractive index perturbation are involved. Some results for

  19. Calculation of neutron fluence to dose equivalent conversion coefficients using GEANT4; Calculo de coeficientes de fluencia de neutrons para equivalente de dose individual utilizando o GEANT4

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Rosane M.; Santos, Denison de S.; Queiroz Filho, Pedro P. de; Mauricio, CLaudia L.P.; Silva, Livia K. da; Pessanha, Paula R., E-mail: rosanemribeiro@oi.com.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    Fluence to dose equivalent conversion coefficients provide the basis for the calculation of area and personal monitors. Recently, the ICRP has started a revision of these coefficients, including new Monte Carlo codes for benchmarking. So far, little information is available about neutron transport below 10 MeV in tissue-equivalent (TE) material performed with Monte Carlo GEANT4 code. The objective of this work is to calculate neutron fluence to personal dose equivalent conversion coefficients, H{sub p} (10)/Φ, with GEANT4 code. The incidence of monoenergetic neutrons was simulated as an expanded and aligned field, with energies ranging between thermal neutrons to 10 MeV on the ICRU slab of dimension 30 x 30 x 15 cm{sup 3}, composed of 76.2% of oxygen, 10.1% of hydrogen, 11.1% of carbon and 2.6% of nitrogen. For all incident energy, a cylindrical sensitive volume is placed at a depth of 10 mm, in the largest surface of the slab (30 x 30 cm{sup 2}). Physic process are included for neutrons, photons and charged particles, and calculations are made for neutrons and secondary particles which reach the sensitive volume. Results obtained are thus compared with values published in ICRP 74. Neutron fluence in the sensitive volume was calculated for benchmarking. The Monte Carlo GEANT4 code was found to be appropriate to calculate neutron doses at energies below 10 MeV correctly. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  1. International key comparison of neutron fluence measurements in mono-energetic neutron fields: C.C.R.I.(3)-K10

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.; Wang, Z.; Rong, C. [China Institute of Atomic Energy (CIAE), Beijing, People' s Republic of China (China); Lovestam, G.; Plompen, A.; Puglisi, N. [EC-JRC-Institute for Reference Materials and Measurements (IRMM), Geel (Belgium); Gilliam, D.M.; Eisenhauer, C.M.; Nico, J.S.; Dewey, M.S. [National Institute of Standards and Technology (NIST), Gaithersburg (United States); Kudo, K.; Uritani, A.; Harano, H.; Takeda, N. [National Metrology Institute of Japan (NMIJ), Tsukuba (Japan); Thomas, D.J.; Roberts, N.J.; Bennett, A.; Kolkowski, P. [National Physical Laboratory (NPL), Teddington (United Kingdom); Moisseev, N.N.; Kharitonov, I.A. [Mendeleyev Institute for Metrology (VNIIM), St Petersburg (Russian Federation); Guldbakke, S.; Klein, H.; Nolte, R.; Schlegel, D. [Physikalisch-Technische Bundesanstalt (PTB), Braunschweig (Germany)

    2007-12-15

    C.C.R.I. Section III (neutron measurements) conducted a unique key comparison of neutron fluence measurements in mono-energetic neutron fields. In contrast to former comparisons, here the fluence measurements were performed with the participants' instruments in the same neutron fields at the P.T.B. accelerator facility. Seven laboratories- the C.I.A.E. (China), I.R.M.M. (E.C.), N.M.I.J. (Japan), N.I.S.T. (USA), N.P.L. (UK), P.T.B. (Germany) and the V.N.I.I.M. (Russia)-employed their primary standard reference methods or transfer instruments carefully calibrated against their primary standards, to determine the fluence of 0.144 MeV, 1.2 MeV, 5.0 MeV and 14.8 MeV neutrons and reported calibration coefficients for a selected neutron monitor and each neutron energy with a detailed uncertainty budget for the measurements. The key comparison reference values (K.C.R.V.) were finally evaluated as the weighted mean values of the neutron fluence at 1 m distance from the target in vacuum per neutron monitor count. The uncertainties of each K.C.R.V. amounted to about 1%. The degree of equivalence (D.o.E.), defined as the deviation of the result reported by the laboratories for each energy from the corresponding K.C.R.V., and the associated expanded uncertainty are also reported. The deviations between the results of two laboratories each with the corresponding expanded uncertainties complete the documentation of the degrees of equivalence. (authors)

  2. Absolute measurement of thermal neutron fluence and its application for fission track dating

    International Nuclear Information System (INIS)

    Ganzawa, Yoshihiro; Honda, Teruyuki; Nozaki, Tetsuya.

    1988-01-01

    The absolute measurements of thermal neutron fluence for fission track dating have been developed after the proceeding results of Honda et al. (1987). The 2,200 m/sec activation cross section of 197 Au (98.8 barn) is corrected to 87.4 barn (σa) by the three factors of the neutron temperature, Maxwellian distribution of thermal neutrons and non 1/v correction factor for the above absolute measurement. The calibrated factor (B th ) of standard glasses (SRM613, SRM962a, CN-1 and CN-2) and zeta-a (ζa) values for fission track dating are determined on the basis of these experimental results. The values of B th , (7.47 ± 0.29) x 10 9 for SRM613, (7.43 ± 0.34) x 10 9 for SRM962a, (2.50 ± 0.06) x 10 9 for CN-1 and (2.74 ± 0.06) x 10 9 for CN-2 closely agree with those reported previously by Honda et al. (1987). Further, the ζa values of 392.3 ± 16.5 for SRM962a and SRM613, 131.4 ± 3.1 for CN-1 and 144.1 ± 3.3 for CN-2 calculated from B th , effective thermal neutron fission cross-section σf (497.4 barn), isotopic abundance ratio 235 U/ 239 U, I (7.2527 x 10 -3 ) and spontaneous fission decay constant of 238 U, λ f (6.85 x 10 -17 a -7 ) show close agreement with ζ b values (392.5 ± 10.0, 131.6 ± 3.3, 140.1 ± 3.5) derived from the absolute age of Fish Canyon Tuff (27.9 ± 0.7 Ma) respectively. The fission track dating of zircons separated from Oligocene-Miocene tuff distributed in Eastern Hokkaido have been carried out by the external detector method using ζ a . The obtained ages are 28.6 ± 0.7 Ma (1 - 2) and 23.3 ± 0.7 Ma (3 - 2). These results agree well with the geologic age supported from Ashoro Fossil Fauna, K-Ar ages of volcanic rocks and stratigraphy in this area. (author)

  3. Assessment of the effects of neutron fluence on Swedish nuclear pressure vessels

    International Nuclear Information System (INIS)

    Rao, S.

    1980-11-01

    Nuclear pressure vessels are subject to neutron irradiation during service causing embrittlement. This is one important factor in the overall problem of reactor vessel integrity. At present the irradiation effects are mainly assessed by the Charpy V-notch test. Two measures of embrittlement are defined: the increase of the ductile/brittle transition temperature and the decrease in the upper-shelf energy. The object of the present work is to assess these changes for the Swedish nuclear pressure vessels. On the basis of data from irradiations carried out in other countries and Swedish surveillance programmes, the expected end of life embrittlement is estimated for Swedish vessels. The results show that the embrittlement of most reactor vessels is expected to be quite small. Oskarshamn 1 and PWR-vessels, however, will probably show moderate changes, the former due to the higher copper content, and the latter due to the high end of life fluences. Some of the vessel materials which exhibit marginal properties in the upper-shelf energy, as measured by the Charpy V-notch impact test, are identified. It is recommended that fracture mechanics analyses be applied in these cases. (author)

  4. Calculation of neutron fluence in the region of the pressure vessel for the history of different reactors by using the Monte-Carlo-method

    International Nuclear Information System (INIS)

    Barz, H.U.; Bertram, W.

    1992-01-01

    Embrittlement of pressure vessel material caused by neutron irradiation is a very important problem for VVER-440 reactors. For the estimation of the fracture risk highly reliable neutron fluence values are necessary. For this reason a special theoretical determination of space dependent neutron fluences has been performed mainly on the basis of Monte-Carlo calculations. The described method allows the accurate calculation of neutron fluences near the pressure vessel in the height of the core region for all reactor histories and loading cycles in an efficient manner. To illustrate the accuracy of the suggested method a comparison with experimental results was done. The calculated neutron fluence values can be used for planning the loading schemes of each reactor according to the safety requirements against brittle fracture. (orig.)

  5. DOUBLE-EXPONENTIAL FITTING FUNCTION FOR EVALUATION OF COSMIC-RAY-INDUCED NEUTRON FLUENCE RATE IN ARBITRARY LOCATIONS.

    Science.gov (United States)

    Li, Huailiang; Yang, Yigang; Wang, Qibiao; Tuo, Xianguo; Julian Henderson, Mark; Courtois, Jérémie

    2017-12-01

    The fluence rate of cosmic-ray-induced neutrons (CRINs) varies with many environmental factors. While many current simulation and experimental studies have focused mainly on the altitude variation, the specific rule that the CRINs vary with geomagnetic cutoff rigidity (which is related to latitude and longitude) was not well considered. In this article, a double-exponential fitting function F=(A1e-A2CR+A3)eB1Al, is proposed to evaluate the CRINs' fluence rate varying with geomagnetic cutoff rigidity and altitude. The fitting R2 can have a value up to 0.9954, and, moreover, the CRINs' fluence rate in an arbitrary location (latitude, longitude and altitude) can be easily evaluated by the proposed function. The field measurements of the CRINs' fluence rate and H*(10) rate in Mt. Emei and Mt. Bowa were carried out using a FHT-762 and LB 6411 neutron prober, respectively, and the evaluation results show that the fitting function agrees well with the measurement results. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  6. Standard Test Method for Measuring Neutron Fluence and Average Energy from 3H(d,n)4He Neutron Generators by Radioactivation Techniques 1

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This test method covers a general procedure for the measurement of the fast-neutron fluence rate produced by neutron generators utilizing the 3H(d,n)4He reaction. Neutrons so produced are usually referred to as 14-MeV neutrons, but range in energy depending on a number of factors. This test method does not adequately cover fusion sources where the velocity of the plasma may be an important consideration. 1.2 This test method uses threshold activation reactions to determine the average energy of the neutrons and the neutron fluence at that energy. At least three activities, chosen from an appropriate set of dosimetry reactions, are required to characterize the average energy and fluence. The required activities are typically measured by gamma ray spectroscopy. 1.3 The measurement of reaction products in their metastable states is not covered. If the metastable state decays to the ground state, the ground state reaction may be used. 1.4 The values stated in SI units are to be regarded as standard. No oth...

  7. Irradiation embrittlement of some 15Kh2MFA pressure vessel steels under varying neutron fluence rates

    Energy Technology Data Exchange (ETDEWEB)

    Valo, M; Bars, B [Technical Research Centre of Finland, Espoo (Finland); Ahlstrand, A [Imatran Voima Oy (IVO), Helsinki (Finland)

    1994-12-31

    Irradiation sensitivity of two forging materials was measured with Charpy-V and fracture mechanic tests, and with different fluence, fluence rate and irradiation time values. Irradiation sensitivity of the materials was found to be less or equal to the current Russian standard, and appears to be well described by the fluence parameter only. A slight additional effect on embrittlement from a long term low fluence irradiation is noticed, but it stays within the total scatter band of data. 7 refs., 17 figs., 4 tabs.

  8. Approaches for accounting and prediction of fast neutron fluence on WWER pressure vessels and results of validation of calculational procedure

    International Nuclear Information System (INIS)

    Borodkin, P.G.; Khrennikov, N.N.; Ryabinin, Yu.A.; Adeev, V.A.

    2015-01-01

    A description is given of the universal procedure for calculation of fast neutron fluence (FNF) on WWER vessels. Approbation of the calculation procedure was carried out by comparing the calculation results for this procedure and measurements on the outer surface of the WWER-440 and WWER-1000 vessels. In addition, an estimation of the uncertainty of the settlement procedure was made in accordance with the requirements of regulatory documents. The developed procedure is applied at Kola NPP for independent fast neutron fluence estimates on the WWER-440 reactor vessels when planning core loads taking into account the introduction of new fuels. The results of the pilot operation of the procedure for calculating FNF at the Kola NPP were taken into account when improving the procedure and its application to the calculations of FNF on the WWER-1000 vessels [ru

  9. BPW34 Commercial p-i-n Diodes for High-Level 1-MeV Neutron Equivalent Fluence Monitoring

    CERN Document Server

    Ravotti, F; Moll, M; Saigne, F

    2008-01-01

    The BPW34 p-i-n diode was characterized at CERN in view of its utilization as radiation monitor at the LHC to cover the broad 1-MeV neutron equivalent fluence (Phieq) range expected for the LHC machine and experiments during operation. Electrical measurements for both forward and reverse bias were used to characterize the device and to understand its behavior under irradiation. When the device is powered forward, a sensitivity to fast hadrons for Phieq > 2 times1012 cm-2 has been observed. With increasing particle fluences the forward I- V characteristics of the diode shifts towards higher voltages. At Phieq > 3times1013 cm-2, the forward characteristic starts to bend back assuming a thyristor-like behavior. An explanation for this phenomenon is given in this article. Finally, detailed radiation-response curves for the forward bias-operation and annealing studies of the diode's forward voltage are presented for proton, neutron and gamma irradiation.

  10. Neutron Fluence Evaluation of Reactor Internal Structure Using 3D Transport Calculation Code, RAPTOR-M3G

    International Nuclear Information System (INIS)

    Maeng, YoungJae; Lim, MiJoung; Kim, KyungSik; Cho, YoungKi; Yoo, ChoonSung; Kim, ByoungChul

    2015-01-01

    Age-related degradation mechanisms are including the irradiation-assisted stress corrosion cracking(IASCC), void swelling, stress relaxation, fatigue, and etc. A lot of Baffle Former Bolts(BFBs) was installed at the former plate ends between baffle and barrel structure. These would undergo severe experiences, which are high temperature and pressure, bypass water flow and neutron exposure and have some radioactive limitation in inspecting their integrity. The objectives of this paper is to evaluate fast neutron fluence(n/cm 2 , E>1.0MeV) for PWR internals using 3D transport calculation code, RAPTOR-M3G, and to figure out a strategy to manage the effects of aging in PWR internals. One of age-related degradation mechanisms, IASCC, which is affected by fast neutron exposure rate, has been currently issued for PWR internals and has 2 x 10 21 (n/cm 2 ) of the threshold value by MRP-175. Because a lot of BFBs was installed around the internal components, closer inspections are required. As part of an aging management for Kori unit 2, 3D transport calculation code, RAPTOR-M3G, was applied for determining fast neutron fluence at baffle, barrel and former plates regions. As a result, the fast neutron fluence exceeds the screening or threshold values of IASCC in all of baffle, barrel and former plate region. And the most significant region is the baffle because it is located closest to the core. In addition, some regions including former plate tend to be more damaged because of less moderate ability than water. In conclusion, Ice's has been progressed for PWR internals of Kori unit 2. Several regions of internal components were damaged by fast neutron exposure and increase in size as time goes by

  11. Rapid method of calculating the fluence and spectrum of neutrons from a critical assembly and the resulting dose

    International Nuclear Information System (INIS)

    Bessis, J.

    1977-01-01

    The proposed calculation method is unsophisticated but rapid. The first part (computer code CRITIC), which is based on the Fermi age equation, evaluates the number of neutrons per fission emitted from a moderated critical assembly and their energy spectrum. The second part (computer code NARCISSE), which uses the current albedo for concrete, evaluates the product of neutron reflection on the walls and calculates the fluence resulting at any point in the room and its energy distribution by 21 groups. The results obtained are shown to compare satisfactorily with those obtained through the use of a Monte Carlo program

  12. Measured thermal and fast neutron fluence rates, ATR Cycle 100-BC, April 23, 1993--May 13, 1993

    International Nuclear Information System (INIS)

    Smith, L.D.; Murray, R.K.; Rogers, J.W.

    1993-07-01

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 100-BC which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order. This report contains fluence rate values corresponding to the particular elevations (relative to the 80 ft. core elevation) where the measurements were taken. The data in this report consists of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, (3) plots of both the thermal and fast neutron fluence rates, and (4) a magnetic record (3.5 inch diskette) containing a listing of only the fast neutron fluence rates, their assigned elevations and proper header identification of all monitor positions contained herein. The fluence rates reported are for the average power levels given in the table of power history and distribution. All open-quotes Hclose quotes holder monitor wires for this cycle are 54 inches long. All open-quotes SRclose quotes holder monitor wires for this cycle are 55 inches long. This length allows measurement of the full core region and makes the first count elevation 24.73 inches above core midplane. Due to the safety rod problems in the west lobe, open-quotes BRclose quotes holders were used in the W-1, 2, 3, and 4 positions. All open-quotes BRclose quotes holder monitor wires for this cycle are 56.25 inches long. The distance from the end of the wires to the first count position was 4.25 inches for all wires counted from this cycle. The results from the measurements in the W-1, 2, 3, 4 monitor positions indicate that the safety rod followers were rotated to a different azimuthal orientation relative to the normal orientation. The results indicate that the rotation was counterclockwise from their normal orientation. This is the same condition observed starting with Cycle 99-B

  13. Three-dimensional Monte Carlo calculations of the neutron and γ-ray fluences in the TFTR diagnostic basement and comparisons with measurements

    International Nuclear Information System (INIS)

    Liew, S.L.; Ku, L.P.; Kolibal, J.G.

    1985-10-01

    Realistic calculations of the neutron and γ-ray fluences in the TFTR diagnostic basement have been carried out with three-dimensional Monte Carlo models. Comparisons with measurements show that the results are well within the experimental uncertainties

  14. Determination of neutron fluence and radiation brittleness temperature of WWER-440 and WWER-1000 pressure vessels in Kozloduy NPP

    International Nuclear Information System (INIS)

    Ilieva, K.; Apostolov, T.; Belousov, S.; Petrova, T.; Antonov, S.; Ivanov, K.; Prodanova, R.

    1993-01-01

    In Units 1-4 of Kozloduy NPP (WWER-440/230), the weld 4 of RPV undergoes the most severe irradiation embrittlement. Neither witness-samples, nor detectors are designed for these reactors. Transport calculations of fast neutron fluence on WWER-440 RPV and ex-vessel measurements by threshold activation detectors are the primary means for adequate assessment of metal state and for prognosis concerning the reactor life span. In WWER-1000 reactors (Units 5-6) the maximum neutron fluence occurs on the weld 3. The systematical observation of metal state is performed through witness-samples and threshold activation detectors ( 54 Fe (n,p), 63 Cu (n,α), 93 Nb (n,n')) placed above the reactor top edge and at the first vessel ring level. There are big differences in energy spectrum and integral neutron flux falling onto the weld 3, the RPV base metal and the staff detectors. This requires additional neutron measurements in the air gap between the RPV and the thermal insulation. (author)

  15. Determination of the Neutron Fluence, the Beam Characteristics and the Backgrounds at the CERN-PS TOF Facility

    CERN Multimedia

    Leal, L C; Kitis, G; Guber, K H; Quaranta, A; Koehler, P E

    2002-01-01

    In the scope of our programme we propose to start in July 2000 with measurements on elements of well known cross sections, in order to check the reliability of the whole experimental installation at the CERN-TOF facility. These initial exploratory measurements will provide the key-parameters required for the further experimentation at the CERN-TOF neutron beam. The neutron fluence and energy resolution will be determined as a function of the neutron kinetic energy by reproducing standard capture and fission cross sections. The measurements of capture cross sections on elements with specific cross section features will allow to us to disentangle the different components of backgrounds and estimate their level in the experimental area. The time-energy calibration will be determined and monitored with a set of monoenergetic filters as well as by the measurements of elements with resonance-dominated cross sections. Finally, in this initial phase the behaviour of several detectors scheduled in successive measureme...

  16. Spectral representations of neutron-star equations of state

    International Nuclear Information System (INIS)

    Lindblom, Lee

    2010-01-01

    Methods are developed for constructing spectral representations of cold (barotropic) neutron-star equations of state. These representations are faithful in the sense that every physical equation of state has a representation of this type and conversely every such representation satisfies the minimal thermodynamic stability criteria required of any physical equation of state. These spectral representations are also efficient, in the sense that only a few spectral coefficients are generally required to represent neutron-star equations of state quiet accurately. This accuracy and efficiency is illustrated by constructing spectral fits to a large collection of 'realistic' neutron-star equations of state.

  17. Spectral distortion due to scattered cold neutrons in beryllium filter

    International Nuclear Information System (INIS)

    Sakamoto, Yukio; Inoue, Kazuhiko

    1980-01-01

    Polycrystalline beryllium filters are used to discriminate the cold neutrons from the thermal neutrons with energies above Bragg cut-off energy. The cold neutron scattering cross section is very small, but the remaining cross section is not zero. Then the neutrons scattered once from the filter in the cold neutron energy region have chance of impinging on the outlet of filter. Those neutrons are almost upscattered and develop into thermal neutrons; thus the discriminated cold neutrons include a small spectral distortion due to the thermal neutrons. In the present work we have evaluated the effect on the cold neutron spectrum due to the repeatedly scattered and transmitted neutrons by using a Monte Carlo calculation method. (author)

  18. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR

    International Nuclear Information System (INIS)

    Martinez C, E.

    2011-01-01

    One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-θ and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-θ, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, θ and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm 2 s, at a height H 4 (239.07 cm) and angle 32.236 o in the core shroud and 4.00 E + 09 n/cm 2 s at a height H 4 and angle 35.27 o in the inner wall of the reactor vessel, positions that are consistent to within ±10% over the ones reported in the literature. (Author)

  19. Radiation clusters formation and evolution in FCC metals at low-temperature neutron irradiation up to small damage fluences

    International Nuclear Information System (INIS)

    Kozlov, A.V.; Shcherbakov, E.N.; Asiptsov, O.I.; Skryabin, L.A.; Portnykh, I.A.

    2006-01-01

    Methods of transmission electron microscopy and precision size measurements are used to study the formation of radiation-induced clusters in FCC metals (Ni, Pt, austenitic steels EhI-844, ChS-68) irradiated with fast neutron (E>0.1 MeV) fluences from 7 x 10 21 up to 3.5 x 10 22 m -2 at a temperature of 310 K. Using statistical thermodynamic methods the process of radiation clusters formation and evolution is described quantitatively. The change in the concentration of point defects under irradiation as well as size variations of irradiated specimens on annealing are calculated [ru

  20. A novel wide range, real-time neutron fluence monitor based on commercial off the shelf gallium arsenide light emitting diodes

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, B., E-mail: bhaskar.mukherjee@uk-essen.de [Westdeutsches Protonentherapiezentrum Essen (WPE) gGmbH, Hufelandstrasse 55, D-45147 Essen (Germany); Hentschel, R. [Strahlenklinik, University Hospital Essen (Germany); Lambert, J. [Westdeutsches Protonentherapiezentrum Essen (WPE) gGmbH, Hufelandstrasse 55, D-45147 Essen (Germany); Deya, W. [Strahlenklinik, University Hospital Essen (Germany); Farr, J. [Westdeutsches Protonentherapiezentrum Essen (WPE) gGmbH, Hufelandstrasse 55, D-45147 Essen (Germany)

    2011-10-01

    Displacement damage produced by high-energy neutrons in gallium arsenide (GaAs) light emitting diodes (LED) results in the reduction of light output. Based on this principle we have developed a simple, cost effective, neutron detector using commercial off the shelf (COTS) GaAs-LED for the assessment of neutron fluence and KERMA at critical locations in the vicinity of the 230 MeV proton therapy cyclotron operated by Westdeutsches Protonentherapiezentrum Essen (WPE). The LED detector response (mV) was found to be linear within the neutron fluence range of 3.0x10{sup 8}-1.0x10{sup 11} neutron cm{sup -2}. The response of the LED detector was proportional to neutron induced displacement damage in LED; hence, by using the differential KERMA coefficient of neutrons in GaAs, we have rescaled the calibration curve for two mono-energetic sources, i.e. 1 MeV neutrons and 14 MeV neutrons generated by D+T fusion reaction. In this paper we present the principle of the real-time GaAs-LED based neutron fluence monitor as mentioned above. The device was calibrated using fast neutrons produced by bombarding a thick beryllium target with 14 MeV deuterons from a TCC CV 28 medical cyclotron of the Strahlenklinik University Hospital Essen.

  1. Specific heat of Nb3Sn and V2Zr compounds irradiated with high fluences fast neutrons

    International Nuclear Information System (INIS)

    Kar'kin, A.E.; Mirmel'shtejn, A.V.; Arkhipov, V.E.; Goshchitskij, B.N.

    1987-01-01

    Specific heat of Nb 3 Sn (structure A15) and V 2 Zr (C15) specimens irradiated with high fluences of bast neutrons has been measured. It is shown that in these compounds the temperature reduction of superconducting transition T c under neutron irradiation is accompanied with high decrease of N(E F ). Phonon spectrum of the irradiated V 2 Zr (amorphous phase) on the whole is harder, than at an initial state, for irradiated Nb 3 Sn state (disordered crystalline structure) phonon spectrum is differ weakly from initial one. General regularities of parameter change of electron and phonon subsystems for A15 compounds investigated here and earlier (V 3 Si, Mo 3 Si, Mo 3 Ge) have been analysed

  2. NGI-9 pulsed neutron generator with a fluence to 1010 n/s

    International Nuclear Information System (INIS)

    Allakhverdov, A.Sh.; Ogarkin, V.I.; Silicheva, G.P.; Timofeev, Yu.I.

    1975-01-01

    A neutron pulse generator with 14 MeV energy used for the activation analysis, is described. Its functional diagram and the technical characteristics are presented. The studies of the generator that resulted in determination of the effect of the accelerating voltage amplitude, the delay in the ion source firing with respect to the pulse of the accelerating voltage, the amount of operating ion sources and the energy imparted to them on the neutron flux magnitude are conducted. It is confirmed by the studies that the neutron generator operating in the nominal regime makes it possible to obtain a neutron flux of 5x10 9 -10 10 neutr./s. The dependence of the neutron flux variation on the frequency of pulse sequence for various ion sources is shown

  3. Analysis of influence of fast neutron fluence irradiated to Beryllium element of The RSG-GAS reactor

    International Nuclear Information System (INIS)

    Sri Kuntjoro

    2010-01-01

    Analysis of influence fast neutron fluence irradiated to the RSG-GAS beryllium reflector have been done. Methods of analysis was carried out by measuring fluxes neutron in beryllium element and block position that function as reflector.The calculation done for determination it is there any influence of neutron as long as beryllium in the core. Besides that, visualization done to make sure it there is any deformation at beryllium as effect of irradiation. Fluxes and fluences of beryllium element measurement result in 200 kW reactor power are 2.30E+07 n/cm 2 .sec and 4.19E+17 n/cm 2 in position E-2, 3.70E+07 n/cm 2 s and 6.74E+17 n/cm 2 in position J-8, 2.19E+12 n/cm 2 s and 3.99E+22 n/cm 2 in position. Measurement results in the position B-3 are 2.12E+12 n/cm 2 s and 3.86E+22 n/cm 2 in position G-10 respectively. Other result are fluxes and fluence in beryllium block, those are 5,02E+07 n/cm 2 s and 9,15E+17 n/cm 2 in position (5-6), and 2,32E+07 n/cm 2 s and 4,23E+17 n/cm 2 in position (C-D). Deformation (L/L) results for beryllium element are 1,12E-08 in position E-2, 1,84E-08 in position J-8, 1,60E-03 in position B-3, and 1,55E-03 in position G-10. In beryllium block deformation results are 2,52E-08 in position (5-6) and 1,13E-08 in position (C-D). Those results are shown unseen deformation in beryllium element and beryllium block and demonstrably by visual observation in reactor hot cell. (author)

  4. Swelling in several commercial alloys irradiated to very high neutron fluence

    International Nuclear Information System (INIS)

    Gelles, D.S.; Pintler, J.S.

    1984-01-01

    Swelling values have been obtained from a set of commercial alloys irradiated in EBR-II to a peak fluence of 2.5 x 10 23 n/cm 2 (E > 0.1 MeV) or approx. 125 dpa covering the range 400 to 650 0 C. The alloys can be ranked for swelling resistance from highest to lowest as follows: the martensitic and ferritic alloys, the niobium based alloys, the precipitation strengthened iron and nickel based alloys, the molybdenum alloys and the austenitic alloys

  5. Neutron fluence in a 18 MeV Electron Accelerator for Therapy; Fluencia de neutrones en un Acelerador de Electrones de 18 MeV para terapia

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L C [Instituto Nacional de Investigaciones Nucleares, Direccion de Innovacion Tecnologica, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    An investigation was made on the theoretical fundamentals for the determination of the neutron fluence in a linear electron accelerator for radiotherapy applications and the limit values of leakage neutron radiation established by guidelines and standards in radiation protection for these type of accelerators. This investigation includes the following parts: a) Exhaustive bibliographical review on the topics mentioned above, in order to combine and to update the necessary basic information to facilitate the understanding of this subject; b) Analysis of the accelerator operation and identification of its main components, specially in the accelerator head; c) Study of different types of targets and its materials for the Bremsstrahlung production which is based on the electron initial energy, the thickness of the target, and its angular distribution and energy, which influences in the neutron generation by means of the photonuclear and electro disintegration reactions; d) Analysis of the neutron yield based on the target type and its thickness, the energy of electrons and photons; e) Analysis of the neutron energy spectra generated in the accelerator head, inside and outside the treatment room; f) Study of the dosimetry fundamentals for neutron and photon mixed fields, the dosimeter selection criteria and standards applied for these applications, specially the Panasonic U D-809 thermoluminescent dosemeter and C R-39 nuclear track dosimeter; g) Theoretical calculation of the neutron yield using a simplified geometric model for the accelerator head with spherical cell, which considers the target, primary collimator, flattener filter, movable collimators and the head shielding as the main components for radiation production. The cases with W and Pb shielding for closed movable collimators and an irradiation field of 20 x 20 cm{sup 2} were analyzed and, h) Experimental evaluation of the leakage neutron radiation from the patient and head planes, observing that the

  6. Development of a High Fluence Neutron Source for Nondestructive Characterization of Nuclear Waste

    International Nuclear Information System (INIS)

    Pickrell, Mark M.

    1999-01-01

    We are addressing the need to measure nuclear wastes, residues, and spent fuel in order to process these for final disposition. For example, TRU wastes destined for the WIPP must satisfy extensive characterization criteria outlined in the Waste Acceptance Criteria, the Quality Assurance Program Plan, and the Performance Demonstration Plan. Similar requirements exist for spent fuel and residues. At present, no nondestructive assay (NDA) instrumentation is capable of satisfying all of the PDP test cycles (particularly for Remote-Handled TRU waste). One of the primary methods for waste assay is by active neutron interrogation. The objective of this project is to improve the capability of all active neutron systems by providing a higher intensity neutron source (by about a factor of 1,000) for essentially the same cost, power, and space requirements as existing systems. This high intensity neutron source is an electrostatically confined (IEC) plasma device. The IEC is a symmetric sphere that was originally developed in the 1960s as a possible fusion reactor. It operates as DT neutron generator. Although it is not likely that this device will scale to fusion reactor levels, previous experiments1 have demonstrated a neutron yield of 2 x 1010 neutrons/second on a table-top device that can be powered from ordinary laboratory circuits (9 kilowatts). Subsequently, the IEC physics has been extensively studied at the University of Illinois and other locations. We have established theoretically the basis for scaling the output up to 1 x 1011 neutrons/second. In addition, IEC devices have run for cumulative times approaching 10,000 hours, which is essential for practical application to NDA. They have been operated in pulsed and continuous mode. The essential features of the IEC plasma neutron source, compared to existing sources of the same cost, size and power consumption, are: Table 1: Present and Target Operating Parameters for Small Neutron Generators Parameter Present IEC

  7. Development and application of a detector for absolute measurement of neutron fluence rate in MeV region

    International Nuclear Information System (INIS)

    Silva Dias, M. da.

    1988-01-01

    The development and performance of the DTS (Dual Thin Scintillator) for the absolute measurement of the neutron fluence rate between 1 and 15 MeV is decribed. The DTS detector consists of a pair of organic scintillators in a dual configuration, where the incident produces a proton-recoil which is detected in a 2Π geometry therefore avoiding the effect of the escape of protons. Thin scintillators are used resulting in small multiple scattering corrections. The theoretical caluclations of detector efficiency and proton-recoil spectrum were performed by means of a Monte Carlos code - CARLO DTS. The calculated efficiency was compared to the experimental one at two neutron energies namely 2.446 MeV and 14.04 MeV applying the Time Correlated Associated Particle technique. The theoretical and experimental efficiencies agreed within the experimental uncertainties of 1.44% and 0.77%, respectively. The performance of the DTS has been verified in an absolute 235 U(n,f) cross section measurement between 1 and 6 MeV neutron energy. The cross section results were compared to those obtained replacing the DTS detector by the NBS (National Bureau of Standards, USA) Black Neutron Detector. The agreement was excellent in the overlapping energy interval of the two experiments (between 1 and 3 MeV), within the estimated uncertainly in the range of 1,0 to 1,7%. The agreement with the most recent evaluation from the ENDF/B-VI was excellent in almost all the energy range between 1 and 6 MeV. The 235 U(n,f) cross section, average over the 252 Cf fission neutron spectrum has been evaluated. The result including the cross section values of the present work was 1220 mb, in excellent agreement with the average value among the most recent measurements, 1227 +- 12 mb, and with the value 1213 mb, using the ENDF/B-VI data. (author) [pt

  8. Radiation annealing mechanisms of low-alloy reactor pressure vessel steels dependent on irradiation temperature and neutron fluence

    International Nuclear Information System (INIS)

    Pachur, D.

    1982-01-01

    Heat treatment after irradiation of reactor pressure vessel steels showed annealing of irradiation embrittlement. Depending on the irradiation temperature, the embrittlement started to anneal at about 220 0 C and was completely annealed at 500 0 C with 4 h of annealing time. The annealing behavior was normally measured in terms of the Vickers hardness increase produced by irradiation relative to the initial hardness as a function of the annealing temperature. Annealing results of other mechanical properties correspond to hardness results. During annealing, various recovery mechanisms occur in different temperature ranges. These are characterized by activation energies from 1.5 to 2.1 eV. The individual mechanisms were determined by the different time dependencies at various temperatures. The relative contributions of the mechanisms showed a neutron fluence dependence, with the lower activation energy mechanisms being predominant at low fluence and vice versa. In the temperature range where partial annealing of a mechanism took place during irradiation, an increase in activation energy was observed. Trend curves for the increase in transition temperature with irradiation, for the relative increase of Vickers hardness and yield strength, and for the relative decrease of Charpy-V upper shelf energy are interpreted by the behavior of different mechanisms

  9. Spectral correction factors for conventional neutron dosemeters used in high-energy neutron environments

    International Nuclear Information System (INIS)

    Lee, K.W.; Sheu, R.J.

    2015-01-01

    High-energy neutrons (>10 MeV) contribute substantially to the dose fraction but result in only a small or negligible response in most conventional moderated-type neutron detectors. Neutron dosemeters used for radiation protection purpose are commonly calibrated with 252 Cf neutron sources and are used in various workplace. A workplace-specific correction factor is suggested. In this study, the effect of the neutron spectrum on the accuracy of dose measurements was investigated. A set of neutron spectra representing various neutron environments was selected to study the dose responses of a series of Bonner spheres, including standard and extended-range spheres. By comparing 252 Cf-calibrated dose responses with reference values based on fluence-to-dose conversion coefficients, this paper presents recommendations for neutron field characterisation and appropriate correction factors for responses of conventional neutron dosemeters used in environments with high-energy neutrons. The correction depends on the estimated percentage of high-energy neutrons in the spectrum or the ratio between the measured responses of two Bonner spheres (the 4P6-8 extended-range sphere versus the 6'' standard sphere). (authors)

  10. Rhodium self-powered detector for monitoring neutron fluence, energy production, and isotopic composition of fuel

    International Nuclear Information System (INIS)

    Sokolov, A.P.; Pochivalin, G.P.; Shipovskikh, Yu.M.; Garusov, Yu.V.; Chernikov, O.G.; Shevchenko, V.G.

    1993-01-01

    The use of self-powered detectors (SPDs) with a rhodium emitter customarily involves monitoring of neutron fields in the core of a nuclear reactor. Since current in an SPD is generated primarily because of the neutron flux, which is responsible for the dynamics of particular nuclear transformations, including fission reactions of heavy isotopes, the detector signal can be attributed unambiguously to energy release at the location of the detector. Computation modeling performed with the KOMDPS package of programs of the current formation in a rhodium SPD along with the neutron-physical processes that occur in the reactor core makes it possible to take account of the effect of the principal factors characterizing the operating conditions and the design features of the fuel channel and the detector, reveal quantitative relations between the generated signal and individual physical parameters, and determine the metrological parameters of the detector. The formation and transport of changed particles in the sensitive part of the SPC is calculated by the Monte Carlo method. The emitter activation, neutron transport, and dynamics of the isotopic composition in the fuel channel containing the SPD are determined by solving the kinetic equation in the multigroup representation of the neutron spectrum, using the discrete ordinate method. In this work the authors consider the operation of a rhodium SPD in a bundle of 49 fuel channels of the RBMK-1000 reactor with a fuel enrichment of 2.4% from the time it is inserted into a fresh channel

  11. Neutronics equations: Positiveness; compactness; spectral theory; time asymptotic behavior

    International Nuclear Information System (INIS)

    Mokhtar-Kharroubi, M.

    1987-12-01

    Neutronics equations are studied: the continuous model (with and without delayed neutrons) and the multigroup model. Asymptotic descriptions of these equations (t→+∞) are obtained, either by the Dunford method or by using semigroup perturbation techniques, after deriving the spectral theory for the equations. Compactness problems are reviewed, and a general theory of compact injection in neutronic functional space is derived. The effects of positiveness in neutronics are analyzed: the irreducibility of the transport semigroup, and the properties of the main eigenvalue (existence, nonexistence, frame, strict dominance, strict monotony in relation to all the parameters). A class of transport operators whose real spectrum can be completely described is shown [fr

  12. COLLI-PTB, Neutron Fluence Spectra for 3-D Collimator System by Monte-Carlo

    International Nuclear Information System (INIS)

    Schlegel-Bickmann, Dietrich

    1995-01-01

    1 - Description of program or function: For optimizing collimator systems (shieldings) for fast neutrons with energies between 10 KeV and 20 MeV. Only elastic and inelastic neutron scattering processes are involved. Isotropic angular distribution for inelastic scattering in the center of mass system is assumed. 2 - Method of solution: The Monte Carlo method with importance sampling technique, splitting and Russian Roulette is used. The neutron attenuation and scattering kinematics is taken into account. 3 - Restrictions on the complexity of the problem: Energy range from 10 KeV to 20 MeV. For the output spectra any bin width is possible. The output spectra are confined to 40 equidistant channels

  13. Measurement of low neutron fluences with polycarbonate foils electrochemically etched with methyl alcohol-KOH solution

    International Nuclear Information System (INIS)

    Kumamoto, Y.

    1982-01-01

    Electrochemical etching of polycarbonate foils was performed using a KOH solution with a high concentration of methyl alcohol under different conditions of field strength, frequency, temperature and etching time. These studies showed that the highest neutron sensitivity relative to the inherent background in the foil was obtained under the following etching conditions: 52 kV/cm, 1 kHz, 30 0 C, 30 min in a solution of 45 gm KOH + 80 cc CH 3 OH + 20 cc H 2 O. Under these conditions, 100 mrem of neutrons from a Ra-Be source gave 70 pits per cm 2 while background was 7 +- 3 pits per cm 2 (10 +- 5 mrem). The pit diameters were about 90 μm. This sensitivity (giving lowest measurable dose of 15 mrem) is quite sufficient for personnel neutron dosimetry applications and the size of the pits is large enough for easy counting using a microscope at magnification of 40X. (author)

  14. Neutron Spectra, Fluence and Dose Rates from Bare and Moderated Cf-252 Sources

    Energy Technology Data Exchange (ETDEWEB)

    Radev, Radoslav P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-04-01

    A new, stronger 252Cf source (serial number SR-CF-3050-OR) was obtained from Oak Ridge National Laboratory (ORNL) in 2014 to supplement the existing 252Cf sources which had significantly decayed. A new instrument positioning track system was designed and installed by Hopewell Designs, Inc. in 2011. The neutron field from the new, stronger 252Cf source in the modified calibration environment needed to be characterized as well as the modified neutron fields produced by the new source and seven different neutron moderators. Comprehensive information about our 252Cf source, its origin, production, and isotopic content and decay characteristics needed to be compiled as well. This technical report is intended to address these issues.

  15. 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

  16. Research Reactor Application for Materials under High Neutron Fluence. Proceedings of an IAEA Technical Meeting (TM-34779)

    International Nuclear Information System (INIS)

    2011-05-01

    Research reactors (RRs) have played, and continue to play, a key role in the development of the peaceful uses of nuclear energy and technology. The role of the IAEA is to assist Member States in the effective utilization of these technologies in various domains of research such as fundamental and applied science, industry, human health care and environmental studies, as well as nuclear energy applications. In particular, material testing reactors (MTRs), serve as unique tools in scientific and technological development and they have quite a wide variety of applications. Today, a large range of different RR designs exist when compared with power reactors and they also have different operating modes, producing high neutron fluxes, which may be steady or pulsed. Recently, an urgent need has arisen for the development of new advanced materials, for example in the nuclear industry, where RRs offer capacities for irradiation programmes. Besides the scientific and research activities and commercial applications, RRs are also used extensively for educational training activities for scientists and engineers. This report is a compilation of outputs of an IAEA Technical Meeting (TM-34779) held on Research Reactor Application for Materials under High Neutron Fluence. The overall objective of the meeting was to review typical applications of small and medium size RRs, such as material characterization and testing, neutron physics and beam research, neutron radiography and imaging as well as isotope production and other types of non-nuclear applications. Several issues were discussed during the meeting, in particular: (1) recent development of irradiation facilities, specific irradiation programmes and their implementation; (2) effective and optimal RR operation regimes for irradiation purposes; (3) sharing of best practices and existing technical knowledge; and (4) fostering of advanced or innovative technologies, e.g. information exchange and effective collaboration. This

  17. Swelling of Fe-Mn and Fe-Cr-Mn alloys at high neutron fluence

    International Nuclear Information System (INIS)

    Garner, F.A.; Brager, H.R.

    1986-06-01

    Swelling data on neutron-irradiated simple Fe-Cr-Mn and Fe-Mn alloys, as well as commercial Fe-Cr-Mn base alloys are now becoming available at exposure levels approaching 50 dpa. The swelling rate decreases from the ∼1%/dpa found at lower exposures, probably due to the extensive formation of ferritic phases. As expected, commercial alloys swell less than the simple alloys

  18. Study of the response of a piezoceramic motor irradiated in a fast reactor up to a neutron fluence of 2.77E+17 n/cm2

    International Nuclear Information System (INIS)

    Pillon, Mario; Monti, Chiara; Mugnaini, Giampiero; Neri, Carlo; Rossi, Paolo; Carta, Mario; Fiorani, Orlando; Santagata, Alfonso

    2015-01-01

    Highlights: • Piezoceramic motors are compliant with magnetic field, temperature and vacuum. • We studied the response of a piezoceramic motor during the irradiation with neutrons. • The response was studied using 1 MeV neutrons up to a neutron fluence of 2.77E+17 n/cm 2 . • Neutron irradiation produces a shift of the optimal resonance frequency and a decrease of the motor speed. • The performance changes do not affect the proper operation of the motor. - Abstract: A piezoceramic motor has been identified as the potential apparatus for carrying out the rotation of the scanning head of a laser radar system used for viewing the first wall of the ITER vessel. This diagnostic is simply referred to as IVVS (In Vessel Viewing System). The choice fell on a piezoceramic motor due to the presence of strong magnetic fields (up 8 T) and of the high vacuum and temperature conditions. To be compliant with all the ITER environmental conditions it was necessary to qualify the piezo-motor under gamma and neutron irradiation. In this paper are described the procedures and tests that have been performed to verify the compatibility of the operation of the motor adopted in the presence of a fast neutron fluence which was gradually increased over time in order to reach a total value of 2.77 × 10 17 n/cm 2 . Such neutron fluence was obtained by irradiating the motor in a position close to the core of the fast nuclear reactor TAPIRO, in operation at the ENEA Casaccia Research Centre, Italy. The neutron spectrum in this position has been identified as representative of that found in the rest position of the IVVS head during ITER operation. The cumulative neutron fluence reached corresponds to that it is expected to be reached during the entire life of ITER for the IVVS in the rest position without any shield. This work describes the experimental results of this test; the methodology adopted to determine the total neutron fluence achieved and the methodology adopted for the

  19. Neutron Fluence and Energy Reconstruction with the LNE-IRSN/MIMAC Recoil Detector MicroTPC at 27 keV

    Energy Technology Data Exchange (ETDEWEB)

    Maire, D.; Lebreton, L.; Querre, Ph. [Institute for Radioprotection and Nuclear Safety - IRSN, site of Cadarache, 13115 Saint Paul lez Durance (France); Bosson, G.; Guillaudin, O.; Muraz, J.F.; Riffard, Q.; Santos, D. [Laboratoire de Physique Subatomique et de Cosmologie - LPSCCNRSIN2P3/ UJF/INP, 38000 Grenoble (France)

    2015-07-01

    The French Institute for Radiation protection and Nuclear Safety (IRSN), designated by the French Metrology Institute (LNE) for neutron metrology, is developing a time projection chamber using a Micromegas anode: microTPC. This work is carried out in collaboration with the Laboratory of Subatomic Physics and Cosmology (LPSC). The aim is to characterize the energy distribution of neutron fluence in the energy range 8 keV - 5 MeV with a primary procedure. The time projection chambers are gaseous detectors able to measure charged particles energy and to reconstruct their track if a pixelated anode is used. In our case, the gas is used as a (n, p) converter in order to detect neutrons down to few keV. Coming from elastic collisions with neutrons, recoil protons lose a part of their kinetic energy by ionizing the gas. The ionization electrons are drifted toward a pixelated anode (2D projection), read at 50 MHz by a self-triggered electronic system to obtain the third track dimension. The neutron energy is reconstructed event by event thanks to proton scattering angle and proton energy measurements. The scattering angle is deduced from the 3D track. The proton energy is obtained by charge collection measurements, knowing the ionization quenching factor (i.e. the part of proton kinetic energy lost by ionizing the gas). The fluence is calculated thanks to the detected events number and the simulation of the detector response. The μTPC is a new reliable detector able to measure energy distribution of the neutron fluence without unfolding procedure or prior neutron calibration contrary to usual gaseous counters. The microTPC is still being developed and measurements have been carried out at the AMANDE facility, with neutrons energies going from 8 keV to 565 keV. After the context and the μ-TPC working principle presentation, measurements of the neutron energy and fluence at 27 keV and 144 keV are shown and compared to the complete detector response simulation. This work

  20. The effect of incremental gamma-ray doses and incremental neutron fluences upon the performance of self-biased sup 1 sup 0 B-coated high-purity epitaxial GaAs thermal neutron detectors

    CERN Document Server

    Gersch, H K; Simpson, P A

    2002-01-01

    High-purity epitaxial GaAs sup 1 sup 0 B-coated thermal neutron detectors advantageously operate at room temperature without externally applied voltage. Sample detectors were systematically irradiated at fixed grid locations near the core of a 2 MW research reactor to determine their operational neutron dose threshold. Reactor pool locations were assigned so that fast and thermal neutron fluxes to the devices were similar. Neutron fluences ranged between 10 sup 1 sup 1 and 10 sup 1 sup 4 n/cm sup 2. GaAs detectors were exposed to exponential fluences of base ten. Ten detector designs were irradiated and studied, differentiated between p-i-n diodes and Schottky barrier diodes. The irradiated sup 1 sup 0 B-coated detectors were tested for neutron detection sensitivity in a thermalized neutron beam. Little damage was observed for detectors irradiated at neutron fluences of 10 sup 1 sup 2 n/cm sup 2 and below, but signals noticeably degraded at fluences of 10 sup 1 sup 3 n/cm sup 2. Catastrophic damage was appare...

  1. Neutron spectral characterization of the PCA-PV benchmark facility

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Kam, F.B.K.; Fabry, A.

    1980-01-01

    The Pool Critical Assembly (PCA) at the Oak Ridge National Laboratory is being used to generate the PCA-PV benchmark neutron field. A configuration consisting of steel blocks and water gaps is used to simulate the thermal shield pressure vessel configurations in power reactors. The distances between the steel blocks can be changed so that the penetration of neutrons through water and steel can be determined and compared for many different configurations. Easy access and low flux levels make it possible to conduct extensive measurements using active and passive neutron dosimetry, which are impossible to perform in commercial reactors. The clean core and simple geometry facilitates neutron transport calculations which can be validated in detail by comparison with measurements. A facility which has the same configuration of water and steel as the PCA-PV facility but contains test specimens for materials testing, will be irradiated in the higher fluxes at the Oak Ridge Research Reactor. Using the results from the PCA-PV facility, the correlation between neutron flux-fluences and radiation damage in steel can be established. This facility is being discussed in a separate paper

  2. Comparisons of calculated and measured spectral distributions of neutrons from a 14-MeV neutron source inside the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Barnes, J.M.; Alsmiller, R.G. Jr.; Emmett, M.B.; Drischler, J.D.

