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

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...

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

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.

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

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.)

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

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.

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

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)

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)

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.

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)

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

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

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)

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.

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.)

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).

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).

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-...

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

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%.

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

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)

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)

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.

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).

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.)

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)

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

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)

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)

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.

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)

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)

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

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

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)

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.

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

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)

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)

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.

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

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)

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.

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

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.)

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.)

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.)

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

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.

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

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)

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...

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)

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)

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.)

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.)

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)

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.)

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%.

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.

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

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

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

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

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)

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

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).

SU-F-BRE-11: Neutron Measurements Around the Varian TrueBeam Linac

Energy Technology Data Exchange (ETDEWEB)

Maglieri, R; Seuntjens, J; Kildea, J [McGill University, Montreal, QC (Canada); Liang, L; DeBlois, F [Jewish General Hospital, Montreal, QC (Canada); Evans, M [Montreal General Hospital, Montreal, QC (Canada); Licea, A [Canadian Nuclear Safety Comission, Ottawa, Ontario (Canada); Dubeau, J; Witharana, S [Detec, Gatineau, QC (Canada)

2014-06-15

Purpose: With the emergence of flattening filter free (FFF) photon beams, several authors have noted many advantages to their use. One such advantage is the decrease in neutron production by photonuclear reactions in the linac head. In the present work we investigate the reduction in neutrons from a Varian TrueBeam linac using the Nested Neutron Spectrometer (NNS, Detec). The neutron spectrum, total fluence and source strength were measured and compared for 10 MV with and without flattening filter and the effect of moderation by the room and maze was studied for the 15 MV beam. Methods: The NNS, similar to traditional Bonner sphere detectors but operated in current mode, was used to measure the neutron fluence and spectrum. The NNS was validated for use in high dose rate environments using Monte Carlo simulations and calibrated at NIST and NRC Canada. Measurements were performed at several positions within the treatment room and maze with the linac jaws closed to maximize neutron production. Results: The measurements showed a total fluence reduction between 35-40% in the room and maze when the flattening filter was removed. The neutron source strength Qn was calculated from in-room fluence measurements and was found to be 0.042 × 10{sup 2} n/Gy, 0.026 × 10{sup 2} n/Gy and 0.59 × 101{sup 2} n/Gy for the 10 MV, the 10 MV FFF and 15 MV beams, respectively. We measured ambient equivalent doses of 11 mSv/hr, 7 mSv/hr and 218 mSv/hr for the 10 MV, 10 MV FFF and 15 MV by the head. Conclusion: Our measurements revealed a decrease in total fluence, neutron source strength and equivalent dose of approximately 35-40% across the treatment room for the FFF compared to FF modes. This demonstrates, as expected, that the flattening filter is a major component of the neutron production for the TrueBeam. The authors greatly acknowledge support form the Canadian Nuclear Commission and the Natural Sciences and Engineering Research Council of Canada through the CREATE program. Co

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.)

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...

Advisory Committee for the calibration standards of ionizing radiation measurement. Section 3. Neutron measurements

International Nuclear Information System (INIS)

1980-01-01

Section III (Neutron measurements) of the Comite Consultatif pour les Etalons de Mesure des Rayonnements ionisants held its fourth meeting in April 1979. After discussing the final report on the fast neutron fluence rate intercomparison, it requested BIPM to submit it for publication in Metrologia. Section III studied the state-of-the-art of an international comparison of a 252 Cf (10 7 s -1 ) source which is in progress. A new fast neutron fluence rate intercomparison is scheduled for 1980; the energies and the methods to be used have been investigated. Finally, Section III studied carefully and rewrote a proposal for a 14-MeV neutron dosimetry facility at BIPM, and added a Recommendation for CCEMRI [fr

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.

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

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

Neutron measurements on the JET tokamak by means of bubble detectors

International Nuclear Information System (INIS)

Gherendi, M.; Craciunescu, T.; Pantea, A.; Zoita, V.; Edlington, T.; Kiptily, V.; Popovichev, S.; Murari, A.

2009-01-01

Full text: The bubble detectors (superheated fluid detectors - SHFDs) are based on suspensions of superheated fluid droplets which vaporise into bubbles when nucleated by radiation interactions. The active detecting medium is in the form of microscopic (20-50 μm) droplets suspended within an elastic polymer. The bubble detectors are of interest for neutron detection in nuclear fusion devices due to some particular characteristics: - High neutron detection efficiency (counts/unit fluence) that ranges from about 4x10 -2 to 4x10 -5 ; - Almost flat, threshold-type energy response over a broad energy range (10's keV to 10's MeV); - The possibility of having any energy threshold within the above-mentioned energy range; - Practically zero sensitivity to gamma-radiation; - Good spatial resolution (sub-centimetre resolution in the image plane). A series of the neutron measurements have been carried out by means of bubble detectors on the JET tokamak, at Culham Science Centre, Abingdon, UK, during the experimental campaigns C17-C26 (2007-2009). The neutron field parameters (yield, fluence, energy distribution) at a specific location outside the JET Torus Hall have been measured using three types of bubble detectors (BD-PND, DEFENDER, and BDS). The bubble detector measurement location is situated at the end of a vertical collimated line of sight, behind the TOFOR neutron time-of-flight spectrometer. The field-of-view is defined by a variable pre-collimator located on top of the JET tokamak. This paper reports only on the neutron fluence measurements. Spatial (radial and toroidal) distributions of the neutron fluence have been obtained with a two-dimensional array having up to 30 bubble detectors. The operation of the bubble detector array as a neutron pinhole camera having a radial resolution at the JET vacuum chamber mid-plane of about 55 mm was demonstrated in measurements using various openings of the pre-collimator. (authors)

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)

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

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.

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

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 ...

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...

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

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)

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

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

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)

无

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.

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)

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.

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 ...

Benchmark referencing of neutron dosimetry measurements

International Nuclear Information System (INIS)

Eisenhauer, C.M.; Grundl, J.A.; Gilliam, D.M.; McGarry, E.D.; Spiegel, V.

1980-01-01

The concept of benchmark referencing involves interpretation of dosimetry measurements in applied neutron fields in terms of similar measurements in benchmark fields whose neutron spectra and intensity are well known. The main advantage of benchmark referencing is that it minimizes or eliminates many types of experimental uncertainties such as those associated with absolute detection efficiencies and cross sections. In this paper we consider the cavity external to the pressure vessel of a power reactor as an example of an applied field. The pressure vessel cavity is an accessible location for exploratory dosimetry measurements aimed at understanding embrittlement of pressure vessel steel. Comparisons with calculated predictions of neutron fluence and spectra in the cavity provide a valuable check of the computational methods used to estimate pressure vessel safety margins for pressure vessel lifetimes

Neutron temperature measurements in a cryogenic hydrogenous moderator

International Nuclear Information System (INIS)

Ball, R.M.; Hoovler, G.S.; Lewis, R.H.

1995-01-01

Benchmarkings of neutronic calculations are most successful when there is a direct correlation between a measurement and an analytic result. In the thermal neutron energy region, the fluence rate as a function of moderator temperature and position within the moderator is an area of potential correlation. The measurement can be done by activating natural lutetium. The two isotopes of the element lutetium have widely different cross sections and permit the discrimination of flux shape and energy distributions at different reactor conditions. The 175 Lu has a 1/v dependence in the thermal energy region, and 176 Lu has a resonance structure that approximates a constant cross section in the same region. The saturation activation of the two isotopes has been measured in an insulated moderator container at the center of a thermal heterogeneous reactor designed for space nuclear propulsion. The measurements were made in a hydrogenous (polyethylene) moderator at three temperatures (83, 184, and 297 K) and five locations within the moderator. Simultaneously, the reactivity effect of the change in the moderator temperature was determined to be positive with an increase in temperature. The plot of activation shows the variation in neutron fluence rate and current with temperature and explains the positive reactivity coefficient. A neutron temperature can be inferred from a postulated Maxwell-Boltzmann distribution and compared with Monte Carlo or other calculations

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)

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)

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

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

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.

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

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)

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

Measurements of {sup 60}Co in spoons activated by neutrons during the JCO criticality accident at Tokai-mura in 1999

Energy Technology Data Exchange (ETDEWEB)

Gasparro, J.; Hult, M. E-mail: mikael.hult@irmm.jrc.be; 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

2004-07-01

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 {sup 60}Co more than 2 years after the criticality event. The measurements show that neutron fluence determined from {sup 60}Co activity is in agreement with previous measurements using the short-lived radionuclides {sup 51}Cr and {sup 59}Fe. Limits on contamination of the samples with {sup 60}Co are evaluated and shown to not greatly affect the utility of neutron fluence determinations using {sup 60}Co activation.

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%

Neutron monitoring measurements for the CIT [Compact Ignition Tokamak] materials irradiations in the ATR I1 position

International Nuclear Information System (INIS)

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

1989-12-01

Measurements were performed to help characterize the neutron environments in which the Compact Ignition Tokamak (CIT) materials were irradiated. These materials were irradiated in a lead shield plug assembly at the ATR I1 position. Neutron monitor materials were placed in the capsules in proximity with the CIT specimens. The neutron monitors sensed the neutrons through reactions that have different neutron energy region responses. By measuring the radioactivity of the neutron monitors it was possible to determine the neutron fluence rates (n/cm 2 /sec) and fluences (n/cm 2 ) at the locations of the monitors. It was also possible to determine the axial and radial gradients of the neutron environments near the specimens. This report presents the results obtained from these measurements for both the CIT number-sign 1 (ORNL 64-2) and CIT number-sign 2 (ORNL 64-1) capsules. In general, ASTM methods and procedures were used in all neutron monitoring associated activities. 7 refs., 9 figs., 10 tabs

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

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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)

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)

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

Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method covers procedures for measuring reaction rates by the activation reaction 58Ni(n,p)58Co. 1.2 This activation reaction is useful for measuring neutrons with energies above approximately 2.1 MeV and for irradiation times up to about 200 days in the absence of high thermal neutron fluence rates (for longer irradiations, see Practice E 261). 1.3 With suitable techniques fission-neutron fluence rates densities above 107 cm−2·s−1 can be determined. 1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E 261. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 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. Note—The burnup corrections were com...

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)

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

Sci-Sat AM: Brachy - 04: Neutron production around a radiation therapy linac bunker - monte carlo simulations and physical measurements.

Science.gov (United States)

Khatchadourian, R; Davis, S; Evans, M; Licea, A; Seuntjens, J; Kildea, J

2012-07-01

Photoneutrons are a major component of the equivalent dose in the maze and near the door of linac bunkers. Physical measurements and Monte Carlo (MC) calculations of neutron dose are key for validating bunker design with respect to health regulations. We attempted to use bubble detectors and a 3 He neutron spectrometer to measure neutron equivalent dose and neutron spectra in the maze and near the door of one of our bunkers. We also ran MC simulations with MCNP5 to measure the neutron fluence in the same region. Using a point source of neutrons, a Clinac 1800 linac operating at 10 MV was simulated and the fluence measured at various locations of interest. We describe the challenges faced when measuring dose with bubble detectors in the maze and the complexity of photoneutron spectrometry with linacs operating in pulsed mode. Finally, we report on the development of a userfriendly GUI for shielding calculations based on the NCRP 151 formalism. © 2012 American Association of Physicists in Medicine.

Study of the environmental neutron spectrum at Zacatecas city

International Nuclear Information System (INIS)

Vega C, H.R.

2003-01-01

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

Study of the environmental neutron spectrum at Zacatecas city

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

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

Prospects for a new cold neutron beam measurement of the neutron lifetime

Energy Technology Data Exchange (ETDEWEB)

Dewey, M., E-mail: mdewey@nist.go [National Institute of Standards and Technology, Gaithersburg, MD 20899 (United States); Coakley, K., E-mail: kevin.coakley@nist.go [National Institute of Standards and Technology, Boulder, CO 80305 (United States); Gilliam, D., E-mail: david.gilliam@nist.go [National Institute of Standards and Technology, Gaithersburg, MD 20899 (United States); Greene, G., E-mail: greenegl@ornl.go [Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); Physics Division, Oak Ridge National Lab, Building 6010, Oak Ridge, TN 37831 (United States); Laptev, A., E-mail: alaptev@nist.go [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Nico, J., E-mail: jnico@nist.go [National Institute of Standards and Technology, Gaithersburg, MD 20899 (United States); Snow, W., E-mail: wsnow@indiana.ed [Indiana University/IUCF, Bloomington, IN 47408 (United States); Wietfeldt, F., E-mail: few@tulane.ed [Tulane University, New Orleans, LA 70118 (United States); Yue, A., E-mail: ayue@nist.go [Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States)

2009-12-11

In the most accurate cold neutron beam determination of the neutron lifetime based on the absolute counting of decay protons, the largest uncertainty was attributed to the absolute determination of the capture flux of the cold neutron beam. Currently an experimental effort is underway at the National Institute of Standards and Technology (NIST) that will significantly reduce this contribution to the uncertainty in the lifetime determination. The next largest source of uncertainty is the determination of the absolute count rate of decay protons, which contributes to the experimental uncertainty approximately at the 1 s level. Experience with the recent neutron radiative decay experiment, which used the neutron lifetime apparatus, has provided valuable insights into ways to reduce other uncertainties. In addition, the cold neutron fluence rate at NIST is presently 1.5 times greater than in the 2003 measurement, and there is the prospect for a significantly higher rate with the new guide hall expansion. This paper discusses an approach for achieving a determination of the neutron lifetime with an accuracy of approximately 1 s.