    1985-12-01

    A recent paper presented neutron spectral distributions (energy greater than or equal to0.91 MeV) measured at various locations around the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The neutron source for the series of measurements was a small D-T generator placed at various positions in the TFTR vacuum chamber. In the present paper the results of neutron transport calculations are presented and compared with these experimental data. The calculations were carried out using Monte Carlo methods and a very detailed model of the TFTR and the TFTR test cell. The calculated and experimental fluences per unit energy are compared in absolute units and are found to be in substantial agreement for five different combinations of source and detector positions

  3. Development of a TPC for energy and fluence references in low energies neutronic fields (from 8 keV to 5 MeV)

    International Nuclear Information System (INIS)

    Maire, Donovan

    2015-01-01

    In order to judge the measurement reliability, metrology requires to measure quantities with their uncertainties, in relation to a reference through a documented and unbroken chain of calibrations. In neutron radiation field, instrument response has to be known as a function of the neutron energy. Then detector calibrations are required using reference neutron fields. In France, primary reference neutron fields are held by the LNE-IRSN, at the Laboratory for Neutron Metrology and Dosimetry (LMDN). In order to improve reference neutron field characterization, the LNE-IRSN MIMAC μTPC has been developed. This detector is a Time Projection Chamber (TPC), using a gas at low pressure (30 mbar abs. to 1 bar abs.). Nuclear recoils are generated by neutron elastic scattering onto gas atoms. By measuring the nuclear recoil energy and scattering angle, the μTPC detector is able to measure the energy distribution of the neutron fluence between 8 keV and 5 MeV. The main challenge was to perform accurate spectrometry of neutron fields in the keV range, following a primary procedure. First of all, a metrological approach was followed in order to master every physical process taking part in the neutron detection. This approach led to develop the direct and inverse models, representing the detector response function and its inverse function respectively. Using this detailed characterization, the energy distribution of the neutron fluence has been measured for a continuous neutron field of 27 keV. The reconstructed energy is 28,2 ± 4,5 keV, the difference between μTPC integral fluence measurement and other measurement methods is less than 6%. The LNE-IRSN MIMAC μTPC system becomes the only one system able to measure simultaneously energy and fluence at energies lower than 100 keV, following a primary procedure. The project goal is then reached. These measurements at energies lower than 100 keV shows also a non-linearity between the ionization charge and the ion kinetic energy

  4. An application of low leakage loading pattern to reduce fast neutrons. Fluence on WWER-440 reactor pressure vessel in Kozloduy NPP

    International Nuclear Information System (INIS)

    Haralampieva, Tz.; Antonov, A.; Monev, M.

    2001-01-01

    The neutron exposure of a reactor pressure vessel (RPV) is one of the key factors that have to be quantified and assess reliably to provide plant life assurance and for an extension to operational life. This paper summarizes the principal methods that are used in core design optimisation for WWER-440 reactors in NPP-Kozloduy in order to reduce flux of fast neutrons at the RPV. Results of fast neutron fluence changes during the all last cycles of units 1-4 with WWER-440 reactors are considered (Authors)

  5. A Project for High Fluence 14 MeV Neutron Source

    CERN Document Server

    Pillon, Mario; Pizzuto, Aldo; Pietropaolo, Antonino

    2014-01-01

    The international community agrees on the importance to build a large facility devoted to test and validate materials to be used in harsh neutron environments. Such a facility, proposed by ENEA , reconsiders a previous study known as “Sorgentina” but takes into account new technological development so far attained. The “New Sorgentina” Fusion Source (NSFS) project is based upon an intense D - T 14 MeV neutron source achievable with T and D ion beams impinging on 2 m radius rotating target s . NSFS produces about 1 x10 13 n cm - 2 s - 1 over about 50 cm 3 . The NSFS facility will use the ion source and accelerating system technology developed for the Positive Ion Injectors (PII) used to heat the plasma in the fusion experiments,. NSFS, to be intended as an European facility, may be realized in a few years, once provided a preliminary technological program devote to study the operation of the ion source in continuous mode, target h eat loading/ removal, target and tritium handling, inventory as well as ...

  6. Neutron fluence and energy reconstruction with the IRSN recoil detector μ-TPC at 27 keV, 144 keV and 565 keV

    Energy Technology Data Exchange (ETDEWEB)

    Maire, D.; Lebreton, L.; Richer, J.P. [IRSN, PRP-HOM, SDE, LMDN, 13115 Saint Paul-Lez-Durance (France); Bosson, G.; Bourrion, O.; Guillaudin, O.; Riffard, Q.; Santos, D. [CNRS/IN2P3-UJF-INPG, LPSC, 38000 Grenoble (France)

    2015-07-01

    The French Institute for Radioprotection and Nuclear Safety (IRSN), associated to the French Metrology Institute (LNE), is developing a time projection chamber using a Micromegas anode: μ-TPC. This work is carried out in collaboration with the Laboratory of Subatomic Physics and Cosmology (LPSC). The aim is to characterize with a primary procedure the energy distribution of neutron fluence in the energy range 8 keV - 1 MeV. The time projection chambers are gaseous detectors, which are able to measure charged particles energy and to reconstruct their track if a pixelated anode is used. In our case, the gas is used as a (n, p) converter in order to detect neutrons down to few keV. Coming from elastic collisions with neutrons, recoil protons lose a part of their kinetic energy by ionizing the gas. The ionization electrons are drifted toward a pixelated anode (2D projection), read at 50 MHz by a self-triggered electronic system to obtain the third track dimension. The neutron energy is reconstructed event by event thanks to proton scattering angle and proton energy measurements. The scattering angle is deduced from the 3D track. The proton energy is obtained by charge collection measurements, knowing the ionization quenching factor (i.e. the part of proton kinetic energy lost by ionizing the gas). The fluence is calculated thanks to the detected events number and the simulated detector response. The μ-TPC is a new reliable detector which enables to measure energy distribution of the neutron fluence without deconvolution or neutron calibration contrary to usual gaseous counters. The μ-TPC is still being developed and measurements have been carried out at the AMANDE facility, with neutrons energies going from 8 keV to 565 keV. After the context and the μ-TPC working principle presentation, measurements of the neutron energy and fluence at 27.2 keV, 144 keV and 565 keV are shown and compared to the complete detector simulation. This work shows the first direct

  7. Determination of uranium and thorium in semiconductor memory materials by high fluence neutron activation analysis

    International Nuclear Information System (INIS)

    Dyer, F.F.; Emery, J.F.; Northcutt, K.J.; Scott, R.M.

    1981-01-01

    Uranium and thorium were measured by absolute neutron activation analysis in high-purity materials used to manufacture semiconductor memories. The main thrust of the study concerned aluminum and aluminum alloys used as sources for thin film preparation, evaporated metal films, and samples from the Czochralski silicon crystal process. Average levels of U and Th were found for the source alloys to be approx. 65 and approx. 45 ppB, respectively. Levels of U and Th in silicon samples fell in the range of a few parts per trillion. Evaporated metal films contained about 1 ppB U and Th, but there is some question about these results due to the possibility of contamination

  8. Comparison of Calculated and Measured Neutron Fluence in Fuel/Cladding Irradiation Experiments in HFIR

    International Nuclear Information System (INIS)

    Ellis, Ronald James

    2011-01-01

    A recently-designed thermal neutron irradiation facility has been used for a first series of irradiations of PWR fuel pellets in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. Since June 2010, irradiations of PWR fuel pellets made of UN or UO 2 , clad in SiC, have been ongoing in the outer small VXF sites in the beryllium reflector region of the HFIR, as seen in Fig. 1. HFIR is a versatile, 85 MW isotope production and test reactor with the capability and facilities for performing a wide variety of irradiation experiments. HFIR is a beryllium-reflected, light-water-cooled and -moderated, flux-trap type reactor that uses highly enriched (in 235 U) uranium (HEU) as the fuel. The reactor core consists of a series of concentric annular regions, each about 2 ft (0.61 m) high. A 5-in. (12.70-cm)-diam hole, referred to as the flux trap, forms the center of the core. The fuel region is composed of two concentric fuel elements made up of many involute-shaped fuel plates: an inner element that contains 171 fuel plates, and an outer element that contains 369 fuel plates. The fuel plates are curved in the shape of an involute, which provides constant coolant channel width between plates. The fuel (U 3 O 8 -Al cermet) is nonuniformly distributed along the arc of the involute to minimize the radial peak-to-average power density ratio. A burnable poison (B 4 C) is included in the inner fuel element primarily to reduce the negative reactivity requirements of the reactor control plates. A typical HEU core loading in HFIR is 9.4 kg of 235 U and 2.8 g of 10 B. The thermal neutron flux in the flux trap region can exceed 2.5 x 10 15 n/cm 2 · s while the fast flux in this region exceeds 1 x 10 15 n/cm 2 · s. The inner and outer fuel elements are in turn surrounded by a concentric ring of beryllium reflector approximately 1 ft (0.30 m) thick. The beryllium reflector consists of three regions: the removable reflector, the semi-permanent reflector, and the

  9. Neutron activation system for spectral measurements of pulsed ion diode neutron production

    International Nuclear Information System (INIS)

    Hanson, D.L.; Kruse, L.W.

    1980-02-01

    A neutron energy spectrometer has been developed to study intense ion beam-target interactions in the harsh radiation environment of a relativistic electron beam source. The main component is a neutron threshold activation system employing two multiplexed high efficiency Ge(Li) detectors, an annihilation gamma coincidence system, and a pneumatic sample transport. Additional constraints on the neutron spectrum are provided by total neutron yield and time-of-flight measurements. A practical lower limit on the total neutron yield into 4π required for a spectral measurement with this system is approx. 10 10 n where the neutron yield is predominantly below 4 MeV and approx. 10 8 n when a significant fraction of the yield is above 4 MeV. Applications of this system to pulsed ion diode neutron production experiments on Hermes II are described

  10. Application of low-cost Gallium Arsenide light-emitting-diodes as kerma dosemeter and fluence monitor for high-energy neutrons

    International Nuclear Information System (INIS)

    Mukherjee, B.; Simrock, S.; Khachan, J.; Rybka, D.; Romaniuk, R.

    2007-01-01

    Displacement damage (DD) caused by fast neutrons in unbiased Gallium Arsenide (GaAs) light emitting diodes (LED) resulted in a reduction of the light output. On the other hand, a similar type of LED irradiated with gamma rays from a 60 Co source up to a dose level in excess of 1.0 kGy (1.0 x 10 5 rad) was found to show no significant drop of the light emission. This phenomenon was used to develop a low cost passive fluence monitor and kinetic energy released per unit mass dosemeter for accelerator-produced neutrons. These LED-dosemeters were used to assess the integrated fluence of photoneutrons, which were contaminated with a strong Bremsstrahlung gamma-background generated by the 730 MeV superconducting electron linac driving the free electron laser in Hamburg (FLASH) at Deutsches Elektronen-Synchrotron. The applications of GaAs LED as a routine neutron fluence monitor and DD precursor for the electronic components located in high-energy accelerator environment are highlighted. (authors)

  11. The vessel fluence; Fluence cuve

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This book presents the proceedings of the technical meeting on the reactors vessels fluence. They are grouped in eight sessions: the industrial context and the stakes of the vessels control; the organization and the methodology for the fluence computation; the concerned physical properties; the reference computation methods; the fluence monitoring in an industrial context; vessels monitoring under irradiation; others methods in the world; the research and development programs. (A.L.B.)

  12. Evaluation of the fluence to dose conversion coefficients for high energy neutrons using a voxel phantom coupled with the GEANT4 code

    CERN Document Server

    Paganini, S

    2005-01-01

    Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from Galactic cosmic radiation. Crews of future high-speed commercial flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the male adult voxels phantom MAX, developed in the Nuclear Energy Department of Pernambuco Federal University in Brazil, has been coupled with the Monte Carlo simulation code GEANT4. This toolkit, distributed and upgraded from the international scientific community of CERN/Switzerland, simulates thermal to ultrahigh energy neutrons transport and interactions in the matter. The high energy neutrons are pointed as the component that contribute about 70% of the neutron effective dose that represent the 35% to 60% total dose at aircraft altitude. In this research calculations of conversion coefficients from fluence to effective dose are performed for neutrons of energies from 100 MeV ...

  13. Interministerial decree of 10 February 1988 fixing the derived limits of the air concentration and the annual intake limit and the values of the quality factor and the neutron fluence rate

    International Nuclear Information System (INIS)

    1988-01-01

    This decree establishes the derived concentration limits in the air and annual inhalation limits for the radioisotopes and the values of the quality factors and the conversion factors fluence/dose equivalent for neutrons and protons

  14. Spectral correction factors for conventional neutron dose meters used in high-energy neutron environments improved and extended results based on a complete survey of all neutron spectra in IAEA-TRS-403

    International Nuclear Information System (INIS)

    Oparaji, U.; Tsai, Y. H.; Liu, Y. C.; Lee, K. W.; Patelli, E.; Sheu, R. J.

    2017-01-01

    This paper presents improved and extended results of our previous study on corrections for conventional neutron dose meters used in environments with high-energy neutrons (E n > 10 MeV). Conventional moderated-type neutron dose meters tend to underestimate the dose contribution of high-energy neutrons because of the opposite trends of dose conversion coefficients and detection efficiencies as the neutron energy increases. A practical correction scheme was proposed based on analysis of hundreds of neutron spectra in the IAEA-TRS-403 report. By comparing 252 Cf-calibrated dose responses with reference values derived from fluence-to-dose conversion coefficients, this study provides recommendations for neutron field characterization and the corresponding dose correction factors. Further sensitivity studies confirm the appropriateness of the proposed scheme and indicate that (1) the spectral correction factors are nearly independent of the selection of three commonly used calibration sources: 252 Cf, 241 Am-Be and 239 Pu-Be; (2) the derived correction factors for Bonner spheres of various sizes (6''-9'') are similar in trend and (3) practical high-energy neutron indexes based on measurements can be established to facilitate the application of these correction factors in workplaces. (authors)

  15. Genetic Algorithms: A New Method for Neutron Beam Spectral Characterization

    International Nuclear Information System (INIS)

    David W. Freeman

    2000-01-01

    A revolutionary new concept for solving the neutron spectrum unfolding problem using genetic algorithms (GAs) has recently been introduced. GAs are part of a new field of evolutionary solution techniques that mimic living systems with computer-simulated chromosome solutions that mate, mutate, and evolve to create improved solutions. The original motivation for the research was to improve spectral characterization of neutron beams associated with boron neutron capture therapy (BNCT). The GA unfolding technique has been successfully applied to problems with moderate energy resolution (up to 47 energy groups). Initial research indicates that the GA unfolding technique may well be superior to popular unfolding methods in common use. Research now under way at Kansas State University is focused on optimizing the unfolding algorithm and expanding its energy resolution to unfold detailed beam spectra based on multiple foil measurements. Indications are that the final code will significantly outperform current, state-of-the-art codes in use by the scientific community

  16. Ionization chamber with build-up cup spectral sensitivity to megavoltage (0.5-20 MeV) photon fluences in free air

    International Nuclear Information System (INIS)

    Gorlachev, G.E.

    2002-01-01

    In-air measurements of photon beam properties, used in radiation therapy, is common practice for determining radiation output dependence from the field size, known as head scatter factors (HSF). PMMA and brass build-up caps are most popular miniphantoms for providing electron equilibrium. Discrepancies up to 2% in HSF measurements by different combinations of detectors and equilibrium caps have been published. One of the main reasons of those discrepancies is the detector system spectral sensitivity and differences in primary and scatter radiation spectra. In the light of new model based dose calculation methods direct radiation fluence measurement is of great interest. So, understanding of detector spectral sensitivity is important task for modern dosimetry of radiation therapy. In the present study Monte Carlo (MC) method was employed to calculate ionization chamber response to monoenergetic photon fluences, normalized to water kerma units. Simulation was done using EGS4 package. Electron transport was performed with ESTEPE equal to 4%. PEGS cross sections were generated for maximal energy 20 MeV with cutoff kinetic energy 10 KeV both for photons and electrons. Scanditronix RK-05 ionization chamber was chosen as a prototype. Eight cylindrical miniphantoms, representing four materials (PMMA, Al, Cu, Pb) and two front wall thickness, were simulated. Results are presented. Miniphantom front wall thicknesses in each case are shown in the figure. Diameter depends on the material and equal respectively: PMMA - 4, Al - 2.5, Cu - 1.5, and PB - 1.5 cm. Ionization chamber outer diameter is equal to 0.7 cm. Detector sensitivity has considerable energy dependence. Two effects explain it. First is the radiation attenuation in the miniphantom. Second is pair production, which dominates in high atomic number miniphantoms for energies above 5 MeV. Depending on the miniphantom material detector response changes from 1.5 to 5 times in the energy range from 0.5 to 20 MeV. Correct

  17. Neutron spectral modulation as a new thermal neutron scattering technique. Pt. 1

    International Nuclear Information System (INIS)

    Ito, Y.; Nishi, M.; Motoya, K.

    1982-01-01

    A thermal neutron scattering technique is presented based on a new idea of labelling each neutron in its spectral position as well as in time through the scattering process. The method makes possible the simultaneous determination of both the accurate dispersion relation and its broadening by utilizing the resolution cancellation property of zero-crossing points in the cross-correlated time spectrum together with the Fourier transform scheme of the neutron spin echo without resorting to the echoing. The channel Fourier transform applied to the present method also makes possible the determination of the accurate direct energy scan profile of the scattering function with a rather broad incident neutron wavelength distribution. Therefore the intensity sacrifice for attaining high accurarcy is minimized. The technique is used with either a polarized or unpolarized beam at the sample position with no precautions against beam depolarization at the sample for the latter case. Relative time accurarcy of the order of 10 -3 to 10 -4 may be obtained for the general dispersion relation and for the quasi-elastic energy transfers using correspondingly the relative incident neutron wavelength spread of 10 to 1% around an incident neutron energy of a few meV. (orig.)

  18. Comparison of pressure vessel neutron fluences for the Balakovo-3 reactor with measurements and investigation of the influence of neutron cross sections and number of groups on the results

    Energy Technology Data Exchange (ETDEWEB)

    Barz, H U; Boehmer, B; Konheiser, J; Stephan, I

    1998-10-01

    The general methodical questions of experimental and theoretical determination of neutron fluences have been described in connection with the measurements and 3-D Monte Carlo calculation for the Rovno-3 reactor. The same calculation and measurement methods were applied for the Balakovo-3 reactor. In the first part, the results of the comparison for Balakovo will be given and discussed. However, for this reactor the main attention was focussed on investigations of the accuracy of the calculation. In this connection an important question is the influence of neutron data on the results. With this respect not only the source of the data but also the number of energy groups is important. (orig.)

  19. New development of neutron spectral modulation data analysis

    International Nuclear Information System (INIS)

    Ito, Y.

    1988-01-01

    A study is made on procedures for obtaining desired scattering function information. The neutron spectral modulation technique incorporates both the low (including DC) and high frequency Fourier components in its incident spectrum. Lake's procedure increases the Fourier components of the doconvoluted scattering function by using the existing Fourier components as nucleus, thereby bridges the Fourier gap and extends the Fourier region. Since the Lake's procedure takes care of the missing Fourier components, a single measurement using an appropriate NSM modulation suffices to recover the S(W) line shape. Deep modulation depth is not essential to reproduce the scattering function. This should be contrasted to the previous NSM treatment as well as to the neutron spin echo method, both of which require the several repeat of measurements with the varying modulation frequency under the high degree of beam polarization condition. Although the computer simulation of the present paper does not include the statistical fluctuation encountered in the experimental data, these analyses show a great promise of the NSM method, which can now be used with much flexibility in the field of both cold and ultracold neutron scattering experiment. (N.K.)

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

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

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

  1. Neutron fluence at the reactor pressure vessel wall - a comparison of French and German procedures and strategies in PWRs

    International Nuclear Information System (INIS)

    Tricot, N.; Jendrich, U.

    2003-01-01

    While the neutrons within the core may take part in the chain reaction, those neutrons emitted from the core are basically lost for the energy production. This 'neutron leakage' represents a loss of fuel efficiency and causes neutron embrittlement of the reactor pressure vessel (RPV) wall. The latter raises safety concerns, needs to be monitored closely and may necessitate mitigating measures. There are different strategies to deal with these two undesirable effects: The neutron emission may be reduced to some extent all around the core or just at the 'hot spots' of RPV embrittlement by tailored core loading patterns. A higher absorption rate of neutrons may also be achieved by a larger water gap between the core and the RPV. In this paper the inter-relations between the distribution of neutron flux, core geometry, core loading strategy, RPV embrittlement and its surveillance are discussed at first. Then the different strategies followed by the German and French operators are described. Finally the conclusions will highlight the communalities and differences between these strategies as different approaches to the same problem of safety as well as economy. (authors)

  2. Measurements of the absolute neutron fluence spectrum emitted at 00 and 900 from the Little-Boy replica

    International Nuclear Information System (INIS)

    Roberts, J.H.; Gold, R.; Preston, C.C.

    1986-01-01

    Nuclear research emulsions (NRE) have been used to characterize the neutron spectrum emitted by the Little-Boy replica. NRE were irradiated at the Little-Boy surface, as well as approximately 2 m from the center of the Little-Boy replica, using polar angles of 0 0 , 30 0 , 60 0 , and 90 0 . For the NRE exposed at 2 m, neutron background was determined using shadow shields of borated polyethylene. Emulsion scanning to date has concentrated exclusively on the 2-m, 0 0 and 2-m, 90 0 locations. Approximately 5000 proton-recoil tracks have been measured in NRE irradiated at each of these locations. At the 2-m, 90 0 location, the NRE neutron spectrum extends from 0.37 MeV up to 8.2 MeV; whereas the NRE neutron spectrum at the 2-m, 0 0 location is much softer and extends only up to 2.7 MeV. NRE neutron spectrometry results at these two locations are compared with both liquid scintillator neutron spectrometry and Monte Carlo calculations. (author)

  3. Spectral effects in low-dose fission and fusion neutron irradiated metals and alloys

    International Nuclear Information System (INIS)

    Heinisch, H.L.; Atkin, S.D.; Martinez, C.

    1986-04-01

    Flat miniature tensile specimens were irradiated to neutron fluences up to 9 x 10 22 n/m 2 in the RTNS-II and in the Omega West Reactor. Specimen temperatures were the same in both environments, with runs being made at both 90 0 C and 290 0 C. The results of tensile tests on AISI 316 stainless steel, A302B pressure vessel steel and pure copper are reported here. The radiation-induced changes in yield strength as a function of neutron dose in each spectrum are compared. The data for 316 stainless steel correlate well on the basis of displacements per atom (dpa), while those for copper and A302B do not. In copper the ratio of fission dpa to 14 MeV neutron dpa for a given yield stress change is about three to one. In A302B pressure vessel steel this ratio is more than three at lower fluences, but the yield stress data for fission and 14 MeV neutron-irradiated A302B steel appears to coalesce or intersect at the higher fluences

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  5. Thermal and epithermal neutron fluence rate gradient measurements by PADC detectors in LINAC radiotherapy treatments-field

    Energy Technology Data Exchange (ETDEWEB)

    Barrera, M. T., E-mail: mariate9590@gmail.com; Barros, H.; Pino, F.; Sajo-Bohus, L. [Universidad Simón Bolívar, Nuclear Physics Laboratory, Sartenejas, Caracas (Venezuela, Bolivarian Republic of); Dávila, J. [Física Médica C. A. and Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    LINAC VARIAN 2100 is where energetic electrons produce Bremsstrahlung radiation, with energies above the nucleon binding energy (E≈5.5MeV). This radiation induce (γ,n) and (e,e’n) reactions mainly in the natural tungsten target material (its total photoneutron cross section is about 4000 mb in a energy range from 9-17 MeV). These reactions may occur also in other components of the system (e.g. multi leaf collimator). During radiation treatment the human body may receive an additional dose inside and outside the treated volume produced by the mentioned nuclear reactions. We measured the neutron density at the treatment table using nuclear track detectors (PADC-NTD). These covered by a boron-converter are employed, including a cadmium filter, to determine the ratio between two groups of neutron energy, i.e. thermal and epithermal. The PADC-NTD detectors were exposed to the radiation field at the iso-center during regular operation of the accelerator. Neutron are determined indirectly by the converting reaction {sup 10}B(n,α){sup 7}Li the emerging charged particle leave their kinetic energy in the PADC forming a latent nuclear track, enlarged by chemical etching (6N, NaOH, 70°C). Track density provides information on the neutron density through calibration coefficient (∼1.6 10{sup 4} neutrons /track) obtained by a californium source. We report the estimation of the thermal and epithermal neutron field and its gradient for photoneutrons produced in radiotherapy treatments with 18 MV linear accelerators. It was obsered that photoneutron production have higher rate at the iso-center.

  6. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core; Determinacion de la fluencia neutronica en las soldaduras del 'core shroud' del nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M A; Xolocostli M, J V; Gomez T, A M; Palacios H, J C [ININ, 52750 Ocoyoacac, Estado de mexico (Mexico)

    2006-07-01

    With the purpose of defining the inspection frequency, in function of the embrittlement of the materials that compose the welding of the 'Core Shroud' or encircling of the core of a BWR type reactor, is necessary to know the neutron fluence received for this welding. In the work the calculated values of neutron fluence accumulated maxim (E > 1 MeV) during the first 8 operation cycles of the reactor are presented. The calculations were carried out according to the NRC Regulatory Guide 1.190, making use of the DORT code, which solves the transport equation in discreet ordinate in two dimensions (xy, r{theta}, and rz). The results in 3D were obtained applying the Synthesis method according to the guide before mentioned. Results are presented for the horizontal welding H3, H4, and H5, showing the corresponding curves to the fluence accumulated to the cycle 8 and a projection for the cycle 14 is presented. (Author)

  7. Comparison of four NDT methods for indication of reactor steel degradation by high fluences of neutron irradiation

    Czech Academy of Sciences Publication Activity Database

    Tomáš, Ivan; Vértesy, G.; Pirfo Barroso, S.; Kobayashi, S.

    2013-01-01

    Roč. 265, DEC (2013), s. 201-209 ISSN 0029-5493 Institutional support: RVO:68378271 Keywords : neutron irradiation * steel degradation * nuclear reactor pressure vessel * magnetic NDT * magnetic minor hysteresis loops * Magnetic Barkhausen Emission Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 0.972, year: 2013 http://www.sciencedirect.com/science/article/pii/S0029549313004664

  8. Spectral distribution measurements of neutrons in paraffin borax mixtures

    International Nuclear Information System (INIS)

    El-Khatib, A.M.; Gaber, M.; Abou El-Khier, M.A.

    1987-01-01

    Neutron fluxes from a compact D-T neutron source has been measured in paraffin-borax mixtures by using activation foil detectors with successive threshold energies. The absorbed doses, backscattering coefficients and build-up factors were determined as well. The contribution of thermal and intermediate neutron dose is much lower, compared to that of fast neutrons. Among the used mediums, paraffin loaded with 4% borax concentration was found to be the best absorbing medium against neutrons at near depths within the blocks, while at a depth around 12 cm the neutron absorption (or scattering) is independent on the type of the used medium. (author)

  9. Study of the response of a piezoceramic motor irradiated in a fast reactor up to a neutron fluence of 2.77E+17 n/cm{sup 2}

    Energy Technology Data Exchange (ETDEWEB)

    Pillon, Mario, E-mail: mario.pillon@enea.it [Associazione EURATOM-ENEA sulla Fusione, ENEA C.R. Frascati, via E. Fermi, 45, 00044 Frascati, Rome (Italy); Monti, Chiara; Mugnaini, Giampiero; Neri, Carlo; Rossi, Paolo [Associazione EURATOM-ENEA sulla Fusione, ENEA C.R. Frascati, via E. Fermi, 45, 00044 Frascati, Rome (Italy); Carta, Mario; Fiorani, Orlando; Santagata, Alfonso [ENEA C.R. CASACCIA, via Anguillarese, 301, 00123 S. Maria di Galeria, Rome (Italy)

    2015-10-15

    Highlights: • Piezoceramic motors are compliant with magnetic field, temperature and vacuum. • We studied the response of a piezoceramic motor during the irradiation with neutrons. • The response was studied using 1 MeV neutrons up to a neutron fluence of 2.77E+17 n/cm{sup 2}. • Neutron irradiation produces a shift of the optimal resonance frequency and a decrease of the motor speed. • The performance changes do not affect the proper operation of the motor. - Abstract: A piezoceramic motor has been identified as the potential apparatus for carrying out the rotation of the scanning head of a laser radar system used for viewing the first wall of the ITER vessel. This diagnostic is simply referred to as IVVS (In Vessel Viewing System). The choice fell on a piezoceramic motor due to the presence of strong magnetic fields (up 8 T) and of the high vacuum and temperature conditions. To be compliant with all the ITER environmental conditions it was necessary to qualify the piezo-motor under gamma and neutron irradiation. In this paper are described the procedures and tests that have been performed to verify the compatibility of the operation of the motor adopted in the presence of a fast neutron fluence which was gradually increased over time in order to reach a total value of 2.77 × 10{sup 17} n/cm{sup 2}. Such neutron fluence was obtained by irradiating the motor in a position close to the core of the fast nuclear reactor TAPIRO, in operation at the ENEA Casaccia Research Centre, Italy. The neutron spectrum in this position has been identified as representative of that found in the rest position of the IVVS head during ITER operation. The cumulative neutron fluence reached corresponds to that it is expected to be reached during the entire life of ITER for the IVVS in the rest position without any shield. This work describes the experimental results of this test; the methodology adopted to determine the total neutron fluence achieved and the methodology adopted

  10. The desing study of high voltage plasma focus for a large fluence neutron source by using a water capacitor bank

    International Nuclear Information System (INIS)

    Ueno, Isao; Kobata, Tadasuke

    1983-01-01

    A new possibility for high intensity neutron source (HINS) would be opened by the plasma focus device if we have a high voltage capacitor bank. A scaling law of neutron yield for D-T gas discharge in plasma focus device is obtained after Imshennik, Filippov and Filippova. The resulting scaling law shows the realizability of the D-T HINS by the use of plasma focus, provided that the device is operated under a high voltage condition. Until now, it has been difficult to construct the high voltage capacitor bank of long life, for example with V 0 =300kV, C 0 =200μF and L 0 --5nH necessary in the level of HINS. It becomes possible to design this capacitor bank by using the coaxial water capacitor which has been developed for the electron and ion beam accelerator. The size of a capacitor designed for V 0 =300kV, C 0 =1μF is phi5m x 22m. Two hundred capacitors are used in parallel in order to get the 200μF. (author)

  11. Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

    International Nuclear Information System (INIS)

    Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

    2009-01-01

    The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

  12. Poster - 25: Neutron Spectral Measurements around a Scanning Proton Beam

    Energy Technology Data Exchange (ETDEWEB)

    Kildea, John; Enger, Shirin; Maglieri, Robert; Mirzakhanian, Lalageh; Dahlgren, Christina Vallhagen; Dubeau, Jacques; Witharana, Sanjeeva [Medical Physics Unit, McGill University Health Centre, Medical Physics Unit, McGill University, Medical Physics Unit, McGill University, Medical Physics Unit, McGill University, Skandion Clinic, Detec Inc., Gatineau, Quebec, Detec Inc., Gatineau, Quebec (Canada)

    2016-08-15

    We describe the measurements of neutron spectra that we undertook around a scanning proton beam at the Skandion proton therapy clinic in Uppsala, Sweden. Measurements were undertaken using an extended energy range Nested Neutron Spectrometer (NNS, Detec Inc., Gatineau, QC) operated in pulsed and current mode. Spectra were measured as a function of location in the treatment room and for various Bragg peak depths. Our preliminary unfolded data clearly show the direct, evaporation and thermal neutron peaks and we can show the effect on the neutron spectrum of a water phantom in the primary proton beam.

  13. Development of a modular directional and spectral neutron detection system using solid-state detectors

    Energy Technology Data Exchange (ETDEWEB)

    Weltz, A., E-mail: weltza3@gmail.com; Torres, B.; McElwain, L.; Dahal, R.; Huang, J.; Bhat, I.; Lu, J.; Danon, Y.

    2015-08-21

    A detection system using room-temperature, microstructured solid-state thermal neutron detectors with very low leakage current has been developed at Rensselaer Polytechnic Institute (RPI) with the ability to provide positional and spectral information about an unknown neutron source. The Directional and Spectral Neutron Detection System (DSNDS) utilizes a set of small-but-scalable, zero-bias solid-state thermal neutron detectors which have demonstrated high thermal neutron efficiency and adequate gamma insensitivity. The DSNDS can gather spectral information about an unknown neutron source with a relatively small number of detectors, simplifying the detector electronics and minimizing cost; however, the DSNDS is modular in design, providing the capability to increase the detection efficiency and angular resolution. The system used in this paper was comprised of a stack of five high-density polyethylene (HDPE) disks with a thickness of 5 cm and a diameter of 30 cm, the middle disk containing 16 detectors positioned as one internal (moderated) and one external (unmoderated) ring of solid-state neutron detectors. These two detector rings provide the ability to determine the directionality of a neutron source. The system gathers spectral information about a neutron source in two ways: by measuring the relative responses of the internal ring of detectors as well as measuring the ratio of the internal-to-external detector responses. Experiments were performed with variable neutron spectra: a {sup 252}Cf spontaneous fission neutron source which was HDPE moderated, HDPE reflected, lead (Pb) shielded, and bare in order to benchmark the system for spectral sensitivity. Simulations were performed in order to characterize the neutron spectra corresponding to each of the source configurations and showed agreement with experimental measurements. The DSNDS demonstrates the ability to determine the relative angle of the source and the hardness of the neutron spectrum. By using the

  14. Results of neutron measurements in the spectral position of the Juelich FKS steel irradiation capsules

    International Nuclear Information System (INIS)

    Schneider, W.

    1986-10-01

    This is a report on the planning and results of neutron monitoring in the capsules of the Juelich steel irradiation for the research project on component safety (FKS). The table of results and their discussion is provided specifically for the spectral positions (for characterising the neutron spectrum) in each of the types of irradiation capsules used. The results are given for the reaction rates of the neutron measurement reactions used (activation or fission reactions), for the neutron flux densities and fluxes derived from them related to the actual or at least plausible neutron spectra and finally for the radiation damage (or exposure) of the irradiated material calculated from them, expressed as the atomic displacement figure (dpa) and its percentage in sections of the neutron spectrum. (orig.) [de

  15. Evaluation of spectral unfolding techniques for neutron spectroscopy

    International Nuclear Information System (INIS)

    Sunden, Erik Andersson; Conroy, S.; Ericsson, G.; Johnson, M. Gatu; Giacomelli, L.; Hellesen, C.; Hjalmarsson, A.; Ronchi, E.; Sjoestrand, H.; Weiszflog, M.; Kaellne, J.; Gorini, G.; Tardocchi, M.

    2008-01-01

    The precision of the JET installations of MAXED, GRAVEL and the L-curve version of MAXED has been evaluated by using synthetic neutron spectra. We have determined the number of counts needed for the detector systems NE213 and MPR to get an error below 10% of the MAXED unfolded neutron spectra is determined to be ∼10 6 and ∼10 4 , respectively. For GRAVEL the same number is ∼10 7 and ∼3·10 4 for NE213 and MPR, respectively

  16. Browns Ferry Unit 3 cavity neutron spectral analysis

    International Nuclear Information System (INIS)

    Martin, G.C.; Till, H.A.

    1982-01-01

    The General Electric Company at the Vallecitos Nuclear Center (GE-VNC) has performed neutron dosimetry measurements in the Browns Ferry Unit 3 reactor (BF3) cavity using multiple dosimeter and spectrum unfolding techniques. These measurements are the first in a BWR cavity and comprise an important part in a general program related to verification of pressure vessel integrity and to validation of calculations. Determinations of BF3 cavity neutron flux densities at five key locations at full power (1098 MWe) during core cycle 2 (November 1978 to August 1979) are presented

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

    CERN Document Server

    Unholzer, S; Klein, H; Seidel, K

    2002-01-01

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

  18. Spectral investigation of neutron radiation in three-sectional concrete labyrinth from a californium-252 source

    International Nuclear Information System (INIS)

    Belogorlov, E.A.; Britvich, G.I.; Getmanov, V.B.

    1985-01-01

    Construction of labyrinths in points of communication output from the storage-ring under construction is accompanied by numerous difficulties due to a considerable number of gas and cryogenic pipelines, which require large cross sections at the minimal length of the pipelines proper for their location. It results in unfavourable for radiation attenuation ratios between cross section and length of the labyrinth separate sections. Neutron spectra in a model concrete labyrinth, at the entrance to which a neutron source with fission spectrum (californium-252) and the same source in a polyethylene moderator are located, are measured. On the basis of the spectra obtained the formation of fluence and equivalent dose along the labyrinth geometric axis is analyzed. Conditions permitting actually to reduce radiation dose in the labyrinth (dead end provision, the use of cover materials, construction of diaphragms and shielding plates) are simulated

  19. Neutron dosimetry at the intense neutron source (INS)

    International Nuclear Information System (INIS)

    Dierckx, R.

    1977-01-01

    The neutron monitoring consists of two parts: the spectral characterization and the fluence determination. The experimental measurements are combined with theoretical calculations. The following methods are proposed for determining the spectra: a telescope (np) spectrometer, a telescope 6 Li(nα)T spectrometer, spectrometers needing unfolding, time-of-flight technique, and multiple foil technique

  20. Homogenization of the critically spectral equation in neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Allaire, G. [CEA Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie]|[Paris-6 Univ., 75 (France). Lab. d' Analyse Numerique; Bal, G. [Electricite de France (EDF), 92 - Clamart (France). Direction des Etudes et Recherches

    1998-07-01

    We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)

  1. Homogenization of the critically spectral equation in neutron transport

    International Nuclear Information System (INIS)

    Allaire, G.; Paris-6 Univ., 75; Bal, G.

    1998-01-01

    We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)

  2. The Continued Spectral Evolution of the Neutron Star RX J0720.4-3125

    NARCIS (Netherlands)

    Vink, Jacco; de Vries, Cor P.; Méndez, Mariano; Verbunt, Frank

    2004-01-01

    We observed the isolated neutron star RX J0720.4-3125 with Chandra's Low Energy Transmission Grating Spectrometer, following the XMM-Newton discovery of the long-term spectral evolution of this source. The new observation shows that the spectrum of RX J0720.4-3125 has continued to change in the

  3. Neutron spectral characterization of the NRC-HSST experiments

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Kam, F.B.K.

    1979-01-01

    Irradiation experiments are being performed for the US Nuclear Regulatory Commission (NRC) Heavy Section Steel Technology (HSST) program. Results of dosimetry performed in the second experiment have been previously reported. Similar procedures were followed in the third experiment. The experiences gained in these two experiments have led to modifications in the composition and distribution of foil dosimeters which monitor the neutron flux-spectra in the irradiated steel specimens. It is expected that in the new experiments much higher accuracies than previously possible can be achieved in the determination of irradiation damage parameters

  4. Neutron spectral characterization of the NRC-HSST experiments

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F. W.; Kam, F. B.K.

    1979-01-01

    Irradiation experiments are being performed for the US Nuclear Regulatory Commission (NRC) Heavy Section Steel Technology (HSST) program. Results of dosimetry performed in the second experiment have been previously reported. Similar procedures were followed in the third experiment. The experiences gained in these two experiments have led to modifications in the composition and distribution of foil dosimeters which monitor the neutron flux-spectra in the irradiated steel specimens. It is expected that in the new experiments much higher accuracies than previously possible can be achieved in the determination of irradiation damage parameters.

  5. Neutron spectrum perturbations due to scattering materials and their effect on the average neutron energy, the spectral index, and the hardness parameter

    International Nuclear Information System (INIS)

    Wright, H.L.; Meason, J.L.; Wolf, M.; Harvey, J.T.

    1976-01-01

    Measurements have been performed on the perturbing effect of a number of scattering materials by the 'free-field' neutron leakage spectrum from a Godiva Type Critical Assembly (White Sands Missile Range Fast Burst Reactor). The results of these measurements are interpreted in relation to some of the general parameters characterizing a neutron environment, namely, the average neutron energy >10 KeV, the spectral index and the hardness parameter. Three neutron spectrum measurements have been performed, each under different experimental configurations of scattering materials. Results from these measurements show the following with relation to the spectral index: (1) The neutron environment on the core surface and at 12-inches from the core surface (free-field) yield a spectral index of 6.8, (2) The neutron environment behind a 4.75-inch Plexiglas plate yield 4.6 for the spectral index and (3) The neutron environment behind a 2-inch aluminum plate yield 6.7 for the spectral index. It is concluded that the core surface and the 12-inch from core surface neutron environment are identical with the 'free-field' neutron environment at 20-inches when considering only those neutrons with energy >10 KeV. On the other hand, it appears that the 4.75 inches of Plexiglas severely perturbs the 'free-field' neutron environment, i.e., a much harder neutron spectrum >10 KeV. In the situation where 2-inches of aluminum is used as the perturbing medium, essentially no change in the neutron spectrum >10 KeV is noted

  6. Multiscale simulation of neutron induced damage in tritium breeding blankets with different spectral shifters

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong Hee; Joo, Han Gyu, E-mail: joohan@snu.ac.kr

    2013-10-15

    Highlights: • A multiscale defect simulation system tailored for neutron damage estimation is introduced. • The new recoil spectrum code can use the most recent ENDF-B/VII nuclear data. • The high energy cascades are broken into subcascades using the INCAS model. • OKMC simulation provides data for shear stress estimation using dislocation dynamics formula. • Demonstration is made with a fusion blanket design having different spectral shifters. -- Abstract: A multiscale material defect simulation established to evaluate neutron induced damages on metals is applied to an estimation of material degradation in helium cooled molten lithium blankets in which four different spectral shifter materials are examined as a means of maximizing the tritium breeding ratio through proper shaping of the neutron spectrum. The multiscale system consists of a Monte Carlo neutron transport code, a recoil spectrum generation code, a molecular dynamics code, a high energy cascade breakup model, an object kinetic Monte Carlo code, and a simple formula as the shear stress estimator. The average recoil energy of the primary knock-on atoms, the total concentration of the defects, average defect sizes, and the increase in shear stress after a certain irradiation time are calculated for each spectral shifter. Among the four proposed materials of B4C, Be, Graphite and TiC, B4C reveals the best shielding performance in terms of neutron radiation hardening. The result for the increase in shear stress after 100 days of irradiation indicates that the increased shear stress is 1.5 GPa for B4C which is about 40% less than that of the worst one, the graphite spectral shifter. The other damage indicators show consistent trends.

  7. The spectral element approach for the solution of neutron transport problems

    International Nuclear Information System (INIS)

    Barbarino, A.; Dulla, S.; Ravetto, P.; Mund, E.H.

    2011-01-01

    In this paper a possible application of the Spectral Element Method to neutron transport problems is presented. The basic features of the numerical scheme on the one-dimensional diffusion equation are illustrated. Then, the AN model for neutron transport is introduced, and the basic steps for the construction of a bi-dimensional solver are described. The AN equations are chosen for their structure, involving a system of coupled elliptic-type equations. Some calculations are carried out on typical benchmark problems and results are compared with the Finite Element Method, in order to evaluate their performances. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  9. The spectral element method for static neutron transport in AN approximation. Part I

    International Nuclear Information System (INIS)

    Barbarino, A.; Dulla, S.; Mund, E.H.; Ravetto, P.

    2013-01-01

    Highlights: ► Spectral elements methods (SEMs) are extended for the neutronics of nuclear reactor cores. ► The second-order, A N formulation of neutron trasport is adopted. ► Results for classical benchmark cases in 2D are presented and compared to finite elements. ► The advantages of SEM in terms of precision and convergence rate are illustrated. ► SEM consitutes a promising approach for the solution of neutron transport problems. - Abstract: Spectral elements methods provide very accurate solutions of elliptic problems. In this paper we apply the method to the A N (i.e. SP 2N−1 ) approximation of neutron transport. Numerical results for classical benchmark cases highlight its performance in comparison with finite element computations, in terms of accuracy per degree of freedom and convergence rate. All calculations presented in this paper refer to two-dimensional problems. The method can easily be extended to three-dimensional cases. The results illustrate promising features of the method for more complex transport problems

  10. Radiance and particle fluence

    International Nuclear Information System (INIS)

    Papiez, L.; Battista, J.J.

    1994-01-01

    The International Commission on Radiological Units and Measurements (ICRU) has defined fluence in terms of the number of the radiation particles crossing a small sampling sphere. A second definition has been proposed in which the length of track segments contained within any sampling volume are used to calculate the incident fluence. This approach is often used in Monte Carlo simulations of individual particle tracks, allowing the fluence to be scored in small volumes of any shape. In this paper we stress that the second definition generalizes the classical (ICRU) concept of fluence. We also identify the assumptions inherent in the two definitions of fluence and prove their equivalence for the case of straight-line particle trajectories. (author)

  11. Design, Fabrication and Test Report on a Verification Capsule (05M-06K) for the Control of a Neutron Irradiation Fluence of Specimens in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H.; Cho, M. S.; Son, J. M.; Shin, Y. T.; Park, S. J.; Choi, M. H.; Lee, D. S.

    2007-02-15

    As a part of a project for a capsule development and utilization for an irradiation test, a verification capsule (05M-06K) was designed, fabricated and tested for the development of new instrumented capsule technology for a more precise control of the irradiation fluence of a specimen, irrespective of the reactor operation condition. The basic structure of the 05M-06K capsule was based on the 04M-22K mock-up capsule which was successfully designed and out-pile tested to confirm the various key technologies necessary for the fluence control of a specimen. 21 square and round shaped specimens made of STS 304 were inserted into the capsule. The capsule was constructed in 5 stages with specimens and an independent electric heater at each stage. Each of the five specimens which were accommodated in the 1st stage (top) of the capsule can be taken out of the HANARO core during a normal reactor operation. The specimen is extracted by a specimen extraction mechanism using a steel wire. During the out-pile test, the temperatures of the specimens were measured by 12 thermocouples installed in the capsule. The capsule was successfully out-pile tested in a single channel test loop. The obtained results will be used for a safety evaluation of the new irradiation capsule for controlling the irradiation fluence of specimens in HANARO.

  12. Accuracy of helium accumulation fluence monitor for fast reactor dosimetry

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

    A helium (He) accumulation fluence monitor (HAFM) has been developed for fast reactor dosimetry. In order to evaluate the measurement accuracy of neutron fluence by the HAFM method, the HAFMs of enriched boron (B) and beryllium (Be) were irradiated in the Fast Neutron Source Reactor `YAYOI`. The number of He atoms produced in the HAFMs were measured and compared with the calculated values. As a result of this study, it was confirmed that the neutron fluence could be measured within 5 % by the HAFM method, and that met the required accuracy for fast reactor dosimetry. (author)

  13. Inference of core barrel motion from neutron noise spectral density. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, J.C.; Shahrokhi, F.; Kryter, R.C.

    1977-03-15

    A method was developed for inference of core barrel motion from the following statistical descriptors: cross-power spectral density, autopower spectral density, and amplitude probability density. To quantify the core barrel motion in a typical pressurized water reactor (PWR), a scale factor was calculated in both one- and two-dimensional geometries using forward, variational, and perturbation methods of discrete ordinates neutron transport. A procedure for selection of the proper frequency band limits for the statistical descriptors was developed. It was found that although perturbation theory is adequate for the calculation of the scale factor, two-dimensional geometric effects are important enough to rule out the use of a one-dimensional approximation for all but the crudest calculations. It was also found that contributions of gamma rays can be ignored and that the results are relatively insensitive to the cross-section set employed. The proper frequency band for the statistical descriptors is conveniently determined from the coherence and phase information from two ex-core power range neutron monitors positioned diametrically across the reactor vessel. Core barrel motion can then be quantified from the integral of the band-limited cross-power spectral density of two diametrically opposed ex-core monitors or, if the coherence between the pair is greater than or equal to 0.7, from a properly band-limited amplitude probability density function. Wide-band amplitude probability density functions were demonstrated to yield erroneous estimates for the magnitude of core barrel motion.

  14. Three-dimensional RAMA fluence methodology benchmarking

    International Nuclear Information System (INIS)

    Baker, S. P.; Carter, R. G.; Watkins, K. E.; Jones, D. B.

    2004-01-01

    This paper describes the benchmarking of the RAMA Fluence Methodology software, that has been performed in accordance with U. S. Nuclear Regulatory Commission Regulatory Guide 1.190. The RAMA Fluence Methodology has been developed by TransWare Enterprises Inc. through funding provided by the Electric Power Research Inst., Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP). The purpose of the software is to provide an accurate method for calculating neutron fluence in BWR pressure vessels and internal components. The Methodology incorporates a three-dimensional deterministic transport solution with flexible arbitrary geometry representation of reactor system components, previously available only with Monte Carlo solution techniques. Benchmarking was performed on measurements obtained from three standard benchmark problems which include the Pool Criticality Assembly (PCA), VENUS-3, and H. B. Robinson Unit 2 benchmarks, and on flux wire measurements obtained from two BWR nuclear plants. The calculated to measured (C/M) ratios range from 0.93 to 1.04 demonstrating the accuracy of the RAMA Fluence Methodology in predicting neutron flux, fluence, and dosimetry activation. (authors)

  15. Fast fluence measurement for JOYO irradiation field using niobium dosimeter

    International Nuclear Information System (INIS)

    Ito, Chikara

    2004-03-01

    Neutron fluence and spectrum are key parameters in various irradiation tests and material surveillance tests so they need to be evaluated accurately. The reactor dosimetry test has been conducted by the multiple foil activation method, and a niobium dosimeter has been developed for measurement of fast neutron fluence in the experimental fast reactor JOYO. The inelastic scattering reaction of 93 Nb has a low threshold energy, about 30 keV, and the energy distribution of reaction cross section is similar to the displacement cross section for iron. Therefore, a niobium dosimeter is suitable for evaluation of the fast neutron fluence and the displacement per atom for iron. Moreover, a niobium dosimeter is suited to measure neutron fluence in long-term irradiation test because 93 Nb, which is produced by the reaction, has a long half-life (16.4 years). This study established a high precision measurement technique using the niobium reaction rate. The effect of self-absorption was decreased by the solution and evaporation to dryness of niobium dosimeter. The dosimeter weight was precisely measured using the inductively coupled plasma mass spectrometer. This technique was applied to JOYO dosimetry. The fast neutron fluences (E > 0.1 MeV) found by measuring the reaction rate in the niobium dosimeter were compared with the values evaluated using the multiple foil activation method. The ratio of measured fast neutron fluences by means of niobium dosimeter and multiple foil activation method range from 0.97 to 1.03 and agree within the experimental uncertainty. The measurement errors of fast neutron fluence by niobium dosimeter range from 4.5% (fuel region) to 10.1% (in-vessel storage rack). As a result of this study, the high precision measurement of fast neutron fluence by niobium dosimeters was confirmed. The accuracy of fast reactor dosimetry will be improved by application of niobium dosimeters to the irradiation tests in the JOYO MK-III core. (author)

  16. s-process studies in the light of new experimental cross sections: Distribution of neutron fluences and r-process residuals

    International Nuclear Information System (INIS)

    Kaeppeler, F.; Beer, H.; Wisshak, K.; Clayton, D.D.; Macklin, R.L.; Ward, R.A.