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%)

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

Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Titanium

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method covers procedures for measuring reaction rates by the activation reactions 46Ti(n,p) 46Sc + 47Ti(n, np)46Sc. Note 1—Since the cross section for the (n,np) reaction is relatively small for energies less than 12 MeV and is not easily distinguished from that of the (n,p) reaction, this test method will refer to the (n,p) reaction only. 1.2 The reaction is useful for measuring neutrons with energies above approximately 4.4 MeV and for irradiation times up to about 250 days (for longer irradiations, see Practice E 261). 1.3 With suitable techniques, fission-neutron fluence rates above 109 cm–2·s–1 can be determined. However, in the presence of a high thermal-neutron fluence rate, 46Sc depletion should be investigated. 1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E 261. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 This standard does not purport to address all...

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)

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

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

A novel track density measurement method by thermal neutron activation of DYECETs

International Nuclear Information System (INIS)

Sohrabi, M.; Mahdi, Sh.

1995-01-01

A novel track density evaluation method based on thermal neutron activation of some elements of dyed electrochemically etched tracks (DYECETs) of charged particles in detectors like polycarbonate (PC) followed by measurements of gamma activity of the activated detectors is introduced. In this method, the tracks of charged particles like fast neutron-induced recoils in PC detectors were electrochemically etched, dyed by ''QuicDYECET'' methods as recently introduced by us, activated by thermal neutrons and counted for gamma activity determination to be correlated with track density. The activities of elements such as bromine-82 ( 82 Br) and sodium-24 ( 24 Na) on dyes such as Eosin Yellowish, Eosin Bluish, etc. determined by a hyper-pure germanium detector, were found to be in good correlation with DYECET density and thus particle fluence or dose. The effects of different types of dyes and their elements, dye concentration, neutron fluences and ECE durations on the DYECET density responses were studied. This new development is a method of scientific interest, potentially possessing some interesting features, as an alternative method for ECE track density determination using a gamma activity measuring system. It also has the drawback of being applicable only in centres having thermal neutron facilities. The results of the above studies are presented and discussed. (Author)

Resistivity measurements on the neutron irradiated detector grade silicon materials

Energy Technology Data Exchange (ETDEWEB)

Li, Zheng

1993-11-01

Resistivity measurements under the condition of no or low electrical field (electrical neutral bulk or ENB condition) have been made on various device configurations on detector grade silicon materials after neutron irradiation. Results of the measurements have shown that the ENB resistivity increases with neutron fluence ({Phi}{sub n}) at low {phi}{sub n} (<10{sup 13} n/cm{sup 2}) and saturates at a value between 300 and 400 k{Omega}-cm at {phi}{sub n} {approximately}10{sup 13} n/cm{sup 2}. Meanwhile, the effective doping concentration N{sub eff} in the space charge region (SCR) obtained from the C-V measurements of fully depleted p{sup +}/n silicon junction detectors has been found to increase nearly linearly with {phi}{sub n} at high fluences ({phi}{sub n} > 10{sup 13} n/cm{sup 2}). The experimental results are explained by the deep levels crossing the Fermi level in the SCR and near perfect compensation in the ENB by all deep levels, resulting in N{sub eff} (SCR) {ne} n or p (free carrier concentrations in the ENB).

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

Neutron leakage measurements from a medical linear accelerator

International Nuclear Information System (INIS)

Palta, J.R.; Hogstrom, K.R.; Tannanonta, C.

1984-01-01

The McCall method has been used to measure neutron leakage from the Mevatron 77, 18- and 15-MV photon beams. Gold foil activation has been used employing a beta counting technique for the 18-MV beam and a gamma counting technique for both the 18- and 15-MV beam. The two counting techniques were used to evaluate their relative merit. The measurements were made at various locations in the patient-treatment plane for different field sizes. The results show that the thermal-neutron dose equivalent contributes only about 1%--2% of the total neutron dose equivalent. At 100 cm, the neutron dose equivalent for the 18-MV beam is approximately six times that of the 15-MV beam, slightly exceeding the 0.1% of the useful beam criteria used by some of the regulatory agencies. In light of the uncertainty in fluence to dose equivalent conversion factors, the increased dose equivalent above 0.1% is insignificant

Activation measurements for thermal neutrons. Part J. Evaluation of thermal neutron transmission factors

International Nuclear Information System (INIS)

Egbert, Stephen D.

2005-01-01

In order to relate thermal neutron activation measurements in samples to the calculated free-in-air thermal neutron activation levels given in Chapter 3, use is made of sample transmission factors. Transmission factors account for the modification of the fluence and activation at each sample's in situ location. For the purposes of this discussion, the transmission factor (TF) is defined as the ratio of the in situ sample activation divided by the free-in-air (FIA) activation at a height of 1 m above ground at the same ground range. The procedures for calculation of TF's and example results are presented in this section. (author)

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

Intercomparison measurements with albedo neutron dosimeters

International Nuclear Information System (INIS)

Alberts, W.G.; Kluge, H.

1994-01-01

Since the introduction of the albedo dosimeter as the official personal neutron dosimeter the dosimetry services concerned have participated in intercomparison measurements at the PTB. Their albedo dosimeters were irradiated in reference fields produced by unmoderated and D 2 O-moderated 252 Cf neutron sources in the standard irradiation facility of the PTB. Six fields with fluences different in energy and angle distribution could be realised in order to determine the response of the albedo dosimeter. The dose equivalent values evaluated by the services were compared with the reference values of the PTB for the directional dose equivalent H'(10). The results turned out to be essentially dependent on the evaluation method and the choice of the calibration factors. (orig.) [de

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

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)

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

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

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.

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.

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

Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Copper

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 This test method covers procedures for measuring reaction rates by the activation reaction 63Cu(n,α)60Co. The cross section for 60Co produced in this reaction increases rapidly with neutrons having energies greater than about 5 MeV. 60Co decays with a half-life of 1925.27 days (±0.29 days)(1) and emits two gamma rays having energies of 1.1732278 and 1.332492 MeV (1). The isotopic content of natural copper is 69.17 % 63Cu and 30.83 % 65Cu (2). The neutron reaction, 63Cu(n,γ)64Cu, produces a radioactive product that emits gamma rays which might interfere with the counting of the 60Co gamma rays. 1.2 With suitable techniques, fission-neutron fluence rates above 109 cm−2·s−1 can be determined. The 63Cu(n,α)60Co reaction can be used to determine fast-neutron fluences for irradiation times up to about 15 years (for longer irradiations, see Practice E261). 1.3 Detailed procedures for other fast-neutron detectors are referenced in Practice E261. 1.4 This standard does not purport to address all of the...

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)

Activation measurements for fast neutrons. Part E. Evaluation of fast neutron 63Ni transmission factors

International Nuclear Information System (INIS)

Egbert, Stephen D.

2005-01-01

The 63 Ni measurements for fast neutrons in copper samples are compared to the calculated free-in-air 63 Ni neutron activation given in Chapter 3 by use of transmission factors. Transmission factors were calculated to account for the modification of the fluence and activation at each sample's in situ location. For the purposes of this discussion, the transmission factor (TF) is defined as the ratio of the in situ sample activation divided by the untilted free-in-air (FIA) activation at a height of 1 m above ground at the same ground range. Examples of the application of TF's will be provided in this section. (author)

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

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

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

Monte Carlo calculations of thermal neutron capture in gadolinium: a comparison of GEANT4 and MCNP with measurements.

Science.gov (United States)

Enger, Shirin A; Munck af Rosenschöld, Per; Rezaei, Arash; Lundqvist, Hans

2006-02-01

GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S(alpha,beta)] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S(alpha,beta). The location of the thermal neutron peak calculated with MCNP without S(alpha,beta) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.

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.)

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.

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.

Measurement of angular distribution of cosmic-ray muon fluence rate

International Nuclear Information System (INIS)

Lin, Jeng-Wei; Chen, Yen-Fu; Sheu, Rong-Jiun; Jiang, Shiang-Huei

2010-01-01

In this work a Berkeley Lab cosmic ray detector was used to measure the angular distribution of the cosmic-ray muon fluence rate. Angular response functions of the detector at each measurement orientation were calculated by using the FLUKA Monte Carlo code, where no energy attenuation was taken into account. Coincidence counting rates were measured at ten orientations with equiangular intervals. The muon angular fluence rate spectrum was unfolded from the measured counting rates associated with the angular response functions using both the MAXED code and the parameter adjusting method.

Measurements of DT and DD neutron yields by neutron activation on TFTR

International Nuclear Information System (INIS)

Barnes, C.W.; Larson, A.R.; LeMunyan, G.

1994-01-01

A variety of elemental foils have been activated by neutron fluence from TFTR under conditions with the DT neutron yield per shot ranging from 10 12 to over 10 18 , and with the DT/(DD+DT) neutron ratio varying from 0.5% (from triton burnup) to unity. Linear response over this large dynamic range is obtained by reducing the mass of the foils and increasing the cooling time, all while accepting greatly improved counting statistics. Effects on background gamma-ray lines from foil-capsule-material contaminants. and the resulting lower limits on activation foil mass, have been determined. DT neutron yields from dosimetry standard reactions on aluminum, chromium, iron, nickel, zirconium, and indium are in agreement within the ±9% (one-sigma,) accuracy of the measurements: also agreeing are yields from silicon foils using the ACTL library cross-section. While the ENDF/B-V library has too low a cross-section. Preliminary results from a variety of other threshold reactions are presented. Use of the 115 In(n,n) 115m In reaction (0.42 times as sensitive to DT neutrons as DD neutrons) in conjunction with pure-DT reactions allows a determination of the DT/(DD+DT) ratio in trace tritium or low-power tritium beam experiments

Measurements of DT and DD neutron yields by neutron activation on TFTR

International Nuclear Information System (INIS)

Barnes, C.W.; Larson, A.R.; LeMunyan, G.

1995-03-01

A variety of elemental foils have been activated by neutron fluence from TFTR under conditions with the DT neutron yield per shot ranging from 10 12 to over 10 18 , and with the DT/(DD+DT) neutron ratio varying from 0.5% (from triton burnup) to unity. Linear response over this large dynamic range is obtained by reducing the mass of the foils and increasing the cooling time, all while accepting greatly improved counting statistics. Effects on background gamma-ray lines from foil-capsule-material contaminants, and the resulting lower limits on activation foil mass, have been determined. DT neutron yields from dosimetry standard reactions on aluminum, chromium, iron, nickel, zirconium, and indium are in agreement within the ±9% (one-sigma) accuracy of the measurements; also agreeing are yields from silicon foils using the ACTL library cross-section, while the ENDF/B-V library has too low a cross-section. Preliminary results from a variety of other threshold reactions are presented. Use of the 115 In(n.n') 115m In reaction (0.42 times as sensitive to DT neutrons as DD neutrons) in conjunction with pure-DT reactions allows a determination of the DT/(DD+DT) ratio in trace tritium or low-power tritium beam experiments

In-wire measurement of the neutron dose rate on patients with 238Pu pacemakers implanted

International Nuclear Information System (INIS)

Piesch, E.; Burgkhardt, B.; Kollmeier, W.

1975-01-01

In-vivo measurements of the neutron dose on Medtronic pacemakers have been performed by using a proportional counter and a scintillation counter. The paper discusses the technique of free air and phantom calibration and the method of in-vivo measurement of the neutron fluence and the estimation of the dose equivalent. The neutron dose equivalent rate measured on seven patients with 238 Pu pacemakers implanted were found to be (5.6+-0.1) mRem/h at the surface of the pacemaker in 1.25 cm distance from the center of the source corresponding to a neutron emission rate of 940 ns -1 . The results are in good agreement with results of other methods reported by different authors. (Auth.)

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

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)

Advisory Committee for the Calibration Standards of Ionizing Radiation Measurement: Section 3. Neutron measurements

International Nuclear Information System (INIS)

1982-01-01

Section III (Mesures neutroniques) of the Comite Consultatif pour les Etalons de Mesure des Rayonnements Ionisants held its fifth meeting in May 1981. Recent work carried out at BIPM in the field of neutron measurements was reported. The status of a full-scale 252 Cf neutron source intercomparison (10 7 s - 1 ) and of several restricted comparisons was discussed. Intercomparisons of fast neutron fluence rates are in progress ( 115 In(n,n') 115 Insup(m); NB/Zr) or will take place in the near future ( 115 n(n,#betta#) 116 Insup(m); 235 U and 238 U fission chambers). An intercomparison of neutron dosimetry standards by circulating tissue-equivalent ion chambers will be prepared and organized by BIPM. Finally, there was a broad exchange of information on work in progress at the various laboratories represented at the meeting [fr

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.

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

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

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)

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)

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.

Neutron spectrometry and dosimetry measurement at workplaces for calibration of individual PGP-DIN dosemeters

International Nuclear Information System (INIS)

Itie, C.; Muller, H.; Asselineau, B.; Medioni, R.; Crovisier, P.; Valier-Bradier, P.; Groetz, J.E.; Piot, J.