    1981-08-01

    A best set of neutron-capture cross sections has been evaluated for the most important s-process isotopes. With this data base, s-process studies have been carried out using the traditional model which assumes a steady neutron flux and an exponential distribution of neutron irradiations. The calculated sigmaN-curve is in excellent agreement with the empirical sigmaN-values of pure s-process nuclei. Simultaneously, good agreement is found between the difference of solar and s-process abundances and the abundances of pure r-process nuclei. We also discuss the abundance pattern of the iron group elements where our s-process results complement the abundances obtained from explosive nuclear burning. The results obtained from the traditional s-process model such as seed abundances, mean neutron irradiations, or neutron densities are compared to recent stellar model calculations which assume the He-burning shells of red giant stars as the site for the s-process. (orig.) [de

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

    International Nuclear Information System (INIS)

    Rataj, J.

    1993-10-01

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

  18. Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

    Science.gov (United States)

    Harkness, Ira; Zhu, Ting; Liang, Yinong; Rauch, Eric; Enqvist, Andreas; Jordan, Kelly A.

    2018-01-01

    Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

  19. Proposal of a New Method for Neutron Dosimetry Based on Spectral Information Obtained by Application of Artificial Neural Networks

    International Nuclear Information System (INIS)

    Fehrenbacher, G.; Schuetz, R.; Hahn, K.; Sprunck, M.; Cordes, E.; Biersack, J.P.; Wahl, W.

    1999-01-01

    A new method for the monitoring of neutron radiation is proposed. It is based on the determination of spectral information on the neutron field in order to derive dose quantities like the ambient dose equivalent, the dose equivalent, or other dose quantities which depend on the neutron energy. The method uses a multi-element system consisting of converter type silicon detectors. The unfolding procedure is based on an artificial neural network (ANN). The response function of each element is determined by a computational model considering the neutron interaction with the dosemeter layers and the subsequent transport of produced ions. An example is given for a multi-element system. The ANN is trained by a given set of neutron spectra and then applied to count responses obtained in neutron fields. Four examples of spectra unfolded using the ANN are presented. (author)

  20. Study of the distribution of neutron fluence in a treatment room with proton; Estudio de la distribucion de fluencia neutronica en una sala de tratamientos con protonterapia

    Energy Technology Data Exchange (ETDEWEB)

    Lagares, J. I.; Arce, P.; Sansaloni, F.; Terron, J. A.; Exposito, M. R.; Nieto-Camero, J. J.; Korf, A.; Loubser, M.; Sanchez-Doblado, F.

    2011-07-01

    The aim of this study is to determine the energy distribution of flow in the treatment room. With this information we seek the most appropriate location in which to place an active detector, based on digital memories, to calculate the neutron dose in the patient peripheral.

  1. Spatially adaptive hp refinement approach for PN neutron transport equation using spectral element method

    International Nuclear Information System (INIS)

    Nahavandi, N.; Minuchehr, A.; Zolfaghari, A.; Abbasi, M.

    2015-01-01

    Highlights: • Powerful hp-SEM refinement approach for P N neutron transport equation has been presented. • The method provides great geometrical flexibility and lower computational cost. • There is a capability of using arbitrary high order and non uniform meshes. • Both posteriori and priori local error estimation approaches have been employed. • High accurate results are compared against other common adaptive and uniform grids. - Abstract: In this work we presented the adaptive hp-SEM approach which is obtained from the incorporation of Spectral Element Method (SEM) and adaptive hp refinement. The SEM nodal discretization and hp adaptive grid-refinement for even-parity Boltzmann neutron transport equation creates powerful grid refinement approach with high accuracy solutions. In this regard a computer code has been developed to solve multi-group neutron transport equation in one-dimensional geometry using even-parity transport theory. The spatial dependence of flux has been developed via SEM method with Lobatto orthogonal polynomial. Two commonly error estimation approaches, the posteriori and the priori has been implemented. The incorporation of SEM nodal discretization method and adaptive hp grid refinement leads to high accurate solutions. Coarser meshes efficiency and significant reduction of computer program runtime in comparison with other common refining methods and uniform meshing approaches is tested along several well-known transport benchmarks

  2. Simulations of inspiraling and merging double neutron stars using the Spectral Einstein Code

    Science.gov (United States)

    Haas, Roland; Ott, Christian D.; Szilagyi, Bela; Kaplan, Jeffrey D.; Lippuner, Jonas; Scheel, Mark A.; Barkett, Kevin; Muhlberger, Curran D.; Dietrich, Tim; Duez, Matthew D.; Foucart, Francois; Pfeiffer, Harald P.; Kidder, Lawrence E.; Teukolsky, Saul A.

    2016-06-01

    We present results on the inspiral, merger, and postmerger evolution of a neutron star-neutron star (NSNS) system. Our results are obtained using the hybrid pseudospectral-finite volume Spectral Einstein Code (SpEC). To test our numerical methods, we evolve an equal-mass system for ≈22 orbits before merger. This waveform is the longest waveform obtained from fully general-relativistic simulations for NSNSs to date. Such long (and accurate) numerical waveforms are required to further improve semianalytical models used in gravitational wave data analysis, for example, the effective one body models. We discuss in detail the improvements to SpEC's ability to simulate NSNS mergers, in particular mesh refined grids to better resolve the merger and postmerger phases. We provide a set of consistency checks and compare our results to NSNS merger simulations with the independent bam code. We find agreement between them, which increases confidence in results obtained with either code. This work paves the way for future studies using long waveforms and more complex microphysical descriptions of neutron star matter in SpEC.

  3. Measurements of thermal neutron fluence in the bunker of a cyclotron for PET isotope production; Medidas de fluencia de neutrones termicos en el bunker de un ciclotron de produccion de isotopos para PET

    Energy Technology Data Exchange (ETDEWEB)

    Mendez Villafane, R.; Sansoloni florit, F.; Lagares gonzalez, J. L.; Llop Roig, J.; Guerrero Araque, J. E.; Muniz Gutierrez, J. L.; Perez Morales, J. M.

    2011-07-01

    To measure the neutron spectrum has been used spectrometry system based on Bonner spheres with Au flakes as thermal neutron detector at its center while the results are still pending and will be analyzing another job.

  4. Determination of the Spectral Index in the Fission Spectrum Energy Regime

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Amy Sarah [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-05-16

    Neutron reaction cross sections play a vital role in tracking the production and destruction of isotopes exposed to neutron fluence. They are central to the process of reconciling the initial and final atom inventories. Measurements of irradiated samples by radiochemical methods in tangent with an algorithm are used to evaluate the fluence a sample is exposed to over the course of the irradiation. This algorithm is the Isotope Production Code (IPC) created and used by the radiochemistry data assessment team at Los Alamos National Laboratory (LANL). An integral result is calculated by varying the total neutron fluence seen by a sample. A sample, irradiated in a critical assembly, will be exposed to a unique neutron flux defined by the neutron source and distance of the sample from the source. Neutron cross sections utilized are a function of the hardness of the neutron spectrum at the location of irradiation. A spectral index is used an indicator of the hardness of the neutron spectrum. Cross sections fit forms applied in IPC are collapsed from a LANL 30-group energy structure. Several decades of research and development have been performed to formalize the current IPC cross section library. Basis of the current fission spectrum neutron reaction cross section library is rooted in critical assembly experiments performed from the 1950’s through the early 1970’s at LANL. The focus of this report is development of the spectral index used an indicator of the hardness of the neutron spectrum in the fission spectrum energy regime.

  5. Unfolding of neutron spectra from Godiva type critical assemblies

    International Nuclear Information System (INIS)

    Harvey, J.T.; Meason, J.L.; Wright, H.L.

    1976-01-01

    The results from three experiments conducted at the White Sands Missile Range Fast Burst Reactor Facility are discussed. The experiments were designed to measure the ''free-field'' neutron leakage spectrum and the neutron spectra from mildly perturbed environments. SAND-II was used to calculate the neutron spectrum utilizing several different trial input spectra for each experiment. Comparisons are made between the unfolded neutron spectrum for each trial input on the basis of the following parameters: average neutron energy (above 10 KeV), integral fluence (above 10 KeV), spectral index and the hardness parameter, phi/sub eq//phi

  6. Extended use of alanine irradiated in experimental reactor for combined gamma- and neutron-dose assessment by ESR spectroscopy and thermal neutron fluence assessment by measurement of (14)C by LSC.

    Science.gov (United States)

    Bartoníček, B; Kučera, J; Světlík, I; Viererbl, L; Lahodová, Z; Tomášková, L; Cabalka, M

    2014-11-01

    Gamma- and neutron doses in an experimental reactor were measured using alanine/electron spin resonance (ESR) spectrometry. The absorbed dose in alanine was decomposed into contributions caused by gamma and neutron radiation using neutron kerma factors. To overcome a low sensitivity of the alanine/ESR response to thermal neutrons, a novel method has been proposed for the assessment of a thermal neutron flux using the (14)N(n,p) (14)C reaction on nitrogen present in alanine and subsequent measurement of (14)C by liquid scintillation counting (LSC). Copyright © 2014 Elsevier Ltd. All rights reserved.

  7. Fluence map segmentation

    International Nuclear Information System (INIS)

    Rosenwald, J.-C.

    2008-01-01

    The lecture addressed the following topics: 'Interpreting' the fluence map; The sequencer; Reasons for difference between desired and actual fluence map; Principle of 'Step and Shoot' segmentation; Large number of solutions for given fluence map; Optimizing 'step and shoot' segmentation; The interdigitation constraint; Main algorithms; Conclusions on segmentation algorithms (static mode); Optimizing intensity levels and monitor units; Sliding window sequencing; Synchronization to avoid the tongue-and-groove effect; Accounting for physical characteristics of MLC; Importance of corrections for leaf transmission and offset; Accounting for MLC mechanical constraints; The 'complexity' factor; Incorporating the sequencing into optimization algorithm; Data transfer to the treatment machine; Interface between R and V and accelerator; and Conclusions on fluence map segmentation (Segmentation is part of the overall inverse planning procedure; 'Step and Shoot' and 'Dynamic' options are available for most TPS (depending on accelerator model; The segmentation phase tends to come into the optimization loop; The physical characteristics of the MLC have a large influence on final dose distribution; The IMRT plans (MU and relative dose distribution) must be carefully validated). (P.A.)

  8. Fast neutron detection by means of an organic solid state track detector

    International Nuclear Information System (INIS)

    Doerschel, B.; Streubel, G.

    1980-01-01

    Solid state track detectors consisting of cellulose triacetate foils are appropriate for measuring the fast neutron fluence without applying external radiators. Detector sensitivity has been determined as a function of neutron energy by performing irradiations with various neutron sources and monoenergetic neutrons of different energies. A comparison with theoretical results given in the literature for a simple model of track recording has shown sufficient agreement. The measuring errors and the influence of spectral changes in the neutron field on detector response are discussed for the studied method of fluence measurement. By means of these errors the measuring range has been determined for well defined irradiation conditions, taking into account spectral changes in the neutron field. (author)

  9. The spectral energy distributions of isolated neutron stars in the resonant cyclotron scattering model

    Science.gov (United States)

    Tong, Hao; Xu, Renxin

    2013-03-01

    The X-ray dim isolated neutron stars (XDINSs) are peculiar pulsar-like objects, characterized by their very well Planck-like spectrum. In studying their spectral energy distributions, the optical/UV excess is a long standing problem. Recently, Kaplan et al. (2011) have measured the optical/UV excess for all seven sources, which is understandable in the resonant cyclotron scattering (RCS) model previously addressed. The RCS model calculations show that the RCS process can account for the observed optical/UV excess for most sources. The flat spectrum of RX J2143.0+0654 may due to contribution from bremsstrahlung emission of the electron system in addition to the RCS process.

  10. Analysis of neutron reflectivity data: maximum entropy, Bayesian spectral analysis and speckle holography

    International Nuclear Information System (INIS)

    Sivia, D.S.; Hamilton, W.A.; Smith, G.S.

    1991-01-01

    The analysis of neutron reflectivity data to obtain nuclear scattering length density profiles is akin to the notorious phaseless Fourier problem, well known in many fields such as crystallography. Current methods of analysis culminate in the refinement of a few parameters of a functional model, and are often preceded by a long and laborious process of trial and error. We start by discussing the use of maximum entropy for obtained 'free-form' solutions of the density profile, as an alternative to the trial and error phase when a functional model is not available. Next we consider a Bayesian spectral analysis approach, which is appropriate for optimising the parameters of a simple (but adequate) type of model when the number of parameters is not known. Finally, we suggest a novel experimental procedure, the analogue of astronomical speckle holography, designed to alleviate the ambiguity problems inherent in traditional reflectivity measurements. (orig.)

  11. Influence of target-scattered neutrons on cross-section measurements

    International Nuclear Information System (INIS)

    Lesiecki, H.; Cosack, M.; Siebert, B.R.L.

    1985-01-01

    Monoenergetic neutrons produced with accelerators are usually accompanied by degraded and secondary neutrons which arise from reactions of source neutrons in the material of the target construction. A Monte Carlo code was written which takes into account the kinematics and the angular source strength of the neutron producing reaction and the interactions of the neutrons with the material in the immediate vicinity of their production. The calculation of the spectral distribution of the neutron fluence is compared with the result of a time-of-flight measurement. (author)

  12. Application of spectral decomposition analysis to in vivo quantification of aluminum by neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Comsa, D.C. E-mail: comsadc@mcmaster.ca; Prestwich, W.V.; McNeill, F.E.; Byun, S.H

    2004-12-01

    The toxic effects of aluminum are cumulative and result in painful forms of renal osteodystrophy, most notably adynamic bone disease and osteomalacia, but also other forms of disease. The Trace Element Group at McMaster University has developed an accelerator-based in vivo procedure for detecting aluminum body burden by neutron activation analysis (NAA). Further refining of the method was necessary for increasing its sensitivity. In this context, the present study proposes an improved algorithm for data analysis, based on spectral decomposition. A new minimum detectable limit (MDL) of (0.7{+-}0.1) mg Al was reached for a local dose of (20{+-}1) mSv. The study also addresses the feasibility of a new data acquisition technique, the electronic rejection of the coincident events detected by a NaI(Tl) system. It is expected that the application of this technique, together with spectral decomposition analysis, would provide an acceptable MDL for the method to be valuable in a clinical setting.

  13. Heat-to-heat variability of irradiation creep and swelling of HT9 irradiated to high neutron fluence at 400-600{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    Irradiation creep data on ferritic/martensitic steels are difficult and expensive to obtain, and are not available for fusion-relevant neutron spectra and displacement rates. Therefore, an extensive creep data rescue and analysis effort is in progress to characterize irradiation creep of ferritic/martensitic alloys in other reactors and to develop a methodology for applying it to fusion applications. In the current study, four tube sets constructed from three nominally similar heats of HT9 subjected to one of two heat treatments were constructed as helium-pressurized creep tubes and irradiated in FFTF-MOTA at four temperatures between 400 and 600{degrees}C. Each of the four heats exhibited a different stress-free swelling behavior at 400{degrees}C, with the creep rate following the swelling according to the familiar B{sub o} + DS creep law. No stress-free swelling was observed at the other three irradiation temperatures. Using a stress exponent of n = 1.0 as the defining criterion, {open_quotes}classic{close_quotes} irradiation creep was found at all temperatures, but, only over limited stress ranges that decreased with increasing temperature. The creep coefficient B{sub o} is a little lower ({approx}50%) than that observed for austenitic steel, but the swelling-creep coupling coefficient D is comparable to that of austenitic steels. Primary transient creep behavior was also observed at all temperatures except 400{degrees}C, and thermal creep behavior was found to dominate the deformation at high stress levels at 550 and 600{degrees}C.

  14. Basics of Neutrons for First Responders

    Energy Technology Data Exchange (ETDEWEB)

    Rees, Brian G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2018-02-05

    These are slides from a presentation on the basics of neutrons. A few topics covered are: common origins of terrestrial neutron radiation, neutron sources, neutron energy, interactions, detecting neutrons, gammas from neutron interactions, neutron signatures in gamma-ray spectra, neutrons and NaI, neutron fluence to dose (msV), instruments' response to neutrons.

  15. Preliminary design of the Neutron Spectral Shifter that is dedicated to the IFMIF Liquid Breeder Validation Module

    Energy Technology Data Exchange (ETDEWEB)

    Mas, A., E-mail: amassanchez@gmail.com; Mota, F.; Casal, N.; García, A.; Rapisarda, D.; Arroyo, J.M.; Molla, J.; Ibarra, A.

    2014-10-15

    The International Fusion Materials Irradiation Facility (IFMIF) has a D-Li neutron stripping source that provides typical fusion irradiation conditions for material testing. The Liquid Breeder Validation Module (LBVM) is one of the medium flux test modules of the IFMIF that is used to account for some of the DEMO liquid breeder blanket R and D needs. Previous analyses have shown that the main irradiation parameters (He (appm)/dpa and H (appm)/dpa) in the medium flux area of the IFMIF can be improved to fit the expected parameters in the DEMO reactor for functional materials of liquid breeder blankets. Therefore, the design of an additional module, called the Neutron Spectral Shifter (NSS), has been considered to optimize the irradiation conditions of LBVM experiments. The proposed concept consists of supported tungsten plates working as a shifter material inside a steel structure. This design assures the mechanical integrity of the different components and it fulfills the neutronic requirements as well as the cooling capability. This present paper summarizes the work devoted to the design of the LBVM Neutron Spectral Shifter as well as the results of neutronic, thermo-hydraulic, mechanical and safety studies carried out to validate the design.

  16. Spectral-Timing Analysis of Kilohetrz Quasi-Periodic Osciallations in Neutron Star Low-Mass X-ray Binaries

    Science.gov (United States)

    Cackett, Edward; Troyer, Jon; Peille, Philippe; Barret, Didier

    2018-01-01

    Kilohertz quasi-periodic oscillations or kHz QPOs are intensity variations that occur in the X-ray band observed in neutron star low-mass X-ray binary (LMXB) systems. In such systems, matter is transferred from a secondary low-mass star to a neutron star via the process of accretion. kHz QPOs occur on the timescale of the inner accretion flow and may carry signatures of the physics of strong gravity (c2 ~ GM/R) and possibly clues to constraining the neutron star equation of state (EOS). Both the timing behavior of kHz QPOs and the time-averaged spectra of these systems have been studied extensively. No model derived from these techniques has been able to illuminate the origin of kHz QPOs. Spectral-timing is an analysis technique that can be used to derive information about the nature of physical processes occurring within the accretion flow on the timescale of the kHz QPO. To date, kHz QPOs of (4) neutron star LMXB systems have been studied with spectral-timing techniques. We present a comprehensive study of spectral-timing products of kHz QPOs from systems where data is available in the RXTE archive to demonstrate the promise of this technique to gain insights regarding the origin of kHz QPOs. Using data averaged over the entire RXTE archive, we show correlated time-lags as a function of QPO frequency and energy, as well as energy-dependent covariance spectra for the various LMXB systems where spectral-timing analysis is possible. We find similar trends in all average spectral-timing products for the objects studied. This suggests a common origin of kHz QPOs.

  17. The irradiation creep characteristics of graphite to high fluences

    International Nuclear Information System (INIS)

    Kennedy, C.R.; Cundy, M.; Kleist, G.

    1988-01-01

    High-temperature gas-cooled reactors (HTGR) have massive blocks of graphite with thermal and neutron-flux gradients causing high internal stresses. Thermal stresses are transient; however, stresses generated by differential growth due to neutron damage continue to increase with time. Fortunately, graphite also experiences creep under irradiation allowing relaxation of stresses to nominally safe levels. Because of complexity of irradiation creep experiments, data demonstrating this phenomenon are generally limited to fairly low fluences compared to the overall fluences expected in most reactors. Notable exceptions have been experiments at 300/degree/C and 500/degree/C run at Petten under tension and compression creep stresses to fluences greater than 4 /times/ 10 26 (E > 50 keV) neutrons/m 2 . This study complements the previous results by extending the irradiation temperature to 900/degree/C. 2 refs., 3 figs

  18. Evaluation of the Fluence Conversion Factor for 32P in Sulfur

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-03-18

    When 32S is exposed to neutrons it undergoes a 32S(n,p)32P reaction with a neutron cross section as shown in Figure 1. This reaction may be used to characterize the neutron fluence for neutrons greater than 3 MeV.

  19. Spectral flux of the p-7Li(C Q-M neutron source measured by proton recoil telescope

    Directory of Open Access Journals (Sweden)

    Simakov S.P.

    2010-10-01

    Full Text Available The cyclotron-based fast neutron source at NPI produces mono-energetic neutron fields up to 35 MeV neutron energy using the p + 7Li(carbon backing reactions. To be applied for activation cross-section measurements, not only the intensity of neutron peak, but also the contribution of low-energy continuum in the spectra must be well determined. Simulations of the spectral flux from present source at a position of irradiated samples were performed using CYRIC TOF-data validated in the present work against LA150h by calculations with the transport Monte Carlo code MCNPX. Simulated spectra were tested by absolute measurements using a proton-recoil telescope technique. The recoil-proton spectrometer consisted of a shielded scattering chamber with polyethylene and carbon radiators and the ΔE1-ΔE2-E telescope of silicon-surface detectors located to the neutron beam axis at 45° in the laboratory system. Si-detectors were handled by usual data acquisition system. Dead-time – and pulse-overlap losses of events were determined from the count rate of pulse generator registered during duty cycle of accelerator operation. The proton beam charge and data were taken in the list mode for later replay and analysis. The calculations for 7Li(p,n and 12C(p,n reactions reasonably reproduce CYRIC TOF neutron source spectra. The influence of neutron source set-up (proton beam dimensions, 7Li-foil, carbon stopper, cooling medium, target support/chamber and the geometry-arrangement of irradiated sample on the spectral flux is discussed in details.

  20. Extended use of alanine irradiated in experimental reactor for combined gamma-and neutron-dose assessment by ESR spectroscopy and thermal neutron fluence assessment by measurement of C-14 by LSC

    Czech Academy of Sciences Publication Activity Database

    Bartoníček, B.; Kučera, Jan; Světlík, Ivo; Viererbl, L.; Lahodová, Z.; Tomášková, Lenka; Cabalka, M.

    2014-01-01

    Roč. 93, NOV (2014), s. 52-56 ISSN 0969-8043 R&D Projects: GA TA ČR TA02010218 Institutional support: RVO:61389005 Keywords : reactor rediation * alanine/ESR dosimeter * C-14 * LSC simultaneous gamma * neutron dose * assessment Subject RIV: JF - Nuclear Energetics Impact factor: 1.231, year: 2014

  1. Multitaper spectral analysis of cosmic rays Sao Martinho da Serra's muon telescope and Newark's neutron monitor data

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Marlos Rockenbach da; Alarcon, Walter Demetrio Gonzalez; Echer, Ezequiel; Lago, Alisson dal; Lucas, Aline de [National Institute for Space Research - INPE-MCT, Sao Jose dos Campos, SP (Brazil); Vieira, Luis Eduardo Antunes; Guarnieri, Fernando Luis [Universidade do Vale do Paraiba - UNIVAP, Sao Jose dos Campos, SP (Brazil); Schuch, Nelson Jorge [Southern Regional Space Research Center - CRSPE/INPE-MCT, Santa Maria, RS (Brazil); Munakata, Kazuoki, E-mail: marlos@dge.inpe.br, E-mail: gonzalez@dge.inpe.br, E-mail: eecher@dge.inpe.br, E-mail: dallago@dge.inpe.br, E-mail: delucas@dge.inpe.br, E-mail: levieira@univap.br, E-mail: guarnieri@univap.br, E-mail: njschuch@lacesm.ufsm.br, E-mail: kmuna00@gipac.shinshu-u.ac.jp [Physics Department, Shinshu University, Matsumoto (Japan)

    2007-07-01

    In this work we present an analysis on the correction efficiency of atmospheric effects on cosmic ray Sao Martinho da Serra's muon telescope and Newark's neutron monitor data. We use a Multitaper spectral analysis of cosmic rays time series to show the main periodicities present in the corrected and uncorrected data for the atmospheric effects. This kind of correction is very important when intends to study cosmic rays variations of extra-terrestrial origin. (author)

  2. Fast neutron damage of silicon pin photodiodes

    International Nuclear Information System (INIS)

    Dabrowski, W.; Korbel, K.; Skoczen, A.

    1990-01-01

    A Hamamatsu Photonics photodiode S1723 was tested with respect to the fast neutron radiation. The device was irradiated with neutrons of energies in the range of 0.5 MeV to 12 MeV from a Po-Be source. The irradiation was performed in several steps starting from the relatively low fluence of 2.5 x 10 10 n x cm -2 . The following characteristics were measured: leakage current vs bias voltage, capacitance vs bias voltage and vs frequency, noise vs time constant of a quasigaussian shaper and spectral density of noise. Significant changes of the leakage current and of the noise were observed at the fluence of neutrons as low as 2.5 x 10 10 n x cm -2 . 8 figs., 3 tabs., 15 refs. (author)

  3. Multilayer detector for operative estimation of spectral composition of neutron fields

    CERN Document Server

    Dedenko, G L; Kaplun, A A; Kolesnikov, S V; Samosadnyj, A V; Samosadnyj, V T

    2002-01-01

    Paper describes measuring and control equipment to detect and to identify neutron sources. The equipment comprises two multilayer detectors based on sup 3 He-counters of slow neutron and poly ethylene moderator of fast neutrons, as well as, intensifiers-signal shapers, power multichannel intensifies, power unit, 8-bit microprocessor base information acquisition and processing system. Paper contains the results of measurements of energy dependence of sensitivity of neutron recording by detector layers with application of monoenergetic neutron fluxes. Difference of the experimental data and the Monte Carlo method base calculation results is 10% maximum

  4. Helium production measurements for neutron dosimetry and damage correlations

    International Nuclear Information System (INIS)

    Farrar, H. IV; Lippincott, E.P.

    1978-01-01

    Helium accumulation fluence monitors (HAFM's), consisting of miniature vanadium capsules containing small, accurately-known amounts of 10 B or 6 Li, are being used routinely for neutron dosimetry measurements in breeder reactor environments. Additionally, solid wires of Al, Fe and Cu have been irradiated by 14.8-MeV neutrons from the d-T reaction, and measurements of the helium production along these wires have given detailed neutron fluence profiles. Additional materials with relatively high (n,α) cross sections are being tested in a wide variety of neutron environments to select HAFM sets that will provide spectral information by unfolding techniques. The mass spectrometric helium measurement technique has been demonstrated to produce results with better than 2% (1 sigma) absolute accuracy. Intercomparisons with other laboratories have demonstrated good correlations with radiometric and fission chamber dosimetry results

  5. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core; Determinacion de la fluencia neutronica en las soldaduras del 'core shroud' del nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C. [ININ, 52750 Ocoyoacac, Estado de mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2006-07-01

    With the purpose of defining the inspection frequency, in function of the embrittlement of the materials that compose the welding of the 'Core Shroud' or encircling of the core of a BWR type reactor, is necessary to know the neutron fluence received for this welding. In the work the calculated values of neutron fluence accumulated maxim (E > 1 MeV) during the first 8 operation cycles of the reactor are presented. The calculations were carried out according to the NRC Regulatory Guide 1.190, making use of the DORT code, which solves the transport equation in discreet ordinate in two dimensions (xy, r{theta}, and rz). The results in 3D were obtained applying the Synthesis method according to the guide before mentioned. Results are presented for the horizontal welding H3, H4, and H5, showing the corresponding curves to the fluence accumulated to the cycle 8 and a projection for the cycle 14 is presented. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-22

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

  7. Fluence determination by scattering measurements

    CERN Document Server

    Albergo, S; Potenza, R; Tricomi, A; Pillon, M; Angarano, M M; Creanza, D; De Palma, M

    2000-01-01

    An alternative method to determine particle fluence is proposed, which is particularly suitable for irradiations with low-energy charged-particle beams. The fluence is obtained by measuring the elastic scattering produced by a composite thin target placed upstream of the sample. The absolute calibration is performed by comparison with the measured radioactivation of vanadium and copper samples. The composite thin target, made of aluminium, carbon and gold, allows not only the fluence to be measured, but also a continuous monitoring of the beam space distribution. Experimental results with a 27 MeV proton beam are reported and compared with Monte Carlo simulations. (7 refs).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-22

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

  9. A spectral effect on phase evolution in neutron-irradiated aluminum?

    International Nuclear Information System (INIS)

    Farrell, K.

    1994-01-01

    It is proposed that the spatial distribution of transmutation-produced silicon precipitates formed in aluminum by capture of thermal neutrons is systematically modified by the fast neutron portion of the spectrum, and the tensile strength is altered accordingly. The concept is used to rationalize some previously erratic tensile data

  10. Comparison of calculated and measured spectral response and intrinsic efficiency for a boron-loaded plastic neutron detector

    Energy Technology Data Exchange (ETDEWEB)

    Kamykowski, E.A. (Grumman Corporate Research Center, Bethpage, NY (United States))

    1992-07-15

    Boron-loaded scintillators offer the potential for neutron spectrometers with a simplified, peak-shaped response. The Monte Carlo code, MCNP, has been used to calculate the detector characteristics of a scintillator made of a boron-loaded plastic, BC454, for neutrons between 1 and 7 MeV. Comparisons with measurements are made of spectral response for neutron energies between 4 and 6 MeV and of intrinsic efficiencies for neutrons up to 7 MeV. In order to compare the calculated spectra with measured data, enhancements to MCNP were introduced to generate tallies of light output spectra for recoil events terminating in a final capture by {sup 10}B. The comparison of measured and calculated spectra shows agreement in response shape, full width at half maximum, and recoil energy deposition. Intrinsic efficiencies measured to 7 MeV are also in agreement with the MCNP calculations. These results validate the code predictions and affirm the value of MCNP as a useful tool for development of sensor concepts based on boron-loaded plastics. (orig.).

  11. Correlating Fast Fluence to dpa in Atypical Locations

    Directory of Open Access Journals (Sweden)

    Drury Thomas H.

    2016-01-01

    Full Text Available Damage to a nuclear reactor's materials by high-energy neutrons causes changes in the ductility and fracture toughness of the materials. The reactor vessel and its associated piping's ability to withstand stress without brittle fracture are paramount to safety. Theoretically, the material damage is directly related to the displacements per atom (dpa via the residual defects from induced displacements. However in practice, the material damage is based on a correlation to the high-energy (E > 1.0 MeV neutron fluence. While the correlated approach is applicable when the material in question has experienced the same neutron spectrum as test specimens which were the basis of the correlation, this approach is not generically acceptable. Using Monte Carlo and discrete ordinates transport codes, the energy dependent neutron flux is determined throughout the reactor structures and the reactor vessel. Results from the models provide the dpa response in addition to the high-energy neutron flux. Ratios of dpa to fast fluence are calculated throughout the models. The comparisons show a constant ratio in the areas of historical concern and thus the validity of the correlated approach to these areas. In regions above and below the fuel however, the flux spectrum has changed significantly. The correlated relationship of material damage to fluence is not valid in these regions without adjustment. An adjustment mechanism is proposed.

  12. Correlating Fast Fluence to dpa in Atypical Locations

    Science.gov (United States)

    Drury, Thomas H.

    2016-02-01

    Damage to a nuclear reactor's materials by high-energy neutrons causes changes in the ductility and fracture toughness of the materials. The reactor vessel and its associated piping's ability to withstand stress without brittle fracture are paramount to safety. Theoretically, the material damage is directly related to the displacements per atom (dpa) via the residual defects from induced displacements. However in practice, the material damage is based on a correlation to the high-energy (E > 1.0 MeV) neutron fluence. While the correlated approach is applicable when the material in question has experienced the same neutron spectrum as test specimens which were the basis of the correlation, this approach is not generically acceptable. Using Monte Carlo and discrete ordinates transport codes, the energy dependent neutron flux is determined throughout the reactor structures and the reactor vessel. Results from the models provide the dpa response in addition to the high-energy neutron flux. Ratios of dpa to fast fluence are calculated throughout the models. The comparisons show a constant ratio in the areas of historical concern and thus the validity of the correlated approach to these areas. In regions above and below the fuel however, the flux spectrum has changed significantly. The correlated relationship of material damage to fluence is not valid in these regions without adjustment. An adjustment mechanism is proposed.

  13. Spectral nodal methodology for multigroup slab-geometry discrete ordinates neutron transport problems with linearly anisotropic scattering

    Energy Technology Data Exchange (ETDEWEB)

    Oliva, Amaury M.; Filho, Hermes A.; Silva, Davi M.; Garcia, Carlos R., E-mail: aoliva@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: davijmsilva@yahoo.com.br, E-mail: cgh@instec.cu [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional; Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    In this paper, we propose a numerical methodology for the development of a method of the spectral nodal class that will generate numerical solutions free from spatial truncation errors. This method, denominated Spectral Deterministic Method (SDM), is tested as an initial study of the solutions (spectral analysis) of neutron transport equations in the discrete ordinates (S{sub N}) formulation, in one-dimensional slab geometry, multigroup approximation, with linearly anisotropic scattering, considering homogeneous and heterogeneous domains with fixed source. The unknowns in the methodology are the cell-edge, and cell average angular fluxes, the numerical values calculated for these quantities coincide with the analytic solution of the equations. These numerical results are shown and compared with the traditional ne- mesh method Diamond Difference (DD) and the coarse-mesh method spectral Green's function (SGF) to illustrate the method's accuracy and stability. The solution algorithms problems are implemented in a computer simulator made in C++ language, the same that was used to generate the results of the reference work. (author)

  14. Study of spectral response of a neutron filter. Design of a method to adjust spectra

    International Nuclear Information System (INIS)

    Colomb-Dolci, F.

    1999-02-01

    The first part of this thesis describes an experimental method which intends to determine a neutron spectrum in the epithermal range [1 eV -10 keV]. Based on measurements of reaction rates provided by activation foils, it gives flux level in each energy range corresponding to each probe. This method can be used in any reactor location or in a neutron beam. It can determine scepter on eight energy groups, five groups in the epithermal range. The second part of this thesis presents a study of an epithermal neutron beam design, in the frame of Neutron Capture Therapy. A beam tube was specially built to test filters made up of different materials. Its geometry was designed to favour epithermal neutron crossing and to cut thermal and fast neutrons. A code scheme was validated to simulate the device response with a Monte Carlo code. Measurements were made at ISIS reactor and experimental spectra were compared to calculated ones. This validated code scheme was used to simulate different materials usable as shields in the tube. A study of these shields is presented at the end of this thesis. (author)

  15. Gamma-ray production cross sections for 0.9 to 20 MeV neutron interactions with 10B

    International Nuclear Information System (INIS)

    Bywater, R.L. Jr.

    1986-09-01

    Gamma-ray spectral data previously obtained at the 20-meter station of the Oak Ridge Electron Linear Accelerator flight-path 8 were studied to determine cross sections for 0.9- to 20-MeV neutron interactions with 10 B. Data reduction techniques, including those for determination of incident neutron fluences as well as those to compensate for Doppler-broadened gamma-ray-detection responses, are given in some detail in this report. 9 refs., 4 figs., 2 tabs

  16. Influence of neutron scattering and source extent on the measurement of neutron energy spectra at ASDEX

    International Nuclear Information System (INIS)

    Huebner, K.; Baetzner, R.; Roos, M.; Robouch, B.V.; Ingrosso, L.; Wurz, H.

    1987-08-01

    The problem of nuclear emulsion measurements at ASDEX is considered. Besides the application of the VINIA-3DAMC software, this needs a description of the plasma neutron source, a model of the ASDEX structure, and calculation of the response of the nuclear emulsion to the incoming spectral neutron fluence. The latter is essential for comparing the numerical results with measurements at ASDEX. To treat this part, the NEPMC software was developed. The aim of the present work is to demonstrate the feasibility, reliability and usefulness of the method. Therefore simplified treatments for the ASDEX model, the plasma neutron source and the track statistics in the NEPMC software were used. Such calculations are of interest not only for nuclear emulsion measurements as well as any other neutron diagnostics, but also for all problems of neutron shielding for other diagnostics. (orig./GG)

  17. Review of the Palisades pressure vessel accumulated fluence estimate and of the least squares methodology employed

    International Nuclear Information System (INIS)

    Griffin, P.J.

    1998-05-01

    This report provides a review of the Palisades submittal to the Nuclear Regulatory Commission requesting endorsement of their accumulated neutron fluence estimates based on a least squares adjustment methodology. This review highlights some minor issues in the applied methodology and provides some recommendations for future work. The overall conclusion is that the Palisades fluence estimation methodology provides a reasonable approach to a open-quotes best estimateclose quotes of the accumulated pressure vessel neutron fluence and is consistent with the state-of-the-art analysis as detailed in community consensus ASTM standards

  18. Review of the Palisades pressure vessel accumulated fluence estimate and of the least squares methodology employed

    Energy Technology Data Exchange (ETDEWEB)

    Griffin, P.J.

    1998-05-01

    This report provides a review of the Palisades submittal to the Nuclear Regulatory Commission requesting endorsement of their accumulated neutron fluence estimates based on a least squares adjustment methodology. This review highlights some minor issues in the applied methodology and provides some recommendations for future work. The overall conclusion is that the Palisades fluence estimation methodology provides a reasonable approach to a {open_quotes}best estimate{close_quotes} of the accumulated pressure vessel neutron fluence and is consistent with the state-of-the-art analysis as detailed in community consensus ASTM standards.

  19. Special neutron measurement results from the spectral positions of the Juelich FKS steel irradiation capsules

    International Nuclear Information System (INIS)

    Schneider, W.; Kuepper, H.; Pott, G.; Borchardt, G.; Segelhorst, G.; Thoene, L.; Weise, L.

    1986-10-01

    For the German project 'Forschungsvorhaben Komponentensicherheit' (FKS, i.e., Structural Integrity of Components) steel specimen irradiations have been carried out in the Juelich Merlin-type reactor (FRJ-1). The neutron monitoring to these irradiations is described in a German report (Juel-2087). In this context, some special considerations and results are given here, i.e., an experimental investigation of the fast neutron spectrum variation over a thick steel plate (in a special dosimetry test experiment); a comparison of the outcome of this investigation with the results from other FKS participants; and finally, the evaluation of the neutron exposure expressed in displacements per atom (dpa) in the centre of that steel plate. (orig.)

  20. Fluorescence spectra of Rhodamine 6G for high fluence excitation laser radiation

    CERN Document Server

    Hung, J; Olaizola, A M

    2003-01-01

    Fluorescence spectral changes of Rhodamine 6G in ethanol and glycerol solutions and deposited as a film on a silica surface have been studied using a wide range of pumping field fluence at 532 nm at room temperature. Blue shift of the fluorescence spectra and fluorescence quenching of the dye molecule in solution are observed at high excitation fluence values. Such effects are not reported for the film sample. The effects are interpreted as the result of population redistribution in the solute-solvent molecular system induced by the high fluence field and the fluence dependence of the radiationless decay mechanism.

  1. Sync transmission method and apparatus for high frequency pulsed neutron spectral analysis systems

    International Nuclear Information System (INIS)

    Culver, R.B.

    1981-01-01

    An improved synchronization system was developed for high-frequency pulsed-neutron gamma ray well-logging which extends the upper limit of the usable source pulsing frequency. A clock is used to pulse the neutron generator at a given frequency and a scaler generates scaled-down sync pulses at a lower frequency. Radiation from the formations surrounding the borehole is detected and electrical signals related functionally to the radiation are generated. The scaled-down sync pulses and electrical signals are transmitted to the earth's surface via a seven conductor well logging cable. (DN)

  2. Looking for spectral changes occurring during storage of ultra-cold neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Steyerl, A; Malik, S S [Rhode Island Univ., Kingston, RI (United States); Geltenbort, P [Institut Max von Laue - Paul Langevin (ILL), 38 -Grenoble (France)

    1997-04-01

    It seems that the spectrum of ultra-cold neutrons does change. The measured data indicate with 5{sigma} reliability, that a small heating by about 2{center_dot}10{sup -10} eV ({approx} 2 mm of rise height against the earth`s gravity) occurred during the initial {approx} 10{sup 3} wall reflections, and no change thereafter. The reason of this effect is searched for. (author). 3 refs.

  3. Importance of the neutron spectrum for determination of radiation damage

    International Nuclear Information System (INIS)

    Hehn, G.; Stiller, P.; Mattes, M.

    1977-01-01

    Since the radiation effects of neutrons depend strongly on the neutron energy, the correlation between the induced damage and the fluence of the fast neutrons shows appreciable disadvantages. The measured values of changes in material properties resulted in large differences for the same fast neutron fluence, being partly due to different neutron spectra. The uncertainties in damage data led to strong overdesign of important structural components. Different neutron environment at surveillance sample position may give an underestimation of the embrittlement in the reactor pressure vessel, which has to be avoided. The application of damage functions combined with accurately calculated neutron spectra, promise to be a reasonable solution. The damage function has the advantage of a phenomenological quantity that all spectral effects are included. But the correlation quantity has to be determined of high experimental costs. Therefore approximations of its energy distributions are very important. For the keV energy region the kerma function is reasonably good. For the MeV energy region a higher effort is needed to calculate the displacement cross section. The same holds for the low energy part. In all three parts the formation of stable material property levels may vary, so that the final correlation can be determined only by measurements of material properties in different neutron spectra. In material samples the spectra distribution of the displacement production rate was determined at different local positions outside the reactor core of a PWR and a fast breeder showing the most important energy regions of both reactors. (orig.) [de

  4. Inelastic neutron scattering and spectral measurements of advanced cold moderator materials

    International Nuclear Information System (INIS)

    Conrad, H.; Prager, M.; Nuenighoff, K.; Pohl, C.; Kuhs, W.F.

    2004-01-01

    Inelastic neutron scattering with emphasis on energetically low lying modes as well as cold neutron leakage measurements have been performed on four prospective advanced cold moderator materials. Employing the time-of-flight instrument SV29 at the Juelich FRJ-2 reactor, spectra have been obtained from synthetic methane clathrate, tetrahydro-furane (THF) clathrate, 1,3,5-trimethyl-benzene (mesitylene) and light water ice at several temperatures between 2 K and 70 K. Clearly separated excitations at energy transfers of ±1 meV, +2 meV and +3 meV have been observed with synthetic methane clathrate. In mesitylene a wealth of low lying excitations have been observed. In the quenched phase we found lines at 4.7, 7.2, 9.6, 13.6, 15.4, 18.4, 19.0, 23.0, 29.5 and 34.3 meV, respectively. In the annealed phase, we observed significant shifts with the majority of lines. The lowest lying lines now are located at 7.0, 8.5 and 10.5 meV, respectively. In hexagonal ice at T=2 K up to now unreported low lying energy levels were found at energy transfers of 1.8 meV and 2.8 meV. An additional line at about 10 meV could be detected in THF clathrate. Mesitylene, synthetic methane clathrate and water ice, all at T=20 K, have been tested as moderators at the Juelich spallation mock-up JESSICA. The expected gain in neutron leakage current at energies around 2 meV as compared to conventional liquid hydrogen moderators has been observed for methane clathrate and mesitylene. (orig.)

  5. Irradiation induced creep in graphite with respect to the flux effect and the high fluence behaviour

    International Nuclear Information System (INIS)

    Cundy, M.R.

    1984-01-01

    In accelerated irradiation creep tests, performed in the HFR Petten, in a fast neutron flux of about 2x10 4 cm -2 s -1 and at temperatures of 300 and 500 0 C, a fast neutron fluence in excess of 20x10 21 cm -2 (EDN) has been attained so far. As a supplement to this, an analogous creep test was conducted in a fast neutron flux lower by a factor of four which is more typical for the service conditions in a HTR, with a maximum fast fluence of only 4x10 21 cm -2 (EDN). This experiment was aimed at answering the question if, for equal fast fluence, enhanced irradiation creep and Wigner dimensional change would take place in a reduced fast neutron flux. This problem has more generally been addressed to as the ''flux effect'' or the ''equivalent temperature concept''. (orig./IHOE)

  6. Application of damage functions to CTR component fluence limit predictions

    International Nuclear Information System (INIS)

    Simons, R.L.; Doran, D.G.

    1975-01-01

    Material behavior observed under irradiation conditions in test reactors is not always directly applicable to the design of reactor components such as CTR first wall because of differences in the damage effectiveness of test reactor and service spectra. The interpolation and, under some conditions, extrapolation of material property change data from test conditions to different neutron spectra in service conditions can be accomplished using semi-empirical damage functions. The derivation and application of damage functions to CTR conditions is reviewed. Since limited amounts of data are available for applications to CTR design spectra, considerable attention is placed on the effectiveness of various available and proposed neutron sources in determining a damage function and subsequent fluence limit prediction. Neutron sources included in this study were EBR-II, HIFR, LAMPF (Be and Cu targets), high energy deuterons incident on Be (D-Be), and 14 MeV neutrons (D-T)

  7. Field calibration of PADC track etch detectors for local neutron dosimetry in man using different radiation qualities

    Energy Technology Data Exchange (ETDEWEB)

    Haelg, Roger A., E-mail: rhaelg@phys.ethz.ch [Institute for Radiotherapy, Radiotherapie Hirslanden AG, Hirslanden Medical Center, Rain 34, CH-5000 Aarau (Switzerland); Besserer, Juergen [Institute for Radiotherapy, Radiotherapie Hirslanden AG, Hirslanden Medical Center, Rain 34, CH-5000 Aarau (Switzerland); Boschung, Markus; Mayer, Sabine [Division for Radiation Safety and Security, Paul Scherrer Institut, CH-5232 Villigen (Switzerland); Clasie, Benjamin [Department of Radiation Oncology, Massachusetts General Hospital, 30 Fruit Street, Boston, MA 02114 (United States); Kry, Stephen F. [Department of Radiation Physics, The University of Texas M.D. Anderson Cancer Center, 1515 Holcombe Blvd., Houston, TX 77030 (United States); Schneider, Uwe [Institute for Radiotherapy, Radiotherapie Hirslanden AG, Hirslanden Medical Center, Rain 34, CH-5000 Aarau (Switzerland); Vetsuisse Faculty, University of Zurich, Winterthurerstrasse 204, CH-8057 Zurich (Switzerland)

    2012-12-01

    In order to quantify the dose from neutrons to a patient for contemporary radiation treatment techniques, measurements inside phantoms, representing the patient, are necessary. Published reports on neutron dose measurements cover measurements performed free in air or on the surface of phantoms and the doses are expressed in terms of personal dose equivalent or ambient dose equivalent. This study focuses on measurements of local neutron doses inside a radiotherapy phantom and presents a field calibration procedure for PADC track etch detectors. An initial absolute calibration factor in terms of H{sub p}(10) for personal dosimetry is converted into neutron dose equivalent and additional calibration factors are derived to account for the spectral changes in the neutron fluence for different radiation therapy beam qualities and depths in the phantom. The neutron spectra used for the calculation of the calibration factors are determined in different depths by Monte Carlo simulations for the investigated radiation qualities. These spectra are used together with the energy dependent response function of the PADC detectors to account for the spectral changes in the neutron fluence. The resulting total calibration factors are 0.76 for a photon beam (in- and out-of-field), 1.00 (in-field) and 0.84 (out-of-field) for an active proton beam and 1.05 (in-field) and 0.91 (out-of-field) for a passive proton beam, respectively. The uncertainty for neutron dose measurements using this field calibration method is less than 40%. The extended calibration procedure presented in this work showed that it is possible to use PADC track etch detectors for measurements of local neutron dose equivalent inside anthropomorphic phantoms by accounting for spectral changes in the neutron fluence.

  8. Neutron Fluence, Dosimetry and Damage Response Determination in In-Core/Ex-Core Components of the VENUS CEN/SCK LWR Using 3-D Monte Carlo Simulations: NEA's VENUS-3 Benchmark

    International Nuclear Information System (INIS)

    Perlado, J. Manuel; Marian, Jaime; Sanz, Jesus Garcia

    2000-01-01

    Validating state-of-the-art methods used to predict fluence exposure to reactor pressure vessels (RPVs) has become an important issue in identifying the sources of uncertainty in the estimated RPV fluence for pressurized water reactors. This is a very important aspect in evaluating irradiation damage leading to the hardening and embrittlement of such structural components. One of the major benchmark experiments carried out to test three-dimensional methodologies is the VENUS-3 Benchmark Experiment in which three-dimensional Monte Carlo and S n codes have proved more efficient than synthesis methods. At the Instituto de Fusion Nuclear (DENIM) at the Universidad Politecnica de Madrid, a detailed full three-dimensional model of the Venus Critical Facility has been developed making use of the Monte Carlo transport code MCNP4B. The problem geometry and source modeling are described, and results, including calculated versus experimental (C/E) ratios as well as additional studies, are presented. Evidence was found that the great majority of C/E values fell within the 10% tolerance and most within 5%. Tolerance limits are discussed on the basis of evaluated data library and fission spectra sensitivity, where a value ranging between 10 to 15% should be accepted. Also, a calculation of the atomic displacement rate has been carried out in various locations throughout the reactor, finding that values of 0.0001 displacements per atom in external components such as the core barrel are representative of this type of reactor during a 30-yr time span

  9. Microstructural interpretation of the fluence and temperature dependence of the mechanical properties of irradiated AISI 316

    International Nuclear Information System (INIS)

    Johnson, G.D.; Garner, F.A.; Brager, H.R.; Fish, R.L.

    1980-01-01

    The effects of neutron irradiation on the mechanical properties of annealed and 20% cold-worked AISI 316 irradiated in EBR-II were determined for the temperature regime of 370 to 760 0 C for fluences up to 8.4 x 10 22 n/cm 2 (E > 0.1 MeV). At irradiation temperatures below about 500 0 C, both annealed and cold-worked material exhibit a substantial increase in the flow stress with increasing fluence. Furthermore, both materials eventually exhibit the same flow stress, which is independent of fluence. At temperatures in the range of 538 to 650 0 C, the cold-worked material exhibits a softening with increasing fluence. Annealed AISI 316 in this temperature regime exhibits hardening and at a fluence of 2 to 3 x 10 22 n/cm 2 (E > 0.1 MeV) reaches the same value of flow stress as the cold-worked material

  10. Fast reactor fluence dosimetry. Technical progress report, January--November 1976

    International Nuclear Information System (INIS)

    1976-01-01

    The objectives of this task are to: (1) develop and demonstrate the use of 10 B and 6 Li helium accumulation fluence monitors (HAFM's) as a reliable and accurate method of measuring reactor neutron fluence; (2) develop and apply an expanded set of HAFM's which will provide fluence responses in different but overlapping neutron energy ranges; (3) identify, through the precise measurement of spectrum-integrated helium production cross sections, those elements which produce significant helium when used individually or as components of advanced alloys in FTR and LMFBR neutron environments, so that their use might be eliminated, minimized, or controlled; (4) use this information to predict, with confidence, the helium production rate for any alloy or material considered for fast reactor use, and (5) maintain a centralized helium measurements laboratory available to the research community, and upgrade the sample throughput capacity to handle FTR dosimetry requirements

  11. NEW OPTICAL/ULTRAVIOLET COUNTERPARTS AND THE SPECTRAL ENERGY DISTRIBUTIONS OF NEARBY, THERMALLY EMITTING, ISOLATED NEUTRON STARS

    International Nuclear Information System (INIS)

    Kaplan, D. L.; Kamble, A.; Van Kerkwijk, M. H.; Ho, W. C. G.

    2011-01-01

    We present Hubble Space Telescope optical and ultraviolet photometry for five nearby, thermally emitting neutron stars. With these measurements, all seven such objects have confirmed optical and ultraviolet counterparts. Combining our data with archival space-based photometry, we present spectral energy distributions for all sources and measure the 'optical excess': the factor by which the measured photometry exceeds that extrapolated from X-ray spectra. We find that the majority have optical and ultraviolet fluxes that are inconsistent with that expected from thermal (Rayleigh-Jeans) emission, exhibiting more flux at longer wavelengths. We also find that most objects have optical excesses between 5 and 12, but that one object (RX J2143.0+0654) exceeds the X-ray extrapolation by a factor of more than 50 at 5000 A, and that this is robust to uncertainties in the X-ray spectra and absorption. We consider explanations for this ranging from atmospheric effects, magnetospheric emission, and resonant scattering, but find that none is satisfactory.

  12. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    International Nuclear Information System (INIS)

    Johnson, J.O.; Miller, L.F.; Kam, F.B.K.

    1981-05-01

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with 58 Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR

  13. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O.; Miller, L.F.; Kam, F.B.K.

    1981-05-01

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with /sup 58/Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR.

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

    International Nuclear Information System (INIS)

    Nunes, Beatriz Guimaraes

    2012-01-01

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

  15. Effect of non-stationary accretion on spectral state transitions: An example of a persistent neutron star LMXB 4U1636–536

    Science.gov (United States)

    Zhang, Hui; Yu, Wen-Fei

    2018-03-01

    Observations of black hole and neutron star X-ray binaries show that the luminosity of the hard-to-soft state transition is usually higher than that of the soft-to-hard state transition, indicating additional parameters other than mass accretion rate are required to interpret spectral state transitions. It has been found in some individual black hole or neutron star soft X-ray transients that the luminosity corresponding to the hard-to-soft state transition is positively correlated with the peak luminosity of the following soft state. In this work, we report the discovery of the same correlation in the single persistent neutron star low mass X-ray binary (LMXB) 4U 1636–536 based on data from the All Sky Monitor (ASM) on board RXTE, the Gas Slit Camera (GSC) on board MAXI and the Burst Alert Telescope (BAT) on board Swift. We also found such a positive correlation holds in this persistent neutron star LMXB in a luminosity range spanning about a factor of four. Our results indicate that non-stationary accretion also plays an important role in driving X-ray spectral state transitions in persistent accreting systems with small accretion flares, which is much less dramatic compared with the bright outbursts seen in many Galactic LMXB transients.

  16. Fluence-to-absorbed-dose conversion coefficients for neutron beams from 0.001 eV to 100 GeV calculated for a set of pregnant female and fetus models

    International Nuclear Information System (INIS)

    Taranenko, Valery; Xu, X George

    2008-01-01

    Protection of fetuses against external neutron exposure is an important task. This paper reports a set of absorbed dose conversion coefficients for fetal and maternal organs for external neutron beams using the RPI-P pregnant female models and the MCNPX code. The newly developed pregnant female models represent an adult female with a fetus including its brain and skeleton at the end of each trimester. The organ masses were adjusted to match the reference values within 1%. For the 3 mm cubic voxel size, the models consist of 10-15 million voxels for 35 organs. External monoenergetic neutron beams of six standard configurations (AP, PA, LLAT, RLAT, ROT and ISO) and source energies 0.001 eV-100 GeV were considered. The results are compared with previous data that are based on simplified anatomical models. The differences in dose depend on source geometry, energy and gestation periods: from 20% up to 140% for the whole fetus, and up to 100% for the fetal brain. Anatomical differences are primarily responsible for the discrepancies in the organ doses. For the first time, the dependence of mother organ doses upon anatomical changes during pregnancy was studied. A maximum of 220% increase in dose was observed for the placenta in the nine months model compared to three months, whereas dose to the pancreas, small and large intestines decreases by 60% for the AP source for the same models. Tabulated dose conversion coefficients for the fetus and 27 maternal organs are provided

  17. Expected Particle Fluences and Performance of the LHCb Trigger Tracker

    CERN Document Server

    Siegler, M; Needham, M; Steinkamp, O

    2004-01-01

    Monte Carlo simulations of the expected 1 MeV-neutron equivalent fluence in the Trigger Tracker (TT) station of the LHCb detector have been used to investigate the effect of radiation damage on the performance of the detector. The build-up of leakage currents and the corresponding increase in electronic noise has been investigated, as well as the effect of bulk damage on the full-depletion voltage of the sensors and the risk of thermal runaway due to the power generated due to the leakage currents.

  18. Simulation of high fluence swelling behavior in technological materials

    International Nuclear Information System (INIS)

    Garner, F.A.; Powell, R.W.; Diamond, S.; Lauritzen, T.; Rowcliffe, A.F.; Sprague, J.A.; Keefer, D.

    1977-06-01

    The U.S. Breeder Reactor Program is employing charged particle irradiation experiments at accelerated displacement rates to simulate neutron-induced microstructural changes in materials of technological interest. Applications of the simulation technique range from the study of fundamental microstructural mechanisms to the development of predictions of the high fluence swelling behavior of candidate alloys for breeder reactor ducts and fuel cladding. An exact equivalence probably cannot be established between all facets of the microstructural evolution which occurs in the disparate charged-particle and neutron environments. To aid in the correlation of data developed in the two environments an assessment has been made of the factors influencing the simulation process. A series of intercorrelation programs and analysis activities have been conducted to identify and explore the relevant phenomena. The factors found to exert substantial influence on the correlation process fall into two categories, one which deals with those variables which are atypical of the neutron environment and one which deals with the additional factors which arise due to the large differences in displacement rate of the two irradiation environments. While the various simulation techniques have been invaluable in determining the basic mechanisms and parametric dependencies of swelling, the potential of these tools in the confident prediction of swelling at high neutron fluence has yet to be realized. The basic problem lies in the inability of the simulation technique to reproduce the early microstructural development in the period that precedes and encompasses the incubation of voids. The concepts of temperature shift and dose equivalency have also been found to be more complicated than previously imagined. Preconditioning of metals in a neutron environment prior to simulation testing is now being employed in order to provide more appropriate starting microstructures

  19. Frequency domain Monte Carlo simulation method for cross power spectral density driven by periodically pulsed spallation neutron source using complex-valued weight Monte Carlo

    International Nuclear Information System (INIS)

    Yamamoto, Toshihiro

    2014-01-01

    Highlights: • The cross power spectral density in ADS has correlated and uncorrelated components. • A frequency domain Monte Carlo method to calculate the uncorrelated one is developed. • The method solves the Fourier transformed transport equation. • The method uses complex-valued weights to solve the equation. • The new method reproduces well the CPSDs calculated with time domain MC method. - Abstract: In an accelerator driven system (ADS), pulsed spallation neutrons are injected at a constant frequency. The cross power spectral density (CPSD), which can be used for monitoring the subcriticality of the ADS, is composed of the correlated and uncorrelated components. The uncorrelated component is described by a series of the Dirac delta functions that occur at the integer multiples of the pulse repetition frequency. In the present paper, a Monte Carlo method to solve the Fourier transformed neutron transport equation with a periodically pulsed neutron source term has been developed to obtain the CPSD in ADSs. Since the Fourier transformed flux is a complex-valued quantity, the Monte Carlo method introduces complex-valued weights to solve the Fourier transformed equation. The Monte Carlo algorithm used in this paper is similar to the one that was developed by the author of this paper to calculate the neutron noise caused by cross section perturbations. The newly-developed Monte Carlo algorithm is benchmarked to the conventional time domain Monte Carlo simulation technique. The CPSDs are obtained both with the newly-developed frequency domain Monte Carlo method and the conventional time domain Monte Carlo method for a one-dimensional infinite slab. The CPSDs obtained with the frequency domain Monte Carlo method agree well with those with the time domain method. The higher order mode effects on the CPSD in an ADS with a periodically pulsed neutron source are discussed

  20. Optical absorption of neutron-irradiated silica fibers

    International Nuclear Information System (INIS)

    Cooke, D.W.; Farnum, E.H.; Bennett, B.L.

    1996-01-01

    Induced-loss spectra of silica-based optical fibers exposed to high (10 23 n-m -2 ) and low (10 21 n-m -2 ) fluences of neutrons at the Los Alamos Spallation Radiation Effects Facility (LASREF) have been measured. Two types of fibers consisting of a pure fused silica core with fluorine-doped (∼4 mole %) cladding were obtained from Fiberguide Industries and used in the as-received condition. Anhydroguide trademark and superguide trademark fibers contained less than 1 ppm, and 600 to 800 ppm of OH, respectively. The data suggest that presently available silica fibers can be used in plasma diagnostics, but the choice and suitability depends upon the spectral region of interest. Low-OH content fibers can be used for diagnostic purposes in the interval ∼800 to 1400 mn if the exposure is to high-fluence neutrons. For low-fluence neutron exposures, the low-OH content fibers are best suited for use in the interval ∼800 to 2000 nm, and the high-OH content fibers are the choice for the interval ∼400 to 800 nm

  1. Fluence to Dose Equivalent Conversion Coefficients for Evaluation of Accelerator Radiation Environments

    International Nuclear Information System (INIS)

    Thomas, Ralph H.; Zeman, Gary H.

    2001-01-01

    The derivation of a set of conversion functions for the expression of neutron fluence measurements in terms of Effective Dose, E, is described. Four functions in analytical form are presented, covering the neutron energy range from 2.5 10-8 to 10+4 MeV, for the interpretation of fluence measurements in the typical irradiation conditions experienced around high-energy proton accelerators such as the Bevatron. For neutron energies below 200 MeV the analytical functions were modeled after the ISO and ROT conversion coefficients in ICRU 57. For neutron energies above 200 MeV, the analytical function was derived from an analysis of recent published data. Sample calculations using either the analytical expressions or the tabulated conversion coefficients from which the analytical expressions are derived show agreement to better than plus/minus 5%

  2. Unexploded Ordnance identification—A gamma-ray spectral analysis method for Carbon, Nitrogen and Oxygen signals following tagged neutron interrogation

    International Nuclear Information System (INIS)

    Mitra, S.; Dioszegi, I.

    2012-01-01

    A novel gamma-ray spectral analysis method has been demonstrated to optimally extract the signals of the signature elements of explosives, carbon (C), nitrogen (N) and oxygen (O) from 57–155 mm projectiles following tagged neutron interrogation with 14 MeV neutrons. The method was implemented on Monte Carlo simulated, synthetic spectra of Unexploded Ordnance (UXO) that contained high explosive fillers (Composition B, TNT or Explosive D) within steel casings of appropriate thicknesses. The analysis technique defined three broad regions-of-interest (ROI) between 4–7.5 MeV of a spectrum and from a system of three equations for the three unknowns namely C, N and O, the maximum counts from each of these elements were extracted. Unlike conventional spectral analysis techniques, the present method included the Compton continuum under a spectrum. For a neutron output of ∼2×10 7 ns −1 and using four 12.7 cm diameter×12.7 cm NaI(Tl) detectors, the C/N and C/O gamma-ray counts ratios of the explosive fillers were vastly different from that of an inert substance like sand. Conversion of the counts ratios to elemental ratios could further discriminate the different types of explosive fillers. The interrogation time was kept at ten minutes for each projectile.

  3. Measuring neutron spectra in radiotherapy using the nested neutron spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Maglieri, Robert, E-mail: robert.maglieri@mail.mcgill.ca; Evans, Michael; Seuntjens, Jan; Kildea, John [Medical Physics Unit, McGill University, Montreal, Quebec H4A 3J1 (Canada); Licea, Angel [Canadian Nuclear Safety Commission, Ottawa, Ontario K1P 5S9 (Canada)

    2015-11-15

    Purpose: Out-of-field neutron doses resulting from photonuclear interactions in the head of a linear accelerator pose an iatrogenic risk to patients and an occupational risk to personnel during radiotherapy. To quantify neutron production, in-room measurements have traditionally been carried out using Bonner sphere systems (BSS) with activation foils and TLDs. In this work, a recently developed active detector, the nested neutron spectrometer (NNS), was tested in radiotherapy bunkers. Methods: The NNS is designed for easy handling and is more practical than the traditional BSS. Operated in current-mode, the problem of pulse pileup due to high dose-rates is overcome by measuring current, similar to an ionization chamber. In a bunker housing a Varian Clinac 21EX, the performance of the NNS was evaluated in terms of reproducibility, linearity, and dose-rate effects. Using a custom maximum-likelihood expectation–maximization algorithm, measured neutron spectra at various locations inside the bunker were then compared to Monte Carlo simulations of an identical setup. In terms of dose, neutron ambient dose equivalents were calculated from the measured spectra and compared to bubble detector neutron dose equivalent measurements. Results: The NNS-measured spectra for neutrons at various locations in a treatment room were found to be consistent with expectations for both relative shape and absolute magnitude. Neutron fluence-rate decreased with distance from the source and the shape of the spectrum changed from a dominant fast neutron peak near the Linac head to a dominant thermal neutron peak in the moderating conditions of the maze. Monte Carlo data and NNS-measured spectra agreed within 30% at all locations except in the maze where the deviation was a maximum of 40%. Neutron ambient dose equivalents calculated from the authors’ measured spectra were consistent (one standard deviation) with bubble detector measurements in the treatment room. Conclusions: The NNS may

  4. Platinum determination by instrumental neutron activation analysis with special reference to the spectral interference of Sc-47 on the platinum indicator nuclide Au-199

    DEFF Research Database (Denmark)

    Alfassi, Z.B.; Probst, T.U.; Rietz, B.

    1998-01-01

    A method of instrumental neutron activation analysis (INAA) is developed for the determination of platinum by the Au-199 daughter of Pt-199 in the presence of the spectral interference from the Sc-47 daughter of Ca-47. The contributions of the Pt and Ca signals to the integral 157-161 keV peak we...... separated by calculating the number of disintegrations due to the Sc-47 from the signal of her parent Ca-47 at 1297.1 keV gamma-ray. The method was used to calculate the trace concentrations of Pt in air samples, collected on filters, (C) 1998 Elsevier Science B.V....

  5. Neutron spectrometry in mixed fields: characterisation of the Ra-1- reactor workplace

    International Nuclear Information System (INIS)

    Gregori, B.; Carelli, J.; Cruzate, J.; Papadopulos, S.

    2006-01-01

    The characterisation of the neutron spectrum of a workplace is an essential dosimetric tool for improving the assessment of the personal equivalent dose of the workers. In addition, if the operational conditions of the facility are well defined, the set of field spectra obtained may be used as a reference for comparing the performance of different type of neutron detectors. Recently, using a neutron spectrometric system based on a set of moderated spheres with 3 He detector, the characterisation of the neutron spectra in workplaces of the Argentine Reactor No. 1 (R.A. -1) has been carried out. The spectrometric system consists of 12 spheres made of the high density polyethylene d mean δ =0.95 g cm 3 , with diameters between 3'' and 12'' and a proportional counter of 3 He, 4 atm of nominal pressure, Centronic trade mark, located in the centre of the spheres. The neutron response matrix was calculated using the M.C.N.P. -I.V.B. code and E.N.D.F./B-VI library in the energy range between thermal neutron and 100 MeV. The neutron spectrum was unfolded using the M.A.X.E.D. unfolding code. The validation of the spectrometric system was performed at Cea-Cadarache (France) with of 252 Cf, Am Be, and 252 Cf + D 2 O sources. Therefore, in this work, the spectral fluence of the field in the selected points of the facility (R.A.-1) has been presented and the ambient dose equivalent, H *(10), and the personal dose equivalent, Hp(10), have been derived from the neutron fluence, applying ICRP-74 recommended fluence to dose conversion factors. The quantities evaluated have uncertainties less than 15%, which is considered good enough for radiation protection requirements. (authors)

  6. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Parma Edward J.

    2016-01-01

    Full Text Available Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity “bucket” environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters.

  7. Neutron metrology in the HFR

    International Nuclear Information System (INIS)

    Voorbraak, W.P.; Freudenreich, W.E.; Stecher-Rasmussen, F.; Verhagen, H.W.

    1991-10-01

    Neutron fluence rate and gamma dose data are presented for the first series of experiments at the filtered HFR beam HB11 at full reactor power. Measurements were performed on two beagle dogs and one cylindrical phantom. The main results for thermal and epithermal fluence rates, physical neutron dose and gamma dose are presented in the tables 1 and 2. (author). 10 refs.; 9 figs.; 8 tabs

  8. Neutron measurements at BRIT/BARC medical cyclotron facility of RMC, Parel

    International Nuclear Information System (INIS)

    Sathian, Deepa; Sathian, V.; Phandnis, U.V.; Soni, P.S.; Mohite, D.Y.

    2005-01-01

    Neutron leakage and its long distance propagation in the atmosphere from the intense neutron facilities such as high energy accelerators like Cyclotron are very important for the shielding design of the facilities and resulting dose reduction to nearby population, because of strong penetrability of high energy neutrons. The neutron interaction cross sections are highly energy dependent, so different methods are adopted for measuring different energy neutrons. The method also depends on the amount of neutron fluence rate expected at the location. When the fluence rate is very high, the foil activation is the best method for the measurement of neutron fluence rate. In foil activation technique an inactive material is activated by neutrons and the activity is measured and correlated to the neutron fluence rate. In this paper, neutron fluence rate measurement using different activation foils at medical cyclotron room of Radiation Medicine Centre (RMC) is discussed. (author)

  9. Calculation of fluence rate distributions in a pre design clinical facility for BNCT at the LFR

    International Nuclear Information System (INIS)

    Peeters, T.T.J.M.; Freudenreich, W.E.

    1995-12-01

    In a previous study [1], it was demonstrated that the creation of a thermal neutron facility for clinical BNCT in the LFR is feasible. Monte Carlo calculations had shown that the neutron fluence rates and gamma dose rates at the detector position of a model representing a first outline of a clinical facility met all requirements that are necessary for clinical BNCT. In order to gain more information about the neutron fluence rates at several positions, a second step is required. Calculations have been performed for the free beam and for a tumour bearing phantom at 5 cm and 10 cm distance from the irradiation window. Due to thermalization and back scattering, the thermal fluence rates in the tumour at 5 and 10 cm distance from the bismuth shield appeared to be approximately twice as high as the thermal fluence rates in the free beam at the corresponding positions of 5 to 6 cm and 10 to 11 cm from the irradiation window. (orig.)

  10. Correlation of mechanical property changes in neutron-irradiated pressure vessel steels on the basis of spectral effects

    International Nuclear Information System (INIS)

    Heinisch, H.L.

    1991-01-01

    Comparisons are made of tensile data on specimens of A212B and A302B pressure vessel steels irradiated at low temperatures (40-90degC) and to low doses (<0.1 dpa) with 14 MeV D-T fusion neutrons in the Rotating Target Neutron Source (RTNS-II), with fission reactor neutrons in the Omega West Reactor (OWR) and the Oak Ridge Research Reactor (ORR), and with the highly thermal spectrum at the pressure vessel surveillance positions of the High Flux Isotope Reactor (HFIR). For each neutron spectrum, damage cross sections are determined for several defect production functions derived from atomistic computer simulations of collision cascades. Displacements per atom (dpa) and the numbers of freely migrating defects are tested as damage correlation parameters for the tensile data. The data from RTNS-II, OWR and ORR correlate fairly well when compared on the basis of dpa, but the data from HFIR show only about one sixth as many dpa are needed to produce the same radiation-induced yield stress changes as in the other neutron spectra. In the HFIR surveillance position a significant fraction of the displacements is produced by recoils resulting from thermal neutron captures. Having energies of about 400 eV, these recoils are much more efficient per unit energy at producing freely migrating defects than the high energy recoils responsible for most of the displacements in the other neutron spectra considered. A significantly better correlation of data from HFIR with those from the other spectra is achieved when the property changes are compared on the basis of the production of freely migrating self-interstitial defects. (orig./MM)

  11. 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.

  12. Catalogue of neutron spectra

    International Nuclear Information System (INIS)

    Buxerolle, M.; Massoutie, M.; Kurdjian, J.

    1987-09-01

    Neutron dosimetry problems have arisen as a result of developments in the applications of nuclear energy. The largest number of possible irradiation situations has been collected: they are presented in the form of a compilation of 44 neutron spectra. Diagrams show the variations of energy fluence and energy fluence weighted by the dose equivalent/fluence conversion factor, with the logarithm of the corresponding energy. The equivalent dose distributions are presented as percentages for the following energy bins: 0.01 eV/0.5 eV/50 keV/1 MeV/5 MeV/15 MeV. The dose equivalent, the mean energy and the effective energy for the dose equivalent for 1 neutron cm -2 are also given [fr

  13. Neutron nuclear physics under the neutron science project

    Energy Technology Data Exchange (ETDEWEB)

    Chiba, Satoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    The concept of fast neutron physics facility in the Neutron Science Research project is described. This facility makes use of an ultra-short proton pulse (width < 1 ns) for fast neutron time-of-flight works. The current design is based on an assumption of the maximum proton current of 100 {mu}A. Available neutron fluence and energy resolution are explained. Some of the research subjects to be performed at this facility are discussed. (author)

  14. Flux and fluence determination using the material scrapings approach

    International Nuclear Information System (INIS)

    Basha, H.S.; Manahan, M.P.

    1992-01-01

    The conventional approach to flux determination is to use high-purity dosimeters to characterize the neutron field. This paper presents an alternative approach called the scraping method. This method consists of taking scraping samples from an in-service component and using this material to measure the specific activity for various reactions. This approach enables the determination of the neutron flux and fluence incident on any component for which small chips of material can be safely obtained. It offers a capability for determining the neutron flux for components such as reactor internals without destructively removing them from service. The scrapings methodology was benchmarked by comparison with the results obtained using conventional dosimetry data from the San Onofre nuclear generation station Unit 2 (SONGS-2). Additionally, since the goal of any reactor physics analysis is to reduce uncertainty to the extent practical, it is important that the best available cross-section library be used. The fast flux calculated-to-experimental (C/E) ratios at the SONGS-297-deg in-vessel surveillance capsule and the REACTOR-X 90-deg ex-vessel dosimetry positions were studied for several cross-section libraries, including BIGLE-80, SAILOR, and ELXSIR. REACTOR-X is a pressurized water reactor power plant currently operating in the US

  15. Simulated workplace neutron fields

    International Nuclear Information System (INIS)

    Lacoste, V.; Taylor, G.; Rottger, S.

    2011-01-01

    The use of simulated workplace neutron fields, which aim at replicating radiation fields at practical workplaces, is an alternative solution for the calibration of neutron dosemeters. They offer more appropriate calibration coefficients when the mean fluence-to-dose equivalent conversion coefficients of the simulated and practical fields are comparable. Intensive Monte Carlo modelling work has become quite indispensable for the design and/or the characterization of the produced mixed neutron/photon fields, and the use of Bonner sphere systems and proton recoil spectrometers is also mandatory for a reliable experimental determination of the neutron fluence energy distribution over the whole energy range. The establishment of a calibration capability with a simulated workplace neutron field is not an easy task; to date only few facilities are available as standard calibration fields. (authors)

  16. Investigation of Response of Several Neutron Surveymeters by a DT Neutron Generator

    International Nuclear Information System (INIS)

    Kim, Sang In; Jang, In Su; Kim, Jang Lyul; Lee, Jung IL; Kim, Bong Hwan

    2012-01-01

    Several neutron measuring devices were tested under the neutron fields characterized with two distinct kinds of thermal and fast neutron spectrum. These neutron fields were constructed by the mixing of both thermal neutron fields and fast neutron fields. The thermal neutron field was constructed using by a graphite pile with eight AmBe neutron sources. The fast neutron field of 14 MeV was made by a DT neutron generator. In order to change the fraction of fast neutron fluence rate in each neutron fields, a neutron generator was placed in the thermal neutron field at 50 cm and 150 cm from the reference position. The polyethylene neutron collimator was used to make moderated 14 MeV neutron field. These neutron spectra were measured by using a Bonner sphere system with an LiI scintillator, and dosimetric quantities delivered to neutron surveymeters were determined from these measurement results.

  17. Calibration of SPND/Rhodium device for mapping the neutron fluence in the IEA-R1 reactor by means of the activation foil method; Calibracao de um dispositivo de mapeamento de fluxo de neutrons - SNPD/Rodio no reator IEA-R1, por meio do metodo de ativacao de folhas

    Energy Technology Data Exchange (ETDEWEB)

    Ricci Filho, Walter; Dias, Mauro S.; Tondin, Julio B.M.; Koskinas, Marina F. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    2002-07-01

    The IEA-R1 reactor has undergone a modernization to increase its operating power to 5 MW, in order to allow a more efficient production of radioisotopes. The objective of this work is to provide the reactor with flux monitoring device using a rhodium Self-Powered Neutron Detector (SPND). The work presents the results obtained with Rhodium-SPND in several irradiation positions inside the reactor core. A calibration procedure has been performed by means of {sup 197} Au activation foils, with and without cadmium cover, in order do measure the thermal and epithermal neutron fluxes. (author)

  18. Deuterium accumulation in tungsten at high fluences

    Energy Technology Data Exchange (ETDEWEB)

    Zibrov, Mikhail [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany); FOM Institute DIFFER, De Zaale 20, 5612 AJ Eindhoven (Netherlands); Balden, Martin; Matej, Matej [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany); Bystrov, Kirill; Morgan, Thomas [FOM Institute DIFFER, De Zaale 20, 5612 AJ Eindhoven (Netherlands)

    2016-07-01

    The data on the deuterium (D) retention in tungsten (W) at high fluences (≥ 10{sup 27} D/m{sup 2}) are scarce and the existing results are contradictory. Since retention in W is known to be flux-dependent, the laboratory experiments addressing this issue should be carried out in reactor-relevant conditions (high fluxes of low-energy ions). In this work the samples made of polycrystalline W were exposed to D plasmas in the linear plasma generator Pilot-PSI at temperatures ranging from 360 K to 1140 K to fluences in the range of 0.3-8.7 x 10{sup 27} D/m{sup 2}. It was observed that at exposure temperatures of 360 K and 580 K the D retention was only slightly dependent on the ion fluence. In addition, the presence of blister-like structures was found after the exposures, and their density and size distributions were also only weakly dependent on the fluence. In the case of exposure at 1140 K no surface modifications of the samples after plasma exposure were detected and the concentrations of retained D were very small. At all temperatures used the total amounts of retained D were smaller compared to those obtained by other researchers at lower ion flux densities, which indicates that the incident ion flux may play an important role in the total D retention in W.

  19. A fluence device for precise radiation dosimetry

    International Nuclear Information System (INIS)

    Arnott, R.G.T.; Peak, M.J.

    1979-01-01

    An instrument is described which has been designed to ensure precise positioning of samples and sensing devices in three dimensions at all times during irradiation procedures. The system, which is both robust and sensitive, overcomes difficulties experienced when slight variations in the positioning of a sample under irradiation results in large changes in fluence. (UK)

  20. Evaluation of fluence to dose equivalent conversion factors for high energy radiations, (1)

    International Nuclear Information System (INIS)

    Sato, Osamu; Uehara, Takashi; Yoshizawa, Nobuaki; Iwai, Satoshi; Tanaka, Shun-ichi.

    1992-09-01

    Computer code system and basic data have been investigated for evaluating fluence to dose equivalent conversion factors for photons and neutrons up to 10 GeV. The present work suggested that the conversion factors would be obtained by incorporating effective quality factors of charged particles into the HERMES (High Energy Radiation Monte Carlo Elaborate System) code system. The effective quality factors for charged particles were calculated on the basis of the Q-L relationships specified in the ICRP Publication-60. (author)

  1. Formulation of the energetic spectral distribution of in pile neutron flux (energies greater than a few hundred electron volts) (1963); Formulation des repartitions spectrales energetiques de flux neutroniques en pile (energies superieures a quelques centaines d'electrons-volts) (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    In the present report an expression is deduced for the spectral distribution of flux of over a few hundred electron volts; it is valid for most cases of thermal neutron reactors. This expression is: {psi}(E) = K [{psi}{sub o}(E) + h {psi}{sub e}(E)] {psi}{sub o}(E) is the so-called 'homogeneous' constituent of the flux; it corresponds approximately to the case of an infinite homogeneous medium; it is of the type: Y(V - E) e{sup (b{radical}}{sup E)}/E + Y(E-V) F E{sup {upsilon}} e{sup -{beta}}{sup E} The parameters V and F are such that {psi}{sub o}(E) and its derivative are continuous at the junction energy V. {psi}{sub e}(E) is the 'heterogeneous' constituent of the flux; it is of the type: E{sup {upsilon}} e{sup -{beta}}{sup E}. The various parameters of {psi}(E) are on the one hand characteristic of the nature of the reactor moderator, and on the other hand determined by a resonant flux measurement and one, or possibly two, measurements using a fast neutron threshold detector. We have been led furthermore to define an expression for the threshold reaction section which is more exact than the conventional transition function. (author) [French] Il est etabli, dans le present rapport, une formulation {psi}(E) des repartitions spectrales de flux au-dessus de quelques centaines d'electron-s volts, valable dans la majeure partie des cas de piles dites a neutrons thermiques. Cette formulation s'exprime: {psi}(E) = K [{psi}{sub o}(E) + h {psi}{sub e}(E)] {psi}{sub o}(E) est la composante dite 'homogene' du flux; elle correspond a peu pres au cas d'un milieu infini homogene; elle est du type: Y(V - E) e{sup (b{radical}}{sup E)}/E + Y(E-V) F E{sup {upsilon}} e{sup -{beta}}{sup E} les parametres V et F sont tels que {psi}{sub o}(E) et sa derivee soient continues a l'energie de jonction V. {psi}{sub e} est la composante dite 'heterogene' du flux ; elle est du type: E{sup {upsilon}} e{sup -{beta}}{sup E}. Les differents parametres de {psi}(E) sont, d'une part

  2. Fast neutron spectroscopy by gas proton-recoil methods at the light water reactor pressure vessel simulator

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1980-10-01

    Fast neutron spectrum measurements were made in a Light Water Reactor (LWR) Pressure Vessel Simulator (PVS) to provide neutron spectral definition required to appropriately perform and interpret neutron dosimetry measurements related to fast neutron damage in LWR-PV steels. Proton-recoil proportional counter methods using hydrogen and methane gas-filled detectors were applied to obtain the proton spectra from which the neutron spectra were derived. Cylindrical and spherical geometry detectors were used to cover the neutron energy range between 50 keV and 2 MeV. Results show that the neutron spectra shift in energy distribution toward lower energy between the front and back of a PVS. The relative neutron flux densities increase in this energy range with increasing thickness of the steel. Neutron spectrum fine structure shapes and changes are observed. These results should assist in the generation of more accurate effective cross sections and fluences for use in LWR-PV fast neutron dosimetry and materials damage analyses

  3. Neutron dosimetry; Dosimetria de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Fratin, Luciano

    1993-12-31

    A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq {sup 241} Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s{sup -1} and 0,5 {sup {mu}}Sv s{sup -1}. A calibrated 50 nSv s{sup -1} thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the {sup 241} Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold {sup 241} Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,{alpha}) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kV{sub pp} cm{sup -1}, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46{+-} 0,09) 10{sup 4} tracks cm{sup -2} mSv{sup -1} for thermal neutrons, (9{+-}3) 10{sup 2} tracks cm{sup -2} mSV{sup -1} for intermediate neutrons and (26{+-}4) tracks cm{sup -2} mSv{sup -1} for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990`s ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is

  4. Neutron dosimetry; Dosimetria de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Fratin, Luciano

    1994-12-31

    A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq {sup 241} Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s{sup -1} and 0,5 {sup {mu}}Sv s{sup -1}. A calibrated 50 nSv s{sup -1} thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the {sup 241} Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold {sup 241} Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,{alpha}) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kV{sub pp} cm{sup -1}, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46{+-} 0,09) 10{sup 4} tracks cm{sup -2} mSv{sup -1} for thermal neutrons, (9{+-}3) 10{sup 2} tracks cm{sup -2} mSV{sup -1} for intermediate neutrons and (26{+-}4) tracks cm{sup -2} mSv{sup -1} for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990`s ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is

  5. Thermal Neutron Spectral and Spatial Distributions in Light-Water-Moderated Uranium Lattices; Distributions Spectrale et Spatiale des Neutrons Thermiques dans des Reseaux a Uranium et a Eau Leger; Spektral'noe i prostranstvennoe raspredelenie teplovykh nejtronov v uranovykh reshetkakh s vodnym zamedlitelem; Distribuciones Espectral y Espacial de los Neutrones Termicos en los Reticulados de Uranio Moderados por Agua Ligera

    Energy Technology Data Exchange (ETDEWEB)

    Hardy, J.; Volpe, J. J.; Klein, D.; Schmidt, E.; Gelbard, E. [Bettis Atomic Power Laboratory, Pittsburgh, PA (United States)

    1964-02-15

    A review is presented of experiments on thermal neutron behaviour in light-water-moderated uranium lattices. The experiments are of two basic types: thermal utilization measurements, and spatial activation distributions of thermal neutrons in uniformly poisoned water media. The experiments have been designed to test the validity of present theoretical models by distinguishing spectral from spatial effects in a reactor environment. Comparisons have been made with theoretical models involving varying approximations to transport theory and the thermal neutron scattering kernel. The first set of experiments are activation measurements of thermal disadvantage factors in a variety of lattice cells of the TRX assembly, an H{sub 2}O-moderated, slightly enriched uranium reactor with cylindrical fuel rods arranged in a hexagonal array. Comparisons of the thermal utilization data were made with a Monte Carlo description. Two kernels, the Radkowsky and the Nelkin, were used. The Radkowsky kernel is a first order approximation which has been universally used in reactor design. The Nelkin kernel is a more recent and more detailed description of the scattering kernel. Since calculated thermal neutron absorption rates are sensitive to the choice of scattering kernel, particularly if the neutron flux spectrum is non-Maxwellian, another set of activation experiments was designed in the TRX assembly to test the scattering kernel more directly. The experiments were performed under conditions in which the spectral distortion was largely caused by a uniform poison. Measurements were done in a 10-cm diameter, H{sub 2}O-filled tube mounted in the TRX lattice. The tube wall was filled with B{sup 10} powder, effectively decoupling the tube interior from the reactor lattice at energies below 10 keV. Relative thermal activation rates and radial shapes of spectrally sensitive detectors were measured in two cases having, respectively, a near Maxwellian (pure H{sub 2}O) and a spectrum strongly

  6. Study on neutron dosimetry in JNC Tokai Works

    Energy Technology Data Exchange (ETDEWEB)

    Tsujimura, Norio [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan). Tokai Works

    2003-03-01

    The author developed the neutron reference calibration fields using a {sup 252}Cf standard source surrounded with PMMA (polymethylmethacrylates) moderators at the Japan Nuclear Cycle Development Institute (JNC), Tokai Works. The moderators are concentric, annular cylinders made of lead-contained PMMA with a thickness of 13.5, 35.0, 59.5 and 77.0mm, and the {sup 252}Cf source is guided to the geometric center of moderators by the pneumatic system. These fields can provide the moderated neutron spectra very similar to those encountered around the globe-boxes of the fabrication process of MOX (PuO{sub 2}-UO{sub 2} mixed oxide) fuel. The neutron energy spectrum at the reference calibration point was evaluated from the calculations by MCNP4B and the measurements by the INS-type Bonner multi-sphere spectrometer and the hydrogen-filled proportional counters. The calculated neutron spectra were in good agreements with the measured ones. These fields were characterized in terms of the neutron fluence rate, spectral composition and ambient dose equivalent rate, and have served for the response-characterization of various neutron survey instruments. (author)

  7. Neutron detector

    Science.gov (United States)

    Stephan, Andrew C [Knoxville, TN; Jardret,; Vincent, D [Powell, TN

    2011-04-05

    A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.

  8. Neutron measurements in the stray field produced by 158 GeV/c lead ion beams

    International Nuclear Information System (INIS)

    Agosteo, S.; Birattari, C.; Foglio Para, A.; Nava, E.; Silari, M.; Ulrici, L.

    1997-01-01

    This paper discusses measurements carried out at CERN in the stray radiation field produced by 158 GeV/c 208 Pb 82+ ions. The purpose was to test and intercompare the response of several detectors, mainly neutron measuring devices, and to determine the neutron spectral fluence as well as the microdosimetric (absorbed dose and dose equivalent) distributions in different locations around the shielding. Both active instruments and passive dosimeters were employed, including different types of Andersson-Braun rem counters, a tissue equivalent proportional counter, a set of superheated drop detectors, a Bonner sphere system and different types of ion chambers. Activation measurements with 12 C plastic scintillators and with 32 S pellets were also performed to assess the neutron yield of high energy lead ions interacting with a thin gold target. The results are compared with previous measurements and with measurements made during proton runs. (author)

  9. Neutron degradation of UV enhanced optical fibers for fusion installation plasma diagnostics

    International Nuclear Information System (INIS)

    Sporea, D.; Vata, I.; Dudu, D.; Danis, Ana

    2004-01-01

    The remote diagnostics of plasmas in fusion installations requires adequate connection links to transfer the measured signals in media subjected to high electromagnetic disturbances. We evaluated the neutron irradiation induced optical absorption in several commercially available optical fibers, as they were assessed for their possible use in fusion installations. Optical fiber samples were subjected to subsequent irradiation with fluences from 6 x 10 11 to 6 x 10 13 n/cm 2 . Significant radiation induced absorption was observed in the 220-260 nm spectral band, mainly for small core diameter optical fibers (200/220 μm), independent of the cladding material used

  10. Calibration of the JET neutron yield monitors using the delayed neutron counting technique

    International Nuclear Information System (INIS)

    van Belle, P.; Jarvis, O.N.; Sadler, G.; de Leeuw, S.; D'Hondt, P.; Pillon, M.

    1990-01-01

    The time-resolved neutron yield is routinely measured on the JET tokamak using a set of fission chambers. At present, the preferred technique is to employ activation reactions to determine the neutron fluence at a well-chosen position and to relate the measured fluence to the total neutron emission by means of neutron transport calculations. The delayed neutron counting method is a particularly convenient method of performing the activation measurement and the fission cross sections are accurately known. This paper outlines the measurement technique as used on JET

  11. Effects of high thermal and high fast fluences on the mechanical properties of type 6061 aluminum in the HFBR

    International Nuclear Information System (INIS)

    Weeks, J.R.; Czajkowski, C.J.; Tichler, P.R.

    1988-01-01

    The High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is an epithermal, externally moderated (by D 2 O) facility designed to produce neutron beams for research. Type 6061 T-6 aluminum was used for the beam tubes, pressure vessel, fuel cladding, and most other components in the high flux area. The HFBR has operated since 1965. The epithermal, external moderation of the HFBR means that materials irradiated in different areas of the facility receive widely different flux spectra. Thus, specimens from a control rod drive follower tube (CRDF) have received 1.5 /times/ 10 22 n/cm 2 (E > 0.1 MeV) and 3.2 /times/ 10 23 n/cm 2 thermal fluence, while those from a vertical thimble flow shroud received 1.9 /times/ 10 23 n/cm 2 (E > 0.1 MeV) and 1.0 /times/ 10 23 n/cm 2 thermal. These numbers correspond to fast to thermal fluence ratios ranging from 0.05 to 1.9. Irradiations are occurring at approximately 333/degree/K. The data indicate that the increase in tensile strength and decrease in ductility result primarily from the thermal fluence, i.e., the transmutation of aluminum to silicon. These effects appear to be saturating at fluences above approximately 1.8 /times/ 10 23 n/cm 2 thermal at values of 90,000 psi (6700 Kg/mm 2 ) and 9%, respectively. The specimens receiving the highest fluence ratios appear to have less increase in tensile strength and less decrease in ductility than specimens with a lower fast to thermal fluence ratio and the same thermal fluence, suggesting a possible beneficial effect of the high energy neutrons in preventing formation of silicon crystallites. 7 refs., 11 figs., 3 tabs

  12. Determination of fluence-to-dose conversion coefficients by means of artificial neural networks

    International Nuclear Information System (INIS)

    Soto B, T. G.; Rivera P, E.; De Leon M, H. A.; Hernandez D, V. M.; Vega C, H. R.; Gallego, E.; Lorente, A.

    2012-10-01

    In this paper is presented an Artificial Neural Network (Ann) that has been designed, trained and validated to determinate the effective dose e, ambient dose equivalent h(10) and personal dose equivalent hp(10,θ) fluence-to-dose conversion coefficients at different positions, having as only input data 7 count rates obtained with a Bonner Sphere Spectrometer (Bss) system. A set of 211 neutron spectra and the fluence-to-dose conversion coefficients published by the International Atomic Energy Agency were used to train and validate the Ann. This set was divided into 2 subsets, one of 181 elements to train the Ann and the remaining 30 to validate it. The Ann was trained using Bss count rates as input data and the fluence-to-dose conversion coefficients as output data. The network was validated and tested with the set of 30 elements that were not used during the training process. Good results were obtained proving that Ann are a good choice for calculating the fluence-to-dose conversion coefficients having as only data the count rates obtained with a Bss. (Author)

  13. Determination of fluence-to-dose conversion coefficients by means of artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Soto B, T. G.; Rivera P, E.; De Leon M, H. A.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Gallego, E.; Lorente, A., E-mail: tzinnia.soto@gmail.com [Universidad Politecnica de Madrid, Departamento de Ingenieria Nuclear, Jose Gutierrez Abascal No. 2, 28006 Madrid (Spain)

    2012-10-15

    In this paper is presented an Artificial Neural Network (Ann) that has been designed, trained and validated to determinate the effective dose e, ambient dose equivalent h(10) and personal dose equivalent hp(10,{theta}) fluence-to-dose conversion coefficients at different positions, having as only input data 7 count rates obtained with a Bonner Sphere Spectrometer (Bss) system. A set of 211 neutron spectra and the fluence-to-dose conversion coefficients published by the International Atomic Energy Agency were used to train and validate the Ann. This set was divided into 2 subsets, one of 181 elements to train the Ann and the remaining 30 to validate it. The Ann was trained using Bss count rates as input data and the fluence-to-dose conversion coefficients as output data. The network was validated and tested with the set of 30 elements that were not used during the training process. Good results were obtained proving that Ann are a good choice for calculating the fluence-to-dose conversion coefficients having as only data the count rates obtained with a Bss. (Author)

  14. Neutron sources and their characteristics

    International Nuclear Information System (INIS)

    McCall, R.C.; Swanson, W.P.

    1979-03-01

    The significant sources of photoneutrons within a linear-accelerator treatment head are identified and absolute estimates of neutron production per treatment dose are given for typical components. It is found that the high-Z materials within the treatment head do not significantly alter the neutron fluence but do substantially reduce the average energy of the transmitted spectrum. Reflection of neutrons from the concrete treatment room contribute to the neutron fluence, but not substantially to the patient integral dose, because of a further reduction in average energy. The ratio of maximum fluence to the treatment dose at the same distance is given as a function of electron energy. This ratio rises with energy to an almost constant value of 2.1 x 10 5 neutrons cm -2 rad -1 at electron energies above about 25 MeV. Measured data obtained at a variety of accelerator installations are presented and compared with these calculations. Reasons for apparent deviations are suggested. Absolute depth-dose and depth-dose-equivalent distributions for realistic neutron spectra that occur at therapy installations are calculated, and a rapid falloff with depth is found. The ratio of neutron integral absorbed dose to leakage photon absorbed dose is estimated to be 0.04 and 0.2 for 14 to 25 MeV incident electron energy, respectively. Possible reasons are given for lesser neutron production from betatrons than from linear accelerators. Possible ways in which neutron production can be reduced are discussed

  15. Results of 1999 Spectral Gamma-Ray and Neutron Moisture Monitoring of Boreholes at Specific Retention Facilities in the 200 East Area, Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    DG Horton; RR Randall

    2000-01-18

    Twenty-eight wells and boreholes in the 200 East Are% Hanford Site, Washington were monitored in 1999. The monitored facilities were past-practice liquid waste disposal facilities and consisted of six cribs and nineteen ''specific retention'' cribs and trenches. Monitoring consisted of spectral gamma-ray and neutron moisture logging. All data are included in Appendix B. The isotopes {sup 137}Cs, {sup 60}Co, {sup 235}U, {sup 238}U, and {sup 154}Eu were identified on spectral gamma logs from boreholes monitoring the PUREX specific retention facilities; the isotopes {sup 137}Cs, {sup 60}Co, {sup 125}Sb, and {sup 154}Eu were identified on the logs from boreholes at the BC Controlled Area cribs and trenches; and {sup 137}Cs, {sup 60}Co, and {sup 125}Sb were, identified on the logs from boreholes at the BX specific retention trenches. Three boreholes in the BC Controlled Area and one at the BX trenches had previous spectral gamma logs available for comparison with 1999 logs. Two of those logs showed that changes in the subsurface distribution of {sup 137}CS and/or {sup 60}Co had occurred since 1992. Although the changes are not great, they do point to continued movement of contaminants in the vadose zone. The logs obtained in 1999 create a larger baseline for comparison with future logs. Numerous historical gross gamma logs exist from most of the boreholes logged. Qualitative comparison of those logs with the 1999 logs show many substantial changes, most of which reflect the decay of deeper short-lived isotopes, such as {sup 106}Ru and {sup 125}Sb, and the much slower decay of shallower and longer-lived isotopes such as {sup 137}Cs. The radionuclides {sup 137}Cs and {sup 60}Co have moved in two boreholes since 1992. Given the amount of movement and the half-lives of the isotopes, it is expected that they will decay to insignificant amounts before reaching groundwater. However, gamma ray logging cannot detect many of the contaminants of interest such

  16. IMRT fluence map editing to control hot and cold spots

    International Nuclear Information System (INIS)

    Taylor Cook, J.; Tobler, Matt; Leavitt, Dennis D.; Watson, Gordon

    2005-01-01

    Manually editing intensity-modulated radiation therapy (IMRT) fluence maps effectively controls hot and cold spots that the IMRT optimization cannot control. Many times, re-optimizing does not reduce the hot spots or increase the cold spots. In fact, re-optimizing only places the hot and cold spots in different locations. Fluence-map editing provides manual control of dose delivery and provides the best treatment plan possible. Several IMRT treatments were planned using the Varian Eclipse planning system. We compare the effects on dose distributions between fluence-map editing and re-optimization, discuss techniques for fluence-map editing, and analyze differences between fluence editing on one beam vs. multiple beams. When editing a beam's fluence map, it is essential to choose a beam that least affects dose to the tumor and critical structures. Editing fluence maps gives an advantage in treatment planning and provides controlled delivery of IMRT dose

  17. High-altitude cosmic ray neutrons: probable source for the high-energy protons of the earth's radiation belts

    International Nuclear Information System (INIS)

    Hajnal, F.; Wilson, J.