2003-01-01

Measurements to determine new coefficients for individual neutron dosimeters PGP-DIN complying with the ICRP 60 recommendations were performed at two workplaces at the CEA of Valduc: a storage room and a plutonium reprocessing plant. Two spectrometry campaigns were performed allowing a better assessment of doses received by operators working at these workplaces. Neutron energy fluence and ambient dose equivalent rate H * (10) distributions were measured as function of neutron energy by using the ROSPEC device and BONNER spheres spectrometer. The radiation field being mixed neutron and gamma, the gamma component was also evaluated: neutron and photon dose-rate meters were used to evaluate the ambient dose rate equivalent. Individual dosemeters were positioned on an ISO water slab phantom. In addition, calculations were performed using the MCNP simulation code for different configurations. (authors)

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

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)

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

Characteristics of GaAs MESFET inverters exposed to high energy neutrons

International Nuclear Information System (INIS)

Bloss, W.L.; Yamada, W.E.; Young, A.M.; Janousek, B.K.

1988-01-01

GaAs MESFET circuits have been exposed to high energy neutrons with fluences ranging from 1x10/sup 14/ n/cm/sup 2/ to 2x10/sup 15/ m/cm/sup 2/. Discrete transistors, inverters, and ring oscillators were characterized at each fluence. While the MESFETs exhibit significant threshold voltage shifts and transconductance and saturation current degradation over this range of neutron fluences, the authors have observed improvement in the DC characteristics of Schottky Diode FET Logic (SDFL) inverters. This unusual result has been successfully simulated using device parameters extracted from FETs damaged by exposure to high energy neutrons. Although the decrease in device transconductance results in an increase in inverter gate delay, as reflected in ring oscillator frequency measurements, the authors conclude that GaAs ICs fabricated from this logic family will remain functional after exposure to extreme neutron fluences. This is a consequence of the observed improvement in inverter noise margin evident in both measured and simulated circuit performance

Measured Neutron Spectra and Dose Equivalents From a Mevion Single-Room, Passively Scattered Proton System Used for Craniospinal Irradiation

Energy Technology Data Exchange (ETDEWEB)

Howell, Rebecca M., E-mail: rhowell@mdanderson.org [Department of Radiation Physics, The University of Texas M. D. Anderson Cancer Center, Houston, Texas (United States); Burgett, Eric A.; Isaacs, Daniel [Department of Nuclear Engineering, Idaho State University, Pocatello, Idaho (United States); Price Hedrick, Samantha G.; Reilly, Michael P.; Rankine, Leith J.; Grantham, Kevin K.; Perkins, Stephanie; Klein, Eric E. [Department of Radiation Oncology, Washington University, St. Louis, Missouri (United States)

2016-05-01

Purpose: To measure, in the setting of typical passively scattered proton craniospinal irradiation (CSI) treatment, the secondary neutron spectra, and use these spectra to calculate dose equivalents for both internal and external neutrons delivered via a Mevion single-room compact proton system. Methods and Materials: Secondary neutron spectra were measured using extended-range Bonner spheres for whole brain, upper spine, and lower spine proton fields. The detector used can discriminate neutrons over the entire range of the energy spectrum encountered in proton therapy. To separately assess internally and externally generated neutrons, each of the fields was delivered with and without a phantom. Average neutron energy, total neutron fluence, and ambient dose equivalent [H* (10)] were calculated for each spectrum. Neutron dose equivalents as a function of depth were estimated by applying published neutron depth–dose data to in-air H* (10) values. Results: For CSI fields, neutron spectra were similar, with a high-energy direct neutron peak, an evaporation peak, a thermal peak, and an intermediate continuum between the evaporation and thermal peaks. Neutrons in the evaporation peak made the largest contribution to dose equivalent. Internal neutrons had a very low to negligible contribution to dose equivalent compared with external neutrons, largely attributed to the measurement location being far outside the primary proton beam. Average energies ranged from 8.6 to 14.5 MeV, whereas fluences ranged from 6.91 × 10{sup 6} to 1.04 × 10{sup 7} n/cm{sup 2}/Gy, and H* (10) ranged from 2.27 to 3.92 mSv/Gy. Conclusions: For CSI treatments delivered with a Mevion single-gantry proton therapy system, we found measured neutron dose was consistent with dose equivalents reported for CSI with other proton beamlines.

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.

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.

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.

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

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.

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.)

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.

DIANE, a simulation code for the interaction of neutrons with living tissues. Application to low doses of fast neutrons on human tumoral cells; DIANE, un code de simulation de l'interaction des neutrons avec la matiere vivante. Applications aux faibles doses de neutrons rapides sur des cellules tumorales humaines

Energy Technology Data Exchange (ETDEWEB)

Nenot, M.L

2003-07-15

Our work deals with the irradiation of cells and living tissues by 14 MeV neutrons at very low doses (a few 10{sup -2} Gy). Such experiments require an accurate knowledge of the values of neutron dose rates and fluences at the level of cell cultures. We have performed measurements of fluence rates through an activation method applied to gold and copper foils. The fluence rate is deduced from the gamma rays emitted by the irradiated foils. Neutron doses and dose rates have been measured through varied methods: PIN diodes, ionization tissue equivalent chambers, and Geiger-Mueller counters. We have designed the DIANE code to simulate the impact of energetic neutrons on cells. This code can be used with isolated cells or macroscopic tissues, it takes into account the roles of the ionisation electrons produced by recoil nuclei entering the cell. This point is all the more important since recent works have highlighted the impact of very low energy electrons on DNA. (A.C.)

DIANE, a simulation code for the interaction of neutrons with living tissues. Application to low doses of fast neutrons on human tumoral cells

International Nuclear Information System (INIS)

Nenot, M.L.

2003-07-01

Our work deals with the irradiation of cells and living tissues by 14 MeV neutrons at very low doses (a few 10 -2 Gy). Such experiments require an accurate knowledge of the values of neutron dose rates and fluences at the level of cell cultures. We have performed measurements of fluence rates through an activation method applied to gold and copper foils. The fluence rate is deduced from the gamma rays emitted by the irradiated foils. Neutron doses and dose rates have been measured through varied methods: PIN diodes, ionization tissue equivalent chambers, and Geiger-Mueller counters. We have designed the DIANE code to simulate the impact of energetic neutrons on cells. This code can be used with isolated cells or macroscopic tissues, it takes into account the roles of the ionisation electrons produced by recoil nuclei entering the cell. This point is all the more important since recent works have highlighted the impact of very low energy electrons on DNA. (A.C.)

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)

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.

Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 This test method covers procedures measuring reaction rates by the activation reaction 27Al(n,α)24Na. 1.2 This activation reaction is useful for measuring neutrons with energies above approximately 6.5 MeV and for irradiation times up to about 2 days (for longer irradiations, see Practice E261). 1.3 With suitable techniques, fission-neutron fluence rates above 106 cm−2·s−1 can be determined. 1.4 Detailed procedures for other fast neutron detectors are referenced in Practice E261. 1.5 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.

Mercury mass measurement in fluorescent lamps via neutron activation analysis

International Nuclear Information System (INIS)

Viererbl, L.; Vinš, M.; Lahodová, Z.; Fuksa, A.; Kučera, J.; Koleška, M.; Voljanskij, A.

2015-01-01

Mercury is an essential component of fluorescent lamps. Not all fluorescent lamps are recycled, resulting in contamination of the environment with toxic mercury, making measurement of the mercury mass used in fluorescent lamps important. Mercury mass measurement of lamps via instrumental neutron activation analysis (NAA) was tested under various conditions in the LVR-15 research reactor. Fluorescent lamps were irradiated in different positions in vertical irradiation channels and a horizontal channel in neutron fields with total fluence rates from 3×10 8 cm −2 s −1 to 10 14 cm −2 s −1 . The 202 Hg(n,γ) 203 Hg nuclear reaction was used for mercury mass evaluation. Activities of 203 Hg and others induced radionuclides were measured via gamma spectrometry with an HPGe detector at various times after irradiation. Standards containing an Hg 2 Cl 2 compound were used to determine mercury mass. Problems arise from the presence of elements with a large effective cross section in luminescent material (europium, antimony and gadolinium) and glass (boron). The paper describes optimization of the NAA procedure in the LVR-15 research reactor with particular attention to influence of neutron self-absorption in fluorescent lamps. - Highlights: • Mercury is an essential component of fluorescent lamps. • Fluorescent lamps were irradiated in neutron fields in research reactor. • 203 Hg induced radionuclide activity was measured using gamma spectrometry. • Mercury mass in fluorescent lamps can be measured by neutron activation analysis.

Silicon diode measurements for monoenergetic neutrons and critical assemblies (H.P.R.R. and VIPER)

International Nuclear Information System (INIS)

Delafield, H.J.; Reading, A.H.

1981-04-01

The response of the silicon diode (AEI FNDD1) has been measured for monoenergetic neutrons of mean energies 0.56, 2.00 and 3.68 MeV. Using conversion factors from neutron fluence to kerma (ICRU, 1977) it is shown that the theoretical kerma response in muscle tissue is substantially uniform (+- 20%) over the neutron energy range from 250 keV to 17 MeV. Diode measurements were made at the Health Physics Research Reactor at the Oak Ridge National Laboratory, Tennessee, U.S.A., during the 1979 international intercomparison of nuclear accident dosimetry systems. Measurements of kerma in free air and of the surface absorbed dose on the front surface of a phantom were made with the reactor bare, shielded by 20 cm concrete and by 5 cm steel. Further tests were made at the VIPER reactor at AWRE. These diode measurements, covering a range of neutron spectra, were in good agreement (+- 20%) with measurements made by the threshold detector system. (author)

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)

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)

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)

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)

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...

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

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

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.

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.

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.

Mercury mass measurement in fluorescent lamps via neutron activation analysis

Science.gov (United States)

Viererbl, L.; Vinš, M.; Lahodová, Z.; Fuksa, A.; Kučera, J.; Koleška, M.; Voljanskij, A.

2015-11-01

Mercury is an essential component of fluorescent lamps. Not all fluorescent lamps are recycled, resulting in contamination of the environment with toxic mercury, making measurement of the mercury mass used in fluorescent lamps important. Mercury mass measurement of lamps via instrumental neutron activation analysis (NAA) was tested under various conditions in the LVR-15 research reactor. Fluorescent lamps were irradiated in different positions in vertical irradiation channels and a horizontal channel in neutron fields with total fluence rates from 3×108 cm-2 s-1 to 1014 cm-2 s-1. The 202Hg(n,γ)203Hg nuclear reaction was used for mercury mass evaluation. Activities of 203Hg and others induced radionuclides were measured via gamma spectrometry with an HPGe detector at various times after irradiation. Standards containing an Hg2Cl2 compound were used to determine mercury mass. Problems arise from the presence of elements with a large effective cross section in luminescent material (europium, antimony and gadolinium) and glass (boron). The paper describes optimization of the NAA procedure in the LVR-15 research reactor with particular attention to influence of neutron self-absorption in fluorescent lamps.

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.

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

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)

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)

Ultra Low Level Environmental Neutron Measurements Using Superheated Droplet Detectors

Energy Technology Data Exchange (ETDEWEB)

Fernandes, A.C. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Estrada Nacional 10 - km 139.7, 2695-066 Bobadela LRS (Portugal); Centro de Fisica Nuclear, Universidade de Lisboa. Av. Prof. Gama Pinto, 2, 1649- 003 Lisboa (Portugal); Felizardo, M.; Girard, T.A.; Kling, A.; Ramos, A.R. [Centro de Fisica Nuclear, Universidade de Lisboa. Av. Prof. Gama Pinto, 2, 1649- 003 Lisboa (Portugal); Marques, J.G.; Prudencio, M.I.; Marques, R.; Carvalho, F.P. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Estrada Nacional 10 - km 139.7, 2695-066 Bobadela LRS (Portugal)

2015-07-01

Through the application of superheated droplet detectors (SDDs), the SIMPLE project for the direct search for dark matter (DM) reached the most restrictive limits on the spin-dependent sector to date. The experiment is based on the detection of recoils following WIMP-nuclei interaction, mimicking those from neutron scattering. The thermodynamic operation conditions yield the SDDs intrinsically insensitive to radiations with linear energy transfer below ∼150 keVμm{sup -1} such as photons, electrons, muons and neutrons with energies below ∼40 keV. Underground facilities are increasingly employed for measurements in a low-level radiation background (DM search, gamma-spectroscopy, intrinsic soft-error rate measurements, etc.), where the rock overburden shields against cosmic radiation. In this environment the SDDs are sensitive only to α-particles and neutrons naturally emitted from the surrounding materials. Recently developed signal analysis techniques allow discrimination between neutron and α-induced signals. SDDs are therefore a promising instrument for low-level neutron and α measurements, namely environmental neutron measurements and α-contamination assays. In this work neutron measurements performed in the challenging conditions of the latest SIMPLE experiment (1500 mwe depth with 50-75 cm water shield) are reported. The results are compared with those obtained by detailed Monte Carlo simulations of the neutron background induced by {sup 238}U and {sup 232}Th traces in the facility, shielding and detector materials. Calculations of the neutron energy distribution yield the following neutron fluence rates (in 10{sup -8} cm{sup -2}s{sup -1}): thermal (<0.5 eV): 2.5; epithermal (0.5 eV-100 keV): 2.2; fast (>1 MeV): 3.9. Signal rates were derived using standard cross sections and codes routinely employed in reactor dosimetry. The measured and calculated neutron count rates per unit of active mass were 0.15 ct/kgd and 0.33 ct/kg-d respectively. As the major

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

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.

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

Measurement of gamma-ray production cross sections in neutron-induced reactions for Al and Pb

International Nuclear Information System (INIS)

Pavlik, A.; Vonach, H.; Hitzenberger, H.

1995-01-01

The prompt gamma-radiation from the interaction of fast neutrons with aluminum and lead was measured using the white neutron beam of the WNR facility at the Los Alamos National Laboratory. The samples (Al and isotopically enriched 207 Pb and 208 Pb) were positioned at about 20 m or 41 m distance from the neutron production target. The spectra of the emitted gamma-rays were measured with a high-resolution HPGe detector. The incident neutron energy was determined by the time-of-flight method and the neutron fluence was measured with a U fission chamber. From the aluminum gamma-ray spectra excitation functions for prominent gamma-transitions in various residual nuclei (in the range from O to Al) were derived for neutron energies from 3 MeV to 400 MeV. For lead (n,xnγ) reactions were studied for neutron energies up to 200 MeV by analyzing prominent gamma-transitions in the residual nuclei 200,202,204,206,207,208 Pb. The experimental results were compared with nuclear model calculations using the code GNASH. A good overall agreement was obtained without special parameter adjustments

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.)

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

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)

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.

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

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)

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)

Detection and measurement of neutron-irradiated gemstones

International Nuclear Information System (INIS)

Bunnak, S.; Jerachanchai, S.; Chinudomsub, K.; Saiyut, K.

1990-01-01

Color enhance gemstone, neutron-irradiated topaz, was analyzed by gamma spectrometry for examining characteristic and activity. Topaz was irradiated in the wet-tube facility of the Research Reactor TRR/1 which neutron fluence is 2.52x10 17 neutron per square centimeter. After 100 days of decay, topaz was sampling to the qualitative and quantitative analysis using multichannel analyzer of Nuclear Data Model ND65 and hyper pure germanium detector. Calculation and evaluation were done by microcomputer IBM/PC 640 KB RAM. The qualitative analysis showed that the neutron-irradiated topaz has 2 major isotopes, i.e., Ta-182 and Sc-46. Quantitative activity was compared with reference standard source Eu-152 (NBS) and the results were shown in the table 1. The Health Physics Division, OAEP, inspected on 6240.9 gm of the neutron-irradiated topaz using standard release limit 2 nCi/gm (74 Bq/gm). It was found that only 423.9 gm out of the total amount were over the standard release limit

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}.

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

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

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

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)

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.)

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 ...

Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

DESIG: E 263 09 ^TITLE: Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron ^SIGNUSE: Refer to Guide E 844 for guidance on the selection, irradiation, and quality control of neutron dosimeters. Refer to Practice E 261 for a general discussion of the determination of fast-neutron fluence rate with threshold detectors. Pure iron in the form of foil or wire is readily available and easily handled. Fig. 1 shows a plot of cross section as a function of neutron energy for the fast-neutron reaction 54Fe(n,p)54Mn (1). This figure is for illustrative purposes only to indicate the range of response of the 54Fe(n,p)54Mn reaction. Refer to Guide E 1018 for descriptions of recommended tabulated dosimetry cross sections. 54Mn has a half-life of 312.13 days (3) (2) and emits a gamma ray with an energy of 834.845 keV (5). (2) Interfering activities generated by neutron activation arising from thermal or fast neutron interactions are 2.57878 (46)-h 56Mn, 44.95-d (8) 59Fe, and 5.27...

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

Use of a newly developed active thermal neutron detector for in-phantom measurements in a medical LINAC

Energy Technology Data Exchange (ETDEWEB)

Bodogni, R.; Sanchez-Doblado, F.; Pola, A.; Gentile, A.; Esposito, A.; Gomez-ros, J. M.; Pressello, M. C.; Lagares, J. I.; Terron, J. A.; Gomez, F.

2013-07-01

In this work a newly developed active thermal neutron detector, based on a solid state analog device, was used to determine the thermal neutron fluence in selected positions of a simplified human phantom undergoing radiotherapy with a 15 MV LINAC. The results are compared with TLD, the predictions from a Monte Carlo simulation and with measurements indirectly performed with a digital device, located far from the phantom, inside the treatment room. In this work only TLD comparison is presented. Since active neutron instruments are usually affected by systematic deviations when used in a pulsed field with large photon background, the new detector offered in this work may represent an innovative and useful tool for neutron evaluations in accelerator-based radiotherapy. (Author)

Experimental Determination of the Neutron Radiation-Dose Distribution in the Human Phantom

Energy Technology Data Exchange (ETDEWEB)

Stipcic, Neda [Institute Rudjer Bogkovic, Zagreb, Yugoslavia (Serbia)

1967-01-15

The quality of the radiation delivering the radiation dose to the human phantom is quite different from that of the incident neutron beam. This paper describes the experimental investigation of the variation of neutron dose related to the variation of neutron fluence with depth in the human phantom. The distribution of neutron radiation was determined in the human phantom - a cube of paraffin wax 25 cm x 25 cm x 50 cm with a density of 0.92 cm{sup -3}. Po-Be and Ra-Be point sources were used as neutron sources. Neutron fluences were measured using different types of detector: scintillation detector, BF{sub 3} counter, and nuclear-track emulsions. Since the fluence measurements with these three types of detectors were carried out under the same experimental conditions, it was possible to separate and analyse each part of the radiation dose in the paraffin. From the investigations, the distribution of the total radiation dose was obtained as a function of the paraffin depth. The maximum value of this dose distribution is constant with respect to the distance between the source and the paraffin phantom. From the results obtained, some conclusions may be drawn concerning the amount of absorbed radiation dose in the human phantom. (author)

SU-E-T-365: Estimation of Neutron Ambient Dose Equivalents for Radioprotection Exposed Workers in Radiotherapy Facilities Based On Characterization Patient Risk Estimation

Energy Technology Data Exchange (ETDEWEB)

Irazola, L; Terron, J; Sanchez-Doblado, F [Departamento de Fisiologia Medica y Biofisica, Universidad de Sevilla (Spain); Servicio de Radiofisica, Hospital Universitario Virgen Macarena, Sevilla (Spain); Domingo, C; Romero-Exposito, M [Departament de Fisica, Universitat Autonoma de Barcelona, Bellaterra (Spain); Garcia-Fuste, M [Health and Safety Department, ALBA Synchrotron Light Source, Cerdanyola del Valles (Spain); Sanchez-Nieto, B [Instituto de Fisica, Pontificia Universidad Catolica de Chile, Santiago (Chile); Bedogni, R [Laboratori Nazionali di Frascati, Istituto Nazionale di Fisica Nucleare (INFN) (Italy)

2015-06-15

Purpose: Previous measurements with Bonner spheres{sup 1} showed that normalized neutron spectra are equal for the majority of the existing linacs{sup 2}. This information, in addition to thermal neutron fluences obtained in the characterization procedure{sup 3}3, would allow to estimate neutron doses accidentally received by exposed workers, without the need of an extra experimental measurement. Methods: Monte Carlo (MC) simulations demonstrated that the thermal neutron fluence distribution inside the bunker is quite uniform, as a consequence of multiple scatter in the walls{sup 4}. Although inverse square law is approximately valid for the fast component, a more precise calculation could be obtained with a generic fast fluence distribution map around the linac, from MC simulations{sup 4}. Thus, measurements of thermal neutron fluences performed during the characterization procedure{sup 3}, together with a generic unitary spectra{sup 2}, would allow to estimate the total neutron fluences and H*(10) at any point{sup 5}. As an example, we compared estimations with Bonner sphere measurements{sup 1}, for two points in five facilities: 3 Siemens (15–23 MV), Elekta (15 MV) and Varian (15 MV). Results: Thermal neutron fluences obtained from characterization, are within (0.2–1.6×10{sup 6}) cm−{sup 2}•Gy{sup −1} for the five studied facilities. This implies ambient equivalent doses ranging from (0.27–2.01) mSv/Gy 50 cm far from the isocenter and (0.03–0.26) mSv/Gy at detector location with an average deviation of ±12.1% respect to Bonner measurements. Conclusion: The good results obtained demonstrate that neutron fluence and H*(10) can be estimated based on: (a) characterization procedure established for patient risk estimation in each facility, (b) generic unitary neutron spectrum and (c) generic MC map distribution of the fast component. [1] Radiat. Meas (2010) 45: 1391 – 1397; [2] Phys. Med. Biol (2012) 5 7:6167–6191; [3] Med. Phys (2015) 42

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.

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

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

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

International Nuclear Information System (INIS)

Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F.; Vermeeren, L.; Lopez, A. Legrand

2011-01-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10 20 n/cm 2 . A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

Science.gov (United States)

Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

Energy Technology Data Exchange (ETDEWEB)

Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F. [CEA, DEN, Cadarache, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lopez, A. Legrand [CEA, DEN, Saclay, SIREN/LECSI, F-91400 Saclay (France)

2011-03-15

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10{sup 20} n/cm{sup 2}. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

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)

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.

The Production of Low-energy Neutrons in Solar Flares and the Importance of Their Detection in the Inner Heliosphere

Science.gov (United States)

Murphy, R. J.; Kozlovsky, B.; Share, G. H.

2012-09-01

Neutron detectors on spacecraft in the inner heliosphere can observe the low-energy (computer code incorporating updated neutron-production cross sections for the proton and α-particle reactions with heavier elements at all ion energies, especially at low energies (E ion 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.

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)

Method to determine the strength of a neutron source

Energy Technology Data Exchange (ETDEWEB)

Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Chacon R, A.; Mercado, G.A. [UAZ, A.P. 336, 98000 Zacatecas (Mexico); Gallego, E.; Lorente, A. [Depto. Ingenieria Nuclear, Universidad Politecnica de Madrid, (Spain)

2006-07-01

The use of a gamma-ray spectrometer with a 3 {phi} x 3 NaI(Tl) detector, with a moderator sphere has been studied in the aim to measure the neutron fluence rate and to determine the source strength. Moderators with a large amount of hydrogen are able to slowdown and thermalize neutrons; once thermalized there is a probability that thermal neutron to be captured by hydrogen producing 2.22 MeV prompt gamma-ray. The pulse-height spectrum collected in a multicharmel analyzer shows a photopeak around 2.22 MeV whose net area is proportional to total neutron fluence rate and to the neutron source strength. The characteristics of this system were determined by a Monte Carlo study using the MCNP 4C code, where a detailed model of the Nal(Tl) was utilized. As moderators 3, 5, and 10 inches-diameter spheres where utilized and the response was calculated for monoenergetic and isotopic neutrons sources. (Author)

Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic

Energy Technology Data Exchange (ETDEWEB)

Song, P; Holder, J; Young, B; Kalantar, D; Eder, D; Kimbrough, J

2006-11-02

The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the use of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location ({approx}1.7 m from the target) would be {approx}1.4e9/cm{sup 2}. Previous measurements suggest the onset of significant background at a neutron fluence of {approx} 1e8/cm{sup 2}. The radiation damage and operational upsets which starts at {approx}1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor {approx}50.

Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic

International Nuclear Information System (INIS)

Song, P; Holder, J; Young, B; Kalantar, D; Eder, D; Kimbrough, J

2006-01-01

The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the use of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location (∼1.7 m from the target) would be ∼1.4e9/cm 2 . Previous measurements suggest the onset of significant background at a neutron fluence of ∼ 1e8/cm 2 . The radiation damage and operational upsets which starts at ∼1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor ∼50

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)

Proton and Neutron Irradiation Tests of Readout Electronics of the ATLAS Hadronic Endcap Calorimeter

CERN Document Server

Menke, Sven; The ATLAS collaboration

2012-01-01

The readout electronics of the ATLAS Hadronic Endcap Calorimeter will have to withstand the about ten times larger radiation environment of the future high-luminosity LHC (HL-LHC) compared to their design values. The GaAs ASIC which comprises the heart of the readout electronics has been exposed to neutron and proton radiation with fluences up to ten times the total expected fluences for ten years of running of the HL-LHC. Neutron tests where performed at the NPI in Rez, Czech Republic, where a 36 MeV proton beam is directed on a thick heavy water target to produce neutrons. The proton irradiation was done with 200 MeV protons at the PROSCAN area of the Proton Irradiation Facility at the PSI in Villigen, Switzerland. In-situ measurements of S-parameters in both tests allow the evaluation of frequency dependent performance parameters - like gain and input impedance - as a function of the fluence. The linearity of the ASIC response has been measured directly in the neutron tests with a triangular input pulse of...

Proton and Neutron Irradiation Tests of Readout Electronics of the ATLAS Hadronic Endcap Calorimeter

CERN Document Server

INSPIRE-00106910

2012-01-01

The readout electronics of the ATLAS Hadronic Endcap Calorimeter will have to withstand the about ten times larger radiation environment of the future high-luminosity LHC (HL-LHC) compared to their design values. The GaAs ASIC which comprises the heart of the readout electronics has been exposed to neutron and proton radiation with fluences up to ten times the total expected fluences for ten years of running of the HL-LHC. Neutron tests were performed at the NPI in Rez, Czech Republic, where a 36 MeV proton beam is directed on a thick heavy water target to produce neutrons. The proton irradiation was done with 200 MeV protons at the PROSCAN area of the Proton Irradiation Facility at the PSI in Villigen, Switzerland. In-situ measurements of S-parameters in both tests allow the evaluation of frequency dependent performance parameters - like gain and input impedance - as a function of the fluence. The linearity of the ASIC response has been measured directly in the neutron tests with a triangular input pulse of ...