    1992-01-01

    'Full Text:' Several High-altitude cosmic-ray neutron measurements were performed by the NASA Ames Laboratory in the mid-to late-1970s using airplanes flying at about 13km altitude along constant geomagnetic latitudes of 20, 44 and 51 degrees north. Bonner spheres and manganese, gold and aluminium foils were used in the measurements. In addition, large moderated BF-3 counters served as normalizing instruments. Data analyses performed at that time did not provide complete and unambiguous spectral information and field intensities. Recently, using our new unfolding methods and codes, and Bonner-sphere response function extensions for higher energies, 'new' neutron spectral intensities were obtained, which show progressive hardening of neutron spectra as a function of increasing geomagnetic latitude, with substantial increases in the energy region iron, 1 0 MeV to 10 GeV. For example, we found that the total neutron fluences at 20 and 51 degrees magnetic north are in the ratio of 1 to 5.2 and the 10 MeV to 10 GeV fluence ratio is 1 to 18. The magnitude of these ratios is quite remarkable. From the new results, the derived absolute neutron energy distribution is of the correct strength and shape for the albedo neutrons to be the main source of the high-energy protons trapped in the Earth's inner radiation belt. In addition, the results, depending on the extrapolation scheme used, indicate that the neutron dose equivalent rate may be as high as 0.1 mSv/h near the geomagnetic north pole and thus a significant contributor to the radiation exposures of pilots, flight attendants and the general public. (author)

  18. Application of LEPRICON methodology to the unfolding of neutron fluxes in the Arkansas Nuclear One-Unit 1 reactor

    International Nuclear Information System (INIS)

    Maerker, R.E.; Broadhead, B.L.; Williams, M.L.

    1985-01-01

    The LEPRICON (Least-squares EPRI CONsolidation) methodology has been gradually developed over the past few years. The system predicts the absolute neutron fluence levels as a function of energy at specified locations within the pressure vessel of an LWR from the analysis of dosimetry measurements performed at some other readily accessible surveillance location(s). LEPRICON is unique in the field of few-group spectral unfolding in that (1) it solves the extrapolation problem necessitated by the ex-situ measurements; (2) it has the capability of simultaneously unfolding a large number of spectral fluences; (3) it has the capability of simultaneously analyzing a series of benchmark experiments, along with measurements performed in an LWR; (4) it provides state-of-the-art methods for calculating the surveillance dosimeter activities and pressure vessel spectral fluences; (5) it incorporates the basic sensitivity and covariance information necessary for estimates of the uncertainties in the original calculated quantities; and (6) it produces adjustments to the calculated quantities with uncertainties that can be significantly reduced from the original values

  19. Vessel fluence evaluation for SMART using DLC-23 and DLC-185 data libraries

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyo Youn; Cho, Byung Oh; Joo, Han Gyu [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    In this report, it was performed the vessel fluence evaluation for SMART using nuclear cross-section libraries of DLC-23/CASK and DLC-185/BUGLE-96 and it was compared with the results. It was shown that the maximum neutron fluences for the inner surface of SMART vessel, which has 60 years lifetime and 90% capacity factor, resulted from using DLC-23/CASK and DLC-185/BUGLE-96 are 2.88x10{sup 16} n/cm{sup 2} and 1.98 x10{sup 16} n/cm{sup 2}, respectively. It is concluded that the fast neutron fluence at the reactor pressure vessel of SMART is far less than 1.0x10{sup 20} n/cm{sup 2} which is specified by the requirement of 10 CFR 50.61 and the SMART has the preservation of reactor vessel integrity throughout the reactor lifetime. Also, it was shown that the result using DLC-23/CASK has conservatism of about 30% comparing with the result using DLC-185/BUGLE-96. 15 refs., 7 figs., 13 tabs. (Author)

  20. Spectral shift reactor

    International Nuclear Information System (INIS)

    Carlson, W.R.; Piplica, E.J.

    1982-01-01

    A spectral shift pressurized water reactor comprising apparatus for inserting and withdrawing water displacer elements having differing neutron absorbing capabilities for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The displacer elements comprise substantially hollow cylindrical low neutron absorbing rods and substantially hollow cylindrical thick walled stainless rods. Since the stainless steel displacer rods have greater neutron absorbing capability, they can effect greater reactivity change per rod. However, by arranging fewer stainless steel displacer rods in a cluster, the reactivity worth of the stainless steel displacer rod cluster can be less than a low neutron absorbing displacer rod cluster. (author)

  1. Isotopic dependence of GCR fluence behind shielding

    International Nuclear Information System (INIS)

    Cucinotta, Francis A.; Wilson, John W.; Saganti, Premkumar; Hu, Xiaodong; Kim, Myung-Hee Y.; Cleghorn, Timothy; Zeitlin, Cary; Tripathi, Ram K.

    2006-01-01

    In this paper we consider the effects of the isotopic composition of the primary galactic cosmic rays (GCR), nuclear fragmentation cross sections, and isotopic-grid on the solution to transport models used for shielding studies. Satellite measurements are used to describe the isotopic composition of the GCR. For the nuclear interaction data-base and transport solution, we use the quantum multiple scattering theory of nuclear fragmentation (QMSFRG) and high-charge and energy (HZETRN) transport code, respectively. The QMSFRG model is shown to accurately describe existing fragmentation data including proper description of the odd-even effects as function of the iso-spin dependence on the projectile nucleus. The principle finding of this study is that large errors (±100%) will occur in the mass-fluence spectra when comparing transport models that use a complete isotopic-grid (∼170 ions) to ones that use a reduced isotopic-grid, for example the 59 ion-grid used in the HZETRN code in the past; however, less significant errors (<+/-20%) occur in the elemental-fluence spectra. Because a complete isotopic-grid is readily handled on small computer workstations and is needed for several applications studying GCR propagation and scattering, it is recommended that they be used for future GCR studies

  2. Neutrons from medical electron accelerators

    International Nuclear Information System (INIS)

    Swanson, W.P.; McCall, R.C.

    1979-06-01

    The significant sources of photoneutrons within a linear-accelerator treatment head are identified and absolute estimates of neutron production per treatment dose are given for typical components. Measured data obtained at a variety of accelerator installations are presented and compared with these calculations. It is found that the high-Z materials within the treatment head do not significantly alter the neutron fluence, but do substantially reduce the average energy of the transmitted spectrum. Reflected neutrons from the concrete treatment room contribute to the neutron fluence, but not substantially to the patient integral dose, because of a further reduction in average energy. Absolute depth-dose distributions for realistic neutron spectra are calculated, and a rapid falloff with depth is found

  3. Comparison of embrittlement trend curves to high fluence surveillance results

    International Nuclear Information System (INIS)

    Bogaert, A.S.; Gerard, R.; Chaouadi, R.

    2011-01-01

    In the regulatory justification of the integrity of the reactor pressure vessels (RPV) for long term operation, use is made of predictive formulas (also called trend curves) to evaluate the RPV embrittlement (expressed in terms of RTNDT shifts) in function of fluence, chemical composition and in some cases temperature, neutron flux or product form. It has been shown recently that some of the existing or proposed trend curves tend to underpredict high dose embrittlement. Due to the scarcity of representative surveillance data at high dose, some test reactor results were used in these evaluations and raise the issue of representativeness of the accelerated test reactor irradiations (dose rate effects). In Belgium the surveillance capsules withdrawal schedule was modified in the nineties in order to obtain results corresponding to 60 years of operation or more with the initial surveillance program. Some of these results are already available and offer a good opportunity to test the validity of the predictive formulas at high dose. In addition, advanced surveillance methods are used in Belgium like the Master Curve, increased tensile tests, and microstructural investigations. These techniques made it possible to show the conservatism of the regulatory approach and to demonstrate increased margins, especially for the first generation units. In this paper the surveillance results are compared to different predictive formulas, as well as to an engineering hardening model developed at SCK.CEN. Generally accepted property-to-property correlations are critically revisited. Conclusions are made on the reliability and applicability of the embrittlement trend curves. (authors)

  4. Light Water Reactor Sustainability Program BWR High-Fluence Material Project: Assessment of the Role of High-Fluence on the Efficiency of HWC Mitigation on SCC Crack Growth Rates

    Energy Technology Data Exchange (ETDEWEB)

    Sebastien Teysseyre

    2014-04-01

    As nuclear power plants age, the increasing neutron fluence experienced by stainless steels components affects the materials resistance to stress corrosion cracking and fracture toughness. The purpose of this report is to identify any new issues that are expected to rise as boiling water reactor power plants reach the end of their initial life and to propose a path forward to study such issues. It has been identified that the efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report summarizes the data available to support this hypothesis and describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. This program plan includes acquisition of irradiated materials, generation of material via irradiation in a test reactor, and description of the test plan. This plan offers three approaches, each with an estimated timetable and budget.

  5. Neutron irradiation results for the LHCb silicon tracker data readout system components

    CERN Document Server

    Vollhardt, A

    2003-01-01

    This note reports irradiation data for different components of the LHCb Silicon Tracker data readout system, which will be exposed to neutron fluences due to their location in the readout link service box on the tracking station frame. The components were part of a neutron irradiation campaign in April 2003 at the Prospero reactor at CEA Valduc (France) and were exposed to fluences 5 to 100 times higher than the expected fluences at the experiment. For all tested components, minor or no influence on device performance was measured. We therefore consider the tested components to be compatible with the expected neutron fluences at the foreseen locations in the LHCb experiment.

  6. Study of neutron fields around an intense neutron generator.

    Science.gov (United States)

    Kicka, L; Machrafi, R; Miller, A

    2017-12-01

    Neutron fields in the vicinity of the newly built neutron facility, at the University of Ontario Institute of Technology (UOIT), have been investigated in a series of Monte Carlo simulations and measurements. The facility hosts a P-385 neutron generator based on a deuterium-deuterium fusion reaction. The neutron fluence at different locations around the neutron generator facility has been simulated using MCNPX 2.7E Monte Carlo particle transport program. To characterize neutron fields, three neutron sources were modeled with distributions corresponding to different incident deuteron energies of 90kV, 110kV, and 130kV. Measurements have been carried out to determine the dose rate at locations adjacent to the generator using bubble detectors (BDs). The neutron intensity was evaluated and the total dose rates corresponding to different applied acceleration potentials were estimated at various locations. Copyright © 2017 Elsevier Ltd. All rights reserved.

  7. Spectral and Timing Nature of the Symbiotic X-Ray Binary 4U 1954+319: The Slowest Rotating Neutron Star in AN X-Ray Binary System

    Science.gov (United States)

    Enoto, Teruaki; Sasano, Makoto; Yamada, Shin'Ya; Tamagawa, Toru; Makishima, Kazuo; Pottschmidt, Katja; Marcu, Diana; Corbet, Robin H. D.; Fuerst, Felix; Wilms, Jorn

    2014-01-01

    The symbiotic X-ray binary (SyXB) 4U 1954+319 is a rare system hosting a peculiar neutron star (NS) and an M-type optical companion. Its approx. 5.4 hr NS spin period is the longest among all known accretion-powered pulsars and exhibited large (is approx. 7%) fluctuations over 8 yr. A spin trend transition was detected with Swift/BAT around an X-ray brightening in 2012. The source was in quiescent and bright states before and after this outburst based on 60 ks Suzaku observations in 2011 and 2012. The observed continuum is well described by a Comptonized model with the addition of a narrow 6.4 keV Fe-K alpha line during the outburst. Spectral similarities to slowly rotating pulsars in high-mass X-ray binaries, its high pulsed fraction (approx. 60%-80%), and the location in the Corbet diagram favor high B-field (approx. greater than 10(exp12) G) over a weak field as in low-mass X-ray binaries. The observed low X-ray luminosity (10(exp33)-10(exp35) erg s(exp-1)), probable wide orbit, and a slow stellar wind of this SyXB make quasi-spherical accretion in the subsonic settling regime a plausible model. Assuming a approx. 10(exp13) G NS, this scheme can explain the approx. 5.4 hr equilibrium rotation without employing the magnetar-like field (approx. 10(exp16) G) required in the disk accretion case. The timescales of multiple irregular flares (approx. 50 s) can also be attributed to the free-fall time from the Alfv´en shell for a approx. 10(exp13) G field. A physical interpretation of SyXBs beyond the canonical binary classifications is discussed.

  8. Thermoelectric neutron dosimetry: a short introduction

    International Nuclear Information System (INIS)

    Mathieu, F.; Meier, R.; Debrue, J.; Leonard, F.; Schubert, W.

    1977-01-01

    The paper gives a short introduction and state-of-the-art account of an unconventional, non destructive neutron dosimetry method based on monitoring the neutron fluence dependent changes of the thermoelectric properties of base metals and alloys. The basic principles are exposed and illustrated with experimental data obtained during an exploratory irradiation in the BR2 reactor

  9. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  10. Vectorial and plane energy fluences - useful concepts in radiation physics

    International Nuclear Information System (INIS)

    Carlsson, C.A.

    1977-06-01

    The vectorial physical quantities describing the radiation field are defined in this report. The use of these quantities is rare in the radiation dosimetry literature since a knowledge of the directions of motion of the ionizing particle is often uninteresting when determining absorbed doses. However the plane energy fluence rate is a useful quantity in cases with plane irradiation geometries. The plane energy fluence rate is closely related to the vectorial energy fluence rate. The backscattering properties of a medium can be expressed in terms either of its albedo or its reflection-coefficient (backscatter-coefficient). These quantities are discussed in order to derive useful relations between the plane energy fluence and the energy fluence at points on an extended plane surface. Examples are also given of erroneous use of energy fluence instead of vectorial or plane energy fluence. The examples are taken from roentgen diagnostic examinations. To prevent further mistakes it could be valuable if the quantities of vectorial and plane fluences were introduced in text books in radiation dosimetry. Awaiting for this, this report may hopefully be useful. (E.R.)

  11. Photoluminescence in large fluence radiation irradiated space silicon solar cells

    Energy Technology Data Exchange (ETDEWEB)

    Hisamatsu, Tadashi; Kawasaki, Osamu; Matsuda, Sumio [National Space Development Agency of Japan, Tsukuba, Ibaraki (Japan). Tsukuba Space Center; Tsukamoto, Kazuyoshi

    1997-03-01

    Photoluminescence spectroscopy measurements were carried out for silicon 50{mu}m BSFR space solar cells irradiated with 1MeV electrons with a fluence exceeding 1 x 10{sup 16} e/cm{sup 2} and 10MeV protons with a fluence exceeding 1 x 10{sup 13} p/cm{sup 2}. The results were compared with the previous result performed in a relative low fluence region, and the radiation-induced defects which cause anomalous degradation of the cell performance in such large fluence regions were discussed. As far as we know, this is the first report which presents the PL measurement results at 4.2K of the large fluence radiation irradiated silicon solar cells. (author)

  12. Comparison of sources of exit fluence variation for IMRT

    International Nuclear Information System (INIS)

    Gardner, Joseph K; Gordon, J James; Wang Song; Siebers, Jeffrey V; Clews, Luke; Greer, Peter B

    2009-01-01

    The fluence exiting a patient during beam delivery can be used as treatment delivery quality assurance, either by direct comparison with expected exit fluences or by backprojection to reconstruct the patient dose. Multiple possible sources of measured exit fluence deviations exist, including changes in the beam delivery and changes in the patient anatomy. The purpose of this work is to compare the deviations caused by these sources. Machine delivery-related variability is measured by acquiring multiple dosimetric portal images (DPIs) of several test fields without a patient/phantom in the field over a time period of 2 months. Patient anatomy-related sources of fluence variability are simulated by computing transmission DPIs for a prostate patient using the same incident fluence for 11 different computed tomography (CT) images of the patient anatomy. The standard deviation (SD) and maximum deviation of the exit fluence, averaged over 5 mm x 5 mm square areas, is calculated for each test set. Machine delivery fluence SDs as large as 1% are observed for a sample patient field and as large as 2.5% for a picket-fence dMLC test field. Simulations indicate that day-to-day patient anatomy variations induce exit fluence SDs as large as 3.5%. The largest observed machine delivery deviations are 4% for the sample patient field and 7% for the picket-fence field, while the largest difference for the patient anatomy-related source is 8.5%. Since daily changes in patient anatomy can result in substantial exit fluence deviations, care should be taken when applying fluence back-projection to ensure that such deviations are properly attributed to their source. (note)

  13. Verification of neutron irradiation on S/G tube materials

    International Nuclear Information System (INIS)

    Kang, Byoung Hwi; Lee, S. K.; Jang, D. Y.; Jo, K. H.

    2010-12-01

    The fluence monitors were fabricated with metal wires of the purity ≥ 99.9%, whose dimensions were 0.1mm diameter, about 3mm length, and around 150-200 μg mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position. The measured neutron fluences were compared to the calculated ones. (Errors ≤ 35%)

  14. Neutron production from 40 GeV/c mixed proton/pion beam on copper, silver and lead targets in the angular range 30-135 deg

    Energy Technology Data Exchange (ETDEWEB)

    Agosteo, S. [Dipartimento di Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milan (Italy); Istituto Nazionale di Fisica Nucleare, Sezione di Milano, via Celoria 16, 20133 Milan (Italy); Birattari, C. [Istituto Nazionale di Fisica Nucleare, Sezione di Milano, via Celoria 16, 20133 Milan (Italy); Universita degli Studi di Milano, Dipartimento di Fisica, via Celoria 16, 20133 Milan (Italy); Dimovasili, E. [CERN, 1211 Geneva 23 (Switzerland); Foglio Para, A. [Dipartimento di Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milan (Italy); Silari, M. [CERN, 1211 Geneva 23 (Switzerland)]. E-mail: marco.silari@cern.ch; Ulrici, L. [CERN, 1211 Geneva 23 (Switzerland); Vincke, H. [CERN, 1211 Geneva 23 (Switzerland)

    2005-02-01

    The neutron emission from 50 mm thick copper, silver and lead targets bombarded by a mixed proton/pion beam with momentum of 40 GeV/c were measured at the CERN Super Proton Synchrotron. The neutron yield and spectral fluence per incident particle on target were measured with an extended range Bonner sphere spectrometer in the angular range 30-135 deg with respect to the beam direction. Monte Carlo simulations with the FLUKA code were performed to provide a priori information for the unfolding of the experimental data. The spectral fluences show two peaks, an isotropic evaporation component centred at 3 MeV and a high-energy peak sitting around 100-150 MeV. The experimental neutron yields are given in four energy bins: <100 keV, 0.1-20 MeV, 20-500 MeV and 0.5-2 GeV. The total yields show a systematic discrepancy of 30-50%, with a peak of 70% at the largest angles, with respect to the results of the Monte Carlo simulations, which it is believed to be mainly due to uncertainties in the beam normalization factor. Analytic expressions are given for the variation of the integral yield as a function of emission angle and of target mass number.

  15. Neutron production from 40 GeV/c mixed proton/pion beam on copper, silver and lead targets in the angular range 30-135 degree

    CERN Document Server

    Agosteo, S; Dimovasili, E; Foglio-Para, A; Silari, M; Ulrici, L; Vincke, H

    2005-01-01

    The neutron emission from 50 mm thick copper, silver and lead targets bombarded by a mixed proton/pion beam with momentum of 40 GeV/c were measured at the CERN Super Proton Synchrotron. The neutron yield and spectral fluence per incident particle on target were measured with an extended range Bonner sphere spectrometer in the angular range 30-135 degree with respect to the beam direction. Monte Carlo simulations with the FLUKA code were performed to provide a priori information for the unfolding of the experimental data. The spectral fluences show two peaks, an isotropic evaporation component centred at 3 MeV and a high-energy peak sitting around 100-150 MeV. The experimental neutron yields are given in four energy bins: less than 100 keV, 0.1-20 MeV, 20-500 MeV and 0.5-2 GeV. The total yields show a systematic discrepancy of 30-50%, with a peak of 70% at the largest angles, with respect to the results of the Monte Carlo simulations, which it is believed to be mainly due to uncertainties in the beam normaliza...

  16. Neutron production from 40 GeV/c mixed proton/pion beam on copper, silver and lead targets in the angular range 30-135 deg

    International Nuclear Information System (INIS)

    Agosteo, S.; Birattari, C.; Dimovasili, E.; Foglio Para, A.; Silari, M.; Ulrici, L.; Vincke, H.

    2005-01-01

    The neutron emission from 50 mm thick copper, silver and lead targets bombarded by a mixed proton/pion beam with momentum of 40 GeV/c were measured at the CERN Super Proton Synchrotron. The neutron yield and spectral fluence per incident particle on target were measured with an extended range Bonner sphere spectrometer in the angular range 30-135 deg with respect to the beam direction. Monte Carlo simulations with the FLUKA code were performed to provide a priori information for the unfolding of the experimental data. The spectral fluences show two peaks, an isotropic evaporation component centred at 3 MeV and a high-energy peak sitting around 100-150 MeV. The experimental neutron yields are given in four energy bins: <100 keV, 0.1-20 MeV, 20-500 MeV and 0.5-2 GeV. The total yields show a systematic discrepancy of 30-50%, with a peak of 70% at the largest angles, with respect to the results of the Monte Carlo simulations, which it is believed to be mainly due to uncertainties in the beam normalization factor. Analytic expressions are given for the variation of the integral yield as a function of emission angle and of target mass number

  17. Neutron production from 40 GeV/c mixed proton/pion beam on copper, silver and lead targets in the angular range 30-135°

    Science.gov (United States)

    Agosteo, S.; Birattari, C.; Dimovasili, E.; Foglio Para, A.; Silari, M.; Ulrici, L.; Vincke, H.

    2005-02-01

    The neutron emission from 50 mm thick copper, silver and lead targets bombarded by a mixed proton/pion beam with momentum of 40 GeV/c were measured at the CERN Super Proton Synchrotron. The neutron yield and spectral fluence per incident particle on target were measured with an extended range Bonner sphere spectrometer in the angular range 30-135° with respect to the beam direction. Monte Carlo simulations with the FLUKA code were performed to provide a priori information for the unfolding of the experimental data. The spectral fluences show two peaks, an isotropic evaporation component centred at 3 MeV and a high-energy peak sitting around 100-150 MeV. The experimental neutron yields are given in four energy bins: <100 keV, 0.1-20 MeV, 20-500 MeV and 0.5-2 GeV. The total yields show a systematic discrepancy of 30-50%, with a peak of 70% at the largest angles, with respect to the results of the Monte Carlo simulations, which it is believed to be mainly due to uncertainties in the beam normalization factor. Analytic expressions are given for the variation of the integral yield as a function of emission angle and of target mass number.

  18. DT fusion neutron irradiation of ORNL magnesium oxide crystals and BNL--LASL superconductor wires

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1978-01-01

    The DT fusion neutron irradiation of two ORNL magnesium oxide crystals and eleven BNL-LASL superconductor wires is described. The sample position and neutron dose record are given. The maximum neutron fluence on any sample was 2.16 x 10 16 neutrons/cm 2

  19. Contribution to time resolved X-ray fluence and differential spectra measurement method improvement in 5-200 KeV range. Application to pulsed emission sources

    International Nuclear Information System (INIS)

    Vie, M.

    1983-09-01

    Two types of sensors have been developed to measure locally the time-resolved fluence and differential energetic spectrum of pulsed X-ray in the energy range 5 to 200 keV. Rise time of these sensors is very short (10 ns) in order to permit time-resolved measurements. Fluence sensors have been developed by putting filters in front of detector in order to make sensor response independent of X-ray energy and proportional to X-ray fluence. The energetic differential spectrum was calculated by way of a method similar to the ROSS method but using filters separated within a pair defining adjacent spectral width. A detailed analysis of uncertainties affecting calculated fluence and spectrum has been done [fr

  20. Determination of the fluence profile in three dimension for the thermal column of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Herrera A, E.; Urena N, F.; Delfin L, A.; Garcia M, T.

    2006-01-01

    In this work the results of the dosimetric properties of the lithium carbonate are presented (detecting), before the thermal neutrons. The process consists on irradiating samples of lithium carbonate in the installation of the thermal column of the TRIGA Mark III reactor, with a controlled period and with time intervals of 20 hours of irradiation. It is necessary to mention that the detectors were placed in different internal positions of the thermal column. With the purpose of being used these results for future studies, like it is the fluence profile in the thermal column. To use the BNCT technique (Boron Neutron Capture Therapy). Which is a binary technique that requires the simultaneous presence of a neutron flux with appropriate energy and a neutron captor (10B), those which interacting to attack to the tumor cells without producing significant damage to the tissues when both agents are separated. (Author)

  1. Fast neutron dosimetry

    International Nuclear Information System (INIS)

    DeLuca, P.M. Jr.; Pearson, D.W.

    1993-01-01

    Research concentrated on three major areas during the last twelve months: (1) investigations of energy fluence and absorbed dose measurements using crystalline and hot pressed TLD materials exposes to ultrasoft beams of photons, (2) fast neutron kerma factor measurements for several important elements as well as NE-213 scintillation material response function determinations at the intense ''white'' source available at the WNR facility at LAMPF, and (3) kerma factor ratio determinations for carbon and oxygen to A-150 tissue equivalent plastic at the clinical fast neutron radiation facility at Harper Hospital, Detroit, MI. Progress summary reports of these efforts are given in this report

  2. Fluence complexity for IMRT field and simplification of IMRT verification

    International Nuclear Information System (INIS)

    Hanushova, Tereza; Vondarchek, Vladimir

    2013-01-01

    Intensity Modulated Radiation Therapy (IMRT) requires dosimetric verification of each patient’s plan, which is time consuming. This work deals with the idea of minimizing the number of fields for control, or even replacing plan verification by machine quality assurance (QA). We propose methods for estimation of fluence complexity in an IMRT field based on dose gradients and investigate the relation between results of gamma analysis and this quantity. If there is a relation, it might be possible to only verify the most complex field of a plan. We determine the average fluence complexity in clinical fields and design a test fluence corresponding to this amount of complexity which might be used in daily QA and potentially replace patient-related verification. Its applicability is assessed in clinical practice. The relation between fluence complexity and results of gamma analysis has been confirmed for plans but not for single fields. There is an agreement between the suggested test fluence and clinical fields in the average gamma parameter. A critical value of average gamma has been specified for the test fluence as a criterion for distinguishing between poorly and well deliverable plans. It will not be possible to only verify the most complex field of a plan but verification of individual plans could be replaced by a morning check of the suggested test fluence, together with a well-established set of QA tests. (Author)

  3. Probability model for worst case solar proton event fluences

    International Nuclear Information System (INIS)

    Xapsos, M.A.; Summers, G.P.; Barth, J.L.; Stassinopoulos, E.G.; Burke, E.A.

    1999-01-01

    The effects that solar proton events have on microelectronics and solar arrays are important considerations for spacecraft in geostationary orbits, polar orbits and on interplanetary missions. A predictive model of worst case solar proton event fluences is presented. It allows the expected worst case event fluence to be calculated for a given confidence level and for periods of time corresponding to space missions. The proton energy range is from >1 to >300 MeV, so that the model is useful for a variety of radiation effects applications. For each proton energy threshold, the maximum entropy principle is used to select the initial distribution of solar proton event fluences. This turns out to be a truncated power law, i.e., a power law for smaller event fluences that smoothly approaches zero at a maximum fluence. The strong agreement of the distribution with satellite data for the last three solar cycles indicates this description captures the essential features of a solar proton event fluence distribution. Extreme value theory is then applied to the initial distribution of events to obtain the model of worst case fluences

  4. Performance of core modifications to reduce the reactor pressure vessel fluence

    International Nuclear Information System (INIS)

    Kiehlmann, H.D.; Lisdat, R.; Sommer, D.

    1997-01-01

    It's often discussed that nuclear power plants (NPP) are designed for an operation of 40 years equivalent to 32 full power years (FPY) assuming a load factor of 0.8. Such fixed plant life times are subjects of US operating licenses but not, as in most other countries, in the Federal Republic of Germany. Here the operating licenses are issued for an indefinite period. However, the German utilities are continuously upgrading their plants to attain a safety level that meets all current requirements. These upgrading measures also include the replacement of bigger components like e.g. the steam generator. The reactor pressure vessel (RPV), however, has a special status. Unlike most other components of a NPP which most likely will be exchanged during its service life a replacement or annealing treatment of the RPV certainly require more efforts to be economically justified. Thus the embrittlement of the RPV has an essential impact on the life time of a NPP. The end-of-life (EOL) RPV material toughness in essential depends on the steel quality and the accumulated neutron fluence. For a given NPP the reduction of the neutron flux at the inner surface of the RPV is the only way to limit its embrittlement. The resulting modifications for the core loadings in combination with the insertion of additional core components like steel elements are described and the impact on core performance and RPV fluence considered. (UK)

  5. Neutron spectra measurements and neutron flux monitoring for radiation damage purposes

    International Nuclear Information System (INIS)

    Osmera, B.; Petr, J.; Racek, J.; Rumler, C.; Turzik, Z.; Franc, L.; Holman, M.; Hogel, J.; Kovarik, K.; Marik, P.; Vespalec, R.; Albert, D.; Hansen, V.; Vogel, W.

    1979-09-01

    Neutron spectra were measured for the TR-0, WWR-S and SR-0 experimental reactors using the recoil proton method, 6 Li spectrometry, scintillation spectrometry and activation detectors in a variety of conditions. Neutron fluence was also measured and calculated. (M.S.)

  6. Benchmarking of multigroup neutron cross sections libraries on neutron transmission through WWER-440 vessel

    International Nuclear Information System (INIS)

    Ilieva, K.; Belousov, S.; Apostolov, T.

    1998-01-01

    The verification of calculated neutron fluence onto the WWER-440/230 pressure vessel is very topical task in particular referring that some of this type of reactors have been operated the major part of its design lifetime. Since the induced activity from the neutron irradiation onto the elements is a simple response of neutron flux the neutron fluence verification usually is done using the measured activity of radionuclides produced during reactor operation. Calculational and experimental results of 54 Mn induced activity of scraps from inner wall of Unit 1 reactor pressure vessel after 18th cycle and detectors irradiated behind the vessel during the 18th cycle of Unit 1 at Kozloduy NPP as well as neutron flux attenuation through the WWER-440/230 pressure vessel are presented. Neutron cross sections libraries generated on the base of ENDF/B-IV and ENDF/B-VI have been used in the calculations. The comparative analysis of evaluated activities and attenuation coefficient demonstrates the better reliability of the neutron fluence calculations by the libraries based on ENDF/B-VI than by ones on ENDF/B-IV. The extreme rarity of data for the activity of scraps from the WWER-440 reactor vessel and its combination with the data for the detectors irradiated behind the vessel makes them especially attractive for verification of calculational methods of neutron fluence onto the WWER-440 vessel with dummy cassettes loading. (author)

  7. Characterization of the neutron sources storage pool of the Neutron Standards Laboratory, using Montecarlo Techniques

    International Nuclear Information System (INIS)

    Campo Blanco, X.

    2015-01-01

    The development of irradiation damage resistant materials is one of the most important open fields in the design of experimental facilities and conceptual nucleoelectric fusion plants. The Neutron Standards Laboratory aims to contribute to this development by allowing the neutron irradiation of materials in its calibration neutron sources storage pool. For this purposes, it is essential to characterize the pool itself in terms of neutron fluence and spectra due to the calibration neutron sources. In this work, the main features of this facility are presented and the characterization of the storage pool is carried out. Finally, an application is shown of the obtained results to the neutron irradiation of material.

  8. Studsvik thermal neutron facility

    International Nuclear Information System (INIS)

    Pettersson, O.A.; Larsson, B.; Grusell, E.; Svensson, P.

    1992-01-01

    The Studsvik thermal neutron facility at the R2-0 reactor originally designed for neutron capture radiography has been modified to permit irradiation of living cells and animals. A hole was drilled in the concrete shielding to provide a cylindrical channel with diameter of 25.3 cm. A shielding water tank serves as an entry holder for cells and animals. The advantage of this modification is that cells and animals can be irradiated at a constant thermal neutron fluence rate of approximately 10 9 n cm -2 s -1 (at 100 kW) without stopping and restarting the reactor. Topographic analysis of boron done by neutron capture autoradiography (NCR) can be irradiated under the same conditions as previously

  9. Method of measuring neutron spectra in JMTR exclusively used for irradiation and their evaluation

    International Nuclear Information System (INIS)

    Sakurai, Kiyoshi

    1983-01-01

    In the core of the Japan Materials Testing Reactor, about 60 capsules are irradiated. These are the material capsules for irradiating reactor materials, the fuel capsules for irradiating reactor fuel, the RI capsules for producing radioisotopes and so on. In the irradiation experiment using a reactor, the information on the neutron fluence is indispensable, and the neutron fluence in the irradiated specimen part is evaluated with a dosimeter or the nuclear calculation for the core of the JMTR. At the time of irradiating reactor materials, the dosimeter Fe-54 (n,p) Mn-54 is generally used for evaluating the neutron fluence more than 1 MeV. In the case of fuel irradiation, the thermal neutron fluence is evaluated with the dosimeter Co-59 (n,γ) Co-60. It is important to examine in detail neutron spectra by both calculation and experiment in the reactors exclusively used for irradiation such as the JMTR. The neutron irradiation field in the JMTR, neutron spectrum measuring experiment, the neutron flux monitors for standardizing data, the measurement of X-ray and gamma ray, neutron guess spectrum, the compilation of neutron cross section for SAND 2, and the unfolding of neutron spectra are reported. The degree of agreement of the neutron fluence more than 1 MeV by measurement and calculation was +- 10 to 20 %. (Kako, I.)

  10. Determination of the neutron and photon dose equivalent at work places in nuclear facilities of Sweden. An SSI - EURADOS comparison exercise. Part 2: Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Bartlett, D. [National Radiological Protection Board, Chilton (United Kingdom); Drake, P. [Vattenfall AB, Vaeroebacka (Sweden); Lindborg, L. [Swedish Radiation Protection Inst., Stockholm (Sweden); Klein, H. [Physikalisch-Technische Bundesanstalt, Braunschweig (Germany); Schmitz, Th. [Forschungszentrum Juelich GmbH, Juelich (Germany); Tichy, M

    1999-06-01

    Various mixed neutron-photon fields at workplaces in the containment of pressurised water reactors and in the vicinity of transport containers with spent fuel elements were investigated with spectrometers and dosimeters. The spectral neutron fluences evaluated from measurements with multisphere systems were recommended to be used for the calculation of dosimetric reference values for comparison with the readings of the dosemeters applied simultaneously. It turned out that most of the moderator based area dosemeters overestimated, while the TEPC systems generally underestimated the ambient dose equivalent (DE) values of the rather soft neutron fields encountered at these workplaces. The discrepancies can, however, be explained on the basis of energy dependent responses of the instruments used. The ambient DE values obtained with recently developed area dosemeters based on superheated drop detectors and with track etch based personal dosemeters on phantoms, however, were in satisfying agreement with the reference data. Sets of personal dosemeters simultaneously irradiated on a phantom allowed to roughly estimate the directional dependence of the neutron fluence. Hence, personal and limiting dose equivalent quantities could also be calculated. The personal and ambient DE values were always conservative estimates of the limiting quantities. Unexpectedly, discrepancies were observed for photon DE data measured with GM counters and TEPC systems. The up to 50 % higher readings of the GM counters may be explained by a considerable contribution of high energy photons to the total photon dose equivalent, but photon spectrometry is necessary for final clarification.

  11. Determination of the neutron and photon dose equivalent at work places in nuclear facilities of Sweden. An SSI - EURADOS comparison exercise. Part 2: Evaluation

    International Nuclear Information System (INIS)

    Bartlett, D.; Drake, P.; Lindborg, L.; Klein, H.; Schmitz, Th.; Tichy, M.

    1999-06-01

    Various mixed neutron-photon fields at workplaces in the containment of pressurised water reactors and in the vicinity of transport containers with spent fuel elements were investigated with spectrometers and dosimeters. The spectral neutron fluences evaluated from measurements with multisphere systems were recommended to be used for the calculation of dosimetric reference values for comparison with the readings of the dosemeters applied simultaneously. It turned out that most of the moderator based area dosemeters overestimated, while the TEPC systems generally underestimated the ambient dose equivalent (DE) values of the rather soft neutron fields encountered at these workplaces. The discrepancies can, however, be explained on the basis of energy dependent responses of the instruments used. The ambient DE values obtained with recently developed area dosemeters based on superheated drop detectors and with track etch based personal dosemeters on phantoms, however, were in satisfying agreement with the reference data. Sets of personal dosemeters simultaneously irradiated on a phantom allowed to roughly estimate the directional dependence of the neutron fluence. Hence, personal and limiting dose equivalent quantities could also be calculated. The personal and ambient DE values were always conservative estimates of the limiting quantities. Unexpectedly, discrepancies were observed for photon DE data measured with GM counters and TEPC systems. The up to 50 % higher readings of the GM counters may be explained by a considerable contribution of high energy photons to the total photon dose equivalent, but photon spectrometry is necessary for final clarification

  12. Neutron-exposure parameters for the fourth HSST series of metallurgical irradiation capsules

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.; Baldwin, C.A.; Fabry, A.

    1982-01-01

    The neutron exposure parameters for the Heavy Section Steel Technology (HSST) Experiments performed at the Oak Ridge National Laboratory (ORNL) can be determined conservatively to +-10% (1sigma) variance. The neutron exposure parameters used for this study were fluence greater than 1 MeV, fluence greater than 0.1 MeV, and displacements per atom (dpa). Measured reaction rates, calculated neutron transport fluxes, and cross sections values were combined in the logarithmic least square adjustment code LSL

  13. A method for neutron dosimetry in ultrahigh flux environments

    International Nuclear Information System (INIS)

    Ougouag, A.M.; Wemple, C.A.; Rogers, J.W.

    1996-01-01

    A method for neutron dosimetry in ultrahigh flux environments is developed, and devices embodying it are proposed and simulated using a Monte Carlo code. The new approach no longer assumes a linear relationship between the fluence and the activity of the nuclides formed by irradiation. It accounts for depletion of the original ''foil'' material and for decay and depletion of the formed nuclides. In facilities where very high fluences are possible, the fluences inferred by activity measurements may be ambiguous. A method for resolving these ambiguities is also proposed and simulated. The new method and proposed devices should make possible the use of materials not traditionally considered desirable for neutron activation dosimetry

  14. New thermal neutron calibration channel at LNMRI/IRD

    Energy Technology Data Exchange (ETDEWEB)

    Astuto, A.; Lopes, R.T., E-mail: achillesbr@gmail.com [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W. [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ/LNMRI), Rio de Janeiro, RJ (Brazil). Lab. Nacional de Metrologia das Radiacoes Ionizantes

    2015-07-01

    A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four {sup 241}Am-Be sources with 596 GBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence (less than 1%) in the central chamber. (author)

  15. New thermal neutron calibration channel at LNMRI/IRD

    International Nuclear Information System (INIS)

    Astuto, A.; Lopes, R.T.; Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W.

    2015-01-01

    A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four 241 Am-Be sources with 596 GBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence (less than 1%) in the central chamber. (author)

  16. Fast and slow neutrons in an 18-MV photon beam from a Philips SL/75-20 linear accelerator

    International Nuclear Information System (INIS)

    Gur, D.; Rosen, J.C.; Bukovitz, A.G.; Gill, A.W.

    1978-01-01

    Fast- and slow-neutron contamination in an 18-MV photon beam from a Philips SL/75-20 linear accelerator has been measured. Aluminum and indium foils were activated to determine fast- and slow-neutron fluence, which were largely independent of field sizes. Measured fast-neutron fluences were typically 13.9 x 10 4 and 4.4 x 10 4 neutrons/cm 2 /rad of x ray inside and 5 cm outside the field, respectively. Slow-neutron fluences, 1.3 x 10 4 neutrons/cm 2 /rad of x ray, remained relatively constant inside and outside the field. The reported results are about three times higher than neutron fluences recently reported with a betatron operated at the same energy

  17. Broadband Spectral Investigations of Magnetar Bursts

    Science.gov (United States)

    Kırmızıbayrak, Demet; Şaşmaz Muş, Sinem; Kaneko, Yuki; Göğüş, Ersin

    2017-09-01

    We present our broadband (2-250 keV) time-averaged spectral analysis of 388 bursts from SGR J1550-5418, SGR 1900+14, and SGR 1806-20 detected with the Rossi X-ray Timing Explorer (RXTE) here and as a database in a companion web-catalog. We find that two blackbody functions (BB+BB), the sum of two modified blackbody functions (LB+LB), the sum of a blackbody function and a power-law function (BB+PO), and a power law with a high-energy exponential cutoff (COMPT) all provide acceptable fits at similar levels. We performed numerical simulations to constrain the best fitting model for each burst spectrum and found that 67.6% of burst spectra with well-constrained parameters are better described by the Comptonized model. We also found that 64.7% of these burst spectra are better described with the LB+LB model, which is employed in the spectral analysis of a soft gamma repeater (SGR) for the first time here, than with the BB+BB and BB+PO models. We found a significant positive lower bound trend on photon index, suggesting a decreasing upper bound on hardness, with respect to total flux and fluence. We compare this result with bursts observed from SGR and AXP (anomalous X-ray pulsar) sources and suggest that the relationship is a distinctive characteristic between the two. We confirm a significant anticorrelation between burst emission area and blackbody temperature, and find that it varies between the hot and cool blackbody temperatures differently than previously discussed. We expand on the interpretation of our results in the framework of a strongly magnetized neutron star.

  18. Broadband Spectral Investigations of Magnetar Bursts

    Energy Technology Data Exchange (ETDEWEB)

    Kırmızıbayrak, Demet; Şaşmaz Muş, Sinem; Kaneko, Yuki; Göğüş, Ersin, E-mail: demetk@sabanciuniv.edu [Faculty of Engineering and Natural Sciences, Sabancı University, Orhanlı Tuzla, Istanbul 34956 (Turkey)

    2017-09-01

    We present our broadband (2–250 keV) time-averaged spectral analysis of 388 bursts from SGR J1550−5418, SGR 1900+14, and SGR 1806−20 detected with the Rossi X-ray Timing Explorer ( RXTE ) here and as a database in a companion web-catalog. We find that two blackbody functions (BB+BB), the sum of two modified blackbody functions (LB+LB), the sum of a blackbody function and a power-law function (BB+PO), and a power law with a high-energy exponential cutoff (COMPT) all provide acceptable fits at similar levels. We performed numerical simulations to constrain the best fitting model for each burst spectrum and found that 67.6% of burst spectra with well-constrained parameters are better described by the Comptonized model. We also found that 64.7% of these burst spectra are better described with the LB+LB model, which is employed in the spectral analysis of a soft gamma repeater (SGR) for the first time here, than with the BB+BB and BB+PO models. We found a significant positive lower bound trend on photon index, suggesting a decreasing upper bound on hardness, with respect to total flux and fluence. We compare this result with bursts observed from SGR and AXP (anomalous X-ray pulsar) sources and suggest that the relationship is a distinctive characteristic between the two. We confirm a significant anticorrelation between burst emission area and blackbody temperature, and find that it varies between the hot and cool blackbody temperatures differently than previously discussed. We expand on the interpretation of our results in the framework of a strongly magnetized neutron star.

  19. Muon Fluence Measurements for Homeland Security Applications

    Energy Technology Data Exchange (ETDEWEB)

    Ankney, Austin S.; Berguson, Timothy J.; Borgardt, James D.; Kouzes, Richard T.

    2010-08-10

    This report focuses on work conducted at Pacific Northwest National Laboratory to better characterize aspects of backgrounds in RPMs deployed for homeland security purposes. Two polyvinyl toluene scintillators were utilized with supporting NIM electronics to measure the muon coincidence rate. Muon spallation is one mechanism by which background neutrons are produced. The measurements performed concentrated on a broad investigation of the dependence of the muon flux on a) variations in solid angle subtended by the detector; b) the detector inclination with the horizontal; c) depth underground; and d) diurnal effects. These tests were conducted inside at Building 318/133, outdoors at Building 331G, and underground at Building 3425 at Pacific Northwest National Laboratory.

  20. Neutron-induced displacement damage analysis (with particular reference to zirconium)

    International Nuclear Information System (INIS)

    Woo, C.H.

    1978-10-01

    Neutron irradiation produces damage in a solid, initially in the form of atomic displacements. As a first step to understanding the effects of irradiation damage in reactor structural materials, information on the initial atomic displacements is necessary. The computer program DISPKAN, based on an extension and generalization of the program RICE, written at ORNL for such calculations, has been developed and installed on the CDC system at CRNL. Using neutron scattering data from ENDF/B files, DISPKAN performs a displacement and PKO analysis on the initial damage caused by neutrons from a given spectrum. The following quantities are calculated: the displacement rate per unit neutron flux, the PKO spectrum, the displacement spectrum, the fraction of PKO's with energy above T, the fraction of displacements produced by PKO's with energy above T, the average PKO energy, the average number of displacements produced per PKO, and the total number of PKO's produced per atom of the solid per unit fluence. The input and output formats of the program are explained. Sample runs are demonstrated. Results for zirconium, exposed to five neutron spectra typically available to experimentalist, are given to illustrate the spectral dependence of the initial displacement events. (author)

  1. Recoil mixing in high-fluence ion implantation

    International Nuclear Information System (INIS)

    Littmark, U.; Hofer, W.O.

    1979-01-01

    The effect of recoil mixing on the collection and depth distribution of implanted projectiles during high-fluence irradiation of a random solid is investigated by model calculations based on a previously published transport theoretical approach to the general problem of recoil mixing. The most pronounced effects are observed in the maximum implantable amount of projectiles and in the critical fluence for saturation. Both values are significantly increased by recoil mixing. (Auth.)

  2. Development of the neutron reference calibration field using a {sup 252}Cf standard source surrounded with PMMA moderators

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, T.; Kanai, K.; Tsujimura, N. [Japan Nuclear Cycle Development Institute, Ibaraki-ken (Japan)

    2002-07-01

    The authors developed the neutron reference calibration fields using a {sup 252} Cf standard source surrounded with PMMA moderators at the Japan Nuclear Cycle Development (JNC), Tokai Works. The moderators are co-axial, hollow cylinders made of lead-contained PMMA with a thickness of 13.5, 35.0, 59.5 and 77.0mm, and the {sup 252} Cf source is guided to the geometric center of moderators by the pneumatic system. These fields can provide the moderated neutron spectra very similar to those encountered around the globe-boxes of the fabrication process of MOX (PuO{sub 2}-UO{sub 2} mixed oxide) fuel. The neutron energy spectrum at the reference calibration point was evaluated from the calculations by MCNP4B and the measurements by the INS-type Bonner multi-sphere spectrometer and the hydrogen-filled proportional counters. The calculated neutron spectra were in good agreements with the measured ones. These fields were characterized in terms of the neutron fluence rate, spectral composition and ambient dose equivalent rate, and have served for the response-characterization of various neutron survey instruments.

  3. Development of the neutron reference calibration field using a 252Cf standard source surrounded with PMMA moderators

    International Nuclear Information System (INIS)

    Yoshida, T.; Kanai, K.; Tsujimura, N.

    2002-01-01

    The authors developed the neutron reference calibration fields using a 252 Cf standard source surrounded with PMMA moderators at the Japan Nuclear Cycle Development (JNC), Tokai Works. The moderators are co-axial, hollow cylinders made of lead-contained PMMA with a thickness of 13.5, 35.0, 59.5 and 77.0mm, and the 252 Cf source is guided to the geometric center of moderators by the pneumatic system. These fields can provide the moderated neutron spectra very similar to those encountered around the globe-boxes of the fabrication process of MOX (PuO 2 -UO 2 mixed oxide) fuel. The neutron energy spectrum at the reference calibration point was evaluated from the calculations by MCNP4B and the measurements by the INS-type Bonner multi-sphere spectrometer and the hydrogen-filled proportional counters. The calculated neutron spectra were in good agreements with the measured ones. These fields were characterized in terms of the neutron fluence rate, spectral composition and ambient dose equivalent rate, and have served for the response-characterization of various neutron survey instruments

  4. Production of neutrons up to 18 MeV in high-intensity, short-pulse laser matter interactions

    Energy Technology Data Exchange (ETDEWEB)

    Higginson, D. P. [Mechanical and Aerospace Engineering, University of California-San Diego, La Jolla, California 92093 (United States); Lawrence Livermore National Laboratory, Livermore, California 94440 (United States); McNaney, J. M.; Swift, D. C.; Mackinnon, A. J.; Patel, P. K. [Lawrence Livermore National Laboratory, Livermore, California 94440 (United States); Petrov, G. M.; Davis, J. [Naval Research Laboratory, Plasma Physics Division, Washington, DC 20375 (United States); Frenje, J. A. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Jarrott, L. C.; Tynan, G.; Beg, F. N. [Mechanical and Aerospace Engineering, University of California-San Diego, La Jolla, California 92093 (United States); Kodama, R.; Nakamura, H. [Institute of Laser Engineering, Osaka University, 2-5 Yamada-oka, Suita, Osaka 454-0871 (Japan); Lancaster, K. L. [STFC Rutherford Appleton Laboratory, Chilton, Oxon OX11OQX (United Kingdom)

    2011-10-15

    The generation of high-energy neutrons using laser-accelerated ions is demonstrated experimentally using the Titan laser with 360 J of laser energy in a 9 ps pulse. In this technique, a short-pulse, high-energy laser accelerates deuterons from a CD{sub 2} foil. These are incident on a LiF foil and subsequently create high energy neutrons through the {sup 7}Li(d,xn) nuclear reaction (Q = 15 MeV). Radiochromic film and a Thomson parabola ion-spectrometer were used to diagnose the laser accelerated deuterons and protons. Conversion efficiency into protons was 0.5%, an order of magnitude greater than into deuterons. Maximum neutron energy was shown to be angularly dependent with up to 18 MeV neutrons observed in the forward direction using neutron time-of-flight spectrometry. Absolutely calibrated CR-39 detected spectrally integrated neutron fluence of up to 8 x 10{sup 8} n sr{sup -1} in the forward direction.