Sequential measurements of spectrum and dose for cosmic-ray neutrons on the ground

International Nuclear Information System (INIS)

Hirabayashi, N.; Nunomiya, T.; Suzuki, H.; Nakamura, T.

2002-01-01

The earth is continually bathed in high-energy particles that come from outside the solar system, known as galactic cosmic rays. When these particles penetrate the magnetic fields of the solar system and the Earth and reach the Earth's atmosphere, they collide with atomic nuclei in air and secondary cosmic rays of every kind. On the other hand, levels of accumulation of the semiconductor increase recently, and the soft error that the cosmic-ray neutrons cause has been regarded as questionable. There have been long-term measurements of cosmic-ray neutron fluence at several places in the world, but no systematic study on cosmic-ray neutron spectrum measurements. This study aimed to measure the cosmic-ray neutron spectrum and dose on the ground during the solar maximum period of 2000 to 2002. Measurements have been continuing in a cabin of Tohoku University Kawauchi campus, by using five multi-moderator spectrometers (Bonner sphere), 12.7 cm diam by 12.7 cm long NE213 scintillator, and rem counter. The Bonner sphere uses a 5.08 cm diam spherical 3 He gas proportional counter and the rem counter uses a 12.7 cm diam 3 He gas counter. The neutron spectra were obtained by unfolding from the count rates measured with the Bonner sphere using the SAND code and the pulse height spectra measured with the NE213 scintillator using the FORIST code . The cosmic- ray neutron spectrum and ambient dose rates have been measured sequentially from April 2001. Furthermore, the correlation between ambient dose rate and the atmospheric pressure was investigated with a barometer. We are also very much interested in the variation of neutron spectrum following big solar flares. From the sequential measurements, we found that the cosmic-ray neutron spectrum has two peaks at around 1 MeV and at around 100 MeV, and the higher energy peak increases with a big solar flare

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.

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

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)

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

Microhardness measurement in AISI 321 stainless steel with niobium additions before and after fast neutron irradiation

International Nuclear Information System (INIS)

Galli, V.L.; Lucki, G.

1980-01-01

Data about influence of neutron irradiation on the microhardness of stainless steel of type AISI 321 with 0.05 and 0.1wt.% Nb additions are presented. The microhardness measurements were made in the range of 300 to 650 0 C, before and after fast neutron irradiation with fluences about 10 17 n/cm 2 . Our results indicate that radiation damage peaks occur around 480 0 C for the stainless steel of type AISI 321 without Nb addition, around 500 0 C for the composition with 0.05 wt.% Nb addition and around 570 0 C for the composition with 0.1 wt.% Nb addition. Microhardness data are in agreement with those obtained by means of electrical resistivity measurements, performed at the same conditions. (Author) [pt

Measurements of the {sup 235}U(n,f) cross section in the 3 to 30 MeV neutron energy region

Energy Technology Data Exchange (ETDEWEB)

Carlson, A.D.; Wasson, O.A. [National Institute of Standards and Technology, Gaithersburg, MD (United States); Lisowski, P.W. [Los Alamos National Lab., NM (United States)] [and others

1991-12-31

To improve the accuracy of the {sup 235}U(n,f) cross section, measurements have been made of this standard cross section at the target 4 facility at Los Alamos National Laboratory (LANL). The data were obtained at the 20-meter flight path of that facility. The fission reaction rate was determined with a fast parallel plate ionization chamber and the neutron fluence was measured with an annular proton recoil telescope. The measurements provide the shape of the {sup 235}U(n,f) cross section relative to the hydrogen scattering cross section for neutron energies from about 3 to 30 MeV neutron energy. The data have been normalized to the very accurately known value near 14 MeV. The results are in good agreement with the ENDF/B-VI evaluation up to about 15 MeV neutron energy. Above this energy differences as large as 5% are observed.

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

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

The energy spectrum of cosmic-ray induced neutrons measured on an airplane over a wide range of altitude and latitude

International Nuclear Information System (INIS)

Goldhagen, P.; Clem, J. M.; Wilson, J. W.

2004-01-01

Crews of high-altitude aircraft are exposed to radiation from galactic cosmic rays (GCRs). To help determine such exposures, the Atmospheric Ionizing Radiation Project, an international collaboration of 15 laboratories, made simultaneous radiation measurements with 14 instruments on a NASA ER-2 high-altitude airplane. The primary instrument was a sensitive extended-energy multisphere neutron spectrometer. Its detector responses were calculated for energies up to 100 GeV using the radiation transport code MCNPX 2.5.d with improved nuclear models and including the effects of the airplane structure. New calculations of GCR-induced particle spectra in the atmosphere were used to correct for spectrometer counts produced by protons, pions and light nuclear ions. Neutron spectra were unfolded from the corrected measured count rates using the deconvolution code MAXED 3.1. The results for the measured cosmic-ray neutron spectrum (thermal to >10 GeV), total neutron fluence rate, and neutron dose equivalent and effective dose rates, and their dependence on altitude and geomagnetic cut-off agree well with results from recent calculations of GCR-induced neutron spectra. (authors)

Measurement of fast neutron spectrum using CR-39 detectors and a new image analysis program (autoTRAKn)

International Nuclear Information System (INIS)

Paul, Sabyasachi; Tripathy, S.P.; Sahoo, G.S.; Bandyopadhyay, T.; Sarkar, P.K.

2013-01-01

An attempt is made to estimate the neutron spectrum using the CR-39 (Solid state nuclear track) detector and a new image analyzing program. For this purpose the earlier developed program (autoTRAK) is modified by introducing the required features such as angular correction for the recoil particles, fluence-to-dose conversion coefficient, detection sensitivity of CR-39 detectors, etc. to make it applicable for neutron spectrometry and dosimetry. This upgraded program (autoTRAK n ) is tested with a mono-energetic source (D–T) and two other standard neutron sources, viz. 241 Am–Be and 252 Cf. The program is validated by reproducing these standard spectra, and comparing with the spectra reported by other investigators using different measuring techniques. The ratios of dose equivalent (H ⁎ (10)) to fluence (Φ) are also estimated from the spectra and are compared with the reference values for these neutron sources. An additional feature of this program is explored for counting high density overlapping tracks more precisely and effectively compared to other commonly used image analyzing softwares. This method is found to be simple and promising, which can always be used as a supplementary measuring technique. The details of the modified program, reproduction and comparison of the neutron spectra, reproducibility of the methodology and example of overlapping track counting are presented and discussed. -- Highlights: •A novel image analysis technique (autoTRAK n ) is developed to evaluate CR-39 detectors used for neutron spectrometry and dosimetry. •The methodology is tested to reproduce three standard neutron spectra, (a) D–T, (b) 241 Am–Be, and (c) 252 Cf. •A good matching is observed between dosimetric values obtained by the program and the available reference values. •The program autoTRAK n is also observed to be efficient to distinguish high density overlapping tracks without any segregation procedure. •The methodology seems to be simple, which

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)

Low-energy neutron measurements in an iron calorimeter structure irradiated by 200 GeV/c hadrons

Energy Technology Data Exchange (ETDEWEB)

Russ, J S [Carnegie-Mellon University, Pittsburgh, PA (United States); Stevenson, G R; Fasso, A; Nielsen, M C [CERN, Geneva (Switzerland); Furetta, C; Rancoita, P G; Vismara, I [INFN, Milan (Italy)

1989-04-21

Of serious concern in the design of detectors for the new high-luminosity hadron-hadron colliders are the radiation damage effects on silicon and other detectors of low-energy neutrons produced by spallation evaporation or fission processes. Because of the lack of experimental information on the number of neutrons with energies between 0.1 and 10 MeV in the cascades originating from high-energy hadrons, an experiment was carried out using activation detector techniques to measure the neutron fluence in a cascade initiated by 200 GeV hadrons in acalorimeter-like iron structure. It was found that at the maximum of the cascade one produces approximately 3 neutrons per GeV of incident energy: some 70% of these are of energies between 0.1 and 5 MeV, the remainder are fairly uniformly distributed in energy between 5 and several hundred MeV. The number of albedo neutrons leaving the front face of the calorimeter structure was about 0.3 neutrons per GeV of incident energy with in energy distribution similar to those at cascade maximum These data confirm that neutron-induced damage will he of concern in the design of detectors for the new colliders and that further measurements and calculations are necessary for a correct assessment of this damage. (author)

An active pixels spectrometers for neutronic fields metrology

International Nuclear Information System (INIS)

Taforeau, Julien

2013-01-01

The fundamental metrology is responsible for the sustainability of the measurement systems and handles to supply the reference standards. Concerning the metrology of ionizing radiations and, in particular the neutron metrology, detectors standards are used to characterize reference fields, in terms of energy and fluence. The dosimeters or particle detectors are calibrated on these reference fields. This thesis presents the development of a neutron spectrometer neutron candidate to the status of primary standard for the characterization of neutron fields in the range from 5 to 20 MeV. The spectrometer uses the recoil proton telescope as detection principle; the CMOS technology, through three sensor positions, is taking advantage to realize the tracking of protons. A Si(Li) detector handles the measure of the residual proton energy. The device simulations, realized under MCNPX, allow to estimate its performances and to validate the neutron energy reconstruction. An essential step of characterization of the telescope elements and in particular of CMOS sensors is also proposed to guarantee the validity of posterior experimental measurements. The tests realized as well in mono-energy fields as in radionuclide source show the very good performances of the system. The quantification of uncertainties indicates an energy estimation with 1.5 % accuracy and a resolution of less than 6 %. The fluence measurement is performed with an uncertainty about 4 to 6%. (author)

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.

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.

Fast neutron response of coumarin in water and heavy water

International Nuclear Information System (INIS)

Krishnan, D.; Kher, R.K.; Gopakumar, K.; Bhandari, N.S.

1979-01-01

Response of coumarin in aqueous solution has been studied earlier for gamma rays and fast neutrons by fluorescence measurement. For further fast neutron studies, two systems viz coumarin in H 2 0 and coumarin in D 2 0, were irradiated with fast neutrons in SNIF facility in the swimming pool type APSARA reactor at Trombay. Neutron fluence was estimated by measuring induced activity in sulphur pellet and associated gamma radiation was estimated using CaS0 4 :Dy TLD powder. The KERMA values were calculated for H 2 0 and D 2 0, assuming modified fission spectrum for fast neutron in SNIF position, and they were in the ratio of 2:1. Response of a chemical dosimetric system is expected to be proportional to the absorbed dose in the respective system for the same neutron fluence. This was experimentally found to be the case for coumarin in H 2 0 or D 2 0. These results are likely to be true in general for any aqueous chemical system. The limitations of using such a dual system for dosimetry in a mixed field is discussed. (author)

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

Characteristics of SiC neutron sensor spectrum unfolding process based on Bayesian inference

Energy Technology Data Exchange (ETDEWEB)

Cetnar, Jerzy; Krolikowski, Igor [Faculty of Energy and Fuels AGH - University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Ottaviani, L. [IM2NP, UMR CNRS 7334, Aix-Marseille University, Case 231 -13397 Marseille Cedex 20 (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

2015-07-01

This paper deals with SiC detector signal interpretation in neutron radiation measurements in mixed neutron gamma radiation fields, which is called the detector inverse problem or the spectrum unfolding, and it aims in finding a representation of the primary radiation, based on the measured detector signals. In our novel methodology we resort to Bayesian inference approach. In the developed procedure the resultant spectra is unfolded form detector channels reading, where the estimated neutron fluence in a group structure is obtained with its statistical characteristic comprising of standard deviation and correlation matrix. In the paper we present results of unfolding process for case of D-T neutron source in neutron moderating environment. Discussions of statistical properties of obtained results are presented as well as of the physical meaning of obtained correlation matrix of estimated group fluence. The presented works has been carried out within the I-SMART project, which is part of the KIC InnoEnergy R and D program. (authors)

Stability of the Hall sensors performance under neutron irradiation

International Nuclear Information System (INIS)

Duran, I.; Hron, M.; Stockel, J.; Viererbl, L.; Vsolak, R.; Cerva, V.; Bolshakova, I.; Holyaka, R.; Vayakis, G.

2004-01-01

A principally new diagnostic method must be developed for magnetic measurements in steady state regime of operation of fusion reactor. One of the options is the use of transducers based on Hall effect. The use of Hall sensors in ITER is presently limited by their questionable radiation and thermal stability. Issues of reliable operation in ITER like radiation and thermal environment are addressed in the paper. The results of irradiation tests of candidate Hall sensors in LVR-15 and IBR-2 experimental fission reactors are presented. Stable operation (deterioration of sensitivity below one percent) of the specially prepared sensors was demonstrated during irradiation by the total fluence of 3.10 16 n/cm 2 in IBR-2 reactor. Increasing the total neutron fluence up to 3.10 17 n/cm 2 resulted in deterioration of the best sensor's output still below 10% as demonstrated during irradiation in LVR-15 fission reactor. This level of neutron is already higher than the expected ITER life time neutron fluence for a sensor location just outside the ITER vessel. (authors)

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.