  5. Calibration of neutron yield activation measurements at JET using MCNP and furnace neutron transport codes

    International Nuclear Information System (INIS)

    Pillon, M.; Martone, M.; Verschuur, K.A.; Jarvis, O.N.; Kaellne, J.

    1989-01-01

    Neutron transport calculations have been performed using fluence ray tracing (FURNACE code) and Monte Carlo particle trajectory sampling methods (MCNP code) in order to determine the neutron fluence and energy distributions at different locations in the JET tokamak. These calculations were used to calibrate the activation measurements used in the determination of the absolute fusion neutron yields from the JET plasma. We present here the neutron activation response coefficients calculated for three different materials. Comparison of the MCNP and FURNACE results helps identify the sources of error in these neutron transport calculations. The accuracy of these calculations was tested by comparing the total 2.5 MeV neutron yields derived from the activation measurements with those obtained with calibrated fission chambers; agreement at the ±15% level was demonstrate. (orig.)

  6. Activation analysis opportunities using cold neutron beams

    Energy Technology Data Exchange (ETDEWEB)

    Lindstrom, R M; Zeisler, R; Rossbach, M

    1987-05-01

    Guided beams of cold neutrons being installed at a number of research reactors may become increasingly available for analytical research. A guided cold beam will provide higher neutron fluence rates and lower background interferences than in present facilities. In an optimized facility, fluence rates of 10/sup 9/ nxcm/sup -2/xs/sup -1/ are obtainable. Focusing a large area beam onto a small target will further increase the neutron intensity. In addition, the shift to lower neutron energy increases the effective cross sections. The absence of fast neutrons and gamma rays permits detectors to be placed near the sample without intolerable background, and thus the efficiency for counting prompt gamma rays can be much higher than in present systems. Measurements made at the hydrogen cold source of the FRJ-2 (DIDO) reactor at the KFA provide a numerical evaluation of the improvements in PGAA with respect to signal-to-background ratios of important elements and matrices. (author) 15 refs.

  7. Average radiation weighting factors for specific distributed neutron spectra

    International Nuclear Information System (INIS)

    Ninkovic, M.M.; Raicevic, J.J.

    1993-01-01

    Spectrum averaged radiation weighting factors for 6 specific neutron fields in the environment of 3 categories of the neutron sources (fission, spontaneous fission and (α,n)) are determined in this paper. Obtained values of these factors are greater 1.5 to 2 times than the corresponding quality factors used for the same purpose until a few years ago. This fact is very important to have in mind in the conversion of the neutron fluence into the neutron dose equivalent. (author)

  8. Passive neutron dosemeter with activation detector

    Energy Technology Data Exchange (ETDEWEB)

    Valero L, C.; Banuelos F, A.; Guzman G, K. A.; Borja H, C. G.; Hernandez D, V. M.; Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2011-10-15

    A passive neutron dosemeter with {sup 197}Au activation detector has been developed. The area dosemeter was made as a 20.5 {phi} x 20.5 cm{sup 2} polyethylene moderator, with a polyethylene pug where a {sup 197}Au foil can be located either parallel or perpendicular to moderator axis. Using Monte Carlo methods, with the MCNP5 code. With the fluence response and the fluence-to-equivalent dose conversion coefficients from ICRP-74, responses to H*(10) were also calculated, these were compared against responses of commercially available neutron area monitors and dosemeters. (Author)

  9. 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)

  10. 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)

  11. Defect cascades produced by neutron irradiation in YBa2Cu3O7-δ superconductors

    International Nuclear Information System (INIS)

    Frischherz, M.C.; Kirk, M.A.; Farmer, J.

    1994-02-01

    The defect cascades produced by fast neutron irradiation of YBa 2 Cu 3 O 7-δ single crystals were studied by transmission electron microscopy. The visible defects were found to have sizes between 1 and 5 rim. Defect densities were obtained as a function of neutron fluence between 2 and 8x 10 21 m -2 (E>0.1 MeV). The measured defect density scales linearly with fluence and amounts to 1x10 22 m -3 at a neutron fluence of 2x10 2l m -2 . The defect stability was studied at room temperature and through annealing to 400 degrees C

  12. Comprehensive fluence model for absolute portal dose image prediction

    International Nuclear Information System (INIS)

    Chytyk, K.; McCurdy, B. M. C.

    2009-01-01

    Amorphous silicon (a-Si) electronic portal imaging devices (EPIDs) continue to be investigated as treatment verification tools, with a particular focus on intensity modulated radiation therapy (IMRT). This verification could be accomplished through a comparison of measured portal images to predicted portal dose images. A general fluence determination tailored to portal dose image prediction would be a great asset in order to model the complex modulation of IMRT. A proposed physics-based parameter fluence model was commissioned by matching predicted EPID images to corresponding measured EPID images of multileaf collimator (MLC) defined fields. The two-source fluence model was composed of a focal Gaussian and an extrafocal Gaussian-like source. Specific aspects of the MLC and secondary collimators were also modeled (e.g., jaw and MLC transmission factors, MLC rounded leaf tips, tongue and groove effect, interleaf leakage, and leaf offsets). Several unique aspects of the model were developed based on the results of detailed Monte Carlo simulations of the linear accelerator including (1) use of a non-Gaussian extrafocal fluence source function, (2) separate energy spectra used for focal and extrafocal fluence, and (3) different off-axis energy spectra softening used for focal and extrafocal fluences. The predicted energy fluence was then convolved with Monte Carlo generated, EPID-specific dose kernels to convert incident fluence to dose delivered to the EPID. Measured EPID data were obtained with an a-Si EPID for various MLC-defined fields (from 1x1 to 20x20 cm 2 ) over a range of source-to-detector distances. These measured profiles were used to determine the fluence model parameters in a process analogous to the commissioning of a treatment planning system. The resulting model was tested on 20 clinical IMRT plans, including ten prostate and ten oropharyngeal cases. The model predicted the open-field profiles within 2%, 2 mm, while a mean of 96.6% of pixels over all

  13. Fluence scan: an unexplored property of a laser beam

    International Nuclear Information System (INIS)

    Chalupsky, Jaromir; Hajkova, Vera; Burian, Tomas; Juha, Libor; Polcar, Tomas; Gaudin, Jerome; Nagasono, Mitsuru; Yabashi, Makina; Sobierajski, Ryszard; Krzywinski, Jacek

    2013-01-01

    We present an extended theoretical background of so-called fluence scan (f-scan or F-scan) method, which is frequently being used for offline characterization of focused short-wavelength (EUV, soft X-ray, and hard X-ray) laser beams [J. Chalupsky et al., Opt. Express 18, 27836 (2010)]. The method exploits ablative imprints in various solids to visualize iso-fluence beam contours at different fluence and/or clip levels. An f-scan curve (clip level as a function of the corresponding iso-fluence contour area) can be generated for a general non-Gaussian beam. As shown in this paper, fluence scan encompasses important information about energy distribution within the beam profile, which may play an essential role in laser-matter interaction research employing intense non-ideal beams. Here we for the first time discuss fundamental properties of the f-scan function and its inverse counterpart (if-scan). Furthermore, we extensively elucidate how it is related to the effective beam area, energy distribution, and to the so called Liu's dependence [J.M. Liu, Opt. Lett. 7, 196 (1982)]. A new method of the effective area evaluation based on weighted inverse f-scan fit is introduced and applied to real data obtained at the SCSS (SPring-8 Compact SASE Source) facility. (authors)

  14. Neutron dosimetry in EDF experimental surveillance programme for VVER-440 nuclear power plants

    International Nuclear Information System (INIS)

    Brumovsky, M.; Erben, O.; Novosad, P.; Zerola, L.; Hogel, J.; Trollat, C.

    2001-01-01

    Fourteen chains containing experimental surveillance material specimens of the VVER 440/213 nuclear power reactor pressure vessels were irradiated in the surveillance channels of the Nuclear Power Plant Dukovany in the Czech Republic. The irradiation periods were one, two or three cycles. The chains contained different number and types of containers, the omitted ones were replaced by chain elements. All of the containers were instrumented with wire neutron fluence detectors, some of the containers in the chain had spectrometric sets of neutron fluence monitors. For the absolute fluence values evaluation it was taken into account time history of the reactor power and local changes of the neutron flux along the reactor core height, correction factors due to the orientation of monitors with respect to the reactor core centre. Unfolding programs SAND-II or BASA-CF were used. The relative axial fluence distribution was obtained from the O-wire measurements. Neutron fluence values above 0.5 MeV energy and above 1.0 MeV energy in the container axis on the axial positions of the sample centres and fluence values in the geometric centre of the samples was calculated making use the exponential attenuation model of the incident neutron beam. Received fast neutron fluence values can be used as reference values to all VVER-440 type 213 nuclear power plant reactors. (author)

  15. Characterization of Monoenergetic Low Energy Neutron Fields with the {mu}TPC Detector

    Energy Technology Data Exchange (ETDEWEB)

    Golabek, C.; Lebreton, L.; Petit, M. [Laboratoire de Metrologie et de Dosimetrie des Neutrons, IRSN Cadarache, 13115 Saint-Paul-Lez-Durance (France); Billard, J.; Grignon, C.; Bosson, G.; Bourrion, O.; Guillaudin, O.; Mayet, F.; Richer, J.-P.; Santos, D. [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph (France)

    2011-12-13

    The AMANDE facility produces monoenergetic neutron fields from 2 keV to 20 MeV for metrological purposes. To be considered as a reference facility, fluence and energy distributions of neutron fields have to be determined by primary measurement standards. For this purpose, a micro Time Projection Chamber is being developed to be dedicated to measure neutron fields with energy ranging from 2 keV up to 1 MeV. We present simulations showing that such a detector, which allows the measurement of the ionization energy and the 3D reconstruction of the recoil nucleus, provides the determination of neutron energy and fluence of such low energy neutron fields.

  16. Application of imaging plate neutron detector to neutron radiography

    CERN Document Server

    Fujine, S; Kamata, M; Etoh, M

    1999-01-01

    As an imaging plate neutron detector (IP-ND) has been available for thermal neutron radiography (TNR) which has high resolution, high sensitivity and wide range, some basic characteristics of the IP-ND system were measured at the E-2 facility of the KUR. After basic performances of the IP were studied, images with high quality were obtained at a neutron fluence of 2 to 7x10 sup 8 n cm sup - sup 2. It was found that the IP-ND system with Gd sub 2 O sub 3 as a neutron converter material has a higher sensitivity to gamma-ray than that of a conventional film method. As a successful example, clear radiographs of the flat view for the fuel side plates with boron burnable poison were obtained. An application of the IP-ND system to neutron radiography (NR) is presented in this paper.

  17. Neutron stars

    International Nuclear Information System (INIS)

    Irvine, J.M.

    1978-01-01

    The subject is covered in chapters entitled: introduction (resume of stellar evolution, gross characteristics of neutron stars); pulsars (pulsar characteristics, pulsars as neutron stars); neutron star temperatures (neutron star cooling, superfluidity and superconductivity in neutron stars); the exterior of neutron stars (the magnetosphere, the neutron star 'atmosphere', pulses); neutron star structure; neutron star equations of state. (U.K.)

  18. Dosimetric And Fluence Measurements At Hadron Facilities For LHC Radiation Damage Studies

    CERN Document Server

    León-Florián, E

    2001-01-01

    Dosimetry plays an essential role in experiments assessing radiation damage and hardness for the components of detectors to be operated at the future Large Hadron Collider (LHC), CERN (European Laboratory for Particle Physics), Geneva, Switzerland. Dosimetry is used both for calibration of the radiation fields and estimate of fluences and doses during the irradiation tests. The LHC environment will result in a complex radiation field composed of hadrons (mainly neutrons, pions and protons) and photons, each having an energy spectrum ranging from a few keV to several hundreds of MeV or several GeV, even. In this thesis, are exposed the results of measurements of particle fluences and doses at different hadron irradiation facilities: SARA, πE1-PSI and ZT7PS used for testing the radiation hardness of materials and equipment to be used in the future experiments at LHC. These measurements are applied to the evaluation of radiation damage inflicted to various semiconductors (such as silicon) and electronics ...

  19. Fluence dependence of disorder depth profiles in Pb implanted Si

    International Nuclear Information System (INIS)

    Christodoulides, C.E.; Kadhim, N.J.; Carter, G.

    1980-01-01

    The total, depth integrated disorder, induced by Pb implantation into Si at room temperature, initially increases rapidly with implantation fluence and then reaches a quasi saturation level where the increase with fluence is slow. Measurements of the depth distributions of the disorder, using high resolution low angle exit Rutherford Backscattering/Channelling analysis, suggest that the quasi saturation results from overlapping of disordered zones generated deep in the tail of the disorder-depth profiles. The depth of the disordered solid-crystal boundary, xsub(D), increases with ion fluence PHI, according to the relation xsub(D) = x bar + f(PHI).σ, where x bar is the most probable projected depth and σ the projected standard deviation of disorder generation. It is shown that this relationship is consistent with an approximately Gaussian depth distribution of disorder production. (author)

  20. Photon energy-fluence correction factor in low energy brachytherapy

    International Nuclear Information System (INIS)

    Antunes, Paula C.G.; Yoriyaz, Hélio; Vijande, Javier; Giménez-Alventosa, Vicent; Ballester, Facundo

    2017-01-01

    The AAPM TG-43 brachytherapy dosimetry formalism has become a standard for brachytherapy dosimetry worldwide; it implicitly assumes that charged-particle equilibrium (CPE) exists for the determination of absorbed dose to water at different locations. At the time of relating dose to tissue and dose to water, or vice versa, it is usually assumed that the photon fluence in water and in tissues are practically identical, so that the absorbed dose in the two media can be related by their ratio of mass energy-absorption coefficients. The purpose of this work is to study the influence of photon energy-fluence in different media and to evaluate a proposal for energy-fluence correction factors for the conversion between dose-to-tissue (D tis ) and dose-to-water (D w ). State-of-the art Monte Carlo (MC) calculations are used to score photon fluence differential in energy in water and in various human tissues (muscle, adipose and bone) in two different codes, MCNP and PENELOPE, which in all cases include a realistic modeling of the 125 I low-energy brachytherapy seed in order to benchmark the formalism proposed. A correction is introduced that is based on the ratio of the water-to-tissue photon energy-fluences using the large-cavity theory. In this work, an efficient way to correlate absorbed dose to water and absorbed dose to tissue in brachytherapy calculations at clinically relevant distances for low-energy photon emitting seed is proposed. The energy-fluence based corrections given in this work are able to correlate absorbed dose to tissue and absorbed dose to water with an accuracy better than 0.5% in the most critical cases. (author)

  1. Photon energy-fluence correction factor in low energy brachytherapy

    Energy Technology Data Exchange (ETDEWEB)

    Antunes, Paula C.G.; Yoriyaz, Hélio [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Vijande, Javier; Giménez-Alventosa, Vicent; Ballester, Facundo, E-mail: pacrisguian@gmail.com [Department of Atomic, Molecular, and Nuclear Physics and Instituto de Física Corpuscular (UV-CSIC), University of Valencia (Spain)

    2017-07-01

    The AAPM TG-43 brachytherapy dosimetry formalism has become a standard for brachytherapy dosimetry worldwide; it implicitly assumes that charged-particle equilibrium (CPE) exists for the determination of absorbed dose to water at different locations. At the time of relating dose to tissue and dose to water, or vice versa, it is usually assumed that the photon fluence in water and in tissues are practically identical, so that the absorbed dose in the two media can be related by their ratio of mass energy-absorption coefficients. The purpose of this work is to study the influence of photon energy-fluence in different media and to evaluate a proposal for energy-fluence correction factors for the conversion between dose-to-tissue (D{sub tis}) and dose-to-water (D{sub w}). State-of-the art Monte Carlo (MC) calculations are used to score photon fluence differential in energy in water and in various human tissues (muscle, adipose and bone) in two different codes, MCNP and PENELOPE, which in all cases include a realistic modeling of the {sup 125}I low-energy brachytherapy seed in order to benchmark the formalism proposed. A correction is introduced that is based on the ratio of the water-to-tissue photon energy-fluences using the large-cavity theory. In this work, an efficient way to correlate absorbed dose to water and absorbed dose to tissue in brachytherapy calculations at clinically relevant distances for low-energy photon emitting seed is proposed. The energy-fluence based corrections given in this work are able to correlate absorbed dose to tissue and absorbed dose to water with an accuracy better than 0.5% in the most critical cases. (author)

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

    International Nuclear Information System (INIS)

    Kalsi, P.C.

    2000-01-01

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

  3. Introduction of neutron metrology for reactor radiation damage

    International Nuclear Information System (INIS)

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

    1979-01-01

    The background of the procedures for determining irradiation parameters which are of interest in radiation damage experiments is described. The first two chapters outline the concept of damage functions and damage models. The next two chapters give information on methods to determine neutron fluences and neutron spectra. The fifth chapter gives a review of correlation data available for graphite and steels. The last chapter gives guidance how to report the relevant irradiation parameters. Attention is given to the role of the neutron spectrum in deriving values for damage fluence, energy transferred to the lattice, and number of displacements

  4. Dosimetry methods in boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Gambarini, G.; Artuso, E.; Felisi, M.; Regazzoni, V.; Giove, D. [Universita degli Studi di Milano, Department of Physics, Via Festa del Patrono 7, 20122 Milano (Italy); Agosteo, S.; Barcaglioni, L. [Istituto Nazionale di Fisica Nucleare, Milano (Italy); Campi, F.; Garlati, L. [Politecnico di Milano, Energy Department, Piazza Leonardo Da Vinci 32, 20133 Milano (Italy); De Errico, F. [Universita degli Studi di Pisa, Department of Civil and Industrial Engineering, Lungamo Pacinotti 43, 56126 Pisa (Italy); Borroni, M.; Carrara, M. [Fondazione IRCCS Istituto Nazionale Tumori, Medical Physics Unit, Via Venezian 1, 20133 Milano (Italy); Burian, J.; Klupak, V.; Viererbl, L.; Marek, M. [Research Centre Rez, Department of Neutron Physics, 250-68 Husinec-Rez (Czech Republic)

    2014-08-15

    Dosimetry studies have been carried out at thermal and epithermal columns of Lvr-15 research reactor for investigating the spatial distribution of gamma dose, fast neutron dose and thermal neutron fluence. Two different dosimetry methods, both based on solid state detectors, have been studied and applied and the accuracy and consistency of the results have been inspected. One method is based on Fricke gel dosimeters that are dilute water solutions and have good tissue equivalence for neutrons and also for all the secondary radiations produced by neutron interactions in tissue or water phantoms. Fricke gel dosimeters give the possibility of separating the various dose contributions, i.e. the gamma dose, the fast neutron dose and the dose due to charged particles generated during thermal neutron reactions by isotopes having high cross section, like 10-B. From this last dose, thermal neutron fluence can be obtained by means of the kerma factor. The second method is based on thermoluminescence dosimeters. In particular, the developed method draw advantage from the different heights of the peaks of the glow curve of such phosphors when irradiated with photons or with thermal neutrons. The results show that satisfactory results can be obtained with simple methods, in spite of the complexity of the subject. However, the more suitable dosimeters and principally their utilization and analysis modalities are different for the various neutron beams, mainly depending on the relative intensities of the three components of the neutron field, in particular are different for thermal and epithermal columns. (Author)

  5. Dosimetry methods in boron neutron capture therapy

    International Nuclear Information System (INIS)

    Gambarini, G.; Artuso, E.; Felisi, M.; Regazzoni, V.; Giove, D.; Agosteo, S.; Barcaglioni, L.; Campi, F.; Garlati, L.; De Errico, F.; Borroni, M.; Carrara, M.; Burian, J.; Klupak, V.; Viererbl, L.; Marek, M.

    2014-08-01

    Dosimetry studies have been carried out at thermal and epithermal columns of Lvr-15 research reactor for investigating the spatial distribution of gamma dose, fast neutron dose and thermal neutron fluence. Two different dosimetry methods, both based on solid state detectors, have been studied and applied and the accuracy and consistency of the results have been inspected. One method is based on Fricke gel dosimeters that are dilute water solutions and have good tissue equivalence for neutrons and also for all the secondary radiations produced by neutron interactions in tissue or water phantoms. Fricke gel dosimeters give the possibility of separating the various dose contributions, i.e. the gamma dose, the fast neutron dose and the dose due to charged particles generated during thermal neutron reactions by isotopes having high cross section, like 10-B. From this last dose, thermal neutron fluence can be obtained by means of the kerma factor. The second method is based on thermoluminescence dosimeters. In particular, the developed method draw advantage from the different heights of the peaks of the glow curve of such phosphors when irradiated with photons or with thermal neutrons. The results show that satisfactory results can be obtained with simple methods, in spite of the complexity of the subject. However, the more suitable dosimeters and principally their utilization and analysis modalities are different for the various neutron beams, mainly depending on the relative intensities of the three components of the neutron field, in particular are different for thermal and epithermal columns. (Author)

  6. Neutron spectra produced by moderating an isotopic neutron source

    International Nuclear Information System (INIS)

    Carrillo Nunnez, Aureliano; Vega Carrillo, Hector Rene

    2001-01-01

    A Monte Carlo study has been carried out to determine the neutron spectra produced by an isotopic neutron source inserted in moderating media. Most devices used for radiation protection have a response strongly dependent on neutron energy. ISO recommends several neutron sources and monoenergetic neutron radiations, but actual working situations have broad spectral neutron distributions extending from thermal to MeV energies, for instance, near nuclear power plants, medical applications accelerators and cosmic neutrons. To improve the evaluation of the dosimetric quantities, is recommended to calibrate the radiation protection devices in neutron spectra which are nearly like those met in practice. In order to complete the range of neutron calibrating sources, it seems useful to develop several wide spectral distributions representative of typical spectra down to thermal energies. The aim of this investigation was to use an isotopic neutron source in different moderating media to reproduce some of the neutron fields found in practice. MCNP code has been used during calculations, in these a 239PuBe neutron source was inserted in H2O, D2O and polyethylene moderators. Moderators were modeled as spheres and cylinders of different sizes. In the case of cylindrical geometry the anisotropy of resulting neutron spectra was calculated from 0 to 2 . From neutron spectra dosimetric features were calculated. MCNP calculations were validated by measuring the neutron spectra of a 239PuBe neutron source inserted in a H2O cylindrical moderator. The measurements were carried out with a multisphere neutron spectrometer with a 6LiI(Eu) scintillator. From the measurements the neutron spectrum was unfolded using the BUNKIUT code and the UTA4 response matrix. Some of the moderators with the source produce a neutron spectrum close to spectra found in actual applications, then can be used during the calibration of radiation protection devices

  7. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Wilson, J.F.; Sherwood, D.G.

    1982-01-01

    A mechanical spectral shift reactor comprises a reactive core having fuel assemblies accommodating both water displacer elements and neutron absorbing control rods for selectively changing the volume of water-moderator in the core. The fuel assemblies with displacer and control rods are arranged in alternating fashion so that one displacer element drive mechanism may move displacer elements in more than one fuel assembly without interfering with the movement of control rods of a corresponding control rod drive mechanisms. (author)

  8. High temperature ductility of austenitic alloys exposed to thermal neutrons

    International Nuclear Information System (INIS)

    Watanabe, K.; Kondo, T.; Ogawa, Y.

    1982-01-01

    Loss of high temperature ductility due to thermal neutron irradiation was examined by slow strain rate test in vacuum up to 1000 0 C. The results on two heats of Hastelloy alloy X with different boron contents were analyzed with respect to the influence of the temperatures of irradiation and tensile tests, neutron fluence and the associated helium production due to nuclear transmutation reaction. The loss of ductility was enhanced by increasing either temperature or neutron fluence. Simple extrapolations yielded the estimated threshold fluence and the end-of-life ductility values at 900 and 1000 0 C in case where the materials were used in near-core regions of VHTR. The observed relationship between Ni content and the ductility loss has suggested a potential utilization of Fe-based alloys for seathing of the neutron absorber materials

  9. DT fusion neutron irradiation of BPNL niobium nickel and 316 stainless steel at 1750C

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1977-01-01

    The DT fusion neutron irradiation at 175 0 C of 17 niobium wires, one niobium foil, 14 316 stainless steel wires, one 316 stainless steel foil, nine nickel wires, and two nickel foils from BPNL is described. The sample position, beam-on time, neutron dose record, and neutron fluence are given

  10. Neutron spectrometry for protection dosimetry at very low levels

    International Nuclear Information System (INIS)

    Bardell, A.G.; Thomas, D.J.

    1996-01-01

    Dose limits for exposure of members of the public are significantly lower than those for designated radiation workers. The new ICRP 60 recommendation for critical groups of the general public is an effective dose limit of 1 mSv per year which requires a measurement capability at levels down to about 100 nSv h - 1. Radiation protection dosimetry for neutrons at these levels is problematical, nevertheless, operators of nuclear sites are still required to demonstrate acceptably low radiation levels in areas accessible to the public. In addition to the known poor dose equivalent response of available dosemeters, there is an added problem at low levels of inadequate sensitivity. Personal dosemeters are certainly not sufficiently sensitive, and the sensitivity of area survey instruments is such that they can only be used in integral mode Even then, the statistical uncertainties are likely to be large. One further problem concerns the quantity measured. Survey instruments are designed to measure the operational quantity ambient dose equivalent, H*(10), which always tends to be an overestimate of the present limiting quantity effective dose equivalent. If in any situation exposure is near the limit, an estimate of H*(10) may not be sufficient to prove conclusively that levels are less than the statutory limit, and a direct estimate of effective dose equivalent may need to be made. The only way of estimating effective dose equivalent is via an absolute spectral measurement. From such a spectrum any relevant dosimetric quantity can be estimated via tabulated fluence to dose equivalent conversion factors. (Certain quantities also require information about the angular dependence of the field - see later text). Spectrometry at such low neutron fluence levels is difficult, however, there is one instrument available which can perform the required measurements, and this is a well characterised Bonner sphere (BS) set. (author)

  11. Neutron production from 158 GeV/c per nucleon lead ions on thin copper and lead targets in the angular range 30-135 deg

    CERN Document Server

    Agosteo, S; Foglio-Para, A; Gini, L; Mitaroff, W A; Silari, Marco; Ulrici, L

    2002-01-01

    The neutron emission from 5, 10 and 20 mm thick lead and 10 and 20 mm thick copper targets bombarded by a lead ion beam with momentum of 158 GeV/c per nucleon were measured at the CERN Super Proton Synchrotron. The neutron yield and spectral fluence per incident ion on target were measured with an extended range Bonner sphere spectrometer in the angular range 30-135 deg. with respect to beam direction. Monte Carlo simulations with the FLUKA code were performed to establish a guess spectrum for the unfolding of the experimental data. The results have shown that, lacking Monte Carlo radiation transport codes dealing with ions with masses larger than 1 amu, a reasonable prediction can be carried out by scaling the result of a Monte Carlo calculation for protons by the projectile mass number to the power of 0.85-0.95 for a lead target and 0.88-1.03 for a copper target.

  12. A wide-range direction neutron spectrometer

    International Nuclear Information System (INIS)

    Luszik-Bhadra, M.; D'Errico, F.; Hecker, O.; Matzke, M.

    2002-01-01

    A new device is presented which has been developed for measuring the energy and direction of distribution of neutron fluence in fields of broad energy spectra (thermal to 100 MeV) and with a high background of photon, electron and muon radiation. The device was tested in reference fields with different energy and direction distributions of neutron fluence. The direction-integrated fluence spectra agree fairly well with reference spectra. In all cases, the ambient and personal dose equivalent values calculated from measured direction-differential spectra are within 35% of the reference values. Independent measurements of the directional dose equivalent were performed with a directional dose equivalent monitor based on superheated drop detectors

  13. Neutron induced permanent damage in Josephson junctions

    International Nuclear Information System (INIS)

    Mueller, G.P.; Rosen, M.

    1982-01-01

    14 MeV neutron induced permanent changes in the critical current density of Josephson junctions due to displacement damage in the junction barrier are estimated using a worst case model and the binary collision simulation code MARLOWE. No likelihood of single event hard upsets is found in this model. It is estimated that a fluence of 10 18 -10 19 neutrons/cm 2 are required to change the critical current density by 5%

  14. Development and characterization of multi-sensory fluence rate probes

    International Nuclear Information System (INIS)

    Pomerleau-Dalcourt, Natalie; Lilge, Lothar

    2006-01-01

    Multi-sensory fluence rate probes (MSPs) yield several simultaneous measurements of photodynamic therapy (PDT) treatment light fluence from a single interstitial probe. Fluorescent sensors are embedded at desired positions along the axis of the optical fibre. A single fluorescence emission spectrum is obtained and decomposed using a partial least squares (PLS)-based analysis to yield the fluence at each sensor's location. The responsivity, linearity and possible photodegradation of each fluorophore chosen for the MSPs were evaluated using single-sensor probes. The performance of two- and three-sensor MSPs was evaluated experimentally. Individual fluorescence spectra collected from each sensor on the MSP were used to construct the training set necessary for the PLS-based analysis. The MSPs' responsivity, spatial resolution and accuracy were evaluated relative to a single scattering-tip detector. Three-fluorophore MSPs permitted three simultaneous measurements of the fluence rate gradient in a tissue-like phantom, with an average accuracy of 6.7%. No appreciable photodegradation or cross-talk was observed

  15. Nonlocal ultrafast magnetization dynamics in the high fluence limit

    NARCIS (Netherlands)

    Kuiper, K.C.; Malinowski, G.; Dalla Longa, F.; Koopmans, B.

    2011-01-01

    In order to explain a number of recent experimental observations of laser-induced femtosecond demagnetization in the large fluence limit, we discuss the consequences of a recently proposed nonlocal approach. A microscopic description of spin flip scattering is implemented in an effective three

  16. Study of spectral response of a neutron filter. Design of a method to adjust spectra; Etude des moyens de conditionnement de la reponse spectrale d'un filtre a neutrons. Mise au point d'une methode d'ajustement rapide de spectre

    Energy Technology Data Exchange (ETDEWEB)

    Colomb-Dolci, F. [Universite Louis Pasteur, 67 - Strasbourg (France)

    1999-02-01

    The first part of this thesis describes an experimental method which intends to determine a neutron spectrum in the epithermal range [1 eV -10 keV]. Based on measurements of reaction rates provided by activation foils, it gives flux level in each energy range corresponding to each probe. This method can be used in any reactor location or in a neutron beam. It can determine scepter on eight energy groups, five groups in the epithermal range. The second part of this thesis presents a study of an epithermal neutron beam design, in the frame of Neutron Capture Therapy. A beam tube was specially built to test filters made up of different materials. Its geometry was designed to favour epithermal neutron crossing and to cut thermal and fast neutrons. A code scheme was validated to simulate the device response with a Monte Carlo code. Measurements were made at ISIS reactor and experimental spectra were compared to calculated ones. This validated code scheme was used to simulate different materials usable as shields in the tube. A study of these shields is presented at the end of this thesis. (author)

  17. Neutron radiation from medical electron accelerators

    International Nuclear Information System (INIS)

    McCall, R.C.

    1983-01-01

    A method is described using simple gold foils and relatively inexpensive moderators to measure neutron fluences, both fast nd thermal, which then can be converted to dose equivalent using a few simple formulas. The method is sensitive, easy to calibrate, and should work at most accelerators regardless of energy or room geometry

  18. Fluence compensated photoacoustic tomography in small animals (Conference Presentation)

    Science.gov (United States)

    Hussain, Altaf; Pool, Martin; Daoudi, Khalid; de Vries, Liesbeth G.; Steenbergen, Wiendelt

    2017-03-01

    Light fluence inside turbid media can be experimentally mapped by measuring ultrasonically modulated light (Acousto-optics). To demonstrate the feasibility of fluence corrected Photoacoustic (PA) imaging, we have realized a tri-modality (i.e. photoacoustic, acousto-optic and ultrasound) tomographic small animal imaging system. Wherein PA imaging provides high resolution map of absorbed optical energy density, Acousto-optics yields the fluence distribution map in the corresponding PA imaging plane and Ultrasound provides morphological information. Further, normalization of the PA image with the acousto-optically measured fluence map results in an image that directly represents the optical absorption. Human epidermal growth factor receptor 2 (HER2) is commonly found overexpressed in human cancers, among which breast cancers, resulting in a more aggressive tumor phenotype. Identification of HER2-expression is clinically relevant, because cancers overexpressing this marker are amenable to HER2-directed therapies, among which antibodies trastuzumab and pertuzumab. Here, we investigate the feasibility and advantage of acousto-optically assisted fluence compensated PA imaging over PA imaging alone in visualizing and quantifying HER2 expression. For this experiment, nude mice were xenografted with human breast cancer cell lines SKBR3 and BT474 (both HER2 overexpressing), as well as HER2-negative MDA-MB-231. To visualize HER2 expression in these mice, HER2 monoclonal antibody pertuzumab (Perjeta®, Roche), was conjugated to near-infrared dye IRDye 800CW (800CW, LICOR Biosciences) at a ratio of 1∶2 antibody to 800CW. When xenograft tumors measured ≥ 100 mm3, mice received 100 µg 800CW-pertuzumab intravenously. Three days post injection, mice were scanned for fluorescence signal with an IVIS scanner. After fluorescence scans, mice were euthanized and imaged in our PA tomographic imaging system.

  19. Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code

    Science.gov (United States)

    Faghihi, F.; Mehdizadeh, S.; Hadad, K.

    Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.

  20. Fluence-rate effects on irradiation embrittlement and composition and temperature effects on annealing/reirradiation sensitivity

    International Nuclear Information System (INIS)

    Hawthorne, J.R.; Hiser, A.L.

    1988-01-01

    Recent MEA investigation on the effect of neutron fluence rate on radiation-induced embrittlement accrual and the contributions of metallurgical variables to postirradiation annealing and re-irradiation behavior are reviewed. Studies of fluence-rate effects involved experiments in the UBR test reactor and separately, radiation sensitivity determinations for the decommissioned Gundremmingen (KRB-A) vessel material. Annealing-reirradiation studies employed 399 0 C and 454 0 C heat treatments. Material composition is shown to play a major role in postirradiation annealing recovery. Results illustrate effects of variable copper and variable nickel contents on recoveray of steel plate having low phosphorus levels. Composition effects on recovery were also observed for prototypic welds depicting high/low copper and high/low nickel contents and three flux types. The welds, in addition, indicate major differences in re-irradiation sensitivity. The UBR investigations revealed a significant difference in fluence rate sensitivity between the ASTM A 302-B reference plate and a submerged-arc (S/A) Linde 80 weld. Studies of the Gundremmingen reactor vessel, representing a joint USA-FRG-UK undertaking revealed an anomaly in strong vs. weak test orientation radiation sensitivity. (orig./HP)

  1. Preliminary study on 2-dimensional distributions of 10B reaction rate in a water phantom with boron-doped CR-39 for 7Li(p, n)7Be neutrons by 1.95 MeV protons

    International Nuclear Information System (INIS)

    Hasegawa, Y.; Tanaka, K.; Tsuruta, T.

    2000-01-01

    In an Accelerator-based neutron irradiation field using 7 Li(p, n) 7 Be neutrons by 1.95 MeV protons, the distributions of 10 B reaction rates and thermal neutron fluence in a water phantom were measured using Boron-doped CR-39 and Au activation analysis, respectively. Comparing the results of the measurements, we discussed the validity of the evaluation method of 10 B reaction rate using thermal neutron fluence. (author)

  2. A study of rates of (n, f), (n, γ), and (n, 2n) reactions in natU and 232Th produced by the neutron fluence in the graphite set-up (gamma-3) irradiated by 2.33 GeV deuteron beam

    International Nuclear Information System (INIS)

    Adam, J.; Chitra Bhatia; Katovskij, K.

    2011-01-01

    Spallation neutrons produced in a collision of 2.33 GeV deuteron beam with the large lead target are moderated by the thick graphite block surrounding the target and used to activate the radioactive samples of nat U and Th put at the three different positions, identified as holes 'a', 'b' and 'c' in the graphite block. Rates of the (n, f), (n, γ), and (n, 2n) reactions in the two samples are determined using the gamma spectrometry. Ratio of the experimental reaction rates, R(n, 2n)/R(n, f) for the 232 Th and nat U are estimated in order to understand the role of reactions of (n, xn) type in Accelerator Driven Subcritical Systems. For the Th-sample, the ratio is ∼ 54(10)% in case of hole 'a' and ∼ 95(57)% in case of hole 'b' compared to 1.73(20)% for the hole 'a' and 0.710(9)% for the hole 'b' in case of the nat U sample. Also the ratio of fission rates in uranium to thorium, nat U(n, f)/ 232 Th(n, f), is ∼ 11.2(17) in case of hole 'a' and 26.8(85) in hole 'b'. Similarly, ratio 238 U(n, 2n)/ 232 Th(n, 2n) is 0.36(4) for the hole 'a' and 0.20(10) for the hole 'b' showing that 232 Th is more prone to the (n, xn) reaction than 238 U. All the experimental reaction rates are compared with the simulated ones by generating neutron fluxes at the three holes from MCNPX 2.6c and making use of LA150 library of cross sections. The experimental and calculated rates of all the three reactions are in good agreement. The transmutation power of the set-up is estimated using the rates of (n, γ) and (n, 2n) reactions for both the samples in the three holes and compared with some of the results of the 'Energy plus Transmutation' set-up and TARC experiment

  3. Spectral Shifting in Nondestructive Assay Instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nettleton, Anthony Steven [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tutt, James Robert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); LaFleur, Adrienne Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-17

    This project involves spectrum tailoring research that endeavors to better distinguish energies of gamma rays using different spectral material thicknesses and determine neutron energies by coating detectors with various materials.

  4. Fluence-dependent sputtering yield of micro-architectured materials

    Energy Technology Data Exchange (ETDEWEB)

    Matthes, Christopher S.R.; Ghoniem, Nasr M., E-mail: ghoniem@ucla.edu; Li, Gary Z.; Matlock, Taylor S.; Goebel, Dan M.; Dodson, Chris A.; Wirz, Richard E.

    2017-06-15

    Highlights: • Sputtering yield is shown to be transient and heavily dependent on surface architecture. • Fabricated nano- and Microstructures cause geometric re-trapping of sputtered material, which leads to a self-healing mechanism. • Initially, the sputtering yield of micro-architectured Mo is approximately 1/2 the value as that of a planar surface. • The study demonstrates that the sputtering yield is a dynamic property, dependent on the surface structure of a material. • A developed phenomenological model mathematically describes the transient behavior of the sputtering yield as a function of plasma fluence. - Abstract: We present an experimental examination of the relationship between the surface morphology of Mo and its instantaneous sputtering rate as function of low-energy plasma ion fluence. We quantify the dynamic evolution of nano/micro features of surfaces with built-in architecture, and the corresponding variation in the sputtering yield. Ballistic deposition of sputtered atoms as a result of geometric re-trapping is observed, and re-growth of surface layers is confirmed. This provides a self-healing mechanism of micro-architectured surfaces during plasma exposure. A variety of material characterization techniques are used to show that the sputtering yield is not a fundamental property, but that it is quantitatively related to the initial surface architecture and to its subsequent evolution. The sputtering yield of textured molybdenum samples exposed to 300 eV Ar plasma is roughly 1/2 of the corresponding value for flat samples, and increases with ion fluence. Mo samples exhibited a sputtering yield initially as low as 0.22 ± 5%, converging to 0.4 ± 5% at high fluence. The sputtering yield exhibits a transient behavior as function of the integrated ion fluence, reaching a steady-state value that is independent of initial surface conditions. A phenomenological model is proposed to explain the observed transient sputtering phenomenon, and to

  5. Neutron capture prompt gamma-ray activation analysis at the NIST cold neutron research facility

    Energy Technology Data Exchange (ETDEWEB)

    Lindstrom, R M; Zeisler, R; Vincent, D H; Greenberg, R R; Stone, C A; Mackey, E A [National Inst. of Standards and Technology, Gaithersburg, MD (United States); Anderson, D L [Food and Drug Administration, Washington, DC (United States); Clark, D D [Cornell Univ., Ithaca, NY (United States)

    1993-01-01

    An instrument for neutron capture prompt gamma-ray activation analysis (PGAA) has been constructed as part of the Cold Neutron Research Facility at the 20 MW National Institute of Standards and Technology Research Reactor. The neutron fluence rate (thermal equivalent) is 1.5*10[sup 8] n*cm[sup -2]*s[sup -] [sup 1], with negligible fast neutrons and gamma-rays. With compact geometry and hydrogen-free construction, the sensitivity is sevenfold better than an existing thermal instrument. Hydrogen background is thirtyfold lower. (author) 17 refs.; 2 figs.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  7. Divergence of Cs-137 sources fluence used in brachytherapy

    International Nuclear Information System (INIS)

    Vianello, E.A.; Almeida, C.E. de

    1998-01-01

    In this work the experimental determination of correction factor for fluence divergence (kln) of linear Cs-137 sources CDCS J4, with Farmer ionization chamber model 2571 in a central and perpendicular plan to source axis, for distances range from 1 to 7 cm., has been presented. The experimental results were compared to calculating by Kondo and Randolph (1960) isotropic theory and Bielajew (1990) anisotropic theory. (Author)

  8. Study of the effects of neutron irradiation on silicon strip detectors

    International Nuclear Information System (INIS)

    Giubellino, P.; Panizza, G.; Hall, G.; Sotthibandhu, S.; Ziock, H.J.; Ferguson, P.; Sommer, W.F.; Edwards, M.; Cartiglia, N.; Hubbard, B.; Leslie, J.; Pitzl, D.; O'Shaughnessy, K.; Rowe, W.; Sadrozinski, H.F.W.; Seiden, A.; Spencer, E.

    1992-01-01

    Silicon strip detectors and test structures were exposed to neutron fluences up to Φ=6.1x10 14 n/cm 2 , using the ISIS neutron source at the Rutherford Appleton Laboratory (UK). In this paper we report some of our results concerning the effects of displacement damage, with a comparison of devices made of silicon of different resistivity. The various samples exposed showed a very similar dependence of the leakage current on the fluence received. We studied the change of effective doping concentration, and observed a behaviour suggesting the onset of type inversion at a fluence of ∝2.0x10 13 n/cm 2 , a value which depends on the initial doping concentration. The linear increase of the depletion voltage for fluences higher than the inversion point could eventually determine the maximum fluence tolerable by silicon detectors. (orig.)

  9. Study of the effects of neutron irradiation on silicon strip detectors

    Energy Technology Data Exchange (ETDEWEB)

    Giubellino, P.; Panizza, G. (INFN Torino (Italy)); Hall, G.; Sotthibandhu, S. (Imperial Coll., London (United Kingdom)); Ziock, H.J.; Ferguson, P.; Sommer, W.F. (Los Alamos National Lab., NM (United States)); Edwards, M. (Rutherford Appleton Lab., Chilton (United Kingdom)); Cartiglia, N.; Hubbard, B.; Leslie, J.; Pitzl, D.; O' Shaughnessy, K.; Rowe, W.; Sadrozinski, H.F.W.; Seiden, A.; Spencer, E. (Santa Cruz Inst. for Particle Physics, Univ. California, CA (United States))

    1992-05-01

    Silicon strip detectors and test structures were exposed to neutron fluences up to {Phi}=6.1x10{sup 14} n/cm{sup 2}, using the ISIS neutron source at the Rutherford Appleton Laboratory (UK). In this paper we report some of our results concerning the effects of displacement damage, with a comparison of devices made of silicon of different resistivity. The various samples exposed showed a very similar dependence of the leakage current on the fluence received. We studied the change of effective doping concentration, and observed a behaviour suggesting the onset of type inversion at a fluence of {proportional to}2.0x10{sup 13} n/cm{sup 2}, a value which depends on the initial doping concentration. The linear increase of the depletion voltage for fluences higher than the inversion point could eventually determine the maximum fluence tolerable by silicon detectors. (orig.).

  10. Effects of the neutronic irradiation on the impact tests

    International Nuclear Information System (INIS)

    Lapena, J.; Perosanz, F.J.; Hernandez, M.T.

    1993-01-01

    The changes that the Charpy curves suffer when steel is exposed to neutronic fluence are studied. Three steels with different chemical composition were chosen, two of them (JPF and JPJ) being treated at only one neutronic fluence, while the last one (JRQ) was irradiated at three fluences. In this way, it was possible to compare the effect of increasing the neutronic dose, and to study the experimental results as a function of the steel chemical composition. Two characteristic facts have been observed: the displacement of the curve at higher temperatures, and decrease of the upper shelf energy (USE). The mechanical recovery of the materials after two different thermal treatments is also described, and a comparation between the experimental results obtained and the damage prediction formulas given by different regulatory international organisms in the nuclear field is established. Author. 11 refs

  11. Study and development of new dosemeters for thermal neutrons

    International Nuclear Information System (INIS)

    Urena N, F.

    1998-01-01

    An alanine-boron compound, alanine hydroborate, was synthesized and chemically characterized to be used for thermal neutrons fluence measurements. The synthesis of the compound was made by reacting the amino acid alanine with boric acid in three different media: acidic, neutral and alkaline. Physicochemical analysis showed that the alkaline medium is favorable for the synthesis of the alanine hydroborate. The compound was evaluated as a thermal neutron fluence detector by the detection of the free radical yield upon neutron thermal irradiation by Electron Paramagnetic Resonance (EPR). The present work also studies the EPR-signal response of the three preparations to thermal neutron irradiation (φ = 5 x 10 7 n/cm 2 -s). The following EPR signal parameters of the samples were investigated: peak-to-peak signal intensity vs. thermal neutron fluence Φ = φ Δt ; where Δt = 1, 5, 10, 20, 40, 60, 80, 90, 100, 110 and 120 h. , peak-to-peak signal intensity vs. microwave power, signal fading; repeatability, batch homogeneity, stability and zero dose response. It is concluded that these new products could be used in thermal neutron fluence estimations. (Author)

  12. New prospect in neutron radiography

    International Nuclear Information System (INIS)

    Cluzeau, S.

    1991-01-01

    Neutron radiography is a very useful non-destructive testing (NDT) method which frequently complements classical X-ray inspection. Numerous inspections in some fields are currently performed at reactor-based neutron radiography facilities but many other interesting applications in different fields are at present not considered because the objects to be tested cannot be moved to a reactor. It is the goal of the DIANE project, to allow the utilization of this NDT method in industrial facilities by developing a safe and convenient neutron radiography equipment using an ''on-off'' neutron source. As a result of the efforts of the four European partners, a first laboratory demonstration model is currently in operation in Germany and a fully mobile second one is expected by the end of 1992. Good radiographs are obtained with exposure times in the range of a few seconds to ten minutes using an electronic imaging system. The fast neutron generator uses a sealed neutron tube delivering 5.10 11 neutrons.cm -2 .s -1 in 4 π steradian; with a collimator ratio of about 12, the fluence rate onto the object is then close to 1,5.10 5 thermal neutrons.cm -2 .s -1 . (author)

  13. Study and development of new dosemeters for thermal neutrons; Estudio y desarrollo de nuevos dosimetros para neutrones termicos

    Energy Technology Data Exchange (ETDEWEB)

    Urena N, F

    1998-12-31

    An alanine-boron compound, alanine hydroborate, was synthesized and chemically characterized to be used for thermal neutrons fluence measurements. The synthesis of the compound was made by reacting the amino acid alanine with boric acid in three different media: acidic, neutral and alkaline. Physicochemical analysis showed that the alkaline medium is favorable for the synthesis of the alanine hydroborate. The compound was evaluated as a thermal neutron fluence detector by the detection of the free radical yield upon neutron thermal irradiation by Electron Paramagnetic Resonance (EPR). The present work also studies the EPR-signal response of the three preparations to thermal neutron irradiation ({phi} = 5 x 10{sup 7} n/cm{sup 2} -s). The following EPR signal parameters of the samples were investigated: peak-to-peak signal intensity vs. thermal neutron fluence {Phi} = {phi} {Delta}t ; where {Delta}t = 1, 5, 10, 20, 40, 60, 80, 90, 100, 110 and 120 h. , peak-to-peak signal intensity vs. microwave power, signal fading; repeatability, batch homogeneity, stability and zero dose response. It is concluded that these new products could be used in thermal neutron fluence estimations. (Author)

  14. Study on the energy response to neutrons for a new scintillating-fiber-array neutron detector

    CERN Document Server

    Zhang Qi; Wang Qun; Xie Zhong Shen

    2003-01-01

    The energy response of a new scintillating-fiber-array neutron detector to neutrons in the energy range 0.01 MeV<=E sub n<=14 MeV was modeled by combining a simplified Monte Carlo model and the MCNP 4b code. In order to test the model and get the absolute sensitivity of the detector to neutrons, one experiment was carried out for 2.5 and 14 MeV neutrons from T(p,n) sup 3 He and T(d,n) sup 4 He reactions at the Neutron Generator Laboratory at the Institute of Modern Physics, the Chinese Academy of Science. The absolute neutron fluence was obtained with a relative standard uncertainty 4.5% or 2.0% by monitoring the associated protons or sup 4 He particles, respectively. Another experiment was carried out for 0.5, 1.0, 1.5, 2.0, 2.5 MeV neutrons from T(p,n) sup 3 He reaction, and for 3.28, 3.50, 4.83, 5.74 MeV neutrons from D(d,n) sup 3 He reaction on the Model 5SDH-2 accelerator at China Institute of Atomic Energy. The absolute neutron fluence was obtained with a relative standard uncertainty 5.0% by usin...