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)

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)

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)

Development of in-core measurements in the reactor KS-150

International Nuclear Information System (INIS)

Rana, S.B.

1977-01-01

Mapping of the neutron flux density distribution and of the neutron fluence distribution in the KS-150 reactor core was carried out using an in-core measuring system. The system allows the in-service monitoring of important operating properties of the reactor core and fuel elements and consists of a mapping fuel element assembly with built-in SPN detectors, of transmission paths and a computer facility. The measurement of the neutron flux, neutron fluence and temperature fields in the reactor core was carried out during the power start-up of the reactor using self-powered DPZ-1 detectors. The obtained data are given and the axial distribution of neutron flux is graphically represented for different values of burnup at the same configuration of regulating rods, as is the axial distribution of neutron fluence for different configurations of the regulating rods during operation, and the in-service neutron fluence distribution. The maximal fuel temperature of 500.2 degC was found at a distance of 291.2 cm from the upper boundary of the reactor core, at a neutron flux of 1.46x10 14 n/cm 2 s. In comparison with other methods, this method proved easy and quick, the results reliable, reactivity perturbance negligible and the fuel element cost increase a negligible 4%. Neutron flux mapping using in-core self-powered detectors will be performed on a wider scale. (J.P./J.O.)

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.

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)

Development of a simple, low cost, indirect ion beam fluence measurement system for ion implanters, accelerators

Science.gov (United States)

Suresh, K.; Balaji, S.; Saravanan, K.; Navas, J.; David, C.; Panigrahi, B. K.

2018-02-01

We developed a simple, low cost user-friendly automated indirect ion beam fluence measurement system for ion irradiation and analysis experiments requiring indirect beam fluence measurements unperturbed by sample conditions like low temperature, high temperature, sample biasing as well as in regular ion implantation experiments in the ion implanters and electrostatic accelerators with continuous beam. The system, which uses simple, low cost, off-the-shelf components/systems and two distinct layers of in-house built softwarenot only eliminates the need for costly data acquisition systems but also overcomes difficulties in using properietry software. The hardware of the system is centered around a personal computer, a PIC16F887 based embedded system, a Faraday cup drive cum monitor circuit, a pair of Faraday Cups and a beam current integrator and the in-house developed software include C based microcontroller firmware and LABVIEW based virtual instrument automation software. The automatic fluence measurement involves two important phases, a current sampling phase lasting over 20-30 seconds during which the ion beam current is continuously measured by intercepting the ion beam and the averaged beam current value is computed. A subsequent charge computation phase lasting 700-900 seconds is executed making the ion beam to irradiate the samples and the incremental fluence received by the sampleis estimated usingthe latest averaged beam current value from the ion beam current sampling phase. The cycle of current sampling-charge computation is repeated till the required fluence is reached. Besides simplicity and cost-effectiveness, other important advantages of the developed system include easy reconfiguration of the system to suit customisation of experiments, scalability, easy debug and maintenance of the hardware/software, ability to work as a standalone system. The system was tested with different set of samples and ion fluences and the results were verified using

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.

Measurement method of activation cross-sections of reactions producing short-lived nuclei with 14 MeV neutrons

CERN Document Server

Kawade, K; Kasugai, Y; Shibata, M; Iida, T; Takahashi, A; Fukahori, T

2003-01-01

We describe a method for obtaining reliable activation cross-sections in the neutron energy range between 13.4 and 14.9 MeV for the reactions producing short-lived nuclei with half-lives between 0.5 and 30 min. We noted neutron irradiation fields and measured induced activities, including (1) the contribution of scattered low-energy neutrons, (2) the fluctuation of the neutron fluence rate during the irradiation, (3) the true coincidence sum effect, (4) the random coincidence sum effect, (5) the deviation in the measuring position due to finite sample thickness, (6) the self-absorption of the gamma-ray in the sample material and (7) the interference reactions producing the same radionuclides or the ones emitting the gamma-ray with the same energy of interest. The cross-sections can be obtained within a total error of 3.6%, when good counting statistics are achieved, including an error of 3.0% for the standard cross-section of sup 2 sup 7 Al (n, alpha) sup 2 sup 4 Na. We propose here simple methods for measuri...

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

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)

Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

International Nuclear Information System (INIS)

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

2012-10-01

Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131±11 and 124±10 p Sv-cm 2 for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55±4 p Sv-cm 2 for 10 W. (Author)

Estimated neutron dose to embryo and foetus during commercial flight

International Nuclear Information System (INIS)

Chen, J.; Lewis, B. J.; Bennett, L. G. I.; Green, A. R.; Tracy, B. L.

2005-01-01

A study has been carried out to assess the radiation exposure from cosmic-ray neutrons to the embryo and foetus of pregnant aircrew and air travellers in consideration of the radiation exposure from cosmic-ray neutrons to the embryo and foetus. A Monte Carlo analysis was performed to determine the equivalent dose from neutrons to the brain and body of an embryo at 8 weeks and to the foetus at the 3, 6 and 9 month periods. Neutron fluence-to-absorbed dose conversion coefficients for the foetal brain and for the entire foetal body (isotropic irradiation geometry) have been determined at the four developmental stages. The equivalent dose rate to the foetus during commercial flights has been further evaluated considering the fluence-to-absorbed dose conversion coefficients, a neutron spectrum measured at an altitude of 11.3 km and an ICRP-92 radiation-weighting factor for neutrons. This study indicates that the foetus can exceed the annual dose limit of 1 mSv for the general public after, for example, 15 round trips on commercial trans-Atlantic flights. (authors)

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

Comparison of calculations with neutron dosimetry measurements performed at the Oak Ridge Poolside Facility

Energy Technology Data Exchange (ETDEWEB)

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

1981-01-01

The Oak Ridge Poolside Facility (PSF), like the Pool Critical Assembly (PCA), is used for benchmark dosimetry measurements which can serve to validate the transport methods used in calculating the high-energy neutron fluences (> 0.1 MeV) in LWR pressure vessels required to estimate the neutron damage to the pressure vessels in the form of embrittlement. The PSF consists of an arrangement of two water gaps of 4 and 12 cm thickness separated by a simulated thermal shield and followed by a simulated pressure vessel wall and then a void box to represent a reactor cavity. The PSF is driven by the 30 MW ORR reactor, whereas the geometrically similar core of the PCA has a maximum power of only 10 KW. This paper reports the results of some calculated activities and compares them with published PSF measurements performed by HEDL and other laboratories on the so-called Westinghouse surveillance capsule perturbation experiment.

Comparison of calculations with neutron dosimetry measurements performed at the Oak Ridge Poolside Facility

International Nuclear Information System (INIS)

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

1981-01-01

The Oak Ridge Poolside Facility (PSF), like the Pool Critical Assembly (PCA), is used for benchmark dosimetry measurements which can serve to validate the transport methods used in calculating the high-energy neutron fluences (> 0.1 MeV) in LWR pressure vessels required to estimate the neutron damage to the pressure vessels in the form of embrittlement. The PSF consists of an arrangement of two water gaps of 4 and 12 cm thickness separated by a simulated thermal shield and followed by a simulated pressure vessel wall and then a void box to represent a reactor cavity. The PSF is driven by the 30 MW ORR reactor, whereas the geometrically similar core of the PCA has a maximum power of only 10 KW. This paper reports the results of some calculated activities and compares them with published PSF measurements performed by HEDL and other laboratories on the so-called Westinghouse surveillance capsule perturbation experiment

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)

Construction of 144, 565 keV and 5.0 MeV monoenergetic neutron calibration fields at JAERI.

Science.gov (United States)

Tanimura, Y; Yoshizawa, M; Saegusa, J; Fujii, K; Shimizu, S; Yoshida, M; Shibata, Y; Uritani, A; Kudo, K

2004-01-01

Monoenergetic neutron calibration fields of 144, 565 keV and 5.0 MeV have been developed at the Facility of Radiation Standards of JAERI using a 4 MV Pelletron accelerator. The 7Li(p,n)7Be and 2H(d,n)3He reactions are employed for neutron production. The neutron energy was measured by the time-of-flight method with a liquid scintillation detector and calculated with the MCNP-ANT code. A long counter is employed as a neutron monitor because of the flat response. The monitor is set up where the influence of inscattered neutrons from devices and their supporting materials at a calibration point is as small as possible. The calibration coefficients from the monitor counts to the neutron fluence at a calibration point were obtained from the reference fluence measured with the transfer instrument of the primary standard laboratory (AIST), a 24.13 cm phi Bonner sphere counter. The traceability of the fields to AIST was established through the calibration.

Defect cascades produced by neutron irradiation in YBa2Cu3O7-δ

International Nuclear Information System (INIS)

Frischherz, M.C.; Kirk, M.A.; Farmer, J.; Greenwood, L.R.; Weber, H.W.

1992-12-01

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 nm. Defect densities were obtained as a function of neutron fluence between 2 and 8 x 10 21 m -2 (E>0.1 MeV) and compared to damage calculations. The measured defect density was found to scale linearly with fluence and to be 1 x 10 22 m -3 at 2 x 10 21 m -2 . The defect stability was studied at room temperature and through annealing to 400 degrees C. The high fluence regime (∼10 22 m -2 ) was investigated as well

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

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

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

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.)

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

Science.gov (United States)

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

2018-01-01

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

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)

Neutron measuring device

International Nuclear Information System (INIS)

Hatayama, Akiyoshi; Seki, Eiji; Kita, Yoshio; Nishitani, Takeo.

1993-01-01

The device of the present invention concerns measurement for neutrons in a tokamak type thermonuclear device and it can measure total amount of generated neutrons accurately throughout the operation period even if an error is caused in counted values by plasma disruption. That is, the device comprises (1) a means for detecting presence or absence of occurrence of plasma disruption and the time for the initiation of the occurrence, (2) a first data processing means for processing detection signals, (3) a means for detecting neutrons generated in plasmas and (4) a second data processing means for calculating integrated values for the number of neutrons generated from the start to the completion of electric discharge when no disruption occurs and calculating integrated values for the number of generated neutrons from the start of electric discharge to the time at the initiation of occurrence of the disruption when disruption is present. In the thus constituted device, even if an error is caused by frequent occurrence of plasma disruption, total time integrated amount of neutrons generated in the plasmas can be measured accurately. (I.S.)

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

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

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

Studies of the dependence on oxidation thermal processes of effects on the electrical properties of silicon detectors by fast neutron radiation

International Nuclear Information System (INIS)

Li, Zheng; Kraner, H.W.

1991-11-01

High resistivity silicon detectors along with MOS capacitors made on five silicon dioxides with different thermal conditions (975 degree C to 1200 degree C) have been exposed to fast neutron irradiation up to the fluence of a few times 10 14 n/cm 2 . New measurement techniques such as capacitance-voltage (C-V) of MOS capacitors and current-voltage (I-V) of back to back diodes (p + -n - - p + if n - is not inverted to p) or resistors (p + -p-p + if inverted) have been introduced in this study in monitoring the possible type-inversion (n→p) under high neutron fluence. No type-inversion in the material underneath SiO 2 and the p + contact has been observed so far in this work for detectors made on the five oxides up to the neutron fluence of a few times 10 13 n/cm 2 . However, it has been found that detectors made on higher temperature oxides (T≤ 1100 degree C) exhibit less leakage current increase at high neutron fluence (φ ≤ 10 13 n/cm 2 )

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

Characterisation of neutron fields: challenges in assessing the directional distribution

International Nuclear Information System (INIS)

Cauwels, Vanessa; Vanhavere, Filip; Reginatto, Marcel

2014-01-01

The SCK.CEN has carried out neutron field characterisation campaigns at several nuclear reactors. The main goal of these measurement campaigns was to evaluate the performance of different neutron personal dosemeters. To be able to evaluate the performance of neutron personal dosemeters in terms of H p (10), knowledge of the directional distribution is indispensable. This distribution was estimated by placing several personal dosemeters on all six sides of a slab phantom. The interpretation and conversion of this information into a reliable value for H p (10) requires great care. The data were analysed using three methods. In the first approach, a linear interpolation was performed on three perpendicular axes. In the other two approaches, an icosahedron was used to model the angle of incidence of the neutrons and a linear interpolation or a Bayesian analysis was performed. This study describes the limitations and advantages of each of these methods and provides recommendations for their use to estimate the personal dose equivalent H p (10) for neutron dosimetry. Neutron personal dosimetry is complicated by the fact that the neutron dose quantity H p (10) is strongly energy and angular dependent. Instead of simply assuming that the fluence is unidirectional or that the fluence is isotropic, an attempt was made to estimate the directional distribution of the neutron field using a relatively simple measurement procedure. A number of active and passive personal dosemeters were placed on the six faces of a slab phantom and the results were analysed via different algorithms to obtain partial fluences in several directions of incidence. The results from all calculations in this study show the importance of introducing information about the directional distribution of the neutron fluence for the estimation of the personal dose equivalent H p (10). The difference between H p (10) dose estimates carried out using a unidirectional or an isotropic distribution can be of up

Photo-induced current transient spectroscopy for high-resistivity neutron-transmutation-doped silicon

International Nuclear Information System (INIS)

Tokuda, Yutaka; Inoue, Yajiro; Usami, Akira

1987-01-01

Defects in high-resistivity neutron-transmutation-doped (NTD) silicon prior to annealing were studied by photo-induced current transient spectroscopy (PICTS). The thermal-neutron fluence was 9.5 x 10 17 cm -2 to give a resistivity of about 30 Ω after annealing, and the fast-neutron fluence was 9.5 x 10 16 cm -2 . Four traps with thermal emission activation energies of 0.15, 0.41. 0.47 and 0.50 eV were observed in NTD silicon. A trap with the thermal emission activation energy of 0.15 eV was considered to correspond to the divacancy. Although the clustered nature of the defects was observed, PICTS measurements suggest that the material state of high-resistivity NTD silicon is still crystalline and not amorphous. (author)

Experimental evaluation of scattered thermal neutrons from various jig materials for use in fixing detectors for the calibration

International Nuclear Information System (INIS)

Shimizu, Shigeru; Yoshizawa, Michio

2000-05-01

Some jigs to fix detectors are used when radiation measuring instruments are calibrated or reference fluence rates are measured in thermal neutron irradiation fields. In this case, scattered thermal neutrons from the jigs, in particular, which contain hydrogenous materials, may affect the results of the calibration and measurements. In this study, scattered thermal neutrons were measured and calculated to clarify the characteristics of the thermal neutron scattered from various materials which are frequently used for the jigs. A spherical BF 3 -counter of 2-inches in diameter was used in the experiment. Ratios of the fluence of scattered neutrons to primaries (hereinafter, scattering ratio) were evaluated as a function of thickness and size of the materials, as well as the distance from the surface of the materials. The scattering ratios of the jigs that were actually-used in the calibration were also measured in order to select appropriate materials and thickness for the jigs. It was found that the scattering ratios were saturated with increase of thickness and size of the materials. The higher values were observed in the case of PMMA (polymethylmethacrylates) and paraffin since these materials contain more number of hydrogen atoms than the others. The saturated value was obtained 130% for PMMA and paraffin with the thickness of more than 5 cm and the size of 40 cm x 40 cm. The results for the actually-used jigs show that the thinner plate of styrofoam and aluminum reduces the scattering ratio to the value of less than 1%. The obtained data will be useful to improve the accuracy of the calibration of thermal neutron detectors and the measurement of reference fluence rates in thermal neutron irradiation fields. (author)

Experimental evaluation of scattered thermal neutrons from various jig materials for use in fixing detectors for the calibration

Energy Technology Data Exchange (ETDEWEB)

Shimizu, Shigeru; Yoshizawa, Michio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nemoto, Hisashi; Kurosawa, Koji [Institute of Radiation Measurement, Tokai, Ibaraki (Japan)

2000-05-01

Some jigs to fix detectors are used when radiation measuring instruments are calibrated or reference fluence rates are measured in thermal neutron irradiation fields. In this case, scattered thermal neutrons from the jigs, in particular, which contain hydrogenous materials, may affect the results of the calibration and measurements. In this study, scattered thermal neutrons were measured and calculated to clarify the characteristics of the thermal neutron scattered from various materials which are frequently used for the jigs. A spherical BF{sub 3}-counter of 2-inches in diameter was used in the experiment. Ratios of the fluence of scattered neutrons to primaries (hereinafter, scattering ratio) were evaluated as a function of thickness and size of the materials, as well as the distance from the surface of the materials. The scattering ratios of the jigs that were actually-used in the calibration were also measured in order to select appropriate materials and thickness for the jigs. It was found that the scattering ratios were saturated with increase of thickness and size of the materials. The higher values were observed in the case of PMMA (polymethylmethacrylates) and paraffin since these materials contain more number of hydrogen atoms than the others. The saturated value was obtained 130% for PMMA and paraffin with the thickness of more than 5 cm and the size of 40 cm x 40 cm. The results for the actually-used jigs show that the thinner plate of styrofoam and aluminum reduces the scattering ratio to the value of less than 1%. The obtained data will be useful to improve the accuracy of the calibration of thermal neutron detectors and the measurement of reference fluence rates in thermal neutron irradiation fields. (author)

High-energy neutron irradiation of superconducting compounds

International Nuclear Information System (INIS)

Sweedler, A.R.; Snead, C.L.; Newkirk, L.; Valencia, F.; Geballe, T.H.; Schwall, R.H.; Matthias, B.T.; Corenswit, E.

1975-01-01

The effect of high-energy neutron irradiation (E greater than 1 MeV) at ambient reactor temperatures on the superconducting properties of a variety of superconducting compounds is reported. The materials studied include the A-15 compounds Nb 3 Sn, Nb 3 Al, Nb 3 Ga, Nb 3 Ge and V 3 Si, the C-15 Laves phase HfV 2 , the ternary molybdenum sulfide Mo 3 Pb 0 . 5 S 4 and the layered dichalcogenide NbSe 2 . The superconducting transition temperature has been measured for all of the above materials for neutron fluences up to 5 x 10 19 n/cm 2 . The critical current for multifilamentary Nb 3 Sn has also been determined for fields up to 16 T and fluences between 3 x 10 17 n/cm 2 and 1.1 x 10 19 n/cm 2

Activation Measurements for Thermal Neutrons, Part D. U.S. Measurements of 36Cl in Mineral Samples from Hiroshima and Nagasaki

International Nuclear Information System (INIS)

Tore Straume; Alfredo A, Marchetti; Stephen D, Egbert; James A, Roberts; Ping Men; Shoichiro Fujita; Kiyoshi Shizuma; Masaharu Hoshi; G, Rugel; W, Ruhm; G, Korschinek; J. E. McAninch; K. L. Carroll; T. Faestermann; K. Knie; R. E. Martinelli; A. Wallner; C. Wallner

2005-01-01

The present paper presents the 36 Cl measurement effort in the US. A large number of 36 Cl measurements have been made in both granite and concrete samples obtained from various locations and distances in Hiroshima and Nagasaki. These measurements employed accelerator mass spectrometry (AMS) to quantify the number of atoms of 36 Cl per atom of total Cl in the sample. Results from these measurements are presented here and discussed in the context of the DS02 dosimetry reevaluation effort for Hiroshima and Nagasaki atomic-bomb survivors. The production of 36 Cl by bomb neutrons in mineral samples from Hiroshima and Nagasaki was primarily via the reaction 35 Cl(n,γ) 36 Cl. This reaction has a substantial thermal neutron cross-section (43.6 b at 0.025 eV) and the product has a long half-life (301,000 y). hence, it is well suited for neutron-activation detection in Hiroshima and Nagasaki using AMS more than 50 years after the bombings. A less important reaction for bomb neutrons, 39 K(n,α) 36 Cl, typically produces less than 10% of the 36 Cl in mineral samples such as granite and concrete, which contain ∼ 2% potassium. In 1988, only a year after the publication of the DS86 final report (Roesch 1987), it was demonstrated experimentally that 36 Cl measured using AMS should be able to detect the thermal neutron fluences at the large distances most relevant to the A-bomb survivor dosimetry. Subsequent measurements in mineral samples from both Hiroshima and Nagasaki validated the experimental findings. The potential utility of 36 Cl as a thermal neutron detector in Hiroshima was first presented by Haberstock et al. who employed the Munich AMS facility to measure 36 Cl/Cl ratios in a gravestone from near the hypocenter. That work subsequently resulted in an expanded 36 Cl effort in Germany that paralleled the US work. More recently, there have also been 36 Cl measurements made by a Japanese group. The impetus for the extensive 36 Cl and other neutron activation

Validation of MCNP NPP Activation Simulations for Decommissioning Studies by Analysis of NPP Neutron Activation Foil Measurement Campaigns

Directory of Open Access Journals (Sweden)

Volmert Ben

2016-01-01

Full Text Available In this paper, an overview of the Swiss Nuclear Power Plant (NPP activation methodology is presented and the work towards its validation by in-situ NPP foil irradiation campaigns is outlined. Nuclear Research and consultancy Group (NRG in The Netherlands has been given the task of performing the corresponding neutron metrology. For this purpose, small Aluminium boxes containing a set of circular-shaped neutron activation foils have been prepared. After being irradiated for one complete reactor cycle, the sets have been successfully retrieved, followed by gamma-spectrometric measurements of the individual foils at NRG. Along with the individual activities of the foils, the reaction rates and thermal, intermediate and fast neutron fluence rates at the foil locations have been determined. These determinations include appropriate corrections for gamma self-absorption and neutron self-shielding as well as corresponding measurement uncertainties. The comparison of the NPP Monte Carlo calculations with the results of the foil measurements is done by using an individual generic MCNP model functioning as an interface and allowing the simulation of individual foil activation by predetermined neutron spectra. To summarize, the comparison between calculation and measurement serve as a sound validation of the Swiss NPP activation methodology by demonstrating a satisfying agreement between measurement and calculation. Finally, the validation offers a chance for further improvements of the existing NPP models by ensuing calibration and/or modelling optimizations for key components and structures.

Thermal neutron detection by means of an organic solid-state track detector

International Nuclear Information System (INIS)

Doerschel, B.; Streubel, G.

1979-01-01

Thermal neutrons can be detected by means of organic solid-state track detectors if they are combined with radiators in which charged secondary particles are produced in neutron interaction processes. The secondary particles can produce etchable tracks in the detector material. For thermal neutron fluence determination from the track densities, the thermal neutron sensitivity was calculated for cellulose triacetate detectors with LiF radiators, taking into account energy and angular distribution of the alpha particles produced in the LiF radiator. This value is in good agreement with the sensitivity measured during irradiation in different neutron fields if corrections are considered the production of etchable or visuable tracks. Measuring range and measuring accuracy meet the requirements of thermal neutron detection in personnel dosimetry. Possibilities of extending the measuring range are discussed. (author)

Evidos: optimisation of individual monitoring in mixed neutron/photon fields at workplaces of the nuclear fuel cycle

International Nuclear Information System (INIS)

Luszik-Bhadra, M.; Reginatto, M.; Schuhmacher, H.; Lacoste, V.; Muller, M.; Boschung, M.; Fiechtner, A.; Coeck, M.; Vanhavere, F.; Curzio, G.; D'errico, F.; Kyllonen, J.E.; Lindborg, L.; Molinos, C.; Tanner, R.; Derdau, D.; Lahaye, Th.

2005-01-01

Within its 5. Framework Programme, the EC is funding the project EVIDOS ('Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields'). The aim of this project is the optimisation of individual monitoring at workplaces of the nuclear fuel cycle with special regard to neutrons. Various dosemeters for mixed field application - passive and new electronic devices - are tested in selected workplace fields in nuclear installations in Europe. The fields are characterised using a series of spectrometers that provide the energy distribution of neutron fluence (Bonner spheres) and newly developed devices that provide the energy and directional distribution of the neutron fluence. Results from the first measurement campaign, carried out in simulated workplace fields (IRSN, Cadarache. France), and those of a second measurement campaign, carried out at workplaces at a boiling water reactor and at a storage cask with used fuel elements (Kernkraftwerk Kriimmel, Germany), are described. (authors)

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)

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

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

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.

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)

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.

The new remcounter LB6411: Measurement of neutron ambient dose equivalent H*(10) according to ICRP60 with high sensitivity

International Nuclear Information System (INIS)

Klett, A.; Burgkhardt, B.

1996-01-01

Since the International Commission on Radiological Protection has issued in publication ICRP60 new recommendations on radiation protection quantities, in neutron monitoring there is now increasing Interest in commercially available instruments optimized and calibrated for the measurement of ambient dose equivalent H*(10). Therefore within a joint cooperation between the Research Center Karlsruhe and EG ampersand G Berthold the neutron-dose-rate meter LB6411 was newly developed. The detector system with integrated electronics has a 3 He proportional counter tube centered in a moderating sphere. The response between thermal energies and 20 MeV was optimized with the help of extensive MCNP Monte-Carlo calculations. The instrument has extremely high sensitivity of approximately 3 counts per nSv and can be used both as a portable or as a stationary neutron monitor. Fluence responses and angular dependencies had been measured in monoenergetic neutron beams provided by the Physikalisch-Technische Bundesanstalt (PTB) in Braunschweig, Germany. The ambient dose equivalent response of the LB6411 is reported over the whole energy range

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.

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.

Experimental characterization of the neutron spectra generated by a high-energy clinical LINAC

Energy Technology Data Exchange (ETDEWEB)

Amgarou, K., E-mail: khalil.amgarou@uab.e [Institut de Radioprotection et de Surete Nucleaire (IRSN), Laboratoire de Metrologie et de Dosimetrie des Neutrons, F-13115 Saint Paul-Lez-Durance (France); Lacoste, V.; Martin, A. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Laboratoire de Metrologie et de Dosimetrie des Neutrons, F-13115 Saint Paul-Lez-Durance (France)

2011-02-11

The production of unwanted neutrons by electron linear accelerators (LINACs) has attracted a special attention since the early 50s. The renewed interest in this topic during the last years is due mainly to the increased use of such machines in radiotherapy. Specially, in most of developing countries where many old teletherapy irradiators, based on {sup 60}Co and {sup 137}Cs radioactive sources, are being replaced with new LINAC units. The main objective of this work is to report the results of an experimental characterization of the neutron spectra generated by a high-energy clinical LINAC. Measurements were carried out, considering four irradiation configurations, by means of our recently developed passive Bonner sphere spectrometer (BSS) using pure gold activation foils as central detectors. This system offers the possibility to measure neutrons over a wide energy range (from thermal up to a few MeV) at pulsed, intense and complex mixed n-{gamma} fields. A two-step unfolding method that combines the NUBAY and MAXED codes was applied to derive the final neutron spectra as well as their associated integral quantities (in terms of total neutron fluence and ambient dose equivalent rates) and fluence-averaged energies.