  15. Effect of gamma and neutron irradiation on the mechanical properties of Spectralon™ porous PTFE

    Energy Technology Data Exchange (ETDEWEB)

    Gourdin, William H., E-mail: gourdin1@llnl.gov [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA USA (United States); Datte, Philip; Jensen, Wayne; Khater, Hesham; Pearson, Mark [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA USA (United States); Girard, Sylvain [Laboratoire Hubert Curien − UMR CNRS 5516, 18 rue du Pr. Benoît Lauras, F-42000 Saint Etienne (France); Paillet, Philippe; Alozy, Eric [CEA, DAM, DIF, F-91297 Arpajon (France)

    2016-11-15

    Highlights: • The effects of neutrons and gammas on PTFE are equivalent for a given absorbed dose. • A neutron fluence of 10{sup 13} n/cm{sup 2} corresponds to a gamma dose of 200 Gy. • The dose-to-fluence conversion factor is approximately 5 × 10{sup 10} n/(cm{sup 2}-Gy). • Irradiation in a low-oxygen environment enhances loads and elongations. • Mechanical properties of PTFE will deteriorate at a neutron fluence of 10{sup 13} n/cm{sup 2}. - Abstract: We establish a correspondence between the mechanical properties (maximum load and failure elongation) of Spectralon™ porous PTFE irradiated with 14 MeV neutrons and 1.17 and 1.33 MeV gammas from a cobalt-60 source. From this correspondence we infer that the effects of neutrons and gammas on this material are approximately equivalent for a given absorbed dose.

  16. Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 Need for Neutronics Calculations—An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures. 1.2 Methodology—Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is th...

  17. Neutron radiation damage studies on silicon detectors

    International Nuclear Information System (INIS)

    Li, Zheng; Chen, W.; Kraner, H.W.

    1990-10-01

    Effects of neutron radiation on electrical properties of Si detectors have been studied. At high neutron fluence (Φ n ≥ 10 12 n/cm 2 ), C-V characteristics of detectors with high resistivities (ρ ≥ 1 kΩ-cm) become frequency dependent. A two-trap level model describing this frequency dependent effect is proposed. Room temperature anneal of neutron damaged (at LN 2 temperature) detectors shows three anneal stages, while only two anneal stages were observed in elevated temperature anneal. 19 refs., 14 figs

  18. An advanced fusion neutron source facility

    International Nuclear Information System (INIS)

    Smith, D.L.

    1992-01-01

    Accelerator-based 14-MeV-neutron sources based on modifications of the original Fusion Materials Irradiation Facility are currently under consideration for investigating the effects of high-fluence high-energy neutron irradiation on fusion-reactor materials. One such concept for a D-Li neutron source is based on recent advances in accelerator technology associated with the Continuous Wave Deuterium Demonstrator accelerator under construction at Argonne National Laboratory, associated superconducting technology, and advances in liquid-metal technology. In this paper a summary of conceptual design aspects based on improvements in technologies is presented

  19. Microstructure and mechanical properties of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, E.; Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Terai, T.; Tanaka, S.

    1998-01-01

    Microstructure and mechanical properties of the neutron irradiated beryllium with total fast neutron fluences of 1.3 - 4.3 x 10{sup 21} n/cm{sup 2} (E>1 MeV) at 327 - 616degC were studied. Swelling increased by high irradiation temperature, high fluence, and by the small grain size and high impurity. Obvious decreasing of the fracture stress was observed in the bending test and in small grain specimens which had many helium bubbles on the grain boundary. Decreasing of the fracture stress for small grain specimens was presumably caused by crack propagation on the grain boundaries which weekend by helium bubbles. (author)

  20. Polarized neutrons

    International Nuclear Information System (INIS)

    Williams, W.G.

    1988-01-01

    The book on 'polarized neutrons' is intended to inform researchers in condensed matter physics and chemistry of the diversity of scientific problems that can be investigated using polarized neutron beams. The contents include chapters on:- neutron polarizers and instrumentation, polarized neutron scattering, neutron polarization analysis experiments and precessing neutron polarization. (U.K.)

  1. Quality Management for Neutron Transmutation Doping of Silicon Ingot in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ki-Doo; Kim, Ji-Uk; Yun, Hwa-Kyung; Lim, Chul-Hong; Kim, Young-Chil; Kim, Myong-Seop; Park, Sang-Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    By using this doping method, silicon semiconductors with extremely uniform dopant distributions can be produced, and this is the dominant advantage of NTD compared with a conventional chemical doping. Good uniformity of a dopant concentration is usually required for high power applications such as thyristor (SCR), IGBT, IGCT and GTO and for special sensors. Achieving an accurate neutron fluence corresponding to a target resistivity as well as a uniform irradiation is the prime target of a neutron irradiation for NTD. Generally, in order to reach an accurate neutron fluence, a real time neutron flux is monitored by a neutron detector such as a Self-powered Neutron Detector(SPND). And, after an irradiation, the total irradiation fluence is confirmed by measuring the absolute activity of a neutron activation sample that has been irradiated with a silicon ingot, and thus the SPND can be properly calibrated. Excellent irradiation uniformity and a high accuracy for a target neutron dose have been achieved from the early works of NTD. However, to maintain this excellent quality, the neutron irradiation fluence should be continuously modified and controlled. So, in this work, an activity to maintain the irradiation quality is introduced.

  2. Determination of the thermal and epithermal neutron sensitivities of an LBO chamber

    Energy Technology Data Exchange (ETDEWEB)

    Endo, Satoru; Kotani, Kei; Kajimoto, Tsuyoshi; Tanaka, Kenichi [Hiroshima University, Quantum Energy Applications, Graduate School of Engineering, Higashi-Hiroshima (Japan); Sato, Hitoshi; Nakajima, Erika [Ibaraki Prefectural University of Health Science, Radiological Sciences, Ibaraki (Japan); Shimazaki, Takuto [Hiroshima University, Quantum Energy Applications, Graduate School of Engineering, Higashi-Hiroshima (Japan); Delta Kogyo Co., Ltd., Hiroshima (Japan); Suda, Mitsuru; Hamano, Tsuyoshi [National Institute of Radiological Sciences, Chiba-Shi, Chiba (Japan); Hoshi, Masaharu [Hiroshima University, Institute for Peace Science, Hiroshima (Japan)

    2017-08-15

    An LBO (Li{sub 2}B{sub 4}O{sub 7}) walled ionization chamber was designed to monitor the epithermal neutron fluence in boron neutron capture therapy clinical irradiation. The thermal and epithermal neutron sensitivities of the device were evaluated using accelerator neutrons from the {sup 9}Be(d, n) reaction at a deuteron energy of 4 MeV (4 MeV d-Be neutrons). The response of the chamber in terms of the electric charge induced in the LBO chamber was compared with the thermal and epithermal neutron fluences measured using the gold-foil activation method. The thermal and epithermal neutron sensitivities obtained were expressed in units of pC cm{sup 2}, i.e., from the chamber response divided by neutron fluence (cm{sup -2}). The measured LBO chamber sensitivities were 2.23 x 10{sup -7} ± 0.34 x 10{sup -7} (pC cm{sup 2}) for thermal neutrons and 2.00 x 10{sup -5} ± 0.12 x 10{sup -5} (pC cm{sup 2}) for epithermal neutrons. This shows that the LBO chamber is sufficiently sensitive to epithermal neutrons to be useful for epithermal neutron monitoring in BNCT irradiation. (orig.)

  3. Neutron--neutron logging

    International Nuclear Information System (INIS)

    Allen, L.S.

    1977-01-01

    A borehole logging tool includes a steady-state source of fast neutrons, two epithermal neutron detectors, and two thermal neutron detectors. A count rate meter is connected to each neutron detector. A first ratio detector provides an indication of the porosity of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two epithermal neutron detectors. A second ratio detector provides an indication of both porosity and macroscopic absorption cross section of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two thermal neutron detectors. By comparing the signals of the two ratio detectors, oil bearing zones and salt water bearing zones within the formation being logged can be distinguished and the amount of oil saturation can be determined. 6 claims, 2 figures

  4. Investigating multi-objective fluence and beam orientation IMRT optimization

    Science.gov (United States)

    Potrebko, Peter S.; Fiege, Jason; Biagioli, Matthew; Poleszczuk, Jan

    2017-07-01

    Radiation Oncology treatment planning requires compromises to be made between clinical objectives that are invariably in conflict. It would be beneficial to have a ‘bird’s-eye-view’ perspective of the full spectrum of treatment plans that represent the possible trade-offs between delivering the intended dose to the planning target volume (PTV) while optimally sparing the organs-at-risk (OARs). In this work, the authors demonstrate Pareto-aware radiotherapy evolutionary treatment optimization (PARETO), a multi-objective tool featuring such bird’s-eye-view functionality, which optimizes fluence patterns and beam angles for intensity-modulated radiation therapy (IMRT) treatment planning. The problem of IMRT treatment plan optimization is managed as a combined monolithic problem, where all beam fluence and angle parameters are treated equally during the optimization. To achieve this, PARETO is built around a powerful multi-objective evolutionary algorithm, called Ferret, which simultaneously optimizes multiple fitness functions that encode the attributes of the desired dose distribution for the PTV and OARs. The graphical interfaces within PARETO provide useful information such as: the convergence behavior during optimization, trade-off plots between the competing objectives, and a graphical representation of the optimal solution database allowing for the rapid exploration of treatment plan quality through the evaluation of dose-volume histograms and isodose distributions. PARETO was evaluated for two relatively complex clinical cases, a paranasal sinus and a pancreas case. The end result of each PARETO run was a database of optimal (non-dominated) treatment plans that demonstrated trade-offs between the OAR and PTV fitness functions, which were all equally good in the Pareto-optimal sense (where no one objective can be improved without worsening at least one other). Ferret was able to produce high quality solutions even though a large number of parameters

  5. Neutron irradiation control in the neutron transmutation doping process in HANARO using SPND

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Gi-Doo; Kim, Myong-Seop [Korea Atomic Energy Research Institute, Yuseong, Daejeon, 305-353, (Korea, Republic of)

    2015-07-01

    The neutron irradiation control method by using self-powered neutron detector (SPND) is developed for the neutron transmutation doping (NTD) application in HANARO. An SPND is installed at a fixed position of the upper part of the sleeve in HANARO NTD hole for real-time monitoring of the neutron irradiation. It is confirmed that the SPND is significantly affected by the in-core condition and surroundings of the facility. Furthermore, the SPND signal changes about 15% throughout a whole cycle according to the change of the control rod position. But, it is also confirmed that the variation of the neutron flux on the silicon ingots inside the irradiation can is not so big while moving of the control rod. Accordingly, the relationship between the ratio of the neutron flux to the SPND signal output and the control rod position is established. In this procedure, the neutron flux measurement by using zirconium foil is utilized. The real NTD irradiation experiments are performed using the established relationship. The irradiated neutron fluence can be controlled within ±1.3% of the target one. The mean value of the irradiation/target ratio of the fluence is 0.9992, and the standard deviation is 0.0071. Thus, it is confirmed that the extremely accurate irradiation would be accomplished. This procedure can be useful for the SPND application installed at the fixed position to the field requiring the extremely high accuracy. (authors)

  6. Benchmark experiments on neutron streaming through JET Torus Hall penetrations

    Science.gov (United States)

    Batistoni, P.; Conroy, S.; Lilley, S.; Naish, J.; Obryk, B.; Popovichev, S.; Stamatelatos, I.; Syme, B.; Vasilopoulou, T.; contributors, JET

    2015-05-01

    Neutronics experiments are performed at JET for validating in a real fusion environment the neutronics codes and nuclear data applied in ITER nuclear analyses. In particular, the neutron fluence through the penetrations of the JET torus hall is measured and compared with calculations to assess the capability of state-of-art numerical tools to correctly predict the radiation streaming in the ITER biological shield penetrations up to large distances from the neutron source, in large and complex geometries. Neutron streaming experiments started in 2012 when several hundreds of very sensitive thermo-luminescence detectors (TLDs), enriched to different levels in 6LiF/7LiF, were used to measure the neutron and gamma dose separately. Lessons learnt from this first experiment led to significant improvements in the experimental arrangements to reduce the effects due to directional neutron source and self-shielding of TLDs. Here we report the results of measurements performed during the 2013-2014 JET campaign. Data from new positions, at further locations in the South West labyrinth and down to the Torus Hall basement through the air duct chimney, were obtained up to about a 40 m distance from the plasma neutron source. In order to avoid interference between TLDs due to self-shielding effects, only TLDs containing natural Lithium and 99.97% 7Li were used. All TLDs were located in the centre of large polyethylene (PE) moderators, with natLi and 7Li crystals evenly arranged within two PE containers, one in horizontal and the other in vertical orientation, to investigate the shadowing effect in the directional neutron field. All TLDs were calibrated in the quantities of air kerma and neutron fluence. This improved experimental arrangement led to reduced statistical spread in the experimental data. The Monte Carlo N-Particle (MCNP) code was used to calculate the air kerma due to neutrons and the neutron fluence at detector positions, using a JET model validated up to the

  7. Study of natural diamond detector spectrometric properties under neutron irradiation

    CERN Document Server

    Alekseyev, A B; Kaschuck, Y; Krasilnikov, A; Portnov, D; Tugarinov, S

    2002-01-01

    Natural diamond detector (NDD) performance was studied up to a neutron fluence of 10 sup 1 sup 5 neutron/cm sup 2. The variations of the NDD spectrometric response to incident alpha-particles from sup 2 sup 4 sup 1 Am source after exposure to fast neutron fluences up to 3x10 sup 1 sup 6 n/cm sup 2 were examined. No significant variations up to the level of 10 sup 1 sup 4 n/cm sup 2 were observed. Degradation of charge collection efficiency at higher fluences is reported. No remarkable increase of the NDD leakage current and count rate change had been observed up to a neutron fluence of 3x10 sup 1 sup 6 n/cm sup 2. The charge collection efficiency variations of neutron irradiated diamond spectrometer were studied ex situ under gamma-rays, beta-radiation and visible light excitation. Charge collection efficiency restoration up to 75% level and the NDD performance stabilization by extrinsic low-intensity visible light (550 nm

  8. Neutron irradiation effect of thermally-sensitized stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Hide, Kouitiro [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

    1998-03-01

    Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated thermally-sensitized Type 304 Stainless Steels (SSs) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 1.1 x 10{sup 24} n/m{sup 2} (E > 1MeV) at the light water reactor temperature in the Japan Material Test Reactor. The irradiated specimens were examined by slow strain rate stress corrosion cracking tests in 290degC pure water of 0.2 ppm dissolved oxygen concentration and microhardness measurements. The IGSCC susceptibility of the irradiated specimens increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}. From an attempt to correlate the IGSCC susceptibility with the mechanical properties, an excellent correlation was identified between the susceptibility and microhardness increments at the grain boundary relative to the grain center. While intergranular corrosion rate of thermally sensitized SS increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}, that of solution annealed SS did not change. The incremental grain boundary hardening and degradation of intergranular corrosion resistance may presumably be the major factors affecting IGSCC performance. (author)

  9. Neutron induced degradation in nitrided pyrogenic field oxide MOS capacitors

    CERN Document Server

    Vaidya, S J; Shaikh, A M; Chandorkar, A N

    2002-01-01

    Neutron induced oxide charge trapping and generation of interface states in MOS capacitors with pyrogenic and nitrided pyrogenic field oxides have been studied. In order to assess the damage due to neutrons alone, it is necessary to account for the damage produced by the accompanying gamma rays from neutron radiation. This is done by measuring the intensity of gamma radiation accompanying neutrons at different neutron fluences at the irradiation position. MOS capacitor structures were subjected to neutron radiation in a swimming pool type of reactor. Other samples from the same batch were then subjected to an equivalent dose of gamma radiation from a Co sup 6 sup 0 source. The difference in the damage observed was used to characterize the damage caused by neutrons. It is observed that neutrons, though uncharged, are capable of causing ionization damage. This damage is found to be significant when the radiation is performed under biased conditions. Nitridation in different ambients is found to improve the radi...

  10. On-Site Determination and Monitoring of Real-Time Fluence Delivery for an Operating UV Reactor Based on a True Fluence Rate Detector.

    Science.gov (United States)

    Li, Mengkai; Li, Wentao; Qiang, Zhimin; Blatchley, Ernest R

    2017-07-18

    At present, on-site fluence (distribution) determination and monitoring of an operating UV system represent a considerable challenge. The recently developed microfluorescent silica detector (MFSD) is able to measure the approximate true fluence rate (FR) at a fixed position in a UV reactor that can be compared with a FR model directly. Hence it has provided a connection between model calculation and real-time fluence determination. In this study, an on-site determination and monitoring method of fluence delivery for an operating UV reactor was developed. True FR detectors, a UV transmittance (UVT) meter, and a flow rate meter were used for fundamental measurements. The fluence distribution, as well as reduction equivalent fluence (REF), 10th percentile dose in the UV fluence distribution (F 10 ), minimum fluence (F min ), and mean fluence (F mean ) of a test reactor, was calculated in advance by the combined use of computational fluid dynamics and FR field modeling. A field test was carried out on the test reactor for disinfection of a secondary water supply. The estimated real-time REF, F 10 , F min , and F mean decreased 73.6%, 71.4%, 69.6%, and 72.9%, respectively, during a 6-month period, which was attributable to lamp output attenuation and sleeve fouling. The results were analyzed with synchronous data from a previously developed triparameter UV monitoring system and water temperature sensor. This study allowed demonstration of an accurate method for on-site, real-time fluence determination which could be used to enhance the security and public confidence of UV-based water treatment processes.

  11. Spectral stratigraphy

    Science.gov (United States)

    Lang, Harold R.

    1991-01-01

    A new approach to stratigraphic analysis is described which uses photogeologic and spectral interpretation of multispectral remote sensing data combined with topographic information to determine the attitude, thickness, and lithology of strata exposed at the surface. The new stratigraphic procedure is illustrated by examples in the literature. The published results demonstrate the potential of spectral stratigraphy for mapping strata, determining dip and strike, measuring and correlating stratigraphic sequences, defining lithofacies, mapping biofacies, and interpreting geological structures.

  12. Thermal neutron calibration channel at LNMRI/IRD

    International Nuclear Information System (INIS)

    Astuto, A.; Salgado, A.P.; Lopes, R.T.; Leite, S.P.; Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W.

    2014-01-01

    The Brazilian Metrology Laboratory of Ionizing Radiations (LNMRI) standard thermal neutron flux facility was designed to provide uniform neutron fluence for calibration of small neutron detectors and individual dosemeters. This fluence is obtained by neutron moderation from four 241 Am-Be sources, each with 596 GBq, in a facility built with blocks of graphite/paraffin compound and high-purity carbon graphite. This study was carried out in two steps. In the first step, simulations using the MCNPX code on different geometric arrangements of moderator materials and neutron sources were performed. The quality of the resulting neutron fluence in terms of spectrum, cadmium ratio and gamma-neutron ratio was evaluated. In the second step, the system was assembled based on the results obtained on the simulations, and new measurements are being made. These measurements will validate the system, and other intercomparisons will ensure traceability to the International System of Units. The pile construction form using blocks allows distinct arrangements for new studies and possibilities of other LNMRI reference fields. The results can be predicted by the simulation used in this work. Different number of each type of blocks and sources can be used. The main difference observed between the final measurement and simulation results might be due to the difference in composition of paraffin blocks used in assembling the pile. In order to confirm the thermal neutron field and fluence rate in the central chamber (inside the channel) that will be used to irradiate small neutron detectors, it is necessary to use another quantification method such as the gold foils activation with measurement traceability. It will be performed in a future stage. (authors)

  13. Benchmarking the MCNP code for Monte Carlo modelling of an in vivo neutron activation analysis system.

    Science.gov (United States)

    Natto, S A; Lewis, D G; Ryde, S J

    1998-01-01

    The Monte Carlo computer code MCNP (version 4A) has been used to develop a personal computer-based model of the Swansea in vivo neutron activation analysis (IVNAA) system. The model included specification of the neutron source (252Cf), collimators, reflectors and shielding. The MCNP model was 'benchmarked' against fast neutron and thermal neutron fluence data obtained experimentally from the IVNAA system. The Swansea system allows two irradiation geometries using 'short' and 'long' collimators, which provide alternative dose rates for IVNAA. The data presented here relate to the short collimator, although results of similar accuracy were obtained using the long collimator. The fast neutron fluence was measured in air at a series of depths inside the collimator. The measurements agreed with the MCNP simulation within the statistical uncertainty (5-10%) of the calculations. The thermal neutron fluence was measured and calculated inside the cuboidal water phantom. The depth of maximum thermal fluence was 3.2 cm (measured) and 3.0 cm (calculated). The width of the 50% thermal fluence level across the phantom at its mid-depth was found to be the same by both MCNP and experiment. This benchmarking exercise has given us a high degree of confidence in MCNP as a tool for the design of IVNAA systems.

  14. Neutron spectrum unfolding: Pt. 2

    International Nuclear Information System (INIS)

    Matiullah; Wiyaja, D.S.; Berzonis, M.A.; Bondars, H.; Lapenas, A.A.; Kudo, K.; Majeed, A.; Durrani, S.A.

    1991-01-01

    In Part I of this paper, we described the use of the computer code SAIPS in neutron spectrum unfolding. Here in Part II, we present our experimental work carried out to study the shape of the neutron spectrum in different experimental channels of a 5 MW light-water cooled and moderated research reactor. The spectral neutron flux was determined using various fission foils (placed in close contact with mica track detectors) and activation detectors. From the measured activities, the neutron spectrum was unfolded by SAIPS. (author)

  15. Beryllium neutron activation detector for pulsed DD fusion sources

    International Nuclear Information System (INIS)

    Talebitaher, A.; Springham, S.V.; Rawat, R.S.; Lee, P.

    2011-01-01

    A compact fast neutron detector based on beryllium activation has been developed to perform accurate neutron fluence measurements on pulsed DD fusion sources. It is especially well suited to moderate repetition-rate ( 9 Be(n,α) 6 He cross-section, energy calibration of the proportional counters, and numerical simulations of neutron interactions and beta-particle paths using MCNP5. The response function R(E n ) is determined over the neutron energy range 2-4 MeV. The count rate capability of the detector has been studied and the corrections required for high neutron fluence measurements are discussed. For pulsed DD neutron fluencies >3×10 4 cm -2 , the statistical uncertainty in the fluence measurement is better than 1%. A small plasma focus device has been employed as a pulsed neutron source to test two of these new detectors, and their responses are found to be practically identical. Also the level of interfering activation is found to be sufficiently low as to be negligible.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-06

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

  17. Scanning tunneling spectroscopy on neutron irradiated MgB2 thin films

    International Nuclear Information System (INIS)

    Di Capua, Roberto; Salluzzo, Marco; Vaglio, Ruggero; Ferdeghini, Carlo; Ferrando, Valeria; Putti, Marina; Xi Xiaoxing; Aebersold, Hans U.

    2007-01-01

    Neutron irradiation was performed on MgB 2 thin films grown by hybrid physical chemical vapor deposition. Samples irradiated with different neutron fluences, having different critical temperatures, were studied by scanning tunneling spectroscopy in order to investigate the effect of the introduced disorder on the superconducting and spectroscopic properties. A monotonic increase of the π gap with increasing disorder was found

  18. Scanning tunneling spectroscopy on neutron irradiated MgB{sub 2} thin films

    Energy Technology Data Exchange (ETDEWEB)

    Di Capua, Roberto [University of Napoli and CNR-INFM/Coherentia, Via Cinthia, Naples I-80126 (Italy)], E-mail: rdicapua@na.infn.it; Salluzzo, Marco; Vaglio, Ruggero [University of Napoli and CNR-INFM/Coherentia, Via Cinthia, Naples I-80126 (Italy); Ferdeghini, Carlo [CNR-INFM/LAMIA, Via Dodecaneso 33, Genova I-16146 (Italy); Ferrando, Valeria [CNR-INFM/LAMIA, Via Dodecaneso 33, Genova I-16146 (Italy); Pennsylvania State University, University Park, PA 16802 (United States); Putti, Marina [CNR-INFM/LAMIA, Via Dodecaneso 33, Genova I-16146 (Italy); Xi Xiaoxing [Pennsylvania State University, University Park, PA 16802 (United States); Aebersold, Hans U. [Paul Scherrer Institut, Villigen CH-5232 (Switzerland)

    2007-09-01

    Neutron irradiation was performed on MgB{sub 2} thin films grown by hybrid physical chemical vapor deposition. Samples irradiated with different neutron fluences, having different critical temperatures, were studied by scanning tunneling spectroscopy in order to investigate the effect of the introduced disorder on the superconducting and spectroscopic properties. A monotonic increase of the {pi} gap with increasing disorder was found.

  19. Ultracold neutrons

    International Nuclear Information System (INIS)

    Steenstrup, S.

    Briefly surveys recent developments in research work with ultracold neutrons (neutrons of very low velocity, up to 10 m/s at up to 10 -7 eV and 10 -3 K). Slow neutrons can be detected in an ionisation chamber filled with B 10 F 3 . Very slow neutrons can be used for investigations into the dipole moment of neutrons. Neutrons of large wave length have properties similar to those of light. The limit angle for total reflection is governed by the wave length and by the material. Total reflection can be used to filter ultracold neutrons out of the moderator material of a reactor. Total reflection can also be used to store ultracold neutrons but certain problems with storage have not yet been clarified. Slow neutrons can be made to lose speed in a neutron turbine, and come out as ultracold neutrons. A beam of ultracold neutrons could be used in a neutron microscope. (J.S.)

  20. Fast neutron irradiation and thermal properties of doped nonstoichiometric lithium potassium sulphate crystals

    International Nuclear Information System (INIS)

    Kassem, M.E.; Gomaa, N.G.; El-Khatib, A.M.

    1990-01-01

    The influence of point defects introduced by fast neutron irradiations with neutron fluences up to 1.08 x 10 10 n/cm 2 on the thermal properties of pure and doped Li 1.4 K 0.6 SO 4 single crystals are studied in the vicinity of high temperature phase transition at 705 K. The temperature dependence of specific heat is found to be shifted towards lower temperature with the increase of neutron fluence, and can be affected by the presence of Cu 2+ dopant. The change in the value of the specific heat can be attributed to the presence of internal strain generated inside the crystal. (author)

  1. Characteristics of thermal neutron calibration fields using a graphite pile

    International Nuclear Information System (INIS)

    Uchita, Yoshiaki; Saegusa, Jun; Kajimoto, Yoichi; Tanimura, Yoshihiko; Shimizu, Shigeru; Yoshizawa, Michio

    2005-03-01

    The Facility of Radiation Standards of Japan Atomic Energy Research Institute is equipped with thermal neutron fields for calibrating area and personal neutron dosemeters. The fields use moderated neutrons leaked from a graphite pile in which radionuclide sources are placed. In January 2003, we have renewed the pile with some modifications in its size. In accordance with the renewal, we measured and calculated thermal neutron fluence rates, neutron energy distributions and angular distributions of the fields. The thermal neutron fluence rates of the ''inside-pile fields'' and the outside-pile fields'' were determined by the gold foil activation method. The neutron energy distributions of the outside-pile fields were also measured with the Bonner multi-sphere spectrometer system. The contributions of epithermal and fast neutrons to the total dose-equivalents were 9% in the southern outside-pile field and 12% in the western outside-pile field. The personal dose-equivalents, H p,slab (10, α), in the outside-pile fields are evaluated by considering the calculated angular distributions of incoming neutrons. The H p,slab (10, α) was found to be about 40% higher than the value in assuming the unidirectional neutron between the pile and the test point. (author)

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

    International Nuclear Information System (INIS)

    2001-01-01

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

  3. Fast neutron irradiation induced changes in the optical and thermal properties of modified polyvinyl chloride

    Energy Technology Data Exchange (ETDEWEB)

    Abou Taleb, W.M. [Alexandria Univ. (Egypt); Madi, N.K.; Kassem, M.E.; El-Khatib, A.M. [Alexandria Univ. (Egypt). Dept. of Physics

    1996-05-01

    The effect of both dopant and neutron radiation on the optical and thermal properties of polyvinyl chloride (PVC) has been studied. The doped samples with Pb and Cd were irradiated with a 14 MeV-neutron fluence in the range 7-28.8 x 10{sup 9} n/cm{sup 2}. The optical energy gap E{sub op} exhibits a significant dependence on the type of additive and the neutron irradiation fluence. The specific heat at constant pressure C{sub p} showed a nonmonotonical change with radiation fluence. The results of this study show that PVC:Pb behaves as a crystalline structure which is only slightly affected by neutron irradiation, while PVC:Cd is highly affected. (author).

  4. Fast neutron irradiation induced changes in the optical and thermal properties of modified polyvinyl chloride

    International Nuclear Information System (INIS)

    Abou Taleb, W.M.; Madi, N.K.; Kassem, M.E.; El-Khatib, A.M.

    1996-01-01

    The effect of both dopant and neutron radiation on the optical and thermal properties of polyvinyl chloride (PVC) has been studied. The doped samples with Pb and Cd were irradiated with a 14 MeV-neutron fluence in the range 7-28.8 x 10 9 n/cm 2 . The optical energy gap E op exhibits a significant dependence on the type of additive and the neutron irradiation fluence. The specific heat at constant pressure C p showed a nonmonotonical change with radiation fluence. The results of this study show that PVC:Pb behaves as a crystalline structure which is only slightly affected by neutron irradiation, while PVC:Cd is highly affected. (author)

  5. Test of Fibre Bragg Gratings samples under High Fast Neutrons Fluence

    Directory of Open Access Journals (Sweden)

    Cheymol G.

    2018-01-01

    The measurements show that for nearly all gratings the Bragg peak remains visible after the irradiation, and that Radiation Induced Bragg Wavelength Shifts (RI-BWSs vary from few pm (equivalent to an error of less than 1°C for a temperature sensor to nearly 1 nm (equivalent to 100°C depending of the FBG types. High RI-BWSs could indeed be expected when considering the huge refractive index variation and compaction of the bare fibre samples that have been measured by other techniques. Post writing thermal annealing is confirmed as a key parameter in order to obtain a more radiation tolerant FBG. Our results show that specific annealing regimes allow making FGBs suitable to perform temperature measurements in a MTR experiment.

  6. Measurement of peak fluence of neutron beams using Bi-fission ...

    Indian Academy of Sciences (India)

    Home; Journals; Pramana – Journal of Physics; Volume 78; Issue 3 ... useful for personal dosimeters and for flux/dose determination of high-energy particles ... Department of Physics, School of Basic & Applied Sciences, Shobhit University, ...

  7. Measurement of the neutron fluence on the spallation source at Dubna

    Czech Academy of Sciences Publication Activity Database

    Adam, Jindřich; Barashenkov, V. S.; Ganesan, S.; Golovatiouk, S.; Krivopustov, M. I.; Kumar, V.; Kumawat, H.; Palsania, HS.; Pronskikh, V. S.; Sharma, A.; Tsoupko-Sitnikov, V. M.; Vladimirova, NM.; Westmeier, W.

    2005-01-01

    Roč. 70, č. 3 (2005), s. 127-132 ISSN 0932-3902 R&D Projects: GA MŠk(CZ) 1P04LA213 Keywords : beams Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.378, year: 2005

  8. Neutron and gamma-ray transport experiments in liquid air

    International Nuclear Information System (INIS)

    Farley, W.E.

    1976-01-01

    Accurate estimates of neutron and gamma radiations from a nuclear explosion and their subsequent transport through the atmosphere are vital to nuclear-weapon employment studies: i.e., for determining safety radii for aircraft crews, casualty and collateral-damage risk radii for tactical weapons, and the kill range from a high-yield defensive burst for a maneuvering reentry vehicle. Radiation transport codes, such as the Laboratory's TARTNP, are used to calculate neutron and gamma fluences. Experiments have been performed to check and update these codes. Recently, a 1.3-m-radius liquid-air (21 percent oxygen) sphere, with a pulsed source of 14-MeV neutrons at its center, was used to measure the fluence and spectra of emerging neutrons and secondary gamma rays. Comparison of measured radiation dose with TARTNP showed agreement within 10 percent

  9. Experimental characterization of semiconductor-based thermal neutron detectors

    Energy Technology Data Exchange (ETDEWEB)

    Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Bortot, D.; Pola, A.; Introini, M.V.; Lorenzoli, M. [Politecnico di Milano, Dipartimento di Energia, via La Masa 34, 20156 Milano (Italy); INFN—Milano, Via Celoria 16, 20133 Milano (Italy); Gómez-Ros, J.M. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain); Sacco, D. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); INAIL—DIT, Via di Fontana Candida 1, 00040 Monteporzio Catone (Italy); Esposito, A.; Gentile, A.; Buonomo, B. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Palomba, M.; Grossi, A. [ENEA Triga RC-1C.R. Casaccia, via Anguillarese 301, 00060 S. Maria di Galeria, Roma (Italy)

    2015-04-21

    In the framework of NESCOFI@BTF and NEURAPID projects, active thermal neutron detectors were manufactured by depositing appropriate thickness of {sup 6}LiF on commercially available windowless p–i–n diodes. Detectors with different radiator thickness, ranging from 5 to 62 μm, were manufactured by evaporation-based deposition technique and exposed to known values of thermal neutron fluence in two thermal neutron facilities exhibiting different irradiation geometries. The following properties of the detector response were investigated and presented in this work: thickness dependence, impact of parasitic effects (photons and epithermal neutrons), linearity, isotropy, and radiation damage following exposure to large fluence (in the order of 10{sup 12} cm{sup −2})

  10. Calculation of neutron and gamma-ray energy spectra in liquid air and liquid nitrogen due to 14-MeV neutron and californium-252 sources

    International Nuclear Information System (INIS)

    Straker, E.A.; Gritzner, M.L.; Harris, L. Jr.

    1978-01-01

    Calculations of neutron and gamma-ray fluences from 14-MeV neutron and 252 Cf sources in liquid air and liquid nitrogen have been performed. These calculations were made specifically for comparison with experimental data measured at Stohl, Federal Republic of Germany. The discrete-ordinates method was utilized with neutron and gamma-ray cross sections from ENDF/B-IV. One-dimensional calculational models were developed for the sources and tank. Limited comparisons are made with experimental data

  11. Construction of monoenergetic neutron calibration fields using 45Sc(p, n)45Ti reaction at JAEA.

    Science.gov (United States)

    Tanimura, Y; Saegusa, J; Shikaze, Y; Tsutsumi, M; Shimizu, S; Yoshizawa, M

    2007-01-01

    The 8 and 27 keV monoenergetic neutron calibration fields have been developed by using (45)Sc(p, n)(45)Ti reaction. Protons from a 4-MV Pelletron accelerator are used to bombard a thin scandium target evaporated onto a platinum disc. The proton energies are finely adjusted to the resonance to generate the 8 and 27 keV neutrons by applying a high voltage to the target assemblies. The neutron energies were measured using the time-of-flight method with a lithium glass scintillation detector. The neutron fluences at a calibration point located at 50 cm from the target were evaluated using Bonner spheres. A long counter was placed at 2.2 m from the target and at 60 degrees to the direction of the proton beam in order to monitor the fluence at the calibration point. Fluence and dose equivalent rates at the calibration point are sufficient to calibrate many types of the neutron survey metres.

  12. Characterization of thermal neutron fields for calibration of neutron monitors in accordance with great equivalent dose environment H⁎(10)

    International Nuclear Information System (INIS)

    Silva, Larissa P. S. da; Silva, Felipe S.; Fonseca, Evaldo S.; Patrao, Karla C.S.; Pereira, Walsan W.

    2017-01-01

    The Laboratório Brasileiro de Nêutrons do Instituto de Radioproteção e Dosimetria (IRD/CNEN) has developed and built a thermal neutron flux facility to provide neutron fluence for dosimeters (Astuto, 2014). This fluency is obtained by four 16 Ci sources 241 AmBe (α, n) positioned around the channel positioned in the center of the Thermal Flow Unit (UFT). The UFT was built with blocks of paraffin with graphite addition and graphite blocks of high purity to obtain a central field with a homogeneous thermal neutron fluence for calibration purposes with the following measurements: 1.2 x 1.2 x 1.2 m 3 . The objective of this work is to characterize several points, in the thermal energy range, in terms of the equivalent ambient dose quantity H⁎(10) for calibration and irradiation of monitors neutrons

  13. Constant-Fluence Area Scaling for Laser Propulsion

    International Nuclear Information System (INIS)

    Sinko, John E.

    2008-01-01

    A series of experiments was conducted on polyoxymethylene (POM, trade name Delrin registered ) propellants in air at atmospheric pressure. A TEA CO 2 laser with maximum output power up to 20 J was used to deliver 300 ns pulses of 10.6 μm radiation to POM targets. Ablation at a constant fluence and a range of spot areas was achieved by varying combinations of the laser energy and spot size. Relevant empirical scaling laws governing laser propulsion parameters such as the momentum coupling coefficient (C m ) and specific impulse (I sp ) for spot areas within a range of about 0.05-0.25 cm 2 are presented. Experimental measurements of imparted impulse, C m , I sp , and ablated mass per pulse were made using dynamic piezoelectric force sensors and a scientific balance. Finally, Schlieren ICCD imaging of shock waves and vapor plumes was performed and analyzed

  14. Database of episode-integrated solar energetic proton fluences

    Science.gov (United States)

    Robinson, Zachary D.; Adams, James H.; Xapsos, Michael A.; Stauffer, Craig A.

    2018-04-01

    A new database of proton episode-integrated fluences is described. This database contains data from two different instruments on multiple satellites. The data are from instruments on the Interplanetary Monitoring Platform-8 (IMP8) and the Geostationary Operational Environmental Satellites (GOES) series. A method to normalize one set of data to one another is presented to create a seamless database spanning 1973 to 2016. A discussion of some of the characteristics that episodes exhibit is presented, including episode duration and number of peaks. As an example of what can be understood about episodes, the July 4, 2012 episode is examined in detail. The coronal mass ejections and solar flares that caused many of the fluctuations of the proton flux seen at Earth are associated with peaks in the proton flux during this episode. The reasoning for each choice is laid out to provide a reference for how CME and solar flares associations are made.

  15. The Meteoroid Fluence at Mars Due to Comet Siding Spring

    Science.gov (United States)

    Moorhead, Althea V.

    2014-01-01

    Long-period comet C/2013 A1 (Siding Spring) is headed for a close encounter with Mars on 2014 Oct 19. A collision between the comet and the planet has been ruled out, but the comets coma may envelop Mars and its man-made satellites. We present an analytic model of the dust component of cometary comae that describes the spatial distribution of cometary dust and meteoroids and their size distribution. If the coma reaches Mars, we estimate a total incident particle fluence on the planet and its satellites of 0.01 particles per square meter. We compare our model with numerical simulations, data from past comet missions, and recent Siding Spring observations.

  16. Database of episode-integrated solar energetic proton fluences

    Directory of Open Access Journals (Sweden)

    Robinson Zachary D.

    2018-01-01

    Full Text Available A new database of proton episode-integrated fluences is described. This database contains data from two different instruments on multiple satellites. The data are from instruments on the Interplanetary Monitoring Platform-8 (IMP8 and the Geostationary Operational Environmental Satellites (GOES series. A method to normalize one set of data to one another is presented to create a seamless database spanning 1973 to 2016. A discussion of some of the characteristics that episodes exhibit is presented, including episode duration and number of peaks. As an example of what can be understood about episodes, the July 4, 2012 episode is examined in detail. The coronal mass ejections and solar flares that caused many of the fluctuations of the proton flux seen at Earth are associated with peaks in the proton flux during this episode. The reasoning for each choice is laid out to provide a reference for how CME and solar flares associations are made.

  17. Online measurement of fluence and position for protontherapy beams

    International Nuclear Information System (INIS)

    Benati, C.; Boriano, A.

    2004-01-01

    Tumour therapy with proton beams has been used for several decades in many centers with very good results in terms of local control and overall survival. Typical pathologies treated with this technique are located in head and neck, eye, prostate and in general at big depths or close to critical organs. The Experimental Physics Department of the University of Turin and the local Section of INFN, in collaboration with INFN Laboratori Nazionali del Sud Catania and Centre de Protontherapie de Orsay Paris, have developed detector systems that allow the measurement of beam position and fluence, obtained in real time during beam delivery. The Centre in Catania (CATANA: Centro di AdroTerapia ed Applicazioni Nucleari Avanzate) has been treating patients with eye pathologies since spring 2002 using a superconducting cyclotron accelerating protons up to 62 MeV

  18. Online measurement of fluence and position for protontherapy beams

    Energy Technology Data Exchange (ETDEWEB)

    Benati, C.; Boriano, A. [Torino Univ., Torino (Italy). Dipartimento di Fisica Sperimentale; Bourhaleb, F. [TERA Foundation, Novara (Italy)] [and others

    2004-10-01

    Tumour therapy with proton beams has been used for several decades in many centers with very good results in terms of local control and overall survival. Typical pathologies treated with this technique are located in head and neck, eye, prostate and in general at big depths or close to critical organs. The Experimental Physics Department of the University of Turin and the local Section of INFN, in collaboration with INFN Laboratori Nazionali del Sud Catania and Centre de Protontherapie de Orsay Paris, have developed detector systems that allow the measurement of beam position and fluence, obtained in real time during beam delivery. The Centre in Catania (CATANA: Centro di AdroTerapia ed Applicazioni Nucleari Avanzate) has been treating patients with eye pathologies since spring 2002 using a superconducting cyclotron accelerating protons up to 62 MeV.

  19. High fluence effects on ion implantation stopping and range

    International Nuclear Information System (INIS)

    Selvi, S.; Tek, Z.; Oeztarhan, A.; Akbas, N.; Brown, I.G.

    2005-01-01

    We have developed a code STOPPO which can be used to modify the more-widely used ion implantation codes to more accurately predict the mean nuclear and electronic stopping power, preferential sputtering and range of heavy ions in monatomic target materials. In our simulations an effective atomic number and effective atomic mass are introduced into conveniently available analytical stopping cross-sections and a better fitting function for preferential sputtering yield is carefully evaluated for each ion implantation. The accuracy of the code confirmed experimentally by comparison with measured Rutherford backscattering spectrometry (RBS) concentration profiles for 130 keV Zr ions implanted into Be to fluences of 1 x 10 17 , 2 x 10 17 and 4 x 10 17 ions/cm 2 . We find a steady increase in the mean nuclear and electronic stopping powers of the target; the increase in nuclear stopping power is much greater than the increase in electronic stopping power

  20. Online measurement of fluence and position for protontherapy beams

    Science.gov (United States)

    Benati, C.; Boriano, A.; Bourhaleb, F.; Cirio, R.; Cirrone, G. A. P.; Cornelius, I.; Cuttone, G.; Donetti, M.; Garelli, E.; Giordanengo, S.; Guérin, L.; La Rosa, A.; Luparia, A.; Marchetto, F.; Martin, F.; Meyroneinc, S.; Peroni, C.; Pittà, G.; Raffaele, L.; Sabini, M. G.; Valastro, L.

    2004-09-01

    Tumour therapy with proton beams has been used for several decades in many centres with very good results in terms of local control and overall survival. Typical pathologies treated with this technique are located in head and neck, eye, prostate and in general at big depths or close to critical organs. The Experimental Physics Department of the University of Turin and the local Section of INFN, in collaboration with INFN Laboratori Nazionali del Sud Catania and Centre de Protontherapie de Orsay Paris, have developed detector systems that allow the measurement of beam position and fluence, obtained in real time during beam delivery. The centre in Catania (CATANA: Centro di AdroTerapia ed Applicazioni Nucleari Avanzate) has been treating patients with eye pathologies since spring 2002 using a superconducting cyclotron accelerating protons up to 62 MeV.This kind of treatments need high-resolution monitor systems and for this reason we have developed a 256-strip segmented ionisation chamber, each strip being 400 μm wide, with a total sensitive area 13×13 cm2. The Centre de Protontherapie de Orsay (CPO) has been operational since 1991 and features a synchrocyclotron used for eye and head and neck tumours with proton beams up to 200 MeV. The monitor system has to work on a large surface and for this purpose we have designed a pixel-segmented ionisation chamber, each pixel being 5×5 mm2, for a total active area of 16×16 cm2. The results obtained with two prototypes of the pixel and strip chambers demonstrate that the detectors allow the measurement of fluence and centre of gravity as requested by clinical specifications.

  1. Measurement of AC electrical conductivity of single crystal Al2O3 during spallation-neutron irradiation

    International Nuclear Information System (INIS)

    Kennedy, J.C. III; Farnum, E.H.; Sommer, W.F.; Clinard, F.W. Jr.

    1993-01-01

    Samples of single crystal Al 2 O 3 , commonly known as sapphire, and polycrystalline Al 2 O 3 were irradiated with spallation neutrons at the Los Alamos Spallation Radiation Effects Facility (LASREF) under various temperature conditions and with a continuously applied alternating electric field. This paper describes the results of measurements on the sapphire samples. Neutron fluence and flux values are estimated values pending recovery and analysis of dosimetry packages. The conductivity increased approximately with the square root of the neutron flux at fluences less than 3 x 10 21 n/m 2 . The increase in conductivity reached saturated levels as high as 2 x 10 -2 (ohm-m) -1 at fluences as low as 2 x 10 22 n/m 2 . Frequency swept impedance measurements indicated a change in the electrical properties from capacitive to resistive behavior with increasing fluence

  2. Fractal characterization of the silicon surfaces produced by ion beam irradiation of varying fluences

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, R.P. [Department of Physics, University of Allahabad, Allahabad, UP 211002 (India); Kumar, T. [Department of Physics, Central University of Haryana, Jant-Pali, Mahendergarh, Haryana 123029 (India); Mittal, A.K. [Department of Physics, University of Allahabad, Allahabad, UP 211002 (India); K Banerjee Centre of Atmospheric and Ocean Studies, University of Allahabad, Allahabad, UP 211002 (India); Dwivedi, S., E-mail: suneetdwivedi@gmail.com [K Banerjee Centre of Atmospheric and Ocean Studies, University of Allahabad, Allahabad, UP 211002 (India); Kanjilal, D. [Inter-University Accelerator Centre, Aruna Asaf Ali Marg, PO Box 10502, New Delhi 110 067 (India)

    2015-08-30

    Highlights: • Fractal analysis of Si(1 0 0) surface morphology at varying ion fluences. • Autocorrelation function and height–height correlation function as fractal measures. • Surface roughness and lateral correlation length increases with ion fluence. • Ripple pattern of the surfaces is found at higher ion fluences. • Wavelength of the ripple surfaces is computed for each fluence. - Abstract: Si (1 0 0) is bombarded with 200 keV Ar{sup +} ion beam at oblique incidence with fluences ranging from 3 × 10{sup 17} ions/cm{sup 2} to 3 × 10{sup 18} ions/cm{sup 2}. The surface morphology of the irradiated surfaces is captured by the atomic force microscopy (AFM) for each ion fluence. The fractal analysis is performed on the AFM images. The autocorrelation function and height–height correlation function are used as fractal measures. It is found that the average roughness, interface width, lateral correlation length as well as roughness exponent increase with ions fluence. The analysis reveals the ripple pattern of the surfaces at higher fluences. The wavelength of the ripple surfaces is computed for each ion fluence.

  3. Heavy Ion Irradiation Fluence Dependence for Single-Event Upsets of NAND Flash Memory

    Science.gov (United States)

    Chen, Dakai; Wilcox, Edward; Ladbury, Raymond; Kim, Hak; Phan, Anthony; Seidleck, Christina; LaBel, Kenneth

    2016-01-01

    We investigated the single-event effect (SEE) susceptibility of the Micron 16 nm NAND flash, and found the single-event upset (SEU) cross section varied inversely with fluence. The SEU cross section decreased with increasing fluence. We attribute the effect to the variable upset sensitivities of the memory cells. The current test standards and procedures assume that SEU follow a Poisson process and do not take into account the variability in the error rate with fluence. Therefore, heavy ion irradiation of devices with variable upset sensitivity distribution using typical fluence levels may underestimate the cross section and on-orbit event rate.

  4. Change in properties of superconducting magnet materials by fusion neutron irradiation

    International Nuclear Information System (INIS)

    Nishimura, Arata; Nishijima, Shigehiro; Takeuchi, Takao; Nishitani, Takeo

    2007-01-01

    A fusion reactor will generate a lot of high energy neutron and much energy will be taken out of the neutrons by a blanket system. Since some neutrons will stream out of a plasma vacuum vessel through neutral beam injection ports and penetrate a blanket system, a superconducting magnet system, which provides high magnetic field to confirm high energy particles, will be irradiated by a certain amount of neutrons. By developing the new NBI system or by reducing the penetration, the neutron fluence to the superconducting magnet will be able to be reduced. However, it is not easy to achieve the lower streaming and penetration at the present. Therefore, investigations on irradiation behavior of superconducting magnet materials are desired and some novel researches have been performed from 1970s. In general, the critical current of the superconducting wire increases under fast neutron environment comparing with that of the non-irradiated wire, and then decreased to almost zero as an increase of neutron fluence. On the other hand, the critical temperature of the wire starts to get down around 10 22 n/m 2 of neutron fluence and the temperature margin will be decreased during the operation by the neutron irradiation. In this paper, some aspects of irradiated materials will be overviewed and general tendency will be discussed focussing on knock-on effect of fast neutron and long range ordering of A15 compounds

  5. Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Niobium

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This test method describes procedures for measuring reaction rates by the activation reaction 93Nb(n,n′)93mNb. 1.2 This activation reaction is useful for monitoring neutrons with energies above approximately 0.5 MeV and for irradiation times up to about 30 years. 1.3 With suitable techniques, fast-neutron reaction rates for neutrons with energy distribution similar to fission neutrons can be determined in fast-neutron fluences above about 1016cm−2. In the presence of high thermal-neutron fluence rates (>1012cm−2·s−1), the transmutation of 93mNb due to neutron capture should be investigated. In the presence of high-energy neutron spectra such as are associated with fusion and spallation sources, the transmutation of 93mNb by reactions such as (n,2n) may occur and should be investigated. 1.4 Procedures for other fast-neutron monitors are referenced in Practice E 261. 1.5 Fast-neutron fluence rates can be determined from the reaction rates provided that the appropriate cross section information ...