Comparison of neutron and gamma irradiation effects on KU1 fused silica monitored by electron paramagnetic resonance

Energy Technology Data Exchange (ETDEWEB)

Bravo, D. [Department Fisica de Materiales, Universidad Autonoma de Madrid, E-28049 Madrid (Spain)], E-mail: david.bravo@uam.es; Lagomacini, J.C. [Department Fisica de Materiales, Universidad Autonoma de Madrid, E-28049 Madrid (Spain); Leon, M.; Martin, P. [Materiales para Fusion, CIEMAT, Avda. Complutense 22, E-28040 Madrid (Spain); Martin, A. [Department Fisica e Instalaciones, ETS Arquitectura UPM, E-28040 Madrid (Spain); Lopez, F.J. [Department Fisica de Materiales, Universidad Autonoma de Madrid, E-28049 Madrid (Spain); Ibarra, A. [Materiales para Fusion, CIEMAT, Avda. Complutense 22, E-28040 Madrid (Spain)

2009-06-15

Electron paramagnetic resonance (EPR) studies have been carried out on KU1 fused silica irradiated with neutrons at fluences 10{sup 21} and 10{sup 22} n/m{sup 2}, and gamma-ray doses up to 12 MGy. The effects of post-irradiation thermal annealing treatments, up to 850 deg. C, have also been investigated. Paramagnetic oxygen-related defects (POR and NBOHC) and E'-type defects have been identified and their concentration has been measured as a function of neutron fluence, gamma dose and post-irradiation annealing temperature. It is found that neutrons at the highest fluence generate a much higher concentration of defects (mainly E' and POR, both at concentrations about 5 x 10{sup 18} spins/cm{sup 3}) than gamma irradiations at the highest dose (mainly E' at a concentration about 4 x 10{sup 17} spins/cm{sup 3}). Moreover, for gamma-irradiated samples a lower treatment temperature (about 400 deg. C) is required to annihilate most of the observed defects than for neutron-irradiated ones (about 600 deg. C)

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

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)

Measurement of fission cross-section for the {sup 232}Th(n,f){sup 141}Ba reaction induced by neutrons around 14 MeV

Energy Technology Data Exchange (ETDEWEB)

Lan, Chang-Lin; Fang, Kai-Hong [Lanzhou University, School of Nuclear Science and Technology, Lanzhou, Gansu Province (China); Lanzhou University, Engineering Research Center for Neutron Application, Ministry of Education, Lanzhou, Gansu Province (China); Liu, Shuang-Tong; Lv, Tao; Wang, Qiang; Zhang, Zheng-Wei [Lanzhou University, School of Nuclear Science and Technology, Lanzhou, Gansu Province (China); Lai, Cai-Feng [Chinese Academy of Engineering Physics, Institute of Nuclear Physics and Chemistry, Mianyang, Sichuan Province (China)

2016-11-15

The fission cross-section of the {sup 232}Th(n,f){sup 141}Ba reaction induced by neutrons around 14 MeV was measured precisely with the neutron activation and off-line gamma-ray spectrometric technique. Neutron fluence was monitored on-line using the accompanying α-particles from the {sup 3}H({sup 2}H,n){sup 4}He reaction, whereas the neutron energies were measured by the method of cross-section ratios of {sup 90}Zr(n,2n){sup 89}Zr to {sup 93}Nb(n,2n){sup 92m}Nb reactions. The experimentally determined {sup 232}Th(n,f){sup 141}Ba reaction cross-sections were 12.2 ± 0.4 mb at E{sub n} = 14.1 ± 0.3 MeV, 13.0 ± 0.5 mb at E{sub n} = 14.5 ± 0.3 MeV and 13.3 ± 0.5 mb at E{sub n} = 14.7 ± 0.3 MeV, respectively. (orig.)

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

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.

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

Effect of neutron irradiation on single crystal V3Si

International Nuclear Information System (INIS)

Viswanathan, R.; Caton, R.; Cox, D.E.; Guha, A.; Sarachik, M.P.; Smith, F.W.; Testardi, L.R.

1977-01-01

We We have investigated the effect of successive neutron irradiation up to a fluence of approximately 2 x 10 19 n/cm 2 , by measurements of heat capacity, susceptibility, resistivity, acoustic velocity and neutron diffraction in a single crystal V 3 Si. We find that for low level doses (phi t greater than or equal to 3.5 x 10 18 n/cm 2 ) (a) the structural transformation is very sensitive, whereas the suerconducting transition temperature, T/sub c/, is hardly affected, and (b) except for low temperature heat capacity, most of the other measurements show very little change. For the highest fluence of 2 x 10 19 n/cm 2 used to date, the T/sub c/ dropped to 7.5 K with large changes in the linear heat capacity coefficient, magnetic susceptibility and sound velocity. These results are discussed briefly in this paper

Neutron-induced Single Event Upset on the RPC front-end chips for the CMS experiment

Energy Technology Data Exchange (ETDEWEB)

Abbrescia, M.; Colaleo, A.; Iaselli, G.; Loddo, F.; Maggi, M.; Marangelli, B.; Natali, S.; Nuzzo, S.; Pugliese, G.; Ranieri, A.; Romano, F.; Altieri, S.; Belli, G.; Bruno, G.; Guida, R.; Merlo, M.; Ratti, S.P.; Riccardi, C.; Torre, P.; Vitulo, P. E-mail: paolo.vitulo@pv.infn.it; De Bari, A.; Manera, S

2002-05-21

Neutrons from a reactor and from a cyclotron have been used to characterise the CMS Resistive Plate Chambers (RPCs) front-end chip to neutron-induced damaging events. Single Event Upset (SEU) cross-sections have been measured up to 60 MeV for different chip thresholds. Tests at a reactor were done with an integrated fast (E{sub n}>3 MeV) neutron fluence of 1.7x10{sup 10} cm{sup -2} and a thermal neutron fluence of 9.5x10{sup 11} cm{sup -2}. High-energy neutrons from a cyclotron were used up to a fluence of 10{sup 12} cm{sup -2}. Data indicate the existence of a chip SEU sensitivity already at thermal energy and a saturated SEU cross-section from 3 to 60 MeV. Values of the SEU cross-sections from the thermal run well agree with those obtained by another CMS group that uses the same technology (0.8 {mu}m BiCMOS) though with different architecture. Cross-sections obtained with fast neutrons (from 3 MeV to about 10 MeV) are consistently higher by one order of magnitude compared to the thermal one. The average time between consecutive SEU events in each chip of the CMS barrel RPCs can be estimated to be 1 h.

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.

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)

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.

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)

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)

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})

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

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)

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

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...

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.

Simulations of the neutron energy-spectra at the Olympus Gate Environmental Monitoring Station due to historical Bevatron operations

International Nuclear Information System (INIS)

Donahue, R.J.; Thomas, R.H.; Zeman, G.H.

2001-01-01

Offsite neutron fluences resulting from Bevatron operations reached a maximum in 1959, prior to the addition of a permanent concrete roof shield, which was constructed in 1962. From the first operation of the Bevatron measurements of neutron fluence were made at locations around the perimeter of the Lawrence Berkeley National Laboratory (LBNL) campus. Since the late 1950's measurements made at several locations, and particularly at the site of what is now called the Olympus Gate Environmental Monitoring Station, have been routinely reported and published. Early measurements were used to establish the shape of the neutron-energy spectrum from which an energy-averaged fluence-to-dose equivalent conversion coefficient could be derived. This conversion coefficient was then applied to a measured total neutron fluence to obtain the appropriate dose equivalent quantity required by regulation. Recent work by Thomas et al. (2000) have compared the early conversion coefficients used in the sixties with those accepted today and suggest suggested that ''the dose equivalents reported in the late fifties and early sixties were conservative by factors between two and four. In any current review of the historical data, therefore it would be prudent to reduce the reported dose equivalents by at least a factor of two.'' However, that analysis was based on the ''state of the art'' neutron energy-spectra of the '60s. This paper provides a detailed knowledge of the neutron energy spectrum at the site boundary paper thus removing any uncertainty in the analysis of Thomas et al., which might be caused by the use of the early neutron energy-spectra. Detailed Monte Carlo analyses of the interactions of 6.2 GeV protons in thick, medium-A targets are described. In the computer simulations, neutrons produced were allowed to scatter in the atmosphere. Detailed neutron energy spectra were calculated at a distance and elevation corresponding to the location of the Olympus Gate EMS. Both older

Neutron cross section standards for the energy region above 20 MeV

International Nuclear Information System (INIS)

1991-01-01

These proceedings of a specialists' meeting on Neutron cross section standards for the energy region above 20 MeV are divided into 6 sessions bearing on: - session 1: status of the date base for (n-p) scattering (2 conferences) - session 2: status of nucleon-nucleon phase shift calculations (1 conference) - session 3: recent and planned experimental work on n-p cross section measurements and facilities (7 conferences) - session 4: Instruments for utilizing the H (n.n) standard for neutron fluence measurement (4 conferences) - session 5: proposal for other neutron cross-section standards (4 conferences) - session 6: monitor reactions for radiation dosimetry (3 conferences)

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

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

A neutron detector for measurement of total neutron production cross sections

International Nuclear Information System (INIS)

Sekharan, K.K.; Laumer, H.; Kern, B.D.; Gabbard, F.

1976-01-01

A neutron detector has been constructed and calibrated for the accurate measurement of total neutron production cross sections. The detector consists of a polyethylene sphere of 60 cm diameter in which eight 10 BF 3 counters have been installed radially. The relative efficiency of this detector has been determined for average neutron energies from 30 keV to 1.5 MeV by counting neutrons from 7 Li(p, n) 7 Be. By adjusting the radial positions of the BF 3 counters in the polyethylene sphere the efficiency for neutron detection was made nearly constant for this energy range. Measurement of absolute efficiency for the same neutron energy range has been done by counting the neutrons from 51 V(p, n) 51 Cr and 57 Fe(p, n) 57 Co reactions and determining the absolute number of residual nuclei produced during the measurement of neutron yield. Details of absolute efficiency measurements and the use of the detector for determination of neutron production cross sections are given. (Auth.)

Neutron irradiation characteristic tests of oxygen sensors using zirconia solid electrolyte

International Nuclear Information System (INIS)

Hiura, Nobuo; Endou, Yasuichi; Yamaura, Takayuki; Niimi, Motoji; Hoshiya, Taiji; Saito, Junichi; Souzawa, Shizuo; Ooka, Norikazu; Kobiyama, Mamoru.

1997-03-01

In the Department of JMTR of Japan Atomic Energy Research Institute (JAERI), the in-situ measuring technique of oxygen potential has been being developed to study the chemical behavior of high burn-up fuel base-irradiated in the Light Water Reactor. In this test for development of the technique, oxygen sensors using zirconia solid electrolyte stabilized by MgO, CaO and Y 2 O 3 , named MSZ, CSZ and YSZ, respectively, were irradiated by neutrons in the Japan Materials Testing Reactor (JMTR) of JAERI and the characteristics of electromotive force of these sensors under and after irradiation were discussed. From the experimental results, the electromotive force of YSZ sample under irradiation decreased with an increase in irradiation fluence within a range of neutron fluence (E>1 MeV) up to 1 x 10 23 m -2 . The electromotive force of MSZ sensor irradiated with neutron fluences (E>1 MeV) up to 9 x 10 21 m -2 was almost equal to the theoretical value of the electromotive force. It was shown that after irradiation, a decrease in the electromotive force of CSZ sensor was smaller than those of MSZ and YSZ sensors, although the electromotive forces of MSZ, CSZ and YSZ sensors were smaller than the theoretical value. (author)

Final report for the 2nd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 2 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

2008-01-15

This report describes a neutron fluence assessment performed for the Yonggwang Unit 2 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During Cycle 16 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

Final report for the 1st ex-vessel neutron dosimetry installations and evaluations for Kori unit 2 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 2 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 20 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 20.

Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Li, Nam Jin; Hong, Joon Wha

2007-01-15

This report describes a neutron fluence assessment performed for the Yonggwang Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 16 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

2008-01-15

This report describes a neutron fluence assessment performed for the Kori Unit 2 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 21 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 21.

Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

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

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

Neutron metrology in the fuel assemblies of a material test reactor

International Nuclear Information System (INIS)

Voorbraak, W.P.; Paardekoper, A.; Polle, A.N.; Freudenreich, W.E.

1993-08-01

Results are presented of detailed thermal and fast neutron measurements performed in all fuel and control assemblies of the HFR in Petten. The results give information about deviations of a general shape of vertical thermal and fast fluence rate distributions due to material transitions in the reactor core and different control assembly settings. Further it is demonstrated that the ratio of fast and thermal fluence rate at the various monitor positions in the assemblies give useful information for the (relative) local burn-up of the fuel. (orig.)

Neutron spectrum measurement by TOF