  6. Results on Neutron and Gamma Irradiation of Electrolytic Tilmeters

    International Nuclear Information System (INIS)

    Calderon, A.; Calvo, E.; Figueroa, C. F.; Martinez-Rivero, C.; Matorras, F.; Rodrigo, T.; Vila, I.; Virto, A. L.; Alberdi, J.; Arce, P.; Barcala, J. M.; Fernando, A.; Fuentes, J.; Josa, M. I.; Luque, J. M.; Molinero, A.; Navarrate, J.; Valdivieso, P.; Fenyvesi, A.; Molnar, J.

    2004-01-01

    We report on irradiation studies done to a sample of high precision electrolytic tiltmeters with gamma-rays, up to a maximum dose of 150 kGy, an neutrons, up to a maximum fluence 1.5x10''14 cm''2. The effect of the irradiation on their performance is discussed. (Author) 19 refs

  7. Results on Neutron and Gamma Irradiation of Electrolytic Tilmeters

    Energy Technology Data Exchange (ETDEWEB)

    Calderon, A.; Calvo, E.; Figueroa, C. F.; Martinez-Rivero, C.; Matorras, F.; Rodrigo, T.; Vila, I.; Virto, A. L.; Alberdi, J.; Arce, P.; Barcala, J. M.; Fernando, A.; Fuentes, J.; Josa, M. I.; Luque, J. M.; Molinero, A.; Navarrate, J.; Valdivieso, P.; Fenyvesi, A.; Molnar, J.

    2004-07-01

    We report on irradiation studies done to a sample of high precision electrolytic tiltmeters with gamma-rays, up to a maximum dose of 150 kGy, an neutrons, up to a maximum fluence 1.5x10''14 cm''2. The effect of the irradiation on their performance is discussed. (Author) 19 refs.

  8. Epitaxial silicon detectors for particle tracking-Radiation tolerance at extreme hadron fluences

    International Nuclear Information System (INIS)

    Lindstroem, Gunnar; Dolenc, Irena; Fretwurst, Eckhart; Hoenniger, Frank; Kramberger, Gregor; Moll, Michael; Nossarzewska, Elsbieta; Pintilie, Ioana; Roeder, Ralf

    2006-01-01

    Diodes processed on n-type epitaxial silicon with a thickness of 25, 50 and 75 μm had been irradiated with reactor neutrons and high-energy protons (24 GeV/c) up to integrated fluences of Φ eq =10 16 cm -2 . Systematic experiments on radiation-induced damage effects revealed the following results: in contrast to standard and oxygen-enriched float zone (FZ) silicon devices no space charge sign inversion was observed after irradiation. It is shown that the radiation-generated concentration of deep acceptors, dominating the behavior of n-type FZ diodes, is compensated by creation of shallow donors. Thus a positive space charge is maintained throughout the irradiation up to the highest fluence and even during prolonged elevated-temperature annealing cycles. Defect analysis studies using thermally stimulated current measurements attribute the effect to a damage-induced shallow donor at E C -0.23 eV. It is argued that, as in the case of thermal donors, oxygen dimers, out diffusing from the Cz substrate during the diode processing, are responsible precursers. Results from extensive annealing experiments at elevated temperatures are verified by comparison with prolonged room-temperature annealing. These results showed that in contrast to FZ detectors, which always have to be cooled, room-temperature storage during beam off periods of future elementary particle physics experiments would even be beneficial for n-type epi-silicon detectors. A dedicated experiment at CERN-PS had successfully proven this expectation. It was verified, that in such a scenario the depletion voltage for the epi-detector could always be kept at a moderate level throughout the full S-LHC operation (foreseen upgrade of the large hadron collider). Practically no difference with respect to FZ-silicon devices was found in the damage-induced bulk generation current. The charge trapping measured with 90 Sr electrons (mip's) is also almost identical to what was expected. A charge collection efficiency of

  9. Epitaxial silicon detectors for particle tracking-Radiation tolerance at extreme hadron fluences

    Energy Technology Data Exchange (ETDEWEB)

    Lindstroem, Gunnar [Institute for Experimental Physics, University of Hamburg, Hamburg, 22761 (Germany)]. E-mail: gunnar.lindstroem@desy.de; Dolenc, Irena [Jozef Stefan Institute, University of Ljubljana, Ljubljana, 100 (Slovenia); Fretwurst, Eckhart [Institute for Experimental Physics, University of Hamburg, Hamburg, 22761 (Germany); Hoenniger, Frank [Institute for Experimental Physics, University of Hamburg, Hamburg, 22761 (Germany); Kramberger, Gregor [Jozef Stefan Institute, University of Ljubljana, Ljubljana, 100 (Slovenia); Moll, Michael [CERN, Geneva, 1211 (Switzerland); Nossarzewska, Elsbieta [ITME, Institute for Electronocs Materials Technology, Warsaw, 01919 (Poland); Pintilie, Ioana [National Institute of Materials Physics, Bucharest, 077125 (Romania); Roeder, Ralf [CiS Institute for Microsensors gGmbH, Erfurt, 99099 (Germany)

    2006-11-30

    Diodes processed on n-type epitaxial silicon with a thickness of 25, 50 and 75 {mu}m had been irradiated with reactor neutrons and high-energy protons (24 GeV/c) up to integrated fluences of {phi} {sub eq}=10{sup 16} cm{sup -2}. Systematic experiments on radiation-induced damage effects revealed the following results: in contrast to standard and oxygen-enriched float zone (FZ) silicon devices no space charge sign inversion was observed after irradiation. It is shown that the radiation-generated concentration of deep acceptors, dominating the behavior of n-type FZ diodes, is compensated by creation of shallow donors. Thus a positive space charge is maintained throughout the irradiation up to the highest fluence and even during prolonged elevated-temperature annealing cycles. Defect analysis studies using thermally stimulated current measurements attribute the effect to a damage-induced shallow donor at E {sub C}-0.23 eV. It is argued that, as in the case of thermal donors, oxygen dimers, out diffusing from the Cz substrate during the diode processing, are responsible precursers. Results from extensive annealing experiments at elevated temperatures are verified by comparison with prolonged room-temperature annealing. These results showed that in contrast to FZ detectors, which always have to be cooled, room-temperature storage during beam off periods of future elementary particle physics experiments would even be beneficial for n-type epi-silicon detectors. A dedicated experiment at CERN-PS had successfully proven this expectation. It was verified, that in such a scenario the depletion voltage for the epi-detector could always be kept at a moderate level throughout the full S-LHC operation (foreseen upgrade of the large hadron collider). Practically no difference with respect to FZ-silicon devices was found in the damage-induced bulk generation current. The charge trapping measured with {sup 90}Sr electrons (mip's) is also almost identical to what was expected

  10. Calculations on neutron irradiation damage in reactor materials

    International Nuclear Information System (INIS)

    Sone, Kazuho; Shiraishi, Kensuke

    1976-01-01

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

  11. THE PRODUCTION OF LOW-ENERGY NEUTRONS IN SOLAR FLARES AND THE IMPORTANCE OF THEIR DETECTION IN THE INNER HELIOSPHERE

    International Nuclear Information System (INIS)

    Murphy, R. J.; Kozlovsky, B.; Share, G. H.

    2012-01-01

    Neutron detectors on spacecraft in the inner heliosphere can observe the low-energy ( ion –1 ) most important for producing low-energy neutrons from these reactions. We calculate escaping-neutron spectra and neutron-capture line yields from ions propagating in a magnetic loop with various kinetic-energy spectra. This study provides the basis for planning inner-heliospheric missions having a low-energy neutron detector. The MESSENGER spacecraft orbiting Mercury has such a detector. We conclude that a full understanding of ion acceleration, transport, and interaction at the Sun requires observation of both neutrons and gamma rays with detectors of comparable sensitivity. We find that the neutron-capture line fluence at 1 AU is comparable to the 1-10 MeV neutron fluence at 0.5 AU, and therefore as effective for revealing low-energy ion acceleration. However, as the distance from the Sun to the neutron detector decreases, the tremendous increase of the low-energy neutron flux allows exploration of ion acceleration in weak flares not previously observable and may reveal acceleration at other sites not previously detected where low-energy neutrons could be the only high-energy signature of ion acceleration. Also, a measurement of the low-energy neutron spectrum will provide important information about the accelerated-ion spectrum that is not available from the capture line fluence measurement alone.

  12. Prediction analysis of dose equivalent responses of neutron dosemeters used at a MOX fuel facility

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.

    2011-01-01

    To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H p (10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces. (authors)

  13. Neutron-transmutation-doped germanium bolometers

    International Nuclear Information System (INIS)

    Palaio, N.P.; Rodder, M.; Haller, E.E.; Kreysa, E.

    1983-02-01

    Six slices of ultra-pure germanium were irradiated with thermal neutron fluences between 7.5 x 10 16 and 1.88 x 10 18 cm - 2 . After thermal annealing the resistivity was measured down to low temperatures ( 0 exp(δ/T) in the hopping conduction regime. Also, several junction FETs were tested for noise performance at room temperature and in an insulating housing in a 4.2K cryostat. These FETs will be used as first stage amplifiers for neutron-transmutation-doped germanium bolometers

  14. Neutron dosimetry using thermoluminescent material and KBr

    International Nuclear Information System (INIS)

    Sahyun, A.

    1979-01-01

    It is investigated the misture of natural CaF 2 and KBr as thermal neutron dosimeter compared to pure fluorite. Firstly, it was found that a 50 - 50 mixture in weight produces the highest TL sensitivity when irradiated with thermal neutrons. Second, it was shown experimentally that for TL reading within 30 to 40 hours after irradiation, the radioisotopes 42 K, 80 Br and 80 sub(m)Br contribute predominantly, while for periods longer than 4 to 5 days, 82 Br also contributes significantly. It is in question, the minimum detectable fluence is of the order of 10 9 n/cm 2 . (Autor) [pt

  15. Neutron detectors for nuclear reactor control

    International Nuclear Information System (INIS)

    Duchene, Jean; Verdant, Robert.

    1974-01-01

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

  16. Neutron-transmutation-doped germanium bolometers

    Science.gov (United States)

    Palaio, N. P.; Rodder, M.; Haller, E. E.; Kreysa, E.

    1983-01-01

    Six slices of ultra-pure germanium were irradiated with thermal neutron fluences between 7.5 x 10 to the 16th and 1.88 x 10 to the 18th per sq cm. After thermal annealing the resistivity was measured down to low temperatures (less than 4.2 K) and found to follow the relationship rho = rho sub 0 exp(Delta/T) in the hopping conduction regime. Also, several junction FETs were tested for noise performance at room temperature and in an insulating housing in a 4.2 K cryostat. These FETs will be used as first stage amplifiers for neutron-transmutation-doped germanium bolometers.

  17. Neutron reflectometry

    International Nuclear Information System (INIS)

    Van Well, A.A.

    1999-01-01

    Neutron research where reflection, refraction, and interference play an essential role is generally referred to as 'neutron optics'. The neutron wavelength, the scattering length density and the magnetic properties of the material determine the critical angle for total reflection. The theoretical background of neutron reflection, experimental methods and the interpretation of reflection data are presented. (K.A.)

  18. Parameters measurement for the thermal neutron beam in the thermal column hole of Xi’an pulse reactor

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    The distribution of the neutron spectra in the thermal column hole of Xi’an pulse reactor was measured with the time-of-flight method.Compared with the thermal Maxwellian theory neutron spectra,the thermal neutron spectra measured is a little softer,and the average neutron energy of the experimental spectra is about 0.042±0.01 eV.The thermal neutron fluence rate at the front end of thermal column hole,measured with gold foil activation techniques,is about 1.18×105 cm-2 s-1.The standard uncertainty of the measured thermal neutron fluence is about 3%.The spectra-averaged cross section of 197Au(n,γ) determined by the experimental thermal neutron spectra is(92.8±0.93) ×10-24 cm2.

  19. Performance of n-in-p pixel detectors irradiated at fluences up to $5x10^{15} n_{eq}/cm^{2}$ for the future ATLAS upgrades

    CERN Document Server

    INSPIRE-00219560; La Rosa, A.; Nisius, R.; Pernegger, H.; Richter, R.H.; Weigell, P.

    We present the results of the characterization of novel n-in-p planar pixel detectors, designed for the future upgrades of the ATLAS pixel system. N-in-p silicon devices are a promising candidate to replace the n-in-n sensors thanks to their radiation hardness and cost effectiveness, that allow for enlarging the area instrumented with pixel detectors. The n-in-p modules presented here are composed of pixel sensors produced by CiS connected by bump-bonding to the ATLAS readout chip FE-I3. The characterization of these devices has been performed with the ATLAS pixel read-out systems, TurboDAQ and USBPIX, before and after irradiation with 25 MeV protons and neutrons up to a fluence of 5x10**15 neq /cm2. The charge collection measurements carried out with radioactive sources have proven the feasibility of employing this kind of detectors up to these particle fluences. The collected charge has been measured to be for any fluence in excess of twice the value of the FE-I3 threshold, tuned to 3200 e. The first result...

  20. Gain and time resolution of 45 μm thin Low Gain Avalanche Detectors before and after irradiation up to a fluence of 1015 neq/cm2

    International Nuclear Information System (INIS)

    Lange, J.; Cavallaro, E.; Förster, F.; Grinstein, S.; Carulla, M.; Flores, D.; Hidalgo, S.; Merlos, A.; Pellegrini, G.; Quirion, D.; Chytka, L.; Komarek, T.; Nozka, L.; Davis, P.M.; Kramberger, G.; Mandić, I.; Sykora, T.

    2017-01-01

    Low Gain Avalanche Detectors (LGADs) are silicon sensors with a built-in charge multiplication layer providing a gain of typically 10 to 50. Due to the combination of high signal-to-noise ratio and short rise time, thin LGADs provide good time resolutions. LGADs with an active thickness of about 45 μm were produced at CNM Barcelona. Their gains and time resolutions were studied in beam tests for two different multiplication layer implantation doses, as well as before and after irradiation with neutrons up to 10 15 n eq /cm 2 . The gain showed the expected decrease at a fixed voltage for a lower initial implantation dose, as well as for a higher fluence due to effective acceptor removal in the multiplication layer. Time resolutions below 30 ps were obtained at the highest applied voltages for both implantation doses before irradiation. Also after an intermediate fluence of 3×10 14 n eq /cm 2 , similar values were measured since a higher applicable reverse bias voltage could recover most of the pre-irradiation gain. At 10 15 n eq /cm 2 , the time resolution at the maximum applicable voltage of 620 V during the beam test was measured to be 57 ps since the voltage stability was not good enough to compensate for the gain layer loss. The time resolutions were found to follow approximately a universal function of gain for all implantation doses and fluences.

  1. Impurity effects in neutron-irradiated simple oxides: Implications for fusion devices

    International Nuclear Information System (INIS)

    Gonzalez, R.; Chen, Y.; Caceres, D.; Vergara, I.

    2006-01-01

    Radiation damage induced by neutron irradiation was studied in undoped MgO crystals and in MgO doped with either iron, hydrogen or lithium impurities. The oxygen-vacancy concentration produced by irradiation increases with neutron fluence. The net production rates resulting from irradiations with 14.8 MeV neutrons are about twice those produced by fission neutrons. In nominally pure crystals, the oxygen-vacancy concentration incurred by the fission-neutron irradiation is higher in crystals with a larger number of inherent impurities (such as iron) due to trapping of interstitials by impurities. Suppression of these defects is observed in MgO:H crystals and attributed to migration of oxygen vacancies to microcavities filled with H 2 gas. In MgO:Li crystals irradiated with neutron fluences below 10 18 n/cm 2 , most of the oxygen vacancies are camouflaged as hydride ions. Nanoindentation experiments show that hardness increases with neutron fluence and is independent of the presence of lithium in the crystal. Comparison between a neutron-irradiated and a thermochemically reduced crystal containing similar concentrations of oxygen vacancies shows that 70% of the neutron-irradiation hardening is produced by interstitials, 30% by oxygen vacancies and a negligible amount by higher-order point defects

  2. Ge nano-layer fabricated by high-fluence low-energy ion implantation

    International Nuclear Information System (INIS)

    Lu Tiecheng; Dun Shaobo; Hu Qiang; Zhang Songbao; An Zhu; Duan Yanmin; Zhu Sha; Wei Qiangmin; Wang Lumin

    2006-01-01

    A Ge nano-layer embedded in the surface layer of an amorphous SiO 2 film was fabricated by high-fluence low-energy ion implantation. The component, phase, nano-structure and luminescence properties of the nano-layer were studied by means of Rutherford backscattering, glancing incident X-ray diffraction, laser Raman scattering, transmission electron microscopy and photoluminescence. The relation between nano-particle characteristics and ion fluence was also studied. The results indicate that nano-crystalline Ge and nano-amorphous Ge particles coexist in the nano-layer and the ratio of nano-crystalline Ge to nano-particle Ge increases with increasing ion fluence. The intensity of photoluminescence from the nano-layer increases with increasing ion fluence also. Prepared with certain ion fluences, high-density nano-layers composed of uniform-sized nano-particles can be observed

  3. Study of the response of a silicon detector irradiated with 1 MeV neutrons; Etude de la reponse d`un detecteur Si irradie par des neutrons de 1 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Roy, P [Montreal Univ., PQ (Canada). Lab. de Physique Nucleaire

    1994-12-31

    The author studied the response of an n-type silicon detector irradiated with 1 MeV neutrons at fluences ranging from 0.26x10{sup 13} to 11.19x10{sup 13} neutrons/cm{sup 2}. The response of the irradiated detector to {sup 241}Am alpha particles was measured. 13 refs., 7 figs.

  4. THE FLUENCE AND DISTANCE DISTRIBUTIONS OF FAST RADIO BURSTS

    International Nuclear Information System (INIS)

    Vedantham, H. K.; Ravi, V.; Hallinan, G.; Shannon, R. M.

    2016-01-01

    Fast radio bursts (FRB) are millisecond-duration radio pulses with apparent extragalactic origins. All but two of the FRBs have been discovered using the Parkes dish, which employs multiple beams formed by an array of feed horns on its focal plane. In this paper, we show that (i) the preponderance of multiple-beam detections and (ii) the detection rates for varying dish diameters can be used to infer the index α of the cumulative fluence distribution function (the log N –log F function: α = 1.5 for a non-evolving population in a Euclidean universe). If all detected FRBs arise from a single progenitor population, multiple-beam FRB detection rates from the Parkes telescope yield the constraint 0.52 < α < 1.0 with 90% confidence. Searches at other facilities with different dish sizes refine the constraint to 0.5 < α < 0.9. Our results favor FRB searches with smaller dishes, because for α < 1 the gain in field of view for a smaller dish is more important than the reduction in sensitivity. Further, our results suggest that (i) FRBs are not standard candles, and (ii) the distribution of distances to the detected FRBs is weighted toward larger distances. If FRBs are extragalactic, these results are consistent with a cosmological population, which would make FRBs excellent probes of the baryonic content and geometry of the universe.

  5. Solid phase epitaxy of amorphous silicon carbide: Ion fluence dependence

    International Nuclear Information System (INIS)

    Bae, I.-T.; Ishimaru, Manabu; Hirotsu, Yoshihiko; Sickafus, Kurt E.

    2004-01-01

    We have investigated the effect of radiation damage and impurity concentration on solid phase epitaxial growth of amorphous silicon carbide (SiC) as well as microstructures of recrystallized layer using transmission electron microscopy. Single crystals of 6H-SiC with (0001) orientation were irradiated with 150 keV Xe ions to fluences of 10 15 and 10 16 /cm 2 , followed by annealing at 890 deg. C. Full epitaxial recrystallization took place in a specimen implanted with 10 15 Xe ions, while retardation of recrystallization was observed in a specimen implanted with 10 16 /cm 2 Xe ions. Atomic pair-distribution function analyses and energy dispersive x-ray spectroscopy results suggested that the retardation of recrystallization of the 10 16 Xe/cm 2 implanted sample is attributed to the difference in amorphous structures between the 10 15 and 10 16 Xe/cm 2 implanted samples, i.e., more chemically disordered atomistic structure and higher Xe impurity concentration in the 10 16 Xe/cm 2 implanted sample

  6. Deuterium trapping in carbon fiber composites under high fluence

    International Nuclear Information System (INIS)

    Airapetov, A.A.; Begrambekov, L.B.; Kuzmin, A.A.; Shigin, P.A.; Zakharov, A.M.

    2010-01-01

    The paper is devoted to investigation of deuterium trapping in CFC, dance graphite MPG-8 and pyrolytic graphite (PG) under plasma ion- and electron irradiation. Number of specific features of deuterium trapping and retention under plasma ion and electron irradiation is presented and discussed. In particular it is shown that 1) deuterium trapping takes place even when energy of impinging ions approaches zero; 2) deuterium is trapped under irradiation by plasma electrons; 3) under irradiation at equal fluences deuterium trapping is higher, when ion flux is smaller. High energy ion penetrating the surfaces are trapped in the traps created at the expense of their kinetic energy. The process may be named 'kinetic trapping'. Under low energy (smaller than 200 eV) electron and/or ion irradiation the energy of inelastic interaction on the surface provides creation of active centers, which initiate dissociation of deuterium sorbed on the surface, penetration of deuterium atoms into graphite and their trapping in specific low energy traps. The term 'potential trapping' is proposed for this type of trapping. Under high energy irradiation such atoms can fill the traps formed through kinetic mechanism. Origination of moveable deuterium atoms from the layer of surface sorption seems to be time dependent process and it is a reason of increase of trapping along with irradiation time. New features of deuterium trapping and retention in graphite evaluated in this study offer new opportunities for analysis and correct estimation of hydrogen isotope trapping and retention in tokamaks having graphite tiles. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-07

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

  8. Neutron radiography with ultracold neutrons

    International Nuclear Information System (INIS)

    Bates, J.C.

    1981-01-01

    The neutron transmission factor of very thin films may be low if the neutron energy is comparable to the pseudo-potential of the film material. Surprisingly, perhaps, it is relatively easy to obtain neutrons with such low energies in sufficient numbers to produce neutron radiographs. (orig.)

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

    DEFF Research Database (Denmark)

    Fabritsiev, S.A.; Pokrovsky, A.S.; Zinkle, S.J.

    1996-01-01

    -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 Delta rho(tr) which was proportional to thermal neutron fluence and a radiation defect component Delta rho(rd) which...

  10. Neutron personnel dosimetry

    International Nuclear Information System (INIS)

    Griffith, R.V.

    1982-01-01

    The measurement of neutron exposures to personnel is an issue that has received increased attention in the last few years. It is important to consider key aspects of the whole dosimetry system when developing dose estimates. This begins with selection of proper dosimeters and survey instruments, and extends through the calibration methods. One must match the spectral response and sensitivity of the dosimeter to the spectral characteristics of the neutron fields. Threshold detectors that are insensitive to large fractions of neutrons in the lower energy portion of reactor spectra should be avoided. Use of two or more detectors with responses that complement each other will improve measurement quality. It is important to understand the spectral response of survey instruments, so that spectra which result in significant overresponse do not lead to overestimation of dose. Calibration sources that do not match operational field spectra can contribute to highly erroneous results. In those situations, in-field calibration techniques should be employed. Although some detection developments have been made in recent years, a lot can be done with existing technology until fully satisfactory, long term solutions are obtained

  11. Calculated characteristics of subcritical assembly with anisotropic transport of neutrons

    International Nuclear Information System (INIS)

    Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I.

    2003-01-01

    There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5 n . Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)

  12. 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)

  13. 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

  14. On the studies of thermodynamics properties of fast neutron irradiated (LixK1-x)2SO4 crystals

    Science.gov (United States)

    El-Khatib, A. M.; Kassem, M. E.; Gomaa, N. G.; Mahmoud, S. A.

    The effect of fast neutron irradiation on the thermodynamic properties of (LixK1-x)2SO4, (x = 0.1, 0.2,˙˙˙˙˙˙˙˙0.5) has been studied. The measurements were carried out in the vicinity of phase transition. The study reveals that as the lithium content decreases the first high temperature phase Tc = 705 K disappears, while the second one is shifted to lower temperature. It is observed also that the specific heat, Cp, decreases sharply with neutron integrated fluence φ and increases once more. Both entropy and enthalpy changes increase with the increase of neutron integrated fluence.

  15. Pulsed neutron logging - a modern approach to petroleum exploration

    International Nuclear Information System (INIS)

    Navalkar, M.P.

    1978-01-01

    As hydrocarbons have strikingly different properties for neutrons as compared with rocks, neutronic methods are useful to prepare logs of bore holes drilled for petroleum exploration. The earlier neutron logs were based on steady neutron sources, but the superior logs, namely, neutron life time log and neutron induced spectral log are based on pulsed neutron sources. The methods of obtaining these logs, their limitations and the field equipment required are described briefly. As the two logs are complementary, a plea has been made to develop field equipment which can obtain two logs in a single run. (M.G.B.)

  16. Neutron crosstalk between liquid scintillators

    Energy Technology Data Exchange (ETDEWEB)

    Verbeke, J.M., E-mail: verbeke2@llnl.gov; Prasad, M.K., E-mail: prasad1@llnl.gov; Snyderman, N.J., E-mail: snyderman1@llnl.gov

    2015-09-11

    A method is proposed to quantify the fractions of neutrons scattering between liquid scintillators. Using a spontaneous fission source, this method can be utilized to quickly characterize an array of liquid scintillators in terms of crosstalk. The point model theory due to Feynman is corrected to account for these multiple scatterings. Using spectral information measured by the liquid scintillators, fractions of multiple scattering can be estimated, and mass reconstruction of fissile materials under investigation can be improved. Monte Carlo simulations of mono-energetic neutron sources were performed to estimate neutron crosstalk. A californium source in an array of liquid scintillators was modeled to illustrate the improvement of the mass reconstruction.

  17. Neutron crosstalk between liquid scintillators

    International Nuclear Information System (INIS)

    Verbeke, J.M.; Prasad, M.K.; Snyderman, N.J.

    2015-01-01

    A method is proposed to quantify the fractions of neutrons scattering between liquid scintillators. Using a spontaneous fission source, this method can be utilized to quickly characterize an array of liquid scintillators in terms of crosstalk. The point model theory due to Feynman is corrected to account for these multiple scatterings. Using spectral information measured by the liquid scintillators, fractions of multiple scattering can be estimated, and mass reconstruction of fissile materials under investigation can be improved. Monte Carlo simulations of mono-energetic neutron sources were performed to estimate neutron crosstalk. A californium source in an array of liquid scintillators was modeled to illustrate the improvement of the mass reconstruction

  18. Development of an operational neutron spectrometry system dedicated to the characterization of the natural atmospheric radiative environment, implemented at the Pic du Midi

    International Nuclear Information System (INIS)

    Cheminet, Adrien

    2013-01-01

    This PhD Thesis has been achieved thanks to the joint effort between two French organizations, the French Institute for Radiological Protection and Nuclear Safety (IRSN/LMDN, Cadarache) and the French Aerospace Lab (ONERA/ DESP, Toulouse). The aim was to develop an operational neutron spectrometer extended to high energies in order to measure the dynamics of the spectral variations of the natural radiative environment at the summit of the Pic du Midi Observatory in the French Pyrenees. Thereby, the fluence responses of each detector were calculated thanks to Monte Carlo simulations. Afterwards, they were validated by means of experimental campaigns up to high energies (≥20 MeV) nearby reference neutron fields. The systematic uncertainties were deduced after detailed studies of the mathematic reconstruction of the spectra (i.e. unfolding procedure). Then, the system was tested under rocks at the LSBB of Rustrel before being installed at respectively +500 m and +1000 m above sea level for the first environmental campaigns. Finally, the spectrometer has been operating for two years after its deployment at the summit of the Pic du Midi (+2885 m). The continuous data were analysed thanks to an innovative method. Some seasonal and spectral variations were observed. Some Forbush decreases were also recorded after strong solar flares. These data were further analysed thanks to Monte Carlo simulations. The data were made more attractive thanks to several practical applications with personnel dosimetry or reliability of submicron electronics components. (author)

  19. Neutron Skins and Neutron Stars

    OpenAIRE

    Piekarewicz, J.

    2013-01-01

    The neutron-skin thickness of heavy nuclei provides a fundamental link to the equation of state of neutron-rich matter, and hence to the properties of neutron stars. The Lead Radius Experiment ("PREX") at Jefferson Laboratory has recently provided the first model-independence evidence on the existence of a neutron-rich skin in 208Pb. In this contribution we examine how the increased accuracy in the determination of neutron skins expected from the commissioning of intense polarized electron be...

  20. Measurements of 60Co in spoons activated by neutrons during the JCO criticality accident at Tokai-mura in 1999

    International Nuclear Information System (INIS)

    Gasparro, J.; Hult, M.; Komura, K.; Arnold, D.; Holmes, L.; Johnston, P.N.; Laubenstein, M.; Neumaier, S.; Reyss, J.-L.; Schillebeeckx, P.; Tagziria, H.; Van Britsom, G.; Vasselli, R.

    2004-01-01

    Neutron activated items from the vicinity of the place where the JCO criticality accident occurred have been used to determine the fluence of neutrons around the facility and in nearby residential areas. By using underground laboratories for measuring the activation products, it is possible to extend the study to also cover radionuclides with very low activities from long-lived radionuclides. The present study describes γ-ray spectrometry measurements undertaken in a range of underground laboratories for the purpose of measuring 60 Co more than 2 years after the criticality event. The measurements show that neutron fluence determined from 60 Co activity is in agreement with previous measurements using the short-lived radionuclides 51 Cr and 59 Fe. Limits on contamination of the samples with 60 Co are evaluated and shown to not greatly affect the utility of neutron fluence determinations using 60 Co activation

  1. Measurements of 60Co in spoons activated by neutrons during the JCO criticality accident at Tokai-mura in 1999.

    Science.gov (United States)

    Gasparro, J; Hult, M; Komura, K; Arnold, D; Holmes, L; Johnston, P N; Laubenstein, M; Neumaier, S; Reyss, J-L; Schillebeeckx, P; Tagziria, H; Van Britsom, G; Vasselli, R

    2004-01-01

    Neutron activated items from the vicinity of the place where the JCO criticality accident occurred have been used to determine the fluence of neutrons around the facility and in nearby residential areas. By using underground laboratories for measuring the activation products, it is possible to extend the study to also cover radionuclides with very low activities from long-lived radionuclides. The present study describes gamma-ray spectrometry measurements undertaken in a range of underground laboratories for the purpose of measuring (60)Co more than 2 years after the criticality event. The measurements show that neutron fluence determined from (60)Co activity is in agreement with previous measurements using the short-lived radionuclides (51)Cr and (59)Fe. Limits on contamination of the samples with (60)Co are evaluated and shown to not greatly affect the utility of neutron fluence determinations using (60)Co activation.

  2. Experimental characterization of HOTNES: A new thermal neutron facility with large homogeneity area

    Energy Technology Data Exchange (ETDEWEB)

    Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Sperduti, A. [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); ENEA C.R. Frascati, via E. Fermi n. 45, 00044 Frascati, Roma (Italy); Pietropaolo, A.; Pillon, M. [ENEA C.R. Frascati, via E. Fermi n. 45, 00044 Frascati, Roma (Italy); Pola, A. [Politecnico di Milano, Dipartimento di Energia, via La Masa 34, 20156 Milano (Italy); INFN–Milano, Via Celoria 16, 20133 Milano (Italy); Gómez-Ros, J.M. [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)

    2017-01-21

    A new thermal neutron irradiation facility, called HOTNES (HOmogeneous Thermal NEutron Source), was established in the framework of a collaboration between INFN-LNF and ENEA-Frascati. HOTNES is a polyethylene assembly, with about 70 cmx70 cm square section and 100 cm height, including a large, cylindrical cavity with diameter 30 cm and height 70 cm. The facility is supplied by a {sup 241}Am-B source located at the bottom of this cavity. The facility was designed in such a way that the iso-thermal-fluence surfaces, characterizing the irradiation volume, coincide with planes parallel to the cavity bottom. The thermal fluence rate across a given isofluence plane is as uniform as 1% on a disk with 30 cm diameter. Thermal fluence rate values from about 700 cm{sup −2} s{sup −1} to 1000 cm{sup −2} s{sup −1} can be achieved. The facility design, previously optimized by Monte Carlo simulation, was experimentally verified. The following techniques were used: gold activation foils to assess the thermal fluence rate, semiconductor-based active detector for mapping the irradiation volume, and Bonner Sphere Spectrometer to determine the complete neutron spectrum. HOTNES is expected to be attractive for the scientific community involved in neutron metrology, neutron dosimetry and neutron detector testing.

  3. Introduction to neutron metrology for reactor radiation damage

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  4. Correlation of DT and DD fusion neutron damage in silicon surface barrier detector

    International Nuclear Information System (INIS)

    Iida, Toshiyuki; Sueyoshi, Yasuhiro; Sunarno; Takahashi, Akito

    1994-01-01

    In order to examine the correlation of DT and DD fusion neutron damage in Si, a silicon surface barrier detector (Si-SBD) was irradiation with neutrons from a deuteron accelerator. The leakage current increased proportionally with neutron fluence, which determined the neutron damage constant for the Si-SBD. The correlation factor of the DT and DD neutron damage in the Si-SBD was determined from the ratio of the DT and DD neutron damage constants and was found to be 2.3. We also calculated the rate of DT and DD neutron displacement damage for Si by using the TRIM-90 computer program and actual data on neutron reactions in the Si-SBD. The correlation factor of DT and DD neutron damage from the calculation agreed with that from the Si-SBD irradiation experiment. (author)

  5. 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)

  6. Exposure potential to neutrons of the Brazilian workers

    International Nuclear Information System (INIS)

    Martins, Marcelo M.; Almeida, Carlos E. de

    1995-01-01

    Nowadays, there are 222 neutron radioactive sources registered in use in Brazil, in addition to several neutron fields, power and research reactors and neutron generators. Secondary neutron fields can also be generated in particle accelerators by nuclear reactions with its shielding, the experimental set up and the conversion target. These neutron fields are very different, not only in their spectra but also in their fluences. Around 200 radiation workers are monthly monitored since 1983 by an albedo system. Of the evaluated dosimeters 4% only have shown neutron doses, being 13 mSv the maximum measured value. Most of these doses were received by workers of the oil-well logging and research activities users of 241 Am-Be sources. (author). 7 refs., 3 tabs

  7. Neutron irradiation effects on silicon detectors structure, electrical and mechanical characteristics

    International Nuclear Information System (INIS)

    Rabinovich, E.; Golan, G.; Axelevich, A.; Inberg, A.; Oksman, M.; Rosenwaks, I.; Lubarsky, G.; Seidman, A.; Croitoru, N.; Rancoita, P.G.; Rattaggi, M.

    1999-01-01

    Neutron irradiation effects on (p-n) and Schottky-junction silicon detectors were studied. It was shown that neutron interactions with monocrystalline silicon create specific types of microstructure defects with morphology differing according to the level of neutron fluences (Φ). The isolated dislocation loops, formed by interstitial atoms were observed in microstructure images for 10 10 ≤ Φ ≤ 10 12 n/cm 2 . A strong change in the dislocation loops density and a cluster formation was observed for Φ ≥ 10 13 n/cm 2 . A drastic silicon damage was found for fluences over 10 14 n/cm 2 . These fluences created zones enriched with all types of dislocations, covering more than 50 % of the total surface area. A mechanical fragility appeared in that fluence range in a form of microcracks. 10 14 n/cm 2 appears to be a critical value of neutron irradiation because of the radiation damage described above and because the characteristics I f -V f of silicon detectors can be differentiated from those obtained at low fluences. (A.C.)

  8. Neutron dosimetry using aqueous solutions of lithium acetate

    International Nuclear Information System (INIS)

    Rakovan, L.J.

    1996-01-01

    A thermal neutron dosimetry system using the 6 Li(n,α) 3 H reaction and liquid scintillation counting of tritium was developed. Lithium acetate was chosen to supply the 6 Li in the aqueous dosimetry solutions. Neutron irradiations were completed using The Ohio State University Research Reactor. After two sets of samples were irradiated, variables in the system such as the mass of lithium acetate in the solutions and the counting window of the liquid scintillation counter used to analyze the sample were chosen. The system was evaluated by completing two sets of 23 minute irradiations with the reactor at 500 kW, 50 kW, 5 kW, and one irradiation at 500 W. The samples irradiated at 500 W were below the threshold of the system, and could not be used. Prompt analysis was essential due to loss of detectable emissions in the dosimetry solutions over time. The thermal neutron fluences calculated with the data from the samples were compared to the fluences determined from gold wire irradiations. The fluence values differed at most by 6%. The fluence values calculated from the samples were consistently less than those determined from the gold wires

  9. Characterization of materials used for neutron spectra modification

    International Nuclear Information System (INIS)

    Solieman, A.H.M.; Comsan, M.N.H.; Fahmey, M.A.; Morsy, A.A.

    2008-01-01

    Monte Carlo Simulation is used to study the thickness-dependent neutron-spectral-modification after transport in different materials. A collection of significant materials is studied, for choosing of potential candidates in the construction and design of accelerator-based neutron irradiation system suitable for Boron Neutron Capture Therapy (BNCT)

  10. Ultra-fast fluence optimization for beam angle selection algorithms

    Science.gov (United States)

    Bangert, M.; Ziegenhein, P.; Oelfke, U.

    2014-03-01

    Beam angle selection (BAS) including fluence optimization (FO) is among the most extensive computational tasks in radiotherapy. Precomputed dose influence data (DID) of all considered beam orientations (up to 100 GB for complex cases) has to be handled in the main memory and repeated FOs are required for different beam ensembles. In this paper, the authors describe concepts accelerating FO for BAS algorithms using off-the-shelf multiprocessor workstations. The FO runtime is not dominated by the arithmetic load of the CPUs but by the transportation of DID from the RAM to the CPUs. On multiprocessor workstations, however, the speed of data transportation from the main memory to the CPUs is non-uniform across the RAM; every CPU has a dedicated memory location (node) with minimum access time. We apply a thread node binding strategy to ensure that CPUs only access DID from their preferred node. Ideal load balancing for arbitrary beam ensembles is guaranteed by distributing the DID of every candidate beam equally to all nodes. Furthermore we use a custom sorting scheme of the DID to minimize the overall data transportation. The framework is implemented on an AMD Opteron workstation. One FO iteration comprising dose, objective function, and gradient calculation takes between 0.010 s (9 beams, skull, 0.23 GB DID) and 0.070 s (9 beams, abdomen, 1.50 GB DID). Our overall FO time is < 1 s for small cases, larger cases take ~ 4 s. BAS runs including FOs for 1000 different beam ensembles take ~ 15-70 min, depending on the treatment site. This enables an efficient clinical evaluation of different BAS algorithms.

  11. Ultra-fast fluence optimization for beam angle selection algorithms

    International Nuclear Information System (INIS)

    Bangert, M; Ziegenhein, P; Oelfke, U

    2014-01-01

    Beam angle selection (BAS) including fluence optimization (FO) is among the most extensive computational tasks in radiotherapy. Precomputed dose influence data (DID) of all considered beam orientations (up to 100 GB for complex cases) has to be handled in the main memory and repeated FOs are required for different beam ensembles. In this paper, the authors describe concepts accelerating FO for BAS algorithms using off-the-shelf multiprocessor workstations. The FO runtime is not dominated by the arithmetic load of the CPUs but by the transportation of DID from the RAM to the CPUs. On multiprocessor workstations, however, the speed of data transportation from the main memory to the CPUs is non-uniform across the RAM; every CPU has a dedicated memory location (node) with minimum access time. We apply a thread node binding strategy to ensure that CPUs only access DID from their preferred node. Ideal load balancing for arbitrary beam ensembles is guaranteed by distributing the DID of every candidate beam equally to all nodes. Furthermore we use a custom sorting scheme of the DID to minimize the overall data transportation. The framework is implemented on an AMD Opteron workstation. One FO iteration comprising dose, objective function, and gradient calculation takes between 0.010 s (9 beams, skull, 0.23 GB DID) and 0.070 s (9 beams, abdomen, 1.50 GB DID). Our overall FO time is < 1 s for small cases, larger cases take ∼ 4 s. BAS runs including FOs for 1000 different beam ensembles take ∼ 15–70 min, depending on the treatment site. This enables an efficient clinical evaluation of different BAS algorithms.

  12. Relating Solar Energetic Particle Event Fluences to Peak Intensities

    Science.gov (United States)

    Kahler, Stephen W.; Ling, Alan G.

    2018-02-01

    Recently we (Kahler and Ling, Solar Phys. 292, 59, 2017: KL) have shown that time-intensity profiles [I(t)] of 14 large solar energetic particle (SEP) events can be fitted with a simple two-parameter fit, the modified Weibull function, which is characterized by shape and scaling parameters [α and β]. We now look for a simple correlation between an event peak energy intensity [Ip] and the time integral of I(t) over the event duration: the fluence [F]. We first ask how the ratio of F/Ip varies for the fits of the 14 KL events and then examine that ratio for three separate published statistical studies of SEP events in which both F and Ip were measured for comparisons of those parameters with various solar-flare and coronal mass ejection (CME) parameters. The three studies included SEP energies from a 4 - 13 MeV band to E > 100 MeV. Within each group of SEP events, we find a very robust correlation (CC > 0.90) in log-log plots of F versus Ip over four decades of Ip. The ratio increases from western to eastern longitudes. From the value of Ip for a given event, F can be estimated to within a standard deviation of a factor of {≤} 2. Log-log plots of two studies are consistent with slopes of unity, but the third study shows plot slopes of { 10 MeV to {>} 100 MeV. This difference is not explained.

  13. Motion-encoded dose calculation through fluence/sinogram modification

    International Nuclear Information System (INIS)

    Lu, Weiguo; Olivera, Gustavo H.; Mackie, Thomas R.

    2005-01-01

    Conventional radiotherapy treatment planning systems rely on a static computed tomography (CT) image for planning and evaluation. Intra/inter-fraction patient motions may result in significant differences between the planned and the delivered dose. In this paper, we develop a method to incorporate the knowledge of intra/inter-fraction patient motion directly into the dose calculation. By decomposing the motion into a parallel (to beam direction) component and perpendicular (to beam direction) component, we show that the motion effects can be accounted for by simply modifying the fluence distribution (sinogram). After such modification, dose calculation is the same as those based on a static planning image. This method is superior to the 'dose-convolution' method because it is not based on 'shift invariant' assumption. Therefore, it deals with material heterogeneity and surface curvature very well. We test our method using extensive simulations, which include four phantoms, four motion patterns, and three plan beams. We compare our method with the 'dose-convolution' and the 'stochastic simulation' methods (gold standard). As for the homogeneous flat surface phantom, our method has similar accuracy as the 'dose-convolution' method. As for all other phantoms, our method outperforms the 'dose-convolution'. The maximum motion encoded dose calculation error using our method is within 4% of the gold standard. It is shown that a treatment planning system that is based on 'motion-encoded dose calculation' can incorporate random and systematic motion errors in a very simple fashion. Under this approximation, in principle, a planning target volume definition is not required, since it already accounts for the intra/inter-fraction motion variations and it automatically optimizes the cumulative dose rather than the single fraction dose

  14. Neutron Dosimetry

    International Nuclear Information System (INIS)

    Vanhavere, F.

    2001-01-01

    The objective of SCK-CEN's R and D programme on neutron dosimetry is to improve the determination of neutron doses by studying neutron spectra, neutron dosemeters and shielding adaptations. In 2000, R and D focused on the contiued investigation of the bubble detectors type BD-PND and BDT, in particular their sensitivity and temperature dependence; the updating of SCK-CEN's criticality dosemeter, the investigation of the characteristics of new thermoluminescent materials and their use in neutron dosemetry; and the investigation of neutron shielding

  15. Neutron Dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Vanhavere, F

    2001-04-01

    The objective of SCK-CEN's R and D programme on neutron dosimetry is to improve the determination of neutron doses by studying neutron spectra, neutron dosemeters and shielding adaptations. In 2000, R and D focused on the contiued investigation of the bubble detectors type BD-PND and BDT, in particular their sensitivity and temperature dependence; the updating of SCK-CEN's criticality dosemeter, the investigation of the characteristics of new thermoluminescent materials and their use in neutron dosemetry; and the investigation of neutron shielding.

  16. Neutron irradiation effects in pressure vessel steels and weldments

    Energy Technology Data Exchange (ETDEWEB)

    Ianko, L [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power; Davies, L M

    1994-12-31

    This paper deals with the effects of neutron irradiation on the steel and welds used for the pressure vessels which house the reactor cores in light water reactors: irradiation effects on mechanical properties and the shift in ductile-brittle transition temperature, importance of the knowledge of the neutron fluence and of the monitoring and surveillance programmes; empirical and mechanistic modelling of irradiation effects and the necessity of data extension to new operational limits; consequences on the manufacturing and structural design of materials and structures; mitigation of irradiation effects by annealing; international activities and programmes in the field of neutron irradiation effects on PV steels and welds. 37 refs., 22 figs.

  17. Mechanical Properties and Microstructure of Neutron Irradiated Cold-worked Al-1050 and Al-6063 Alloys

    International Nuclear Information System (INIS)

    Munitz, A.; Cotler, A; Talianker, M.

    1998-01-01

    The impact of neutron irradiation on the internal microstructure, mechanical properties and fracture morphology of cold-worked Al-1050 and Al-6063 alloys was studied, using scanning and transmission electron microscopy, and tensile measurements. Specimens consisting of 50 mm long and 6 mm wide gauge sections, were punched out from Al-1050 and Al-6063 23% cold-worked tubes. They were exposed to prolonged neutron irradiation of up to 4.5x10 25 and 8x10 25 thermal neutrons/m 2 (E -3 s -1 . In general, the uniform and total elongation, the yield stress, and the ultimate tensile strength increase as functions of fluence. However, for Al-1050 a decrease in the ultimate tensile strength and yield stress was observed up to a fluence of 1x10 25 thermal neutrons/m 2 which then increase with thermal neutrons fluence. Metallographic examination and fractography for Al-6063 revealed a decrease in the local area reduction of the final fracture necking. This reduction is accompanied by a morphology transition from ductile transgranular shear rupture to a combination of transgranular shear with intergranular dimpled rupture. The intergranular rupture area increases with fluence. In contrast, for Al-1050, fracture morphology remains ductile transgranular shear rupture and the final local area reduction remains almost constant No voids could be observed in either alloy up to the maximum fluence. The dislocation density of cold-worked Al was found to decrease with the thermal neutron fluence. Prolonged annealing of unirradiated cold-worked Al-6063 at 52 degree led to similar results. Thus, it appears that, under our irradiation conditions, whereby the temperature encompassing the samples increases the exposure to this thermal field is the major factor influencing the mechanical properties and microstructure of aluminum alloys

  18. 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.

  19. 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

  20. A {mu}TPC detector for the characterization of low energy neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Golabek, C., E-mail: cedric.golabek@irsn.fr [Laboratoire de Metrologie et de Dosimetrie des Neutrons, IRSN Cadarache, 13115 Saint-Paul-Lez-Durance (France); Billard, J. [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, 53 rue des Martyrs, 38026 Grenoble (France); Allaoua, A. [Laboratoire de Metrologie et de Dosimetrie des Neutrons, IRSN Cadarache, 13115 Saint-Paul-Lez-Durance (France); Bosson, G.; Bourrion, O.; Grignon, C.; Guillaudin, O. [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, 53 rue des Martyrs, 38026 Grenoble (France); Lebreton, L., E-mail: lena.lebreton@irsn.fr [Laboratoire de Metrologie et de Dosimetrie des Neutrons, IRSN Cadarache, 13115 Saint-Paul-Lez-Durance (France); Mayet, F. [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, 53 rue des Martyrs, 38026 Grenoble (France); Petit, M. [Laboratoire de Metrologie et de Dosimetrie des Neutrons, IRSN Cadarache, 13115 Saint-Paul-Lez-Durance (France); Richer, J.-P.; Santos, D. [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, 53 rue des Martyrs, 38026 Grenoble (France)

    2012-06-21

    The AMANDE facility produces monoenergetic neutron fields from 2 keV to 20 MeV for metrological purposes. To be considered as a reference facility, fluence and energy distributions of neutron fields have to be determined by primary measurement standards. For this purpose, a micro Time Projection Chamber is being developed to be dedicated to measure neutron fields with energy ranging from 8 keV up to 1 MeV. In this work we present simulations showing that such a detector, which allows the measurement of the ionization energy and the 3D reconstruction of the recoil nucleus, provides the determination of neutron energy and fluence of these neutron fields